ML093160257

Submittal of Revision 21 to the Final Safety Analysis Report (Updated), 10 CFR 50.59 Evaluation Summary Report, Technical Specifications Bases Changes, and Report Consistent with 10 CFR 54.37(b)

ML093160257
Person / Time
Site:	Nine Mile Point
Issue date:	10/26/2009
From:	Belcher S Constellation Energy Group, Nine Mile Point
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML093160257 (626)

Text

Belcher Sam Belcher P.O. Box 63 P.O.

President-Nine Mile Point Vice President-Nine Point Lycoming, Lycoming, New New York.13093 York.13093 315.349.5200 315.349.5200 315.349.1321 315.349.1321 Fax o .::.:.

Constellation Energy Nine Mile Point Nuclear Constellation Station Energy-October 26, 2009 U. S.S. Nuclear Regulatory Regulatory Commission Washington, DC 20555-0001 ATTENTION: Document Control Control Desk

SUBJECT:

Nine Mile Point Nuclear Station Unit No. 1; Docket No. 50-220 UIiit No.1; Submittal of Revision 21 to the Final Safety Analysis Report (Updated),

10 CFR 50.59 Evaluation Summary Report, Technical Specifications Specifications Bases Changes, and Report Consistent Consistent with 10 CFR 54.37(b) 54.37(b)

Pursuant Pursuant to the requirements of 10 CFR 50.71(e), 10 CFR 50.59(d)(2),50.59(d)(2), and the Nine Nine Mile Point Unit 1 (NMP1)

(NMP1) Technical Specifications (TS) Bases Control Program (TS 6.5.6), Nine Mile Point Nuclear Technical Specifications Nuclear Station, LLC (NMPNS)

(NMPNS) hereby submits the following:

• Revision 221r to the NMPI Revision NMPI Final Safety Safety Analysis Report (Updated) (UFSAR).

• • NMPI Technical Technical Specifications Specifications Bases Changes.

The UFSAR Revision 21 pages pages are contained contained in Attachment Attachment 1. TheThe UFSAR revision contains contains changes changes made made since the submittal submittal of Revision Revision 20 in October October 2007. The revision reflects reflects all changes changes up to April 2009. No 10 CFR 50.59 evaluations evaluations were completed completed for NMP1 between forNMPl between April 2007 2007 and April 2009.

Attachment Attachment 2 contains the revised Technical Specifications Bases pages, Technical Specifications pages, which incorporate incorporate changes made since April 2007.

Consistent Consistent withwith 10 CFR 54.37(b),

54.37(b), Attachment Attachment 3 contains contains aa report describing describing how the effects effects of aging of of newly-identified newly-identified structures, systems, systems, or components components will be be managed, managed, such that the intended intended functions described described in 10 10 CFR 54.454.4 will will be effectively effectively maintained maintained during during the license license renewal renewal period period of extended extended operation.

operation.

Document Control Desk Document Control Desk October 26, 2009 2009 Page 2 Should you have any questions regarding regarding the information in this submittal, please contact T. F. Syrell, (315) 349-5219.

Licensing Director, at (315)

Very truly yours, CERTIFICATION:

CERTIFICATION:

President-Nine Mile Point and that the information contained in this I,I, Sam Belcher, certify that I am Vice President-Nine submittal accurately accurately presents changes changes made since the previous submittal submittal necessary to reflect information information and analysis submitted to the Commission or prepared pursuant to Commission requirement.

V, SLB/RJC Attachments: 1. Final Safety Analysis Report (Updated) Pages Pages

2. Revised Technical Specifications Bases Pages Technical Specifications

3. Report Consistent with 10 CFR 54.37(b) on How Effects of Aging of Newly-Identified Structures, Systems, or Components are Managed Managed cc: S. J. Collins, NRC S.

R. V. Guzman, NRC Resident Inspector, NRC Resident

c..-' , ,;

Control Desk Document Control Desk October October 26, 2009 2009 Page 33 bcc: (w/o Attachments 1 and 2)

(w/o C. W. Fleming, Esquire S. Belcher S. Belcher T. A. Lynch S. Montgomery B. S.

W. C. Byrne W.C.

T. F. Syrell J. J. Dosa M. Fallin R. W. Saunderson NMP1L NMPIL23952395 COMMITMENTS IDENTIFIED IN THIS CORRESPONDENCE:'

COMMITMENTS CORRESPONDENCE:

• 0 None Requirements for Responses -- NOV/Order Posting Requirements NOV/Order No

~,

.' ATTACHMENT ATTACHMENT 1 FINAL SAFETY SAFETY ANALYSIS ANALYSIS REPORT REPORT (UPDATED) PAGES e*

Nine Mile Point Nuclear Nuclear Station, LLC October 26, 2009 October

• ATTACHMENT ATTACHMENT 2 REVISED TECHNICAL TECHNICAL SPECIFICATIONS SPECIFICATIONS BASES PAGES .

• Nine Mile Point Nuclear Nuclear Station, LLC October 26, 2009 October 26,2009

POINT UNIT 1 TECHNICAL MILE POINT NINE MILE TECHNICAL SPECIFICATIONS BASES BASES INSERTION INSTRUCTIONS INSERTION INSTRUCTIONS

• Remove the pages listed in the Remove Remove column and replace them with the pages listed in the Insert Remove column column.

column.

additional page being If there is an additional (-) will be being added to the Technical Specification Bases, dashes (-----)

Likewise, if a page shown in the Remove column. Likewise, page is being removed replacement dashes (--)

removed with no replacement (-----)

will be shown in the Insert column.

REMOVE REMOVE INSERT INSERT LEP-l LEP-1 LEP-l LEP-1 LEP-2 LEP-2 LEP-3 LEP-3 LEP-4 LEP-4 LEP-5 LEP-5 38 38 38 40 40 40 41 41 41a 41a 42 42 49 49 49 100 100 100 1100a OOa 115 115 115 l15a lISa 140 140 140 141 141 142 142 142 150 150 150 167 167 167 167a 167a 169 169 172 172 172 176 176 176 177 177 177 180 180 180 S180a 180a S180b 180b 181 181 181 249 249 250 250 250 250a 250a 258 258 258 258a 258a 273 273 296 296

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NMP1 NMP1 LEP-1 LEP-1 Amendment 203 Amendment (10/01/09) 203(10/01/09)

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NMP1 NMP1 LEP-4 LEP-4 Amendment 203 Amendment 203(10/01/09)

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LEP-5 LEP-5 Amendment 203 Amendment 203(10/01/09)

(10/01/09)

• BASES FOR 3.1.1 AND BASES AND 4.1.1 CONTROL CONTROL ROD SYSTEM inoperable rod patterns The allowable inoperable patterns will be determined using information information obtained obtained in the startup test program supplemented by calculations. During initial startup, the reactivity supplemented reactivity condition condition of the as-built core will be determined.

Also, sub-critical sub-critical patterns of widely separated withdrawn control rods will be observed in the control rod sequences sequences being used. The observations, observations, together with calculated strengths strengths of the strongest control rods in these patterns will comprise a set of allowable allowable separations of malfunctioning malfunctioning rods. During the fuel cycle, similar observations observations mademade during any cold shutdown shutdown can be used to update increase the allowable patterns.

update and/or increase The number of rods permitted to be valved out of service could be many more than the six allowed allowed by thethe specification, particularly specification, particularly late in the operating cycle; however, the occurrence occurrence of more than six could be indicative indicative of a generic problem and the reactor will be shut down.

generic Placing the reactor in the shutdown down. Placing shutdown condition condition inserts the control rods and accomplishes accomplishes the objective objective of the specifications specifications on control rod operability. This operation operation is normally normally expected expected to be accomplished accomplished within ten hours.

Control rod insertion insertion capability demonstrated by inserting capability is demonstrated inserting each each partially partially or fully withdrawn withdrawn control rod at least one one notch and observing that the control rod moves. The control rod may then be returned to its original position. This This ensures that the control rod is not stuck and is free to insert on a scram signal. This surveillance surveillance is not required when thermal power is less than or equal to the low power set point (LPSP) of the RWM, RWM, since notch notch insertion insertion may not be be compatible compatible with the requirements requirements of the RWM.

RWM. The 31 day surveillance test frequency takes into account operating operating experience related to changes experience changes in CRD performance.

performance. This surveillance surveillance requirement allows 31 days after withdrawal of the control rod concurrent with thermalthermal power greater greater than the LPSP of the RWM to perform the surveillance.

The requirement to exercise exercise control rods at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the event power power operation operation is continuing with two or more control rods which are valved valved out of service service or one fully or partially withdrawn withdrawn control rod which can not be be moved, provides a reasonable provide assurance reasonable time to test the control rods and provide remaining assurance of the reliability of the remaining control rods.

control rods.

b. Control Rod Withdrawal (1) Control rod dropout accidents accidents as discussed in Appendix E* can lead to significant core damage. If If coupling integrity is maintained, the possibility of a rod dropout accident is eliminated.

eliminated. The overtravel position feature provides provides a positive positive check as only uncoupled drives may reach this position. Neutron instrumentationinstrumentation response to rod movement provides provides an indirect verification verification that the rod is coupled to its drive. Details of the control rod drive coupling are given in Section IV. B.6. 1*.

IV.B.6.1*.

• FSAR

• FSAR 38 38 AMENDMENT NO. 142, Revision AMENDMENT Revision 25 (A200)

• BASES FOR 3.1.1 BASES 3.1.1 AND

• CONTROL ROD SYSTEM AND 4.1.1 CONTROL:.

6

a. A startup inter-assembly inter-assembly local power peaking factor factor of 1.30 or of 1.30 less.(6))

or less.(

b. An end of o(cycle delayed neutron cycle delayed neutron fraction of 0.005.

c. A beginning beginning of life Doppler reactivity feedback.

d. The Technical insertion rate.

Technical Specification rod scram insertion

e. The maximum possible ft/sec).

possible rod drop velocity (3.11 ftlsec).

f.f. The design accident accident and scram reactivity shape function.

g. The moderator moderator temperature temperature at which criticality occurs.

recognized that these bounds are conservative It is recognized conservative with respect to expected operating conditions. If expected operating If any one of the anyone the above conditions is not satisfied, a more detailed calculation will be done to show compliance compliance with the 280 cal/gm design limit.

In most cases the worth of in-sequence in-sequence rods or rod segments segments will be substantially less than 0.013 t..k. Ak. Further, the the addition of 0.013 Ak t..k worth of reactivity-reactivity- as a result of a rod drop in conjunction with the actual actual values values of the other important accident parameters described accident analysis parameters described- above above would most likely result in a peak fuel enthalpy substantially substantially less than the 280 cal/gm design limit. However, the 0.013 t..k Ak limit is applied in order order to allow room for future reload changes changes and ease of verification without repetitive Technical Specification changes.

Technical Specification Should Should aa control rod drop. accident accident (CRDA) result in a peak fuel energy content content of 280 cal/gm, less than 660 (7 x 7) fuel rods were conservatively conservatively estimated estimated to perforate. For 8 x 8 fuel, less than 850 rods were conservatively conservatively estimated to perforate, which is bounding bounding for GEl GE111 9 x 99 fuel. As noted in UFSAR UFSAR Section XV-C.4.2, XV-C.4.2, CRDA results for banked position withdrawal withdrawal sequence sequence (BPWS) plants have been statistically statistically analyzed and show that, in all cases, the peak fuel enthalpy enthalpy in a CRDA would be much less than the 280 cal/gm design limit. Thus, the CRDA has been deleted from been deleted the standard GE BWR reload package package for BPWS plants. The radiological consequences radiological consequences of a CRDA haVe been shown have to remain remain well within the regulatory limits'.

limits.

AMENDMENT AMENDMENT NO. 142, Revision Revision 21 (A194) 40 40

• BASES FOR 3.1.1 BASES AND 4.1.1 CONTROL 3.1.1 AND CONTROL ROD SYSTEM The RWM provides automatic supervision to assure that out-of-sequence out-of-sequence control rods will not be be withdrawn withdrawn or inserted during startup startup or shutdown, such that only specified specified control rod sequences sequences and relative positions are allowed allowed over the operating range from all rods inserted to 10% 10% RTP. ItIt serves serves as an independent independent backup of the normal withdrawal procedure withdrawal procedure followed by the operator. With the the RWM inoperable RWM inoperable during a reactor startup, the operator operator is still capable of enforcing the prescribed control rod sequence, however, the overall reliability reliability is reduced because a single operator error can result in violating the control rod sequence.

If the RWM becomes inoper?bleinoperable after at least 12 control rods have been withdrawn, startup may continue ifif the RWM function is performedperformed manually and aa required verification verification of control rod rod movement movement by a second licensed licensed operator operator (Reactor Operator or Senior Reactor (Reactor Operator Reactor Operator) or by a qualified member of the technical staff (a qualified qualified shift technical advisor or reactor engineer) is performed. Also, if the RWM is inoperable prior to commencement commencement of startup, or becomes inoperable inoperable during a startup, prior to complete complete withdrawal of the first 12 12 control rods, startup may continue continue if the the RWM function is performed RWM performed manually and a required required verification verification of control movement by aa control rod movement second licensed licensed operator (Reactor Operator Operator or Senior ReactorReactor Operator) or by a qualified member qualified member of the technical staff (a qualified shift technical technical advisor advisor or reactor engineer) is performed, performed, and provided that aa startup with the RWM RWM inoperable inoperable was not performed performed in the last 12 months. In In both procedural control is exercised cases, procedural exercised by verifying all control rod positions after the withdrawal withdrawal of each group, prior to proceeding proceeding to the next group. Allowing substitution substitution of a second licensed licensed operator or other member of the technical other qualified member inoperability recognizes technical staff in case of RWM inoperability recognizes the capability adequately monitor proper capability to adequately sequencing in an alternate manner proper rod sequencing manner without unduly restricting plant operations. Above 10% 10% power, therethere is no requirement operable requirement that the RWM be operable since the control rod drop accidentaccident with out-of-sequence out-of-sequence rods will result in a peak fuel energy energy content of less than 280 cal/gm. The allowed allowed frequency requirements of performing performing a reactor startup with the the RWM inoperable (i.e.,

RWM inoperable (i.e., if if not performed performed in the last 12 months) minimizes the number of reactor startups performed with performed with the the RWM RWM inoperable.

inoperable.

AMENDMENT 142 AMENDMENT 41 Revision Revision 4-(A178),

4 (A179), 22 (A196)

• BASES FOR 3.1.1 AND BASES 0S CONTROL ROD SYSTEM.

AND 4.1.1 CONTROL SYSTEM 0

(4) The source range monitor (SRM) system performs monitor (SRM) automatic safety function. It performs no automatic provide the It does provide the operator with aa visual indication of neutron level which is needed operator knowledgeable and efficient needed for knowledgeable efficient reactor startup at low neutron levels. The results of reactivity accidents are functions of the initial startup initial neutron flux.

The requirement transient begins at or above the initial value of 10-8 of requirement of at least 3 cps assures that any transient of rated power used in the analyses of transients analyses transients from cold conditions. One operable SRM operable SRM channel would be be monitor the approach

. adequate to monitor approach to critical using homogeneous scattered control rods. A homogeneous patterns of scattered operable SRMs is required as an added conservatism.

minimum of three operable conservatism.

c. Scram Insertion Insertion Times Times The revised scram insertion times have been established as the limiting condition for operation operation since since thethe postulated rod drop analysis and associated associated maximum in-sequence control rod worth are based on the revised maximum in-sequence scram insertion times. The specified times are based on design design requirements for control rod scramscram at reactor reactor pressures pressures above 950 psig. For reactor pressures above above 800 psig and below below 950 psig the measured scram times may be longer. The analysis discussed discussed in the next paragraph is still valid since the use of the revised scram insertion times would result in greater margins to safety valves lifting.

insertion times previously selected The insertion selected were based on the large large number number of actual scrams of prototype prototype control rod drive drive mechanisms mechanisms as discussed discussed in Section IV-B.6.3*.

IV-B.6.3*. Rapid control rod insertion following a a demand demand to scram will terminate terminate Station Station transients before before any possibility of damage to the core is approached.

approached. The primary consideration consideration in setting scram time is to permit rapid termination termination of steam generation generation following an isolation transient main-steam-line closure or turbine transient (i.e., main-steam-Iine turbine trip without bypass) such that operation operation of solenoid-actuated solenoid-actuated relief valves will prevent prevent the safety valves from lifting.

Analyses presented in Appendix Appendix E-l*,

E-1*, the Second Supplement and the Technical Second Supplement Supplement to Petition to Increase Technical Supplement Increase Power Power Level were based were based on times which are slower than the proposed revised times.

The scram scram times generated generated at each refueling outage outage when compared compared to previous scram times demonstrate demonstrate that the control rod drive scram function has not deteriorated.

deteriorated.

• FSAR

• FSAR Revision 22 (A Revision (A196) 196) 41a 41 a

• BASES FOR BASES FOR 3.1.1 AND 4.1.1 CONTROL CONTROL ROD SYSTEM

d. Accumulators Control Rod Accumulators The basis basis for this specification was not described in the FSAR and, therefore, is presented in its entirety. Requiring no more than one malfunctioned malfunctioned accumulator in any nine-rod square array is based on a series of XY PDQ-4 quarter quarter core problems of a cold, clean core. The worst one in a nine-rod withdrawal sequence resulted in a keff <1.0--other repeating rod sequence with more rods withdrawn withdrawn resulted in keff >1.0. At reactor pressures pressures in excess of 800 psig, even those control rods with malfunctioned accumulators malfunctioned accumulators will be able to meet required scram insertion insertion times due to the action of reactor pressure. In In addition, addition, they may be normally normally inserted using the control-rod-drive control-rod-drive hydraulic system. Procedural Procedural control will assure that control rods with malfunctioned malfunctioned accumulators will be spaced in a one-in-nine array rather than grouped together.

e. Discharge Volume Scram Discharge Volume The scram The scram discharge discharge volume is required to be OPERABLE OPERABLE so that it will be available needed to accept discharge available when needed discharge water from the control rods during a reactor scram, isolate the reactor coolant system from the containment when required, and to comply with the requirements of the NRC Confirmatory letter of June 24, 1983. The fill/drain test was determined to be an acceptable alternative acceptable alternative to a reactor scram test at approximately approximately 50% ROD DENSITY. Performance Performance of a water fill/drain test during cold shutdown shutdown will verify that the Scram Scram Discharge OPERABLE and instrument Discharge Volume is OPERABLE instrument lines are not plugged. The The volume comparison comparison test of water drained drained equal water water used to fill will demonstrate demonstrate that there is no blockage blockage in the system. By comparing the response of the individual instrument instrument lines during the drain test, partial or complete complete blockage in one line can be be detected.

The SDV Instrumentation/valve Instrumentation/valve response response surveillance surveillance test will be satisfied satisfied anytime a scram occurs (less than or equal to 50%

rod density) or by the fill/drain test not to exceed exceed an operating operating cycle.

AMENDMENT NO. 142, Revision AMENDMENT NO. Revision 22 (A196) 42

• BASES FOR 3.1.2 AND BASES AND 4.1.2 LIQUID POISONPOISON SYSTEM The liquid liquid poison system also has a post-LOCA safety function to buffer the suppression pool pH in order to maintain the bulk pH above 7.0.

necessary to prevent iodine re-evolution This function is necessary methodology. Manual system re-evolution consistent with the Alternative Source Term analysis methodology.

pentaborate initiation is used, and the minimum amount of sodium pentaborate solution required to be injected for suppression pool pH buffering is 1114 suppression 1114 concentration of 9.423 weight percent. This volume consists of the minimum required gallons at a minimum concentration required volume of 1325 gallons minus minus the 197 gallons that are contained below the point where the pump takes suction from the storage tank and minus 14 gallons that are assumed assumed to remain in the pump suction and dischargedischarge piping after injection stops. Operation of a single liquid poison pump can satisfy this this post-LOCA function. This function applies to the power operating condition, and also whenever the reactor coolant system temperature is post-LOCA 212'F except for reactor vessel hydrostatic or leakage testing with the reactor not critical.

greater than 212°F Nearly- all maintenance Nearly maintenance can be completed within a few days. Infrequently, Infrequently, however, major maintenance maintenance might be required. Replacement Replacement of principal principal system components could necessitate necessitate outages of more than 7 days. In spite of the best efforts of the operator to return equipment to service, some maintenance equipment maintenance could require up to 6 months.

The system test specified specified demonstrates demonstrates component component response such as pump starting upon upon manual manual system initiation and is similar to the the operating requirement requirement under accident conditions. The only difference difference is that demineralized demineralized water rather than the boron solution will be be pumped to the reactor vessel. The test interval interval between operating cycles cycles results in a system failure probability 10-66 (Fifth probability of 1.1 xx 10. (Fifth Supplement, p. 115)* and is consistent with practical considerations.

Pump operability operability will be demonstrated demonstrated on aa more frequent basis. A continuity continuity check check of the firing circuit on the explosive explosive valves is provided provided

.by by pilot lights in the control room. Tank level and temperature temperature alarms alarms are provided provided to alert the operator operator of off-normal off-normal conditions.

The functional test and other surveillance surveillance on components, along with the monitoring monitoring instrumentation, instrumentation, gives a a high reliability for liquid poison system operability.

• FSAR

• FSAR AMENDMENT AMENDMENT NO. 142, Revision Revision 21 (A194) 49

• BASES FOR 3.2.4 AND BASES AND 4.2.4 REACTOR REACTOR COOLANT SPECIFIC ACTIVITY The specific activity in the reactor reactor coolant is is an initial condition for evaluation consequences of a main steam line break radiological consequences evaluation of the radiological (MSLB) outside of primary containment. No fuel damage is postulated in the MSLB accident, and radioactive material to the the release of radioactive environment is assumed to end when the main steam isolation valves (MSIVs) close completely. The environment specific iodine activity is limited to to 0.2 IlCi/gm lICi/gm Dose Equivalent 1-131. This limit ensures that the source term assumed Equivalent 1-131. consequences analysis for the assumed in the radiological consequences the MSLB accident is not exceeded, so that any release of radioactivity to the environment environment during a MSLB results in offsite and control room acceptance criteria of 10 CFR 50.67 and Regulatory Guide 1.183. ItIt is also conservative with respect radiation doses that satisfy the acceptance respect to the the value used in the radiological radiological consequences consequences analyses analyses for other postulated small break loss of coolant accidents outside of primary containment and for postulated containment postulated instrument line breaks.

The limits on reactor coolant specific specific activity are applicable applicable in the power operating and hot shutdown conditions, since there is an escape escape path for release of radioactive material from the reactor coolant system to the environment environment in the event of a MSLB outside of primary containment. In the cold shutdown, refueling, refueling, and major maintenance maintenance conditions, no limits are required since the reactor is not pressurized and the potential potential for leakage leakage is reduced.

When the reactor coolant specific activity activity exceeds exceeds the limit of 0.2 IlCilgm pVCi/gm Dose Equivalent 1-131,

• 4.0 IlCilgm, 1-131, but is :::; pVCi/gm, samples must be be analyzed for Dose Equivalent 1-131 analyzed 1-131 at least once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. In addition, the specific specific activity must be restored to the limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

The completion completion time of once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is based on the time needed needed to take and analyze a sample. The 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> completion time to restore the activity level provides aa reasonable reasonable time for temporary temporary coolant coolant activity activity increases (iodine spikes or crud bursts) to be cleaned up up with the normal normal processing processing systems.

The-isotopic The isotopic analyses of reactor reactor water samples will be used to assure that the limit of Specification Specification 3.2.4 is not exceeded exceeded during during normal operation.

operation. The 7 day frequency is dayfrequency adequateadequate to trend changes changes in the iodine activity level. The surveillance surveillance requirement need need only be be performed during performed during the power operating because the level of fission products operating condition because products generated generated in other operating operating conditions conditions is much less.

In addition, the trend of the stack In stack offgas release release rate, which is continuously continuously monitored, is aa goodgood indicator indicator of the trend trend of the iodine iodine activity inin the reactor coolant.

Since Since the concentration concentration of radioactivity radioactivity in the reactor reactor coolant is not continuously continuously measured, coolant sampling would be ineffective ineffective as a means means to rapidly detect detect gross gross fuel element element failures.

failures. However, as discussed discussed in the bases bases for Specification Specification 3.6.2, some capability capability to detect gross fuel element element failures is inherent inherent in the radiation radiation monitors in the offgas system system andand on the main main steam lines.lines.

AMENDMENT AMENDMENT NO. NO. 142, 161, 161, 170, 170, Revision Revision 21 (A194) 100 100

• BASES BASES FOR FOR 3.2.4 3.2.4 AND AND 4.2.4 4.2.4 REACTOR REACTOR COOLANT COOLANT SPECIFIC ACTIVITY ACTIVITY In the In the event event of of aa large large primary primary system break underunder reactor vessel vessel hydrostatic hydrostatic oror leakage leakage test conditions conditions with thethe reactorcoolant reactor coolant temperature>

. temperature > 215 0 F, the 215°F, the reactor reactor not critical, critical, and and primary containment containment integrity established, calculations show the integrity not established, the resultant resultant radiological radiological dose dose at the exclusion exclusion area area boundary boundary to to be conservatively conservatively bounded bounded by the dose calculated calculated for aa main steam line breakbreak outside outside primary primary containment. This dose dose was calculated calculated on the basis the basis of the reactor reactor coolant specific activity limit limit of 0.2 pCi/gm fJCilgm Dose Equivalent 1-131.

Equivalent 1-131.

The reactor reactor coolant coolant sample sample required required by Specification Specification 4.2.4.b 4.2.4.b will be used used to assure assure that that the limit of Specification Specification 3.2.4.d 3.2.4.d is is not exceeded.

exceeded. The sample sample shall be taken during steady steady state conditions conditions to ensure the results results are representative representative of the steady steady state state radioactive radioactive concentration concentration for reactor reactor vessel hydrostatic hydrostatic or leakage leakage test test conditions.

Revision 21 (A(A194) 194) 100a 100a

• . BASES BASES FOR 3.2.7 3.2.7 AND REACTOR COOLANT SYSTEM AND 4.2.7 REACTOR

• SYSTEM ISOLATION VALVES{PRIVATE

ISOLATION VALVES{PRIVATE The list of reactor coolant isolation isolation valves is contained in the procedure governing controlled lists and has been removed from the Technical procedure governing Specifications Specifications per Generic Letter 91-08. Revisions will be processed processed in accordance accordance with Quality Assurance Program requirements. Double isolation valves are provided in lines which connect to the reactor coolant system to assure isolation and minimize reactor coolant loss in the the event of aa line rupture. Closure of the isolation valves also minimizes potential leakage isolation valves containment in the event of a loss-leakage paths from the primary containment of-coolant accident. In whenever fuel is in the reactor vessel and the reactor coolant temperature In addition, whenever temperature is less than or equal to 212°F 212OF (encompassing the cold shutdown and refueling operating conditions), closure of the shutdown cooling system isolation valves ensures that the the reactor vessel water level level does not drop below the top of the active fuel during a vessel draindown draindown event caused by a leak or line break in the the shutdown cooling system. The specified valve requirements assure that isolation isolation is already accomplished accomplished with one valve shut or provide provide redundancy redundancy in an open line with two operative operative valves. Except Except where check valves are used as one or both of a set of double double isolation valves, the the isolation isolation valves shall be capable of automatic automatic initiation. initiation. Valve closure times are selected to minimize coolant losses in the event of the specific line specific line rupturing and are procedurally controlled. controlled. Using the longest closure time on the main-steam-linemain-steam-line valves following aa main-steam-line main-steam-line break break (Section XV-C.1.0)(1), the core is still covered XV-C.1.0)(1"), covered by the time the valves close. Following a specific specific system line break, the cleanup and shutdown cooling closing times will upon initiation initiation from aa low-low level signal limit coolant loss such that the core is not uncovered. uncovered. Feedwater flow would quickly restore coolant coolant levels to prevent clad damage. Closure Closure times are discussed discussed in Section VI-D. VI-D.1.0(1). 1.01*). ItIt is not intended intended that compliance compliance with Technical Specification actions would prevent changes Technical Specification changes in modes modes or other specified conditions that are part of a shutdown of the unit. Accordingly, ifif during aa plant shutdown any shutdown shutdown cooling system isolation valve becomes becomes inoperable for closing while placing shutdown cooling in operation, operation, itit is recommended recommended not to take the action specified specified in 3.2.7.b to isolate one valve in the line having having inoperable valve within 4 hours. This is because, once the line is isolated, the inoperable Specifications preclude unisolating the line unless itit isolated, the Technical Specifications is for the purpose of demonstrating operability of the inoperable demonstrating operability inoperable valve. It It is, therefore, recommended recommended to take the actionaction specified specified in 3.2.7.c within 4 hours (instead hours (instead of the action specified action specified in 3.2.7.b) and proceed with the shutdown actions using shutdown cooling as necessary necessary to reduce reactor coolant temperature temperature to less than 212°F within the following 10 10 hours. Thereafter, the actions actions specified specified in 3.2.7.e and 3.2.7.f would need to be be met. An inoperable shutdown shutdown cooling isolation valve may be opened with the shutdown cooling permissives met (reactor pressure pressure <~ 120 120 psig and temperature < 350 0 F) temperature :-::; 350°F) in order to comply with the shutdown shutdown actions specified specified in 3.2.7.c. During plant operation, the isolation valves in the shutdown shutdown cooling system are normally closed. In lieu of performing performing Type C leak leak rate testing on these isolation valves, aa water seal is provided to prevent containment atmosphere atmosphere leakage through these valves in the event of an accident accident requiring primary primary containment containment isolation. The seal water, supplied supplied from the core spray system, would pressurize pressurize the piping between the inboard and outboard isolation valves. To prevent inadvertent valve opening prevent a spurious or inadvertent defeating the water seal, the motor-operated opening from defeating motor-operated shutdown cooling system isolation valves are required to be de-activated de-activated (power is removed) during normal normal plant operation. operation. Thus, the motor-operated motor-operated shutdown shutdown cooling system isolation valves are considered operable operable when the valves are closed and de-activated de-activated and the water seal is capable capable of of performing performing its function. (1) UFSAR UFSAR AMENDMENT NO. 142,145, AMENDMENT 142, 115, Revision 2 (/\ (A 173), 6 (A! 81),), 24 (A (/\181 (Al197)

97) 115 115

• BASES FOR 3.2.7 AND BASES REACTOR COOLANT AND 4.2.7 REACTOR COOLANT SYSTEM VALVES{PRIVATE}}

SYSTEM ISOLATION VALVES{PRIVATE 0 F), power for When the shutdown shutdown cooling system is placed in service for plant cooldown pressure ;s;_< cooldown (with reactor pressure temperature < 350 120 psig and temperature;s; 350°F), power for motor-operated isolation valves must be restored and the valves opened. Should a loss of coolant the motor-operated accident occur at this time, failure of an coolant accident isolation valve to close upon isolation signal could cause aa loss of the water seal. The risk associated upon receipt of an isolation associated with this potential single failure has been determined determined to be acceptable based on the low probability acceptable based damage event occurring during shutdown probability of a core damage shutdown cooling system operation 2 ). operation(2). Specification 3.2.7.d requires operability Specification operability of the shutdown cooling system isolation valves whenever whenever fuel is in the reactor vessel and the reactor temperature is less than or equal to 212 0 F. If isolation valve becomes coolant temperature 212°F. any isolation becomes inoperable, Specification Specification 3.2.7.e requires that, within 4 hours, at least one valve in each line having having an inoperable inoperable valve is in the mode corresponding corresponding to-the condition. However, if the shutdown to*the isolated condition: shutdown cooling function is needed needed to provide provide core cooling, isolating the shutdown shutdown cooling line is not desirable. Specification 3.2.7.f allows the shutdown desirable. Specification shutdown cooling line to remain remain unisolated unisolated provided action is immediately immediately initiated initiated to suspend operations operations with aa potential for draining the reactor vessel (OPDRVs). If suspending If suspending the OPDRVs would result in closing the shutdown shutdown cooling system isolation isolation valves, an alternative alternative action action is provided to immediately immediately initiate action action to restore the valve(s) to operable status. This allows the shutdown cooling system to remain in service service while actions are being being taken to restore the valve(s). The term "immediately" means that the action should be pursued without delay and in a controlled manner. Either of the actions identified in Specification 3.2.7.f must continue continue until OPDRVs OPDRVs are suspended or the valves are restored to operable operable status. Operation Operation with the shutdown cooling system in service is not considered considered an OPDRVOPDRV so long as system integrity is maintained. System integrity is maintained maintained provided the piping is intact and no maintenance maintenance is being performed performed that has the potential potential for draining the reactor reactor vessel through the the In addition, with the reactor coolant temperature less than or equal to 212 0 F, the water seal function is not required to consider system. In 212°F, consider the the shutdown cooling system isolation isolation valves operable operable since primary containment integrity is not required with reactor coolant temperature less than primary containment 0 F. or equal to 215215°F. The valve valve operability test intervals periods not likely to significantly intervals are based on periods significantly affect operations, and are consistent with testing of other systems. obtained during closure testing are not expected Results obtained expected to differ appreciably appreciably from closure times under under accident conditions as in most cases, flow flow helps to seal the valve. The test interval of once per operating automatic initiation results in a failure probability of 1.1 x 10-operating cycle for automatic 10-77 (Fifth Supplement, p. 115)(3) that aa line line will not isolate. Additional surveillances are in accordance Additional surveillances accordance with the Inservice Inservice Testing Program Program described described in Specification Specification 6.5.4. (2) Letter from G. E. Edison Edison (NRC) to B. B. R. R. Sylvia Sylvia (NMPC) (NMPC) dated March March 20, 1995, 1995, Issuance Issuance of Amendment for Nine Mile Point Nuclear Nuclear Station No. 1 (License Amendment No. 154). Unit No.1 (3) FSAR Revision 24 (A (A197) 197) 115a 115a I

• BASES FOR 3.3.3 BASES 3.3.3 AND AND 4.3.3 LEAKAGE LEAKAGE RATE The primary containment preoperational test pressures are based upon the calculated containment preoperational calculated primary containment pressure response in the event primary containment event loss-of-coolant accident. The peak drywell pressure would be 35psig of a loss-of-coolant 35 psig which would rapidly reduce to 22 psig within 100 seconds seconds following the pipe break. The total time the drywell pressure would be above 22 psig is calculated to be about 10 Following the 10 seconds. Following the pipe break, the suppression pressure rises to 22 psig within 10 seconds, equalizes suppression chamber pressure drywell pressure and thereafter equalizes with drywell thereafter rapidly decay.(1) drywell pressure decay.

decays with the drywell (1) . . . The design pressures of the drywell and suppression chamber chamber are 62 psig and 35 psig, respectively.(2) As pointed out above, the the. pressure pressure response of the drywell response drywell and suppression chamber following an accident suppression chamber accident would be the same after about 10 10 seconds. Based on the the calculated containment pressure calculated primary containment pressure response discussed discussed above and the suppression chamber pressure; primary containment chamber design pressure; containment pressures were chosen. Also, based on the primary preoperational test pressures primary containment containment pressure response and the fact that the drywell drywell and suppression suppression chamber function as aa unit, the primary containment will be testedas primary containment tested* as aa unit rather rather than testing the individual components components separately. The function function of the primary containment containment is to isolate isolate and contain fission products products released from the reactor reactor coolant system following design accidents (DBA). The primary containment basis accidents containment provides an essentially essentially leak tight barrier against against an uncontrolled uncontrolled release of radioactive radioactive material to the environment. The DBA that postulatespostulates the maximum maximum release of radioactive radioactive material within the primary containment containment is aa loss loss accident (LOCA). of coolant accident (LOCA). InIn the analysis analysis of this accident, it iUsis assumed containment integrity is maintained such that release assumed that primary containment release of fission of products to fission products to the environment is the environment is controlled by the controlled by rate of the rate of primary containment leakage. primary containment leakage. The LOCA radiological consequences consequences analysis is based on an alternativealternative source term (AST) methodology methodology (10 (10 CFR 50.67 and Regulatory 1.183). This analysis Guide 1.183). analysis concluded concluded that the calculated total effective effective dose equivalent (T,EDE) (TEDE) values to the control room occupants, the the exclusion exclusion area boundary, and the low population zone are within the TEDE criteria established in 10 CFR 50.67. Primary containment containment leakage at the rate of 1.5%1.5% by weight of the containment containment air per 24 hourshours is assumed assumed in the accident accident analysis. Margin Margin is achieved achieved by establishing establishing the allowable operational leak allowable operational leak rate. The operational operational limit is derived by multiplying multiplying the allowable allowable test leak rate by 0.75 thereby providing a 25% margin margin to allow for leakage leakage deterioration deterioration which may may. occur during the periods between leak rate tests. AMENDMENT NO. 142, 'Revision AMENDMENT Revision 21 (A194) (Al 94) 140

• BASES FOR BASES AND 4.3.3 3.3.3 AND FOR 3.3.3 LEAKAGE RATE 4.3.3 LEAKAGE RATE Closure I

Closure ofof the the containment containment isolation isolation valves valves for for the the purpose purpose of of the the test test is is accomplished accomplished by by the the means means provided provided forfor normal normal operation operation of of the. the valves. The valves. The reactor reactor isis vented vented to to the containment atmosphere the containment atmosphere during ILRT testing. during ILRT testing. The The primary primary containment containment leak leak rate rate test test frequency frequency is is based based on on maintaining maintaining adequate adequate assurance assurance that that the the leak leak rate rate remains remains within within the the specification. specification. TheThe leak leak rate rate test test frequency frequency isis based based on on Option Option BB of of 10 10 CFR CFR 50 50 Appendix Appendix J. J. The The penetration penetration andand air air purge purge piping piping leakage leakage test test frequency, frequency, along along with with the the containment containment leakleak rate rate tests, tests, is is adequate adequate to to allow allow detection detection of of leakage leakage trends. Whenever a double-gasketed penetration (primary containment head equipment hatches and the suppression chamber trends. Whenever a double-gasketed penetration (primary containment head equipment hatches and the suppression chamber access access hatch) hatch) isis broken broken and and remade, remade, the the space space between between the the gaskets gaskets is is pressurized pressurized to to determine determine thatthat the the seals seals are are performing performing properly. properly. TheThe test pressure of 35 psig is consistent with the accident analyses and the maximum preoperational leak rate test pressure of 35 psig is consistent with the accident analyses and the maximum preoperational leak rate test pressure. It is expected test pressure. It is expected that that the the majority majority ofof the the leakage leakage from from valves, valves, penetrations penetrations and and seals seals would would be be into into the the reactor reactor building. building. However, However, it it is is possible possible that that leakage leakage into into other parts of the facility could' occur. Such leakage paths that may affect significantly the consequences of accidents are to other parts of the facility could-occur. Such leakage paths that may affect significantly the consequences of accidents are to be be minimized. minimized. . Leakage Leakage from from airlocks airlocks.isis measured measured under under accident accident pressures pressures inin accordance accordance with with Option Option BB of of 10 10 CFR CFR 5050 Appendix Appendix J. J. AMENDMENT ,NO. l\MENDMENT NO. 142,142,159, Revision 21 159, Revision (A194) 21 (A194) 141 141

• BASES FOR BASES 3.3.3 AND FOR 3.3.3
• LEAKAGE RATE 4.3.3 LEAKAGE AND 4.3.3 RATE The Type The Type A A test test follows follows the the guidelines guidelines stated stated in in ANSI/ANS-56.8(6)

ANSI/ANS-56.8(6) and/or and/or the the Bechtel Topical Report.(4) Bechtel Topical program provides This program Report.(4) This adequate assurance provides adequate assurance that the that the test test results results realistically realistically estimates estimates the the degree of containment degree of following aa loss-of-coolant leakage following containment leakage loss-of-coolant accident. The containment leakage rate is The containment leakage rate is calculated using calculated using the the Absolute Absolute Methodology. Methodology.(8) (8) The specific treatment of selective valve arrangements including the acceptability of the interpretations of 10 CFR 50 Appendix J requirements are The specific treatment of selective valve arrangements including the acceptability of the interpretations of 10 CFR 50 Appendix J requirements are in References given in given References 5, 6, and and 7. Core Core Spray Spray and Containment Spray suction Containment Spray valves will be tested suction valves tested in in accordance with the accordance with the IST 1ST Program. Program.

References:

(1) FSAR, Volume II, FSAR, Volume Appendix E II, Appendix (2) (2) UFSAR, Section VI B.2.1 UFSAR, (3) (3) (Deleted) (Deleted) (4) BN-TOP-1 "Testing Criteria BN-TOP-1 Integrated Leakage Rate Testing of Primary Criteria for Integrated Primary Containment Structures for Nuclear Containment Structures Plants," Revision 1, Nuclear Power Plants," Bechtel Corporation, November November 1, 1, 1972 1972 (5) Evaluation Report dated NRC Safety Evaluation "Regarding Proposed Technical dated May 6, 1988, "Regarding Technical Specifications Requests Related to Specifications and Exemption Requests Appendix J." (6) Mohawk Letter dated July 28, 1988, "Clarifications, Justifications Niagara Mohawk Justifications & Conformance with 10 CFR 50 Appendix J SER."

                                                                                                   & Conformance (7)

(7) November 9, 1988, "Review NRC Letter dated November Appendix J Containment Leakage Rate Testing at Nine Mile "Review of the July 28, 1988 Letter on Appendix Mile Point Unit Point Unit 1."1." (8) ANSI/ANS - 56.8 -- 1994, "Containment System Leakage ANSI/ANS Leakage Testing Requirements." AMENDMENT NO. ,A.MENDMENT NO. ,, 142, 159, Revision 142, 159, Revision g,8, 21 (A194) 21 (A194) 142 142

• BASES FOR 3.3.4 AND BASES
• ISOLATION VALVES CONTAINMENT ISOLATION PRIMARY CONTAINMENT AND 4.3.4 PRIMARY VALVES primary containment The list of primary containment isolation isolation valves is contained contained in the procedure procedure governing governing controlled controlled lists and has been removed from the the Specifications per Generic Letter 91-08. Revisions will be processed in accordance Technical Specifications accordance with Quality Assurance Program requirements.

Double Double isolation penetrating the primary containment and open to the free space of the containment. Closure isolation valves are provided on lines penetrating of one of the valves in each line would be sufficient sufficient to maintain the integrity of the pressure suppression suppression system. Except Except where check valves are used as one or both of a set of double isolation valves, the isolation valves shall be capable capable of automatic initiation. Automatic Automatic initiation initiation is required required to minimize the potential leakage leakage paths from the containment containment in the event of a loss-of-coolant loss-of-coolant accident. Details of the the 1 Section 3.3.3/4.3.3. isolation valves are discussed discussed in Section VI_D(1). VI-D( ). For allowable leakage rate specification, specification, see Section For the design design basis loss-of-coolant accident fuel rod perforation loss-of-coolant accident perforation would not occur until the fuel temperature temperature reached 1700°F which occurs in approximately approximately 100 seconds(2). The required closing times for all primary containment isolation valves are established established to prevent fission product connecting to the primary containment. product release through lines connecting 0 F, For reactor reactor coolant system temperatures temperatures less than 215 215°F, the containment containment could not become pressurized pressurized due loss-of-coolant accident. due to a loss-of-coolant The 215°F limit is based on preventing pressurization of the reactor building and rupture of the blowout panels. The test interval of once per operating cycle for automatic initiation 10-7 that a line will not isolate initiation results in a failure probability of 1.1 x 10-7 isolate (Fifth Supplement, p. 115P).115 )(3). More frequent testing for valve operability operability results in a more reliable system. In addition to routine surveillance as outlined In outlined in Section VI-D.1.0(1) VI-D.1.0(1) each instrument-line instrument-line flow check valve will be tested for operability. All instruments on a given line will be isolated at each instrument. The line will be purged purged by isolating isolating the flow check valve, opening the the bypass valves, and opening opening the drain valve to the equipment drain tank. When purging purging is sufficient to clear the line of non-condensibles non-condensibles and crud the flow-check valve will be cut into service and the bypass bypass valve closed. The main valve will again be opened and the flow-check valve allowed to close. The flow-check valve will be reset by closing the drain valve and opening the bypass valve depressurizing depressurizing part of the system. Instruments will be cut into service after closing the bypass valve. Repressurizing Repressurizing of the individual instruments instruments assures assures that flow-check flow-check valves have operability testing of excess flow check valves may be performed prior to have reset to the open position. Alternatively, operability installation using a test set-up installation set-up that simulates simulates an instrument instrument line break condition. (1) UFSAR (2) (2) Nine Mile Point Nuclear Generation Generation Station Unit 1 Safer/CorecooI/GESTR-LOCA Safer/Corecool/GESTR-LOCA Loss of Coolant Accident Accident Analysis, NEDC-31446P, NEDC-31446P, Supplement Supplement 3, September, 1990. (3) FSAR AMlENDrM1E*rT NO. 112, AMENDMENT 142, 145, Revision 6 (A^18), 145, Revision (A181), 12, 224 (A197), 27 150 150

• BASES BASES FOR 3.4.1 AND AND 4.4.1 LEAKAGE LEAKAGE RATE RATE The secondary secondary containment containment is designed minimize any ground level release of radioactive materials designed to minimize materials that might result from aa seriousserious accident.

secondary containment during reactor operation, when the drywell The reactor building provides secondary drywell is sealed and in service. The reactor building building containment during periods when the reactor is shutdown, the drywell is open, and activities provides primary containment activities are ongoing require secondary ongoing that require containment containment to be in effect. There principal accidents for which credit is taken for reactor building (secondary There are two principal (secondary containment) integrity. These These are a loss of coolant coolant accident accident (LOCA) and a refueling accident involving accident involving "recently irradiated" fuel. The reactor building performs performs no active function in response response to each of these limiting events; however, its leak tightness tightness is required to ensure ensure that the release of radioactive radioactive materials is restricted leakage restricted to those leakage paths paths and associated leakage leakage rates assumed assumed in the accident accident analysis and that fission products entrapped entrapped within the reactor building structure structure will be treated by the Reactor Building Emergency Emergency Ventilation System (RBEVS) (RBEVS) prior to discharge discharge to the environment. In addition to these limiting events, events occurring In occurring during handling of an irradiated fuel cask and operations with a potential the potential for draining the reactor vessel (OPDRVs) can be postulated to cause cause a fission product release. During these events, the reactor building would be the only barrier barrier to aa release to the environment. Thus, reactorreactor building integrity is required during handling of an irradiated irradiated fuel cask and during OPDRVs. The Refueling Refueling Accident Accident analysis analysis is based on an alternative source term (AST) methodology methodology (10 CFR 50.67 and Regulatory Regulatory Guide 1.183). 1.183). This This analysis concluded that the calculated total effective dose equivalent equivalent (TEDE) values to the control room occupants, the exclusion exclusion area area boundary, and the low population population zone are well below the TEDE criteria criteria established in 10 CFR 50.67 without crediting reactor building integrity, operation operation of the RBEVS, or operation of the Control Room Air Treatment Treatment System (CRATS), (CRATS), as long long as the fuel is allowed to decay decay for at least 24 hours following reactor reactor shutdown. As a a result, "recently irradiated" fuel is defined as fuel that has occupied occupied part of a critical reactor core within 24 hours; i.e., reactor fuel that has decayed decayed less than 24 hours following reactor shutdown. shutdown. Therefore, reactor reactor building integrity integrity is not required, and RBEVS RBEVS and CRATS CRATS are not required to. to. be operable, during movement movement of decayeddecayed irradiated fuel that is no. no longer Io.nger considered co.nsidered "recently irradiated." Conversely, Conversely, reactor building integrity is required, and RBEVS and CRA CRATS TS are required to be operable, operable, during movement movement of recently irradiated irradiated fuel assemblies. assemblies. In the answers to Questions Questions 11-3 and IV-5 of the Second Supplement and also in the Fifth Supp/ement*, Seco.nd Supplement Supplement*, the relatio.nships relationships among wind speed, direction, directio.n, pressure distribution outside the building, building internal pressure, and reactor building leakage leakage are discussed. The curve of pressure pressure in Figure 3.4.1 represents the wind directio.n direction which which results in the least building leakage. It It is assumed that when the test is performed, the windwind direction is that which gives the least leakage.

• FSAR
• FSAR AMENDMENT NO. 142, 156, Revisio.n AMENDMENT Revision 21 (A194) 167 167
• BASES FOR BASES AND 4.4.1 3.4.1 AND FOR 3.4.1 LEAKAGE RATE 4.4.1 LEAKAGE RATE IfIf the wind direction the wind direction was not from was not from the direction which the direction gave the which gave the least reactor building least reactor building internal leakage, building building leakage, pressure would internal pressure would not be as not be as negative negative as as Figure Figure 3.4.1 indicates. Therefore, 3.4.1 indicates. Therefore, to to reduce reduce pressure, pressure, thethe fan fan flow flow rate would have rate would have toto be be increased.

increased. ThisThis erroneously erroneously indicates indicates that that reactor reactor building leakage building leakage is is greater greater than than if if wind wind direction direction were accounted for. were accounted for. If If wind wind direction direction were were accounted accounted for,for, another pressure curve another pressure curve could could be be used used which was less which was less negative. negative. ThisThis would would meanmean thatthat less less fan fan flow measured leakage) flow (or measured leakage) would would be be required required to to establish establish building building pressure. pressure. However, for However, for simplicity simplicity itit is is assumed assumed that that the the test test isis conducted during conditions conducted during conditions leading leading toto the the least least leakage leakage while while the accident is the accident is assumed assumed to occur during conditions leading to the greatest reactor to occur during conditions leading to the greatest reactor building leakage. building leakage. As As discussed discussed in in the the Second Second Supplement Supplement and and Fifth Supplement, the Fifth Supplement, the pressure pressure forfor Figure Figure 3.4.1 3.4.1 is is independent independent of the reactor of the reactor building building leakage leakage raterate referenced to zero mph wind speed at a negative differential pressure of 0.25 inch of water. Regardless of the leakage rate at these design referenced to zero mph wind speed at a negative differential pressure of 0.25 inch of water. Regardless of the leakage rate at these design conditions, conditions, the the pressure pressure versus versus wind wind speed speed relationship relationship remains remains unchanged unchanged for for any any given given wind wind direction. direction. By By requiring requiring the the reactor reactor building pressure to building pressure to remain remain within within the the limits limits presented presented in in Figure Figure 3.4.1 and aa reactor 3.4.1 and reactor building building leakage leakage rate rate ofof less less than than 1600 cfm, exfiltration would be prevented. This would assure that the leakage from the primary containment 1600 cfm, exfiltration would be prevented. This would assure that the leakage from the primary containment is directed through the filter is directed through the filter system system and and discharged discharged from from thethe 350-foot 350-foot stack. stack. Revision 21 Revision (A194) 21 (A 194) 167a 167a

• BASES FOR BASES 3.4.2 AND FOR 3.4.2 REACTOR BUILDING 4.4.2 REACTOR AND 4.4.2 BUILDING INTEGRITYINTEGRITY ISOLATIONISOLATION VALVES VALVES.

Isolation Isolation ofof the the reactor reactor building building occurs automatically upon occurs automatically upon high high radiation radiation ofof the the normal normal building exhaust ducts building exhaust ducts or or from from high high radiation radiation at at the the refueling refueling platform platform (See (See 3.6.2). 3.6.2). Isolation Isolation willwill assure assure that that any fission products anyfission products entering entering the the reactor reactor building building willwill be be routed routed to to the the emergency emergency ventilation system ventilation system prior prior to to discharge discharge to the environment to the environment (Section (Section VII-H.3.0 VII-H.3.0 of the FSAR). of the FSAR). The The twotwo principal principal accidents accidents for for which which thethe reactor reactor building building isolation isolation valves valves must must be be operable operable are are aa loss loss of of coolant coolant accident accident (LOCA) (LOCA) and and aa refueling refueling accident accident involving involving "recently "recently irradiated" irradiated" fuel. fuel. InIn addition addition to to these these limiting limiting events, events, events events occurring occurring during during handling handling of of an an irradiated irradiated fuel fuel cask cask andand operations operations withwith aa potential potential forfor draining draining the the reactor reactor vessel vessel (OPDRVs) (OPDRVs) can can bebe postulated postulated to to cause cause aa fission fission product product release. release. During During these these events, events, the the reactor reactor building building would would bebe the the only only barrier barrier to to aa release release toto the the environment. environment. Thus, Thus, thethe reactor reactor building building isolation isolation valves valves areare required required toto bebe .operable operable during during handling handling of of an an irradiated irradiated fuelfuel cask cask and and during during OPDRVs. OPDRVs. . The The Refueling Refueling Accident Accident analysis analysis is is based based on on anan alternative alternative source source termterm (AST) (AST) methodology methodology (10 (10 CFR CFR 50.6750.67 andand Regulatory Regulatory Guide Guide 1.183). 1.183). This This analysis concluded that the calculated total effective dose equivalent (TEDE) values to the analysis concluded that the calculated total effective dose equivalent (TEDE) values to the control room occupants, the exclusion area boundary, control room occupants, the exclusion area boundary, and and the the low low population population zonezone areare well well below below the the TEDE TEDE criteria criteria established established in in 10 10 CFR CFR 50.67 50.67 without without crediting crediting reactor reactor building building integrity integrity or or operation operation of the reactor building emergency ventilation system (RBEVS), as long as the fuel is allowed to decay of the reactor building emergency ventilation system (RBEVS), as long as the fuel is allowed to decay for at least 24 hours following reactor for at least 24 hours following reactor shutdown. shutdown. As As aa result, result, "recently "recently irradiated" irradiated" fuel fuel isis defined defined asas fuel fuel that that has has occupied occupied part part ofof aa critical critical reactor reactor corecore within within 24 24 hours; hours; i.e., i.e., reactor reactor fuel fuel that has decayed less than 24 hours following reactor shutdown. Therefore, reactor building integrity that has decayed less than 24 hours following reactor shutdown. Therefore, reactor building integrity is not required and the reactor building is not required and the reactor building isolation isolation valves valves areare not not required required to to be be operable operable duringduring movement movement of of decayed decayed irradiated irradiated fuelfuel that that isis no no longer longer considered considered "recently "recently irradiated." irradiated." Conversely, reactor building integrity is required and the reactor building isolation valves are required Conversely, reactor building integrity is required and the reactor building isolation valves are required to be operable during movement of to be operable during movement of recently recently irradiated irradiated fuel assemblies. assemblies. AMENDMENT AMENDMENT NQ. NO. 142, 142, Revision Revision 21 (A194) 21 (A 194) 169 169

• BASES BASES FOR 3.4.3 AND 4.4.3 ACCESS CONTROL The secondary containment is designeddesigned to minimize any ground level release of radioactive materials that might result from a serious accident.

secondary containment during reactor operation, when the drywell is sealed and in service. The reactor building The reactor building provides secondary building provides primary containment during periods when the reactor is shutdown, the drywell is open, and activities activities are ongoing that require secondary containment to be in effect. There are two principal accidents accidents for which credit is taken for reactor building (secondary containment) integrity. These are a loss of coolant accident (LOCA) and a refueling accident refueling accident involving "recently irradiated" fuel. The reactor building performs no active function in response to each of these limiting events; however, its leak tightness is required to ensure that the release of radioactive materials materials is restricted to those leakage leakage associated leakage rates assumed in the accident paths and associated accident analysis and that fission products products entrapped entrapped within the reactor building structure will be treated by the Reactor Building Emergency Emergency Ventilation System (RBEVS) prior to dischargedischarge to the environment. limiting events, events In addition to these limiting events occurring occurring during handling of an irradiated fuel cask and operations operations with a potential for draining draining the the reactor vessel (OPDRVs) can be postulated postulated to cause aa fission product release. During these events, the reactor buildingbuilding would be the only only barrier to a release release to the environment. Thus, reactor reactor building integrity is required during during handling handling of an irradiated fuel cask and during during OPDRVs. OPDRVs. The Refueling Refueling Accident analysis is based on an alternative source term (AST) methodology methodology (10 (10 CFR 50.67 and Regulatory Regulatory Guide 1.183). This analysis concluded that the calculated calculated total effective effective dose equivalent equivalent (TEDE) values to the control room occupants, the exclusion area exclusion area boundary, and the low population zone are we" well below the TEDE criteria established in 10 CFR 50.67 without crediting reactor building building integrity, operation of the RBEVS, or operation of the Control Room Air Treatment System (CRATS), (CRA TS), as long as the fuel is allowed to decaydecay for at least 24 hours following reactor shutdown. As a result, "recently irradiated" fuel is defined as fuel that has occupied part of a critical reactor core within 24 hours; i.e., reactor fuel that has decayed decayed less than 24 hours following reactor shutdown. Therefore, reactor building building integrity is not required during movement movement of decayed decayed irradiated irradiated fuel that is no longer considered considered "recently irradiated." Conversely, reactor reactor building building integrity integrity is required required during during movement of recently irradiated fuel assemblies. recently irradiated As discussed discussed in Section VI-F* all access openings of the reactor building access openings building have as aa minimum two doorsdoors in series. Appropriate Appropriate local local alarms alarms and control control room indicators are provided provided to always always insure that reactor building integrity maintained. Surveillance integrity is maintained. Surveillance of the reactor building reactor building access doors access doors provides provides additional assurance assurance that reactor reactor building integrity is maintained. maintained. Maintaining Maintaining closed doors on the pump compartments compartments ensures ensures that suction to the core and containment containment spray pumps is not lost in case of aa gross gross leak from the suppression suppression chamber.

• FSAR
• FSAR AMENDMENT NO.

AMENDMENT NO. 142, 170, Revision 21 (A194) 172 172

• BASES BASES FOR 3.4.4 3.4.4 AND AND 4.4.4 EMERGENCY EMERGENCY VENTILATION VENTILATION SYSTEM emergency ventilation The emergency ventilation system is designed designed to filter and exhaust atmosphere to the stack during secondary exhaust the reactor building atmosphere containment isolation containment isolation conditions. Both emergency emergency ventilation system fans are designed to automatically automatically start upon high radiation radiation in the the reactor building ventilation ventilation duct or at the refueling platform platform and to maintain maintain the reactor building pressure pressure to the design design negative pressure so as to minimize in-leakage. Should one system fail to start, the redundant redundant system is designed designed to start automatically. Each of the two fans fans has 100 percent capacity.

High efficiency efficiency particulate absolute (HEPA) filters are installed before and after the charcoal charcoal adsorbers adsorbers to minimize potential release of particulates particulates to the environment and to prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed adsorbers installed to reduce the potential release release of radioiodine to the environment. The in-place test results should indicate indicate a system leak leak tightness of less than 1 percent bypass percent bypass leakage leakage for the charcoal adsorbers and a HEPA efficiency efficiency of at least 99 percent percent removal of OOPDOP particulates. The laboratory sample laboratory carbon sample radioactive methyl iodide removal test results should indicate aa radioactive removal efficiency derived from applying a safety efficiency of at least 95 percent, which is derived factor of 2 to the charcoal efficiency of 90 percent assumed charcoal filter efficiency assumed in analysis analysis of design design basis accidents. If If the efficiencies efficiencies of the HEPA filters and charcoal adsorbers are as specified, the resulting doses charcoal adsorbers doses will be less than the 10 CFR 50.67 acceptance acceptance criteria for the accidents accidents analyzed. Operation analyzed. Operation of the fans significantly different from the design flow will change significantly different change the removal removal efficiency HEPA filters and efficiency of the HEPA charcoal adsorbers. . Only one of the two emergency emergency ventilation ventilation systems is needed needed to cleanup cleanup the reactor reactor building atmosphere upon containment isolation. If If one one system is found to be inoperable, there immediate threat to the containment system performance there is no immediate operation or refueling performance and reactor operation refueling activities may continue while repairs are being made. IfIf neither . activities neither circuit is operable, the plant is brought to a condition condition where the emergency ventilation system is not required. Pressure Pressure drop across the combined HEPA HEPA filters and charcoal charcoal adsorbers of less than 6 inches of water at the system design design flow rate will indicate indicate that the filters and adsorbers are not clogged by excessive amounts of foreign foreign matter. Pressure Pressure drop should be determined determined at least once per operating cycle to show system performance performance capability. The frequency frequency of tests and sample analysis are necessary necessary to show that the HEPA HEPA filters and charcoal adsorbers can perform charcoal adsorbers perform as evaluated. The charcoal adsorber efficiency test should allow for charcoal sampling to be conducted adsorber efficiency conducted using an ASTM D3803-1989 03803-1989 approved method. If test results are unacceptable, If unacceptable, all adsorbent adsorbent in the system shall be replaced with an adsorbent adsorbent meeting the physical physical property specifications specifications of Table 5~15-1 of ANSI 509-1980. AMENDMENT NO. 112, AMENDMENT 171, Revision 142, 171, Revision 24-(A494), 21 (A194), 26 (A201) (A201) 176 176

• BASES FOR BASES FOR 3.4.4 AND EMERGENCY VENTILATION AND 4.4.4 EMERGENCY VENTILATION SYSTEM The two principal accidents for which principal accidents Emergency Ventilation System (RBEVS) which the Reactor Building Emergency (RBEVS) must be operable operable are a loss of coolant accident accident (LOCA) and a refueling refueling accident involving "recently irradiated" fuel. In addition to these limiting events, events occurring during handling occurring handling of an irradiated fuel cask and operations irradiated operations with a potential postulated to cause potential for draining the reactor vessel (OPDRVs) can be postulated cause a fission product product release.

During these events, the reactor building would be the only barrier to a release to the environment. Thus, the RBEVS is required to be operable operable during handling of an irradiated fuel cask and during OPDRVs. The Refueling Accident analysis source term (AST) methodology analysis is based on an alternative source methodology (10 CFR 50.67 and Regulatory Guide 1.183). This 1.183). This analysis concluded concluded that the calculated calculated total effective equivalent (TEDE) values to the control room occupants, the exclusion area boundary, effective dose equivalent and the low population population zone are well below below the TEDE criteria established integrity or operation crediting reactor building integrity established in 10 CFR 50.67 without crediting operation of of allowed to decay for at least 24 hours following reactor shutdown. As aa result, "recently irradiated" fuel is defined the RBEVS, as long as the fuel is allowed defined as fuel that has occupied part of aa critical reactor core within 24 hours; i.e.,i.e., reactor fuel that has decayed less than 24 hours following reactor has decayed shutdown. Therefore, reactor building integrity is not required and the RBEVS is not required to be operable operable during movement of decayed irradiated during movement considered "recently irradiated." Conversely, reactor building integrity fuel that is no longer considered operable integrity is required and the RBEVS is required to be operable during movement movement of recently irradiated assemblies.. irradiated fuel assemblies adsorber tray removed for the test should meet the same adsorbent replacement charcoal for the adsorber . The replacement HEPA filters found adsorbent quality. Any HEPA defective shall be replaced defective qualified pursuant to ANSI 509-1980. replaced with filters qualified With doors closed and fan in operation, periphery of each sprayed externally along the full linear periphery operation, DOP aerosol shall be sprayed each respective respective door to to detection of DOP in the fan exhaust check the gasket seal. Any detection considered an unacceptable exhaust shall be considered test result and the gaskets repairs and unacceptable test test repeated. If significant painting, fire or chemical If chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated contaminated from the the fumes, chemicals or foreign material, performed as required for operational use. The analysis shall be performed material, the same tests and sample analysis The Knowledgeable staff members operator on duty at the time of the incident. Knowledgeable determination of significant shall be made by the operator determination members should be be consulted prior to making this determination. determination. Demonstration capability and operability of filter cooling is necessary automatic initiation capability Demonstration of the automatic necessary to assure performance capability. If assure system performance If one emergency ventilation system is inoperable, the other system must be verified to be operable daily. This substantiates substantiates the availability operable system and thus reactor operation or refueling of the operable continue during this period. refueling operation may continue AMENDMENT NO. 142, Revision 2-1(*A!94), AMENDMENT 21 (A194), 26 (A201) (A201) 177 177

• BASES BASES FOR 3.4.5 AND AND 4.4.5 CONTROL ROOM AIR TREATMENT SYSTEM SYSTEM The control room air treatment system is designed to filter the control room atmosphere for intak~ air... A roughing !ilte:

intake air used for filter is used recirculation for recirculation flow during normal control room air treatment operation. The control roo.m. room a~r air tr~atment treatment system .IS is deslg~ed designed to maintain the co~trol room control room pressure to the design positive pressure (one-sixteenth inch water) to minimize inleakage Inleakage of unfiltered air. The ~ontrol treatment room air treatment control roor:n system starts automatically automatically upon receipt of a LOCA (high drywell pressure pressure or low-low reactor water level) or Main Steam. Steam L.I~e Line Break (MSLB) (high steam flow main-steam line or high temperature main-steam line tunnel) signal. The system can also be manually manually initiated. Initiated. The Control Room Envelope (CRE) is the area within the confines of the CRE boundary that contains the spaces that control room occupants inhabit to control the unit during normal and accident conditions. The CRE is protected for normal normal operation, natural events, and accident conditions. The CRE boundary is the combination of walls, floor, roof, roof, ducting, doors, penetrations penetrations and equipment that physically form the eRE. CRE. The operability of the CRE boundary boundary must be maintained maintained to ensure thatthe inleakageinleakage of unfiltered air into the CRE will not exceed the inleakage inleakage assumed in the licensing basis analysis of design assumed design basis accident (DBA) consequences consequencesto to CRE occupants. The CRE and its boundary are defined in in Envelope Habitability the Control Room Envelope Habitability Program. The control room air ~reatment treatment system providesprovides protection from smoke and hazardous chemicals to the CRE occupants. occupants. The analysis of hazardous hazardous demonstrates that the toxicity limits are not exceeded chemical releases demonstrates exceeded in the CRE following a hazardous hazardous chemical chemical release (Ref. 1). The The evaluation evaluation of aa smoke challenge challenge demonstrates demonstrates that itit will not result in the inability of the CRE occupants occupants to control the reactor eithereither from the control room or from the remote shutdown panels (Ref. 1). 1). A periodic offsite chemical survey and procedures procedures for controlling controlling onsite chemicals chemicals are essential essential elements of CRE protection protection against hazardous hazardous chemicals. Changes Changes in off offsite, site, mobile, and onsite hazardous hazardous chemical types or quantities are assessed in accordance accordance with the Control Room Envelope Habitability Program. Program. The assessments assessments provide the necessarynecessary justification justification for not installing aa toxic gas monitoring monitoring automatic isolation system. In order order for the control room air treatment treatment system to be considered considered operable, the CRE boundary maintained such boundary must be maintained such that the CRECRE occupant occupant dose from dose from aa large large radioactive release does radioactive release does not exceed the not exceed the calculated calculated dose dose in the licensing in the licensing basis basis consequence consequence analyses analyses for DBAs, DBAs, and that CRE occupants are protected protected from hazardous hazardous chemicals chemicals and smoke. The CRE boundary The CRE boundary may may be be opened intermittently under opened intermittently under administrative controls. This only applies to openings administrative controls. openings in the CRE boundary boundary that can can be be rapidly rapidly restored to the restored to the design design condition, condition, such such as doors, hatches, hatches, floor plugs, and access panels. For entry and exit through doors, the and access the administrative administrative control control ofof the the opening opening is performed by is performed the person(s) by the person(s) entering entering or or exiting the area. exiting the For other area. For other openings, openings, these controls controls should should bebe proceduralized and consist of stationing a dedicated proceduralized and consist of stationing a dedicated individual individual at the opening opening whowho is in continuous continuous communication with the operators operators in the CRE. eRE. This individual This individual will have a method have a method to to rapidly close the opening opening and to restore the CRE boundary eRE boundary to aa condition condition equivalent equivalent to the design condition design condition when aa need need for CRE isolation isolation isis indicated. AMENDMENT AMENDMENT NO. NO. 142, 142, 161, 161, 171, 171, Revision Revision 2-4-(A494ý, 21 (,ll,194), 23 (A195) 180 180

• BASES BASES FOR 3.4.5 AND 4.4.5 CONTROL CONTROL ROOM AIR TREATMENT SYSTEM SYSTEM If If the control room air treatment system is found to be inoperable other than a~

reasonother inoperable for any reason inoperable ~RE an inoperabl.e during th~ CRE du.ring the powe~ operating power operat.ing . condition, condition, there is no immediate threat to the eRE CRE occupants and reactor operation may continue for a limited period penod ~f of ~Ime time while repairs repairs are are being being made. IfIfthe system cannot be repaired within seven days, the reactor is shutdown and brought to a cold shutdown within 36 hours. hours. If the control room air treatment system is found to be inoperable for any reason whenever recently irradiated fuel or .an If an irr~diate~ irradiated fuel cask is beingbeing handled in the reactor building, or during operations draining the reactor vessel (OPDRVs), there IS operations with a potential for draining Immediate thre~t is no immediate t~e. threat to the eRE CRE occupants occupants and these activities may continue for a limited period of time while repairs are being made. If If the system cannot be repaired within seven days, these activities must be immediately suspended. If the unfiltered If unfiltered inleakage of potentially contaminated contaminated air past the eRE CRE boundary boundary and into the CRE can result in CRE occupant radiological dose greater than the calculated greater calculated dose of the licensing basis analyses of DBA consequences consequences (allowed to be up to 5 rem total effective effective dose equivalent TEDE)), or inadequate protection of CRE occupants from hazardous TEDE>>, hazardous chemicals or smoke, the CRE boundary boundary is inoperable. inoperable. If If in the power power operating operating condition, actions must be taken to restore an operable operable CRE boundary boundary within 90 days. During the period period that the CRE boundary is considered inoperable, action must be initiated to implement mitigating initiated implement mitigating actions to lessen lessen the effect on CRE occupants occupants from the potential hazards hazards of a radiological or chemical event or aa challenge from smoke. Actions must be taken within 24 hours to verify that in the event of a DBA, the the mitigating actions will ensure that CRE occupant radiological exposures exposures will not exceed exceed the calculated calculated dose of the licensing basis analyses analyses of DBA consequences, and that CRE occupants consequences, occupants are protected protected from hazardous hazardous chemicals chemicals and smoke. These mitigating actions (i.e., actions that are taken to offset the consequences consequences of the inoperable inoperable CRE eRE boundary) should be preplanned preplanned for implementation implementation upon entry into the condition, regardless regardless of whether whether entry is intentional or unintentional. unintentional. The 24 hour period allowed allowed is reasonable based on the low probability probability of a DBA occurring during this this time period, and the use of mitigating actions. The 90 day period is reasonable based on the determination determination that the mitigating actions will ensure ensure protection protection of eRE CRE occupants occupants within analyzed analyzed limits while limiting the probability probability that CRE occupants implement protective occupants will have to implement protective measures measures that adversely affect their ability to control the reactor and maintain itit in a safe shutdown condition in the event of a DBA. In addition, the 90 day may adversely period reasonable time to diagnose, plan and possibly period is a reasonable possibly repair, and test most problems with the CRE boundary. The testing performed performed for TS 4.4.5.g 4.4.5.g verifies verifies the operability operability of the CRE boundary by testing for unfiltered unfiltered air inleakage inleakage past the CRE boundary and and into the CRE. The details of the testing are specified in specified the Control Control Room Envelope Habitability Program. Envelope Habitability Program. The CRE CRE is considered considered habitable habitable whenwhen the radiological radiological dose dose to CRE occupants occupants calculated in the licensing licensing basis analyses analyses of DBA consequences consequences is no more than 5 rem TEDETEDE and the CRE occupantsoccupants are are protected protected from hazardous hazardous chemicals chemicals and smoke. This surveillance surveillance requirement requirement verifies verifies that thethe unfiltered inleakage into the CRE unfiltered air inleakage eRE is no greater than the flow rate assumed assumed in the licensing licensing basis basis analyses of DBA consequences. consequences. When unfiltered air inleakage unfiltered greater than inleakage is greater than the assumed flow rate during the power power operating condition, condition, TS 3.4.5.f must be entered. The actions actions allow allow time to restore restore the CRE boundary eRE boundary to operable operable status provided mitigating mitigating actions can ensure that ensure that the CRE remains within within the licensing basis basis . habitability habitability limits limits for the occupants occupants following an accident. Compensatory m~asures are discussed Compensatory measures discussed in Regulatory Regulatory Guide Guide 1.196, 1.196, Section Section C.2.7.3, C.2. 7 .3, (Ref. (Ref. 2)2) which which endorses, endorses, with with exceptions, exceptions, NEI NEI 99-03, 99-03, Section Section 8.4 8.4 and and Appendix Appendix FF (Ref. (Ref. 3). 3). These compensatory measures These compensatory measures may may also also be be used used as as mitigating actions as required-by TS 3.4.5.f. Temporary analytical methods may also be used mitigating actions as required by TS 3.4.5.f. Temporary analytical methods may also be used as compensatory measures as compensatory measures to restore operability operability Revision Revision 23 23 (A(A195) 195) 180a 180a

• BASES BASES FOR AND 4.4.5 CONTROL ROOM AIR TREATMENT SYSTEM FOR 3.4.5 AND (Ref.

(Ref. 4). Options Options for restoring the eRE CRE boundary to operable status DBA cons~quence basis DBA: include changing the licensing basis the repairing the analysis, repairing consequence analysis.' eRE CRE boundary, or a combination of these actions. Depending upon the nature test inleakage test nature of the problem and the corrective action, a full scope Inleakage may not be necessary to establish that the eRE CRE boundary has been restored to operable status. High efficiency particulate prevent clogging of the iodine adsorber. The adsorbers to prevent particulate absolute (HEPA) filters are installed before the charcoal adsorbers The charcoal adsorbers are installed to reduce the potential intake of radioiodine to the control room. The in-place test results charcoal should indicate a system leak tightness of less than 1 percent bypass leakage efficiency of at least 99 percent removal of leakage for the charcoal adsorbers and a HEPA efficiency DOP particulates. The laboratory carbon sample test results should indicate a radioactive radioactive methyl iodide removal efficiency of at least 95 percent, which is derived derived from applying a safety factor of 2 to the charcoal filter efficiency of 90 percent assumed assumed in analyses of design basis basis accidents. If If the efficiencies efficiencies of the HEPA filter and charcoal charcoal adsorbers are as specified, adequate radiation protection will be provided such that resulting doses will be less than the allowable levels stated in 10 CFR 50.67. Operation of the fans Significantlysignificantly different different from the design flow will change the removal efficiency efficiency of the HEPA filters and charcoal adsorbers. Pressure drop across the combined HEPA filters and charcoal adsorbers adsorbers of less than 1.5 inches "Ofof water at the system design design flow rate will indicate that the filters and adsorbers adsorbers are not clogged clogged by excessive amounts excessive amounts of foreign matter. Pressure Pressure drop should be determined at least determined once once per operating operating cycle to show system performance performance capability. The frequency of tests and sample analysisanalysis are necessary necessary to show the HEPA filters and charcoal charcoal adsorbers adsorbers can perform perform as evaluated. The The charcoal charcoal adsorber efficiency test should allow for charcoal sampling to be conducted using an ASTM D3803-1989 03803-1989 approved method. If If test results are unacceptable, unacceptable, all adsorbent in the system system shall be replaced with an adsorbent adsorbent meeting meeting the physical property specifications specifications of Table Table 5-1 of ANSI 509-1980. The replacement replacement charcoal for the adsorber tray removed for the test should meet the same adsorbent quality. Any HEPA filters found defective HEPA defective shall be replaced replaced with filters qualified pursuant to ANSI 509-1980. Revision Revision 2323 (A195) (A 195) 18Gb 180b

• BASES FOR 3.4.5 AND BASES AND 4.4.5 CONTROL CONTROL ROOM AIR TREATMENTTREATMENT SYSTEM The two principal accidents accidents for which the Control Room Air Treatment System System (CRATS)

(CRATS) must be operable are a loss of coolant accident accident (LOCA) and a refueling accident involving "recently irradiated" fuel. In addition to these limiting events, events events occurring occurring during during handling handling of an irradiated fuel cask and operations with a potential for draining the reactorreactor vessel (OPDRVs) can be postulated postulated to cause a fission product product release. Thus, the the CRATS CRA TS is required required to be operable during handling handling of an irradiated fuel cask and during OPDRVs. OPDRVs. Refueling Accident analysis is based on an alternative The Refueling alternative source source term (AST) methodology methodology (10(10 CFR 50.67 and Regulatory Guide 1.183). This This analysis concluded that the calculated total effective equivalent (TEDE) values to the control room occupants, the exclusion area effective dose equivalent area boundary, and the low population below the TEDE criteria established in 10 CFR 50.67 without population zone are well below without crediting crediting operation of the CRA TS, as long CRATS, long as the the fuel is allowed to decay for at least 24 hours following reactor shutdown. As a result, "recently irradiated" irradiated" fuel is defined defined as fuel that has occupied occupied part of a critical reactor core within 24 hours; i.e., reactor fuel that has decayed corewithin hours following reactor shutdown. Therefore, the decayed less than 24 hours the CRATS is not required to be operable during movement of decayed irradiated fuel that is no longer considered decayed irradiated irradiated." Conversely, the considered "recently irradiated." the CRATS is required movement of recently irradiated required to be operable during movement irradiated fuel assemblies. Operation Operation of the system for 15 15 minutes every month will demonstrate operability of the filters and adsorber system. If significant If significant painting, fire or chemical release occurs such that the HEPA filter or charcoal adsorber could become become contaminated contaminated from the the fumes, chemicals or foreign materials, the same tests and sample analysis shall be performed performed as required for operational use. The The determination of significant determination significant shall be made made by the operator on duty at the time of the incident. Knowledgeable members should be Knowledgeable staff members be consulted prior to making consulted making this determination.

References:

1.

1. UFSAR, Section III.B.

2. Regulatory Guide 1.196, Revision 0, May 2001.

Regulatory 2001.

3. Habitability Assessment."

NEI 99-03, "Control Room Habitability Assessment." June 2001.2001.

4. Letter from Eric J. Leeds (NRC)

(NRC) to James James W. Davis (NEI)(NEI) dated January 30, 2004, "NEI Draft White Paper, Use of the Generic dated January Generic Letter 91-18 Process Process and Alternative Source Terms in the Context of Control Room Habitability. "" (ADAMS Accession Accession No. ML040160868). ML040160868). AMENDMENT AMENDMENT NO. 142, Revision 21 (A194), 26 (A201) Revision-21-(A494), (A201) 181

• BASES FOR 3.6.2 BASES AND 4.6.2 PROTECTIVE 3.6.2 AND INSTRUMENTATION PROTECTIVE INSTRUMENTATION Each reactor operating condition has a related related reactor instrumentation system operability reactor mode switch position for the safety system. The instrumentation requirements of the related safety system. For example, for each mode switch position is based on the requirements drywell pressure example, the specific high drywell pressure trip systems must be tripped or operable any time core spray, containment containment spray, automatic depressurization isolation.

depressurization or containment isolation* functions are required. In instrumentation systems where two trip systems are required In instrumentation required to initiate action, either operable or one is tripped. either both trip systems are operable Having one trip system already tripped does not decrease Having reliability in terms of initiating the desired action. decrease the reliability action. However, the probability probability of spurious actuation is increased. Certain instrument channels or sensor inputs to instrument channels channels may be bypassed bypassed without affecting affecting safe operation. The basis for allowing bypassing of the specified specified SRM's, IRM's, LPRM's APRM's is discussed LPRM's and APRM's discussed in Volume I (Section VII- VII-C.1.2)*. The high area temperature temperature isolation function for the cleanup system system has one trip system. There There are three instrument channels; each has four sensor inputs. Only two instrument channels channels are required since the area covered covered by anyone any one sensor sensor is also covered covered by a sensor in one of the other two instrument channels. The shutdown shutdown cooling system also has one trip system for high area temperature temperature isolation. However, since the area area of concern is much smaller, only one instrument channel is provided. Four sensors provide input to the the channel. Since the area covered covered is relatively small only three of the four sensors are required to be operable in order to assure isolation when needed. Table 3.6.2b requires that the low-low reactor reactor vessel water level instrumentation instrumentation that initiates isolation isolation of the shutdown cooling system be be operable operable with the reactor mode switch in the Shutdown Shutdown and Refuel positions. Two trip systems must be operable or in the tripped condition. condition in the hot shutdown shutdown condition. However, in the cold shutdown and refueling conditions, only one trip system (with two instrument instrument channels) must be operable operable so long as shutdown cooling system integrityintegrity is maintained. maintained. System integrity is maintained maintained provided provided the piping is intact and maintenance is being no maintenance being performed that has the potential for draining draining the reactor vessel through the system. If If one low-low low-low water level instrument channel in a required Trip System becomes inoperable and cannot cannot be restored or placed placed in the tripped condition within the the allowed time, the associated shutdown shutdown cooling line should be isolated. However, if the shutdown shutdown cooling cooling function is needed to provide provide core cooling, isolating the shutdown shutdown cooling line is not desirable. Table 3.6.2b, Note 0), j), allows the shutdown shutdown cooling line to remain remain unisolated unisolated and the system to remain in service service provided action is immediately initiated initiated to restore the channel to operable operable status. The alternative alternative action action is to . immediately initiate action to isolate immediately isolate the shutdown cooling system, which may require that alternate alternate decay heat heat removal capabilities be be provided. The term "immediately" means that the action pursued without delay action should be pursued delay and in a controlled manner. Either of these actions actions must continue until the channel is restored to operable operable status or the shutdown cooling system is isolated. Manual Manual initiation is available available for scram, reactor reactor isolation isolation and containment containment isolation. In order to manually initiate initiate other systems, each pump and each valve is independently independently initiated from the control room. Containment Containment spray spray raw water cooling is not automatically automatically initiated. Manual initiation of each pump is required as discussed in 3.3.7 3.3.7 above.

• FSAR; Letter, R.R. Schneider to A. Giambusso, dated November 15, 1973
• FSAR; Letter, RR Schneider to A. Giambusso, dated November 15,1973 AMENDMENT AMENDMENT NO. NO. 142, Revision 24 (A197)

(A 197) 249

• BASES BASES FOR FOR 3.6.2 3.6.2 AND AND 4.6.2 PROTECTIVE PROTECTIVE INSTRUMENTATION INSTRUMENTATION

a. The set points included in pOints included in the tables are those used in the tables are transient analysis and the transient and the accident analysis. The the accident The high flow set pointpoint for for main steam the main steam line line is is 105 105 psi psi differential.

differential. This represents a flow This represents approximately 4.4x1 flow of approximately 06 lb/hr. 4.4x1 06 The high flow set Ib/hr. The set point for the the emergency emergency cooling system supply line is cooling system is ::;_ 11.5 differential. This 11.5 psi differential. This represents represents a flow of of approximately 9.8x1 05 lb/hr approximately 9.8x105 Ib/hr at rated rated conditions. conditions. Emergency Cooling Emergency Cooling Initiation Initiation The emergency cooling The emergency cooling initiation logic is separated initiation logic separated into two trip systems which use trip systems one-out-of-two taken twice use a one-out-of-two twice logic logic single trip system actuation of aa single configuration. The actuation configuration. cause aa half system will cause emergency cooling half emergency cooling system initiation. system for initiation. A trip system for the the emergency cooling initiation parameter emergency parameter provides provides the protective action of de-energizing protective action de-energizing one of two DC solenoid of the two solenoid valves valves for for each the two air-operated each of the air-operated condensate condensate returnreturn isolation valves. A high reactorreactor pressure pressure or low-low reactor water level low-low reactor level signal from an instrument channel will de-energize instrument channel corresponding time de-energize its corresponding delay relay after time delay 12 seconds. If either after 12 either of of the two time delay delay relays in a trip system system times out, the two control control circuits associated with that trip system will change circuits associated change state causing one of the the two DC solenoid valves for each each of the two condensate condensate return isolation de-energize. This results in the insertion of aa isolation valves to de-energize. half emergency cooling system initiation cooling system initiation signal signal where the condensate condensate return isolation valves do isolation valves do not open. A full initiation initiation will occur when at least one time delay delay relay each of the two trip systems times out, and all four control relay in each change state to control circuits change de-energize both de-energize both DC solenoid valves for each of the two condensate condensate return thereby opening both valves. It return isolation valves, thereby It is important to recognize important recognize that pulling the fuses for (or. otherwise de-energizing) (or otherwise de-energizing) the DC solenoidsolenoid valves for the condensate condensate return isolation valves will affect the isolation capability on a high steam flow isolation signal. isolation signal. Emergency Emergency Cooling Cooling Isolation Isolation emergency cooling systems (loops) occurs Automatic isolation of the emergency occurs on a high steam flow isolation signal from the four LlP AP transmitters connected to the steam supply lines (two transmitters per steam transmitters connected steam line). Each LlP AP transmitter provides the sensor transmitter provides sensor instrument channel. Automatic isolation of an emergency inputs to its respective instrument emergency cooling system involves closure of both motor-operated steam supply isolation valves and the condensate operated condensate return isolation affected system. [Note that the isolation valve in the affected the requirements of Table 3.6.2c 3.6.2c do not apply to the drain and common loop vent valves since the isolation of these valves is to Requirements also do not apply to the individual prevent bypass leakage. Requirements individual loop vent valves which allow for vent isolation of one one Emergency Cooling loop while maintaining (1) Emergency operable.] For the high steam flow isolation maintaining the other loop operable.] isolation parameter, each emergency emergency cooling system system is required to have two tripped or operable 1rip trip systems, with two operable instrument channels per operable operable trip system. Both instrumentinstrument channels for a given emergency cooling system provide provide isolation trip signals to both of the system's trip systems in a one-out-of-two logic configuration for each trip system. The trip of either trip system will initiate initiate an isolation of the affected system. A trip system for the high steam flow isolation parameter provides the protective action of AMENDMENT NO. 142, AMENDMENT 142,149. 149. Revision W. 4-0. 20 20 250

• BASES FOR 3.6.2 AND BASES AND 4.6.2 PROTECTIVE INSTRUMENTATION PROTECTIVE INSTRUMENTATION closing one of the two steam supply isolation valves and energizing one of the two DC solenoid valves to close the condensate condensate return isolation valve.

The high level in the scram discharge discharge volume is provided provided to assure that there is still sufficient sufficient free volume in the discharge system system to receive the control rod drives discharge. Following a scram, bypassing bypassing is permitted to allow draining of the discharge volume and resetting of the reactor protection system relays. Since all control rods are completely inserted inserted following a scram and since the the bypass of this particular particular scram initiates initiates a control rod block, it permissible to bypass this scram is permissible scram function. The scram trip associated associated with the shutdown shutdown position of the mode mode switch can be reset after 10 seconds. The condenser low-low-low low-low-low vacuum and the main main steam line isolation isolation valve position signals are bypassed in the startup startup and refuel positions of the reactor mode switch switch when the reactor pressure is less than 600 psig. These are bypassed bypassed to allow warmup of thethe main steam lines and to provide a heat sink during startup. AMENDMENT NO. 142, 119, AMENDMENT 149. Revision 4-0, W, 2020 250a 250a

• BASES FOR 3.6.3 AND BASES AND 4.6.3 EMERGENCY EMERGENCY POWER SOURCES SOURCES Other Other than the Station turbine generator, the Station is supplied by four independent independent sources of ac power; power; two 115 kv transmission lines, and two diesel-generators.

diesel-generators. Any Anyoneone of the required power sources will provide required power provide the power power required required for aa LOCA. Engineering Engineering calculations calculations show that a LOCA LOCA concurrent with a loss of offsite powerpower and the single failure of one of the diesel-generators diesel-generators (DG) (DG) results in a loading for the remaining remaining diesel-generator diesel-generator that is below the unit's 2000 hour/year rating. This loading is greater greater than that required during aa Station shutdown condition. The monthly monthly test run paralleled manufacturer's recommendation paralleled with the system is based on the manufacturer's recommendation for these units in this this type of service. The testing during operating cycle will simulate simulate the accident accident conditions conditions under under which operation operation of the diesel-generators diesel-generators is required. major equipment required. The major equipment comprising comprising the maximum maximum diesel-generator diesel-generator loading loading is given in Figure IX-6*. As mentioned mentioned above, aa single diesel-generator diesel-generator is capable of providing the required required power power to equipment equipment following a LOCA. Two fuel oil storage tanks are provided provided with piping interties interties to permit supplying supplying either diesel-generator. A two-day two-day supply will provide adequate adequate time to arrange arrange for fuel makeup makeup ifif needed. needed. The full capacity of both tanks will hold hold a four-day four-day supply. It has been demonstrated It demonstrated in Section XV.B.3.23* that even with complete complete dc loss the reactor can be safely isolated and the emergency cooling system will be operative with makeup water to the emergency emergency cooling system shells maintained maintained manually. Having at least one dc dc . battery system available will permit: automatic automatic makeup makeup to the shells rather than manual, closing of the d-c actuated actuated isolation valve on all lines from the primary system and the suppression chamber, maintenance maintenance of electrical electrical switching functions functions in the Station and providing providing emergency lighting emergency lighting and communications communications power. There are two physically physically separate electrically independent, safety-related separate and electrically safety-related battery battery systems (11 (11 and 12). Each Each system includes one 125-volt station battery, two 100% capacity static battery battery chargers connected in parallel, connected parallel, and one dc power power distribution (battery) board. board. During normal operation, operation, the 125-volt 125-volt dc loads powered from the battery loads are powered battery chargers chargers with the batteries floating on the system. Each battery charger has ample ample power power output capacity capacity for the steady steady state operation operation of connected connected loads required during normal operation, operation, while while at the same same time maintaining maintaining its battery fully charged. In In case of loss of normal power to the battery charger, the dc loads are automatically automatically powered powered from the battery. Both battery systems, each consisting of one battery, at least one battery each.consisting battery charger, and the associated associated dc power distribution distribution (battery) board, are required to be operable operable for all reactor operating conditions except conditions except cold shutdown. In addition, a battery battery system must be operable operable whenever whenever its associated DG is required to be operable since the battery system is a support system system for the DG.DG. A battery battery system shall have a minimum of 106 volts at the battery terminals to be considered considered operable.

 *FSAR
 *FSAR AlMENDMENT NO. 142.

AMENDMENT 142, Revision 28 258

0 BASES FOR BASES 3.6.3 AND FOR 3.6.3 0 EMERGENCY POWER AND 4.6.3 EMERGENCY POWER SOURCES SOURCES 0 If aa battery battery system system becomes becomes inoperable, inoperable, that that battery battery system system must must be be returned returned to anan operable status within 24 hours. If the 24-hour operable status 24-hour allowed allowed outage time cannot be met, then Specification Specification 3.0.1 must be entered 3.0.1 must entered immediately. second paragraph immediately. The second Specification 3.0.1 provides paragraph of Specification provides two options: two

1. Place Place the unit in aa condition consistent the unit consistent with with the individual specification; however, in this case, the individual specification; the individual specification (i.e.,

individual specification (Le., 3.6.3) does not provide does provide anyany action to take take when when aa battery battery system hashas been been inoperable inoperable for more more than 24 To determine 24 hours. To determine required required actions actions and action and action completion completion times, the individual individual specifications specifications for the systems systems supported supported by thethe battery battery system should be system should entered and be entered and reviewed to determine reviewed applicable actions. If determine applicable actions are If no actions are applicable given reactor operating applicable for the given condition, then no operating condition, no actions actions are are required. required. or or

2. Place operational condition in which the specification the unit in an operational Place the specification is is not applicable (i.e.,

not applicable (Le., cold shutdown). Revision 28 28 258a

• BASES BASES 3.6.11 AND 4.6.11 ACCIDENT AND 4.6.11
• MONITORING INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION Accident monitoring instrumentation ensures that sufficient information monitoring instrumentation available on selected information is available parameters to monitor and assess selected plant parameters assess capability is consistent with the recommendations these variables during and following an accident. This capability NUREG-0578, "TMI-2 Lessons recommendations of NUREG-0578, Lessons Recommendations," NUREG-0737, Report and Short-Term Recommendations,"

Learned Task Force Status Report "Clarification of NUREG-0737, "Clarification TMI Action Plan Requirements," ofTMI Requirements," November 1980, NUREG-0661, November Evaluation Report Mark I Containment NUREG-0661, "Safety Evaluation Program," and the NRC Final Rule, "Combustible Containment Long Term Program," "Combustible Gas Control in Containment," made effective effective October 16, 2003 16,2003 (68 FR 54123). surveillance intervals and surveillance Specified surveillance surveillance and maintenance maintenance outage times have been been determined in accordance accordance with GENE-770-06-1, GENE-770-06-1, "Bases for Changes Changes to Surveillance Surveillance Test Intervals Intervals and Allowed Out-Of-Service Out-Of-Service Times for Selected Instrumentation Technical Specifications," Selected Instrumentation as approved approved by the NRC and documented documented in the SER (letter to R. D. Binz IV from C. E. Rossi dated dated July 21,21, 1992). Action 3a and Action 4a of Table Table 3.6.11-2 require that with the number of OPERABLE channels less than the total Number Number of Channels Channels shown in Table 3.6.11-1, 3.6.11-1, a Special prepared and submitted Special Report must be prepared submitted to the NRC within 14 14 days following the event. The term "event""event" refers to the reason that an instrument channel channel is inoperable. For the purpose of applying applying Action 3a and Action 4a of Table 3.6.11-2, 3.6.11-2, removal of a single accident monitoring instrumentation accident monitoring instrumentation channel from service service for the sole purpose of performing performing routine TS required required considered an event requiring preparation surveillances is not considered preparation and submittal of a 14-day 14-day Special Report. If If a single accident monitoring monitoring instrumentation channel instrumentation channel is removed from service for other other activities (e.g., to perform preventive preventive maintenance), or if a channel fails, these if a channel these events preparation and submittal of aa Special Report events require preparation Report in accordance accordance with Actions 3a and 4a. AMENDMENT NO. 142, Revision AMENDMENT A191, 19 Revision 44, 15 (A191), 273

• BASES FOR BASES 3.6.15 AND FOR 3.6.15 AND 4.6.15 4.6.15 MAIN MAIN CONDENSER CONDENSER OFFGASOFFGAS Restricting the Restricting the gross gross radioactivity radioactivity rate rate of of noble noble gases gases from from the the main main condenser condenser provides provides assurance that the total effective assurance that dose equivalent effective dose equivalent to an an individual at the exclusion individual exclusion area area boundary boundary will notnot exceed exceed aa very very small small fraction fraction of of the limits limits of of 10 CFR CFR 50.67 in the the event this this effluent effluent is inadvertently inadvertently discharged directly discharged directly to the environment environment without without treatment. This specification specification implements implements the the requirements requirements ofof General Ger)eral Design Criteria Criteria 60 60 and 64 64 Appendix A to 10 of Appendix 10 CFR Part 50. The The primary primary purpose providing this specification purpose of providing specification is to limit buildup buildup of of fission product product activity activity within the the station systems which station which would would result result if high fuel fuel leakage leakage were to be permitted extended periods.

permitted over extended AMENDMENT NO. 142, 176, Revision 21 (A194) AMENDMENT 296

• ATTACHMENT ATTACHMENT 3 REPORT CONSISTENT CONSISTENT WITH 10 CFR 54.37(b) ON HOW EFFECTS OF AGING AGING OF NEWLY-IDENTIFIED NEWLY-IDENTIFIED STRUCTURES, STRUCTURES, SYSTEMS, COMPONENTS ARE SYSTEMS, OR COMPONENTS MANAGED MANAGED
• Nine Mile Point Nuclear Station, LLC October 26, 2009

ATTACHMENT ATTACHMENT 3 REPORT CONSISTENT REPORT CONSISTENT WITH 10 CFR 54.37(b) 54.37(b) ON ON HOW EFFECTS OF AGING AGING OF NEWLY-IDENTIFIED STRUCTURES, NEWLY-IDENTIFIED STRUCTURES, SYSTEMS, SYSTEMS, OR COMPONENTS COMPONENTS ARE MANAGED MANAGED This report is in lieu of adding a level of detail to the Nine Mile Point Unit 1 Updated Final Safety Analysis Report (UFSAR) that is greater than in the remainder of the UFSAR, including including the License License Renewal Supplement. Renewal Supplement. An entryentry* on the NRC website, "Frequently "Frequently Asked Questions (FAQs) About License Renewal Renewal Inspection Procedure Procedure (IP) 71003, 'Post-Approval

                                                                   'Post-Approval Site Inspection Inspection for License Renewal,"'

Renewal,'" relates to the amount of detail required per 10 CFR 54.37(b). It states, "The "The NRC staff will consider consider it acceptable acceptable if the summary summary information information included in the FSAR update is consistent consistent with the requirements requirements of 10 CFR 54.2l(d), 54.2 1(d), and the guidance provided in Revision of NUREG-l1800, 'Standard Revision' 1 ofNUREG-1800, 'Standard Review Plan for Review of License Renewal Applications Applications for Nuclear Power Power Plants' (SRP-LR), provided Plants' (SRP-LR), provided that the licensee licensee has supplied supplied the technical technical details (as described documented described in RIS 2007-16) in another documented submittal to the NRC." The information in this report is consistent with the technical information information previously previously submitted submitted to the NRC with the Amended Amended License Renewal Application (ALRA). (ALRA). On July 14,2005, 14, 2005, Nine Mile Point Nuclear Nuclear Station, LLC (NMPNS) submitted an ALRA to the NRC to (NMPNS) submitted renew the operating operating licenses for Nine Mile Point Nuclear Nuclear Station Unit 1 (NMP1) (NMP1) and Unit 2 (NMP2) for an additional additional 20 years beyond expiration dates of August 22, 2009 (NMP1) beyond the original expiration (NMP 1) and October 31, 31, 2026 (NMP2). Within the ALRA, system system tables were provided to define the component component types, functions and the Aging Management Management Programs that applied. Lists of individual components components within scope of of license renewal were not required to be provided. Subsequent to the completion of the necessary necessary reviews, audits, responses to Requests for Additional (RAIs), and resolutions of other questions, the NRC published NUREG-1900, Information (RAIs), NUREG-1900, Safety Evaluation Evaluation Report Related to the License RenewalRenewal of Nine Mile Point Nuclear Station, Units 1 and 2, in September of 2006, which documented the NRC staffs review of the information submitted submitted to them through April 21,21, 2006. The renewed operating operating licenses licenses for NMP1 NMPI and NMP2 were issued on October October 31, 2006, extending 31,2006, extending the license for NMPI NMP1 to August 22,2029, 22, 2029, and NMP2 to October 31,2046. 31, 2046. For holders of a renewed renewed operating license, 10 CFR 54.37(b)54.37(b) requires that newly-identified newly-identified Structures, Structures, Systems, or Components (SSCs) be includedincluded in the Final Safety Analysis Report (FSAR) update required by 10 CFR 50.71(e) describing describing how the effects effects of aging will be managed. Newly-identified Newly-identified SSCs are those SSCs that were installed in the plant at the time of the License RenewalRenewal of NMP 1 and NMP2, but ofNMPl were not evaluated as part of the ALRA (as (as discussed in RIS 2007-16). During the period of January 2008 to May 2009, a review of updated drawings and the site component component database revealed approximately approximately 5600 components components installed installed before before October October 31, that had not 31, 2006 that previously been been screened screened for license renewal applicability. applicability. . Of the components that were identified as in-scope and subject to aging management review, 222 components were found to be subject to aging management management requirements requirements and are, therefore, "newly"newly identified" and subject to 10 CFR 54.3 identified" 54.37(b) 7(b) reporting requirements. requirements. The 222 components can be broken down into two groups. The first group of 179 components is already addressed within the tables submitted with the ALRA. The second group consists of 43 components that would not have been addressed under any of the existing tables in the ALRA. The 222 "newly-identified" "newly-identified" components have been assigned assigned to existing Aging Management Management Programs Programs and appropriate aging management strategies strategies have been invoked to adequately detect and manage manage the applicable aging effects effects throughout the period of extended operation and can be verified by NRC inspection. 1 of 8 10f8

ATTACHMENT 3 ATTACHMENT CONSISTENT WiTH REPORT CONSISTENT WITH 10 CFR 54.37(b) 54.37(b) ON HOW EFFECTS OF AGING AGING OF OF NEWLY-IDENTIFIED STRUCTURES, NEWLY-IDENTIFIED STRUCTURES, SYSTEMS, SYSTEMS, OR COMPONENTS COMPONENTS MANAGED ARE MANAGED Although Although the tables provided in the ALRA ALRA are not actually actually being revised, the attached tables show the changes that would have been made had the 43 "newly-identified" "newly-identified" components been included in the ALRA. The table numbers shown herein correlate correlate with those provided in the ALRA. A list of the 222 consistent with the detail provided in the ALRA. The components is not provided in this document consistent changes from the existing ALRA existing ALRA are shown shown in italics in the attached attached tables.

• 2 of 8 20f8
• REPORT CONSISTENT WITH 10 REPORT CONSISTENT ATTAIENT ATTAaENT3 10 CFR NEWLY-IDENTIFIED STRUCTURES, NEWLY-IDENTIFIED 54.37(b) ON CFR 54.37(b) 3 HOW EFFECTS OF ON HOW SYSTEMS, OR COMPONENTS STRUCTURES, SYSTEMS, COMPONENTS AGING OF OF AGING OF ARE MANAGED ARE MANAGED Table 3.3.2.A-1 Table Auxiliary Systems 3.3.2.A-I Auxiliary Systems Circulating Water NMP 1I Circulating Water System Management Evaluation Summary of Aging Management System - Summary Evaluation Components)

(12 Components) NUREG-Aging Effect Aging Effect Aging 1801 Component Component Intended Intended Requiring Management Management Volume 2 Table 1I Type Function Material Material Environment Environment Management Management Program Program Item Item Notes External LBS Copper Alloys Copper Air Air None None None None Surfaces Surfaces (Zinc <15%) (Zinc <15%) Valves Valves LBS Copper Copper Alloys Raw Water Water Loss of O[2en Cy'cie Open Cycle VII.C.I.2.1 VII.C. 1.2.1 3.3.l.A-17 3.3.l .A- 17 A (Zinc <15%) (Zinc <15%) Material Material Cooling Cooling Water Program Program 3 of 8 30f8

• REPORT PURSUANT TO 10 CFR REPORT PURSUANT ATTA~ENT3 ATTIOENT 3 54.37(b) ON CFR 54.37(b) HOW EFFECTS OF AGING ON HOW SYSTEMS, OR STRUCTURES, SYSTEMS, NEWLY-IDENTIFIED STRUCTURES, NEWLY-IDENTIFIED AGING OF COMPONENTS ~

OR COMPONENTS OF ARE MANAGED ARE MANAGED Table 3.3.2.A-7 Table Auxiliary Systems 3.3.2.A-7 Auxiliary Systems NMP 1 Emergency Emergency Diesel Diesel Generator Summary of Generator System - Summary of Aging Management Evaluation Aging Management Evaluation (4 Components) Components) NUREG-NUREG-Aging Effect Aging Effect Aging 1801 Component Intended Intended Requiring Requiring Management Management Volume 22 Table Table 1 Type Type Function Material Material Environment Environment Management Management Program Program Item Item Notes Valves PB Copper Copper Treated Water Treated None None None Alloys Temperature Temperature (Zinc >15%) <140°F

                                       <140°F And Aluminum Aluminum Bronze Bronze 4 of 8 40f8

• REPORT PURSUANT TO 10 CFR REPORT PURSUANT NEWLY-IDENTIFIED STRUCTURES, NEWLY-IDENTIFIED ATTAaENT3 ATTMIIENT 3 HOW EFFECTS OF 54.37(b) ON HOW CFR 54.37(b)

SYSTEMS, OR STRUCTURES, SYSTEMS, OF AGING AGING OF OR COMPONENTS COMPONENTS OF ARE MANAGED ARE MANAGED 3.3.2.A-19 Auxiliary Table 3.3.2.A-19 Auxiliary Systems Systems NMPII Service NMP Service Water Water System Summary of Aging System - Summary Management Evaluation Aging Management Evaluation (14 Components) Components) NUREG-NUREG-Aging Effect Aging 1801 Component Component Intended Intended Requiring Management Management Volume Volume 2 Table 1 Type Function Function Material Material Environment Environment Management Management Program Program Item Item* Item Notes External LBS Cast Cast Air None None None None Surfaces Surfaces Austenitic Austenitic Stainless Stainless Steel Valves Valves LBS Cast Cast Raw Water Loss of Open Cycle 012en Cr.cie VII.C.1.2-a VII.C.l.2-a 3.3.1.A-17 3.3.l.A-17 A Austenitic Material Material Cooling Water Cooling Stainless Stainless System Sr.stem Steel Program Program 5 of 8 50f8

• REPORT PURSUANT REPORT PURSUANT TO TO 10 ATTA~ENT3 ATTAOENT 3 10 CFR CFR 54.37(b)

NEWLY-IDENTIFIED STRUCTURES, NEWLY-IDENTIFIED 54.37(b) ON ON HOW EFFECTS OF HOW EFFECTS STRUCTURES, SYSTEMS, OR COMPONENTS SYSTEMS, OR AGING OF OF AGING COMPONENTS OF

                                                                                                            *O ARE MANAGED ARE   MANAGED Table 3.3.2.A-23 Table            Auxiliary Systems 3.3.2.A-23 Auxiliary NMP 1 Turbine NMP    Turbine Building HVAC HV AC System System - Summary                         Evaluation Management Evaluation Summary of Aging Management Components)

(9 Components) NUREG-NUREG-Aging Effect Aging Effect Aging 1801 Component Intended Intended Requiring Management Management Volume 2 Table 1 Type Function Function Material Material Environment Environment Management Management Program Program Item Item Notes Heat Heat LBS Carbon or Air Loss of Preventive Preventive VII.F.2.1.2 3.3.l.A-05 3.3.l .A-05 A Exchanger Exchanger Low Alloy Material Material Maintenance Maintenance Steel (Yield Steel (Yield Program Program Strength Strength

                     <100 Ksi) 6 of 8 60f8

• PURSUANT TO 10 REPORT PURSUANT REPORT ATTARENT ATTAaENT3 CFR 54.37(b) 10 CFR NEWLY-IDENTIFIED STRUCTURES, NEWLY-IDENTIFIED 3

HOW EFFECTS OF AGING ON HOW 54.37(b) ON SYSTEMS, OR STRUCTURES, SYSTEMS, COMPONENTS OR COMPONENTS AGING OFOF ARE MANAGED ARE MANAGED Table 3.4.2.A-5 Table 3.4.2.A-5 Steam Steam and Power Conversion Conversion System System NMPI1 Condenser NMP Condenser Air Removal Removal and System - Summary Off-Gas System and Off-Gas Summary of Aging Management Evaluation Management Evaluation Components) (3 Components) NUREG-NUREG-Aging Effect Aging 1801 1801 Component Intended Intended Requiring Management Management Volume 2 Table Table 1 Type Function Function Material Environment Environment Management Management Program Program Item Item Item Notes Notes External LBS Cast Cast Air None None None Surfaces Austenitic Stainless Steel Stainless Valves LBS Cast Cast Treated Water Treated Water Loss of One Time 3.4.1 .A-02 3.4.l.A-02 F E Austenitic or Steam, Steam, Material, Material, Inspection Ins12.ection Stainless Stainless Steel temperature > temperature Cracking Program Profffam 482°F, Low 48JOF, Flow Flow Water Chemistry Chemistry' Control Control Program Program 7of8 70f8

• REPORT PURSUANT TO REPORT PURSUANT TO 10 ATTA*ENT 3 54.37(b) ON 10 CFR 54.37(b)

STRUCTURES, SYSTEMS, NEWLY-IDENTIFIED STRUCTURES, NEWLY-IDENTIFIED SYSTEMS, OR AGING OF HOW EFFECTS OF AGING ON HOW OR COMPONENTS COMPONENTS OF ARE ARE MANAGED MANAGED Table Table 3.5.2.A-4 3.S.2.A-4 Structures Structures and Component Component Supports Supports NMPI1 Fuel Handling NMP Handling System Management Evaluation Summary of Aging Management System - Summary Evaluation (1 Component) (1 Component) NUREG-Aging Effect Effect Aging 1801 Component Component Intended Requiring Requiring Management Management Volume 2 Table 1 Type Function Material Environment Environment Management Management Program Program Item Item Notes A uxBridge AuxBridge NSS Aluminum Aluminum Air None None None None Alloy Alloy 8 of 8 80f8

u.s. U.S. NUCLEAR NUCLEAR REGULATORY COMMISSION DOCKET 50-220 LICENSE DPR-63 NINE MILE POINT NINE NUCLEAR STATION NUCLEAR STATION UNIT 1

• FINAL SAFETY ANALYSIS REPORT ANALYSIS (UPDATED)

(UPDATED) OCTOBER 2009 OCTOBER REVISION 21 REVISION

Nine Mile Point Unit 1 UFSAR

• INSERTION INSERTION INSTRUCTIONS The following instructions revision INSTRUCTIONS instructions are for the insertion insertion of the current revision into the Nine Mile Point Unit 1 FSAR (Updated).

(Updated). current Remove pages listed pages listed in in the REMOVE REMOVE column and replace them with the pages listed pages listed in in the INSERT INSERT column. column. Dashes (---)

                                                 --- )   in either in  either   column indicate             required.

indicate no action required. Vertical Vertical bars have been been placed inin the margins of inserted pages inserted pages and tables tables to indicate indicate revision revision locations. locations.

• UFSAR Revision Revision 21 21 FII-I FII-l October 2009 2009

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• REMOVE REMOVE INSERTION INSTRUCTIONS INSERTION LIST OF LIST INSTRUCTIONS OF EFFECTIVE EFFECTIVE PAGES INSERT INSERT PAGES EP EP ii EP i EP EP 1-1 1-1 EP 1-1 i-i EP EP 2-1 2-1 EP 2-1 2-1 EP EP 3-1 3-1 EP 3-1 3-1 EP EP 4-1 4-1 EP 4-1 4-1 EP 5-1 5-1 EP 5-1 5-1 EP EP 6-1 6-1 EP 6-1 6-1 EP 7-1 7-1 EP 7-1 7-1 EP 8-1 thru EP 8-2 8-1 8-2 EP 8-1 8-1 thru EP 8-2 8-2 EP 9-1 9-1 EP 9-1 9-1 EP 10-1 10-i thru EP 10-7 10-7 EP 10-1 thru EP 10-7 10-i thru 10-7 EP 11-1 li-i EP 11-1 il-i EP 12-1 12-1 EP 12-1 12-i EP 13-1 13-1 EP 13-1 13-1 EP 14-1 14-1 EP 14-1 14-i EP 15-1 thru EP 15-1 EP 15-3 15-3 EP 15-1 thru EP 15-3 15-1 thru 15-3 EP 16-1 EP 16-3 16-1 thru EP 16-3 EP 16-1 thru 16-1 thru EP 16-3 16-3 EP 17-1 17-1 thru EP EP 17-2 17-2 EP 17-1 17-1 thru EP 17-2 17-2 EP 18-1 18-1 EP 18-1 18-1 EP A-1 A-I EP A-I A-1 EP B-1 B-I EP B-I B-1 EP C-1 C-i EP C-1 C-I
• UFSAR Revision 21 21 FII-2 October 2009

Nine Mile Mile Point Unit 11 UFSAR

• REMOVE REMOVE INSERTION INSTRUCTIONS INSERTION INSTRUCTIONS VOLUME VOLUME 11 INSERT i/ii i/ii iia/iib iia/iib iii/iv iii/iv thru xa/xb iii/iv thru xa/xb xc/xd xvii/xviia thru xxi/xxiixxi/xxii xvii/xviia xxi/xxii xvii/xviia thru xxi/xxii xxv/xxvi thru xxvia/xxvib xxv/xxvi thru xxvia/xxvib xxxiii/xxxiv xxxiii/xxxiv xxxiva/xxxivb xxxv/xxxvi xxxv/xxxvi xxxv/xxxvi xxxv/xxxvi xxxvii/xxxviii xxxvii/xxxviii xxxvii/xxxviia xxxvii/xxxviia xxxviib/xxxviii xxxvi ib/xxxvii i xlvii/xlviia xlvii/xlviia xlvii/xlviia xlvii/xlviia xlix/1 xlix/l xlix/1 xlix/l la/lb 1-9/10 1-9/10 I-9/9a I-9/9a I-9b/10 1-11/12 thru 1-13/14 1-11/12 thru 1-13/14 I-14a/14b I-14a/14b 1-17/18 1-17/18 I-18a/18b I-18a/18b 1-21/-

1-21/- 1-21/- 1-21/- T 1-2 Sh 1 thru 7 T 1-2 Sh 1 thru 7 11-3/4 11-3/4 11-3/4 11-3/4 II-4a/4b II-4a/4b thru II-4c/4d 11-9/- 11-9/- 11-9/- 11-9/- T T II-5/T II-S/T 11-6 11-6 II-5/T 11-6 T II-S/T 11-6 T 11-9 Sh 1/2 1/2 II-10 Sh 1 & 2 T 11-10 111-9/10 thru III-12a/12b 111-9/10 III-12a/12b 111-9/10 thru III-12a/12b 111-9/10 III-12a/12b 111-39/40 111-39/40 111-39/40 111-39/40 III-1 F 111-1 III-1 F 111-1 IV-19/19a IV-19/19a IV-19/19a IV-19/19a IV-31/32 IV-31/32 IV-31/32 IV-31/32 V-3/4 V-3/4 V-3/3a V-3/3a V-3b/4 V-3b/4 V-21/- V-21/- V-21/- V-21/- T V-I Sh 3/T 3/T V-2 V-2 T V-I Sh 3/- 3/- T V-2 Sh 1/2 1/2 T V-3/T V-4 V-4 T V-3/T V-4 V-4 Revision 21 UFSAR Revision 21 FII-3 FII-3 October 2009

Nine Mile Nine Mile Point Point Unit Unit 1 UFSAR

• REMOVE REMOVE INSERTION INSTRUCTIONS INSERTION INSTRUCTIONS VOLUME 2 VOLUME INSERT i/ii i/ii i/ii i/ii iia/iib iia/iib iii/iv iii/iv thru xa/xb thru iii/iv thru iii/iv thru xa/xb xc/xd xvii/xviia thru xxi/xxii xvii/xviia xxi/xxii xvii/xviia thru xvi.i/xviia thru xxi/xxii xxi/xxii xxv/xxvi thru xxv/xxvi thru xxvia/xxvib xxvia/xxvib xxv /xxvi thru xxv/xxvi thru xxvia/xxvib xxvia/xxvib xxxiii/xxxiv xxxiii/xxxiv xxxiii/xxxiv xxxiii/xxxiv xxxiva/xxxivb xxxiva/xxxivb xxxv/xxxvi xxxv/xxxvi xxxv/xxxvi xxxv/xxxvi xxxvii/xxxviii xxxvii/xxxviii xxxvii/xxxviia xxxvi i/xxxvi ia xxxviib/xxxviii xxxviib/xxxviii xlvii/xlviia xlvii/xlviia xlvii/xlviia xlvii/xlviia xlix/l xlix/l xlix/l xlix/l la/lb la/lb VI-19/20 thru VI-23b/24 VI-19/20 thru VI-23b/24 VI-19/20 VI-29/30 VI-29/30 VI-29/30 VI-29/30 T VI-3a VI-3a Sh 4 VI-3a Sh 4 T VI-3a
• VII-7b/8 thru VII-12a/12b VII-14e/14f VII-15/16 VII-21/22 VII-12a/12b VII-14e/14f thru VII-14g/14h VII-14g/14h VII-19/20 VII-15/16 thru VII-19/20 VII-7b/8 thru VII-12a/12b VII-7b/8 VII-12c/12d VII-12c/12d VII-14e/14f thru VII-14e/14f VII-12a/12b thru VII-14g/14h VII-15/16 thru VII-19/20 VII-15/16 VII-21/21a VII-21/21a VII-21b/22 thru VII-14g/14h VII-19/20 VII-43/44 VII-43/44 VIII-38a/38b VIII-38a/38b VIII-38a/38b VIII-38a/38b T VIII-3 Sh 2 VIII-3 Sh 2 T VIII-3 T VIII-3 Sh 9/10 T VIII-3 VIII-3 Sh 9/109/10 T VIII-3 Sh lOa/lOb 10a/lOb T VIII-3 VIII-3 Sh 10a/10b lOa/lOb T VIII-4 Sh 2 VIII-4 Sh 2 T VIII-4 F IX-6 IX-6 F IX-6 IX-6
• UFSAR Revision 21 21 FII-4 October 2009

Nine Mile Point Unit 1 UFSAR

• REMOVE INSERTION INSERTION INSTRUCTIONS INSTRUCTIONS VOLUME 3 INSERT i/ii i/ii i/ii i/ii iia/iib iia/iib iii/iv iii/iv thru xa/xb iii/iv iii/iv thru xa/xb xc/xd xvii/xviia thru xxi/xxii xxi/xxii xvii/xviia thru xxi/xxii xvii/xviia xxi/xxii xxv/xxvi xxv /xxvi thru xxvia/xxvib xxvia/xxvib xxv/xxvi xxv/xxvi thru xxvia/xxvib xxxiii/xxxiv xxxiii/xxxiv xxxiva/xxxivb xxxv/xxxvi xxxv/xxxvi xxxv/xxxvi xxxv/xxxvi xxxvii/xxxviii xxxvii/xxxviii xxxvii/xxxviia xxxvii/xxxviia xxxviib/xxxviii xxxviib/xxxviii xlvii/xlviia xlvii/xlviia xlvii/xlviia xlvii/xlviia xlix/l xlix/l xlix/l xlix/l la/lb X-9/10 X-9/1O X-9/10 X-10a/10b X-lOa/lOb X-29/30 X-29/30 thru X-33/34 X-29/30 X-29/30 thru X-33/34
• X-41/42 X-41/42 X-43/44 X-43/44 X-55/56 X-55/56 F X-3X-3 X-41/42 X-41/42 X-43/43a X-43/43a X-43b/43c X-43b/43c thru X-43f/44 X-55/56 X-55/56 X-57/58 F X-3 X-3 1OA-i/ii thru 10A-v/vi 10A-i/ii 1OA-v/vi 1OA-i/ii thru 10A-v/vi 10A-i/ii 1OA-v/vi 1OA-5/6 IOA-5/6 thru 1OA-13/14 IOA-13/14 1OA-5/6 thru IOA-13/14 IOA-5/6 1OA-13/14 1OA-14a/14b IOA-14a/14b IOA-17/18 1OA-17/18 1OA-17/18 IOA-17/18 1OA-21/22 IOA-21/22 thru 1OA-25b/26 thru IOA-25b/26 1OA-21/22 thru IOA-25b/26 IOA-21/22 1OA-25b,/26 1OA-51/52 IOA-51/52 1OA-51/52 IOA-51/52 1OA-65/66 IOA-65/66 thru 1OA-67/68 thru IOA-67/68 1OA-65/66 thru IOA-67/67a IOA-65/66 1OA-67/ 67a 1OA-67b/68 IOA-67b/68 1OA-74a/74b IOA-74a/74b 1OA-74a/74b IOA-74a/74b 1OA-77/78 thru IOA-77/78 1OA-79/80 thru IOA-79/80 1OA-77/78 thru IOA-79/80 IOA-77/78 IOA-79/ 80 1OA-80a/80b IOA-80a/80b 1OA-81/82 IOA-81/82 1OA-81/82 IOA-81/82 1OA-91/92 IOA-91/92 1OA-91/92 IOA-91/92 1OA-113/-

IOA-113/- iOA-113/- IOA-113/- 1OB-205/- IOB-205/- IOB-205/- 1OB-205/- IOB-205a/- 1OB-205a/- 1OB-219/220 IOB-219/220 thru IOB-223/- 1OB-223/- IOB-219/220 IOB-223/224 1OB-219/220 thru 1OB-223/224

• XI-5/6 thru XI-9/9a XI-9b/10 F XI-3 XI-3 UFSAR Revision XI-9/9a Revision 2121 XI-5/6 thru XI-9/9a XI-9b/9c XI-9b/9c FII-5 FII-5 F XI-3 XI-3 XI-9/9a thru XI-9d/10 October 2009

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• REMOVE REMOVE INSERTION INSTRUCTIONS INSERTION INSTRUCTIONS VOLUME 3 (Cont'd.)

(Cont'd.) INSERT F XI-5 XI-5 F XI-5 XI-5 XII-15/16 XII-15/16 XII-16a/16b XII-16a/16b XII-21/22 XII-21/21a XII-21/21a XII-21b/22 XllI-1/2 thru XIII-1/2 thru XIII-25/26 XllI-1/2 thru XIII-25/- XIII-1/2 XIII-25/- T XIlI-1 T XIII-l XIII-1 Sh 1/2 T XIII-1 1/2 T XIII-2 XIII-2 XIII-2 T XIII-2 F XIII-1 XlII-1 thru thru F XIII-3 XIII-3 F XIII-l XllI-1 thru F XIII-3 XIII-3 F XIII-3a XIII-3a F XIII-4 XIII-4 thru F XIII-5 XIII-5 XIII-4 thru F XIII-5 F XIII-4 XIII-5

• UFSAR Revision UFSAR Revision 21 21 FII-6 FII-6 October 2009

Nine Mile Point Unit 1 UFSAR

• REMOVE REMOVE INSERTION INSTRUCTIONS INSTRUCTIONS VOLUME 44 INSERT i/ii i/ii i/ii i/ii iia/iib iia/iib iii/iv thru xa/xb iii/iv iii/iv iii/iv thru xa/xb xc/xd xvii/xviia xvii/xviia thru xxi/xxii xxi/xxii xvii/xviia thru xxi/xxii xvii/xviia xxi/xxii xxv/xxvi thru xxvia/xxvib xxv/xxvi xxvia/xxvib xxv /xxvi thru xxvia/xxvib xxv/xxvi xxvia/xxvib xxxiii/xxxiv xxxiii/xxxiv xxxiii/xxxiv xxxiii/xxxiv xxxiva/xxxivb xxxv/xxxvi xxxv/xxxvi xxxv/xxxvi xxxv/xxxvi xxxvii/xxxviii xxxvii/xxxviii xxxvii/xxxviia xxxvii/xxxviia xxxviib/xxxviii xxxviib/xxxviii xlvii/xlviia xlvii/xlviia xlvii/xlviia xlvii/xlviia xlix/l xlix/l xlix/l xlix/l la/lb XV-31/32 thru XV-33/34 XV-33/34 XV-31/32 thru XV-31/32 XV-33/34 XV-33/34 XV-43/44 XV-43/44 thru XV-49/50 XV-49/50 XV-43/44 thru XV-43/44 XV-49/50 XV-49/50 XV-55/56 XV-55/56 thru XV-61b/62 XV-61b/62 XV-55/56 thru XV-55/56 XV-61b/62 XV-61b/62
• XV-67/68 XV-67/68 XV-75/76 XV-79/79a XV-79/79a XV-79b/80 XV-79b/80 XV-81/82 XV-81/82 thru XV-69/70 XV-69/70 XV-75/76 thru XV-77/78 XV-77/78 XV-67/68 thru XV-67/68 XV-70a/70b XV-70a/70b XV-75/76 thru XV-75/76 XV-79/80 XV-79/80 XV-81/-

XV-81/- XV-69/70 XV-69/70 XV-77/78 XV-77/78 T XV-5/T XV-6 XV-6 T XV-5/T XV-6 T XV-6 XV-7/T XV-8 T XV-7/T XV-8 T XV-7/T XV-7a XV-7a T XV-7b/T XV-8 XV-8 XV-21a/T XV-22 T XV-21a/T XV-22 T XV-21a/T XV-21a/T XV-22 XV-22 XV-23/T XV-24 T XV-23/T XV-24 T XV-23/T XV-24 XV-24 XV-25/26 T XV-25/26 T XV-25 Sh 1/2 1/2 T XV-26 Sh 1/2 1/2 XV-27/28 thru T XV-29d/30 T XV-27/28 XV-29d/30 T XV-27/28 thru T XV-29d/-XV-29d/- T XV-30 XV-30 XV-31/32 T XV-31/32 T XV-31 Sh 1/2 1/2 XV-31 T XV-31 Sh 3/T XV-32 XV-32 T XV-33 XV-33 XV-33/34 T XV-33/34 T XV-34 XV-34 T XV-34a XV-34a T XV-34b XV-34b T XV-35/36 XV-35/36 T XV-35 XV-35 T XV-35a thru T XV-35d XV-35a XV-35d T XV-36 XV-36 F XV-56h XV-56h XV-73 F XV-73 F XV-74 XV-74 XVI-23/24 XVI-23/24 XVI-23/24 XVI-23/24 Revision 21 UFSAR Revision 21 FII-7 FII-7 October 2009

Nine Mile Point Unit 1 UFSAR

• REMOVE REMOVE INSERTION INSERTION INSTRUCTIONS VOLUME 4 VOLUME INSTRUCTIONS (Cont/d.)

(Cont'd.) INSERT XVI-25/26 XVI-25/26 XVI-25/25a XVI-25/25a XVI-25b/26 XVI-25b/26 XVI-63/64 thru XVI-69/70 XVI-63/64 XVI-69/70 XVI-63/64 thru XVI-69/70 XVI-63/64 XVI-69/70 T XVI-9a Sh 1/2 1/2 T XVI-9a Sh 1/2 1/2 C.1-7/8 C.1-7/8 C.1-7/7a C.1-7/7a C.1-7b/8 C.1-7b/8 C.1-9/l0 C.1-9/10 thru C.l-ll/12 C.1-I1/12 C.1-9/10 C.1-9/l0 thru C.1-11/12 C.l-ll/12 T C-I C-l Sh 1 thru 10 T C-I C-l Sh 1 thru 1010 UFSAR Revision Revision 2121 FII-8 October 2009

• u.s. NUCLEAR U.S.

COMMISSION DOCKET 50-220 DOCKET 50-220 LICENSE LICENSE DPR-63 REGULATORY NUCLEAR REGULATORY COMMISSION NINE MILE POINT STATION NUCLEAR STATION NUCLEAR UNIT 1

• SAFETY FINAL SAFETY ANALYSIS REPORT ANALYSIS (UPDATED)

OCTOBER 2009 OCTOBER REVISION 21

Nine Mile Point Unit 1 UFSAR LIST OF EFFECTIVE PAGES EFFECTIVE PAGES CONTENTS GENERAL TABLE OF CONTENTS GENERAL Page Page Revision Page Page Revision Number Number Number Number Number Number Number Number i 21 21 xxxi xxxi 16 16 ii ii 21 21 xxxii xxxii 16 16 iii iii 21 21 xxxiii xxxiii 20 20 iv iv 21 21 xxxiv 21 21 v 21 21 xxxiva 19 19 vi vi 21 21 xxxivb 17 17 vii vii 21 21 xxxv 21 21 viii viii 21 21 xxxvi xxxvi 21 21 ix ix 21 21 xxxvii xxxvii 21 21 x 21 21 xxxviia 21 21 xa xa 21 21 xxxvi ib xxxviib 21 21 xb 21 21 xxxviii xxxviii 16 16 xc xc 21 21 xxxix 16 16 xd xd 21 21 xl xl 16 16 xi xi 20 20 xli xli 16 16 xii xii 18 18 xlii xlii 16 16 xiii xiii 16 16 xliii xliii 16 16 xiv 20 20 xliv xliv 19 19 xv xv 20 20 xlv 19 19 xvi XVl 20 20 xlvi xlvi 17 17 xvii xvii 21 21 xlvii xlvii 21 21 xviia 21 21 xlviia 21 21 xviib 21 21 xlviib xlviib 17 17 xviii xviii 20 20 xlviii xlviii 16 16 xix 21 21 xlix xlix 21 21 xx xx 21 21 1 21 21 xxa xxa 19 19 la la 21 21 xxb 17 17 Ib lb 21 21 xxi xxi 20 20 li Ii 16 16 xxii xxii 16 16 lii Iii 16 16 xxiii xxiii 16 16 liii liii 16 16 xxiv 16 16 liv liv 16 16 xxv xxv 21 21 lv Iv 16 16 xxvi xxvi 21 21 Ivi lvi 16 16 xxviaa xxvi 21 21 xxvib 18 18 xxvii xxvii 19 19 xxviii xxviii 16 16 xxix 20 20 xxixa 20 20 xxixb 20 20 xxx 16 16 Revision 21 UFSAR Revision 21 EP i1 October 2009 October 2009

Nine Mile Point Unit 1 UFSAR EFFECTIVE PAGES LIST OF EFFECTIVE PAGES SECTION I SECTION Page Page Revision Page Page Revision Revision Number Number Number Number Number Number Number Number 1-1 I-I 18 18 1-2 1-2 18 18 1-3 1-3 17 17 1-4 17 17 1-5 1-5 17 17 1-6 1-6 17 17 1-7 1-7 18 18 1-8 1-8 17 17 1-9 1-9 21 21 1-9a I-9a 21 21 I-9b I-9b 21 21 1-10 1-10 17 17 1-11 I-l1 21 21 1-12 1-12 21 21 1-13 21 21 1-14 21 21 I-14a I-14a 21 21 I-14b I-14b 21 21 1-15 18 18 1-16 18 18 1-17 21 21 1-18 21 21 I-18a I-18a 21 21 I-18b I-18b 21 21 1-19 18 18 1-20 17 17 1-21 1-21 21 21 T 1-1 I-i 17 17 T 1-2 T 1-2 Sh 1 21 21 T 1-2 T 1-2 Sh 2 21 21 T T 1-2 1-2 Sh 3 20 T 1-2 T 1-2 Sh 4 20 20 T 1-2 1-2 Sh 5 21 21 T 1-2 T 1-2 Sh 6 21 21 T 1-2 1-2 Sh 7 20 F 1-1 I-I 20 20

• UFSAR Revision Revision 21 21 1-1 EP 1-1 October 2009

Nine Mile Point Point Unit 1 UFSAR LIST OF EFFECTIVE PAGES PAGES SECTION II II Page Revision Revision Page Revision Number Number Number Number Number Number 11-1 18 18 11-2 18 18 11-3 21 21 11-4 21 21 II-4a II-4a 21 21 II-4b II-4b 21 21 II-4c II-4c 21 21 II-4d II-4d 21 21 11-5 18 18 11-6 15 15 11-7 15 15 11-8 18 18 11-9 21 21 T 11-1 15 15 T 11-2 15 15 T 11-3 15 15 T 11-4 15 15 T 11-5 21 21 T 11-6 18 18 T 11-7 15 15 T 11-8 Sh 1 15 15 T 11-8 Sh 2 15 15 T 11-9 Sh 1 21 21 T 11-9 Sh 2 21 21 II-10 Sh 1 T 11-10 21 21 T 11-10 II-10 Sh 2 21 21 F 11-1 11-1 14 14 F 11-2 14 14 F 11-3 14 14 F 11-4 14 14 F 11-5 14 14 F 11-6 14 14

• UFSAR Revision 21 21 EP 2-1 2-1 October 2009

Nine Mile Point Point Unit 1 UFSAR LIST OF EFFECTIVE PAGES PAGES SECTION III SECTION III Page Page Revision Revision Page Page Revision Revision Number Number Number Number Number Number Number Number 111-1 III-1 15 15 111-39 111-39 21 21 111-2 111-2 15 15 111-40 111-40 15 15 111-3 15 15 111-41 111-41 15 15 111-4 20 20 111-42 111-42 19 19 III-4a III-4a 20 20 111-43 111-43 19 19 III-4b III-4b 20 20 111-44 111-44 19 19 111-5 111-5 16 16 111-45 111-45 16 16 111-6 17 17 III-1 F 111-1 21 21 111-7 111-7 17 17 F 111-2 16 16 111-8 111-8 16 16 F 111-3 20 20 111-9 111-9 19 19 F 111-4 111-4 20 20 III-I0 III-10 21 21 F 111-5 111-5 20 20 III-II III-ll 21 21 F 111-6 111-6 16 16 111-12 21 21 F 111-7 111-7 20 20 III-12a III-12a 21 21 F 111-8 111-8 16 16 III-12b 21 21 F 111-9 111-9 16 16 111-13 19 19 III-10 F 111-10 14 14 111-14 15 15 F III-11 III-II 14 14 III-15 III-IS 16 16 F 111-12 17 17 111-16 16 16 F 111-13 18 18 111-17 17 17 F 111-14 18 18 111-18 16 16 F 111-15 111-15 17 17 111-19 15 15 F 111-16 18 18 111-20 17 17 F 111-17 16 16 111-21 15 15 F 111-18 14 14 111-22 15 15 F 111-19 14 14 111-23 16 16 F 111-20 14 14 111-24 111-24 17 17 F 111-21 111-21 17 17 111-25 111-25 16 16 F 111-22 14 14 111-26 111-26 15 15 F 111-23 14 14 111-27 111-27 17 17 111-28 111-28 17 17 111-29 111-29 17 17 111-30 111-30 17 17 111-31 111-31 16 16 111-32 111-32 15 15 111-33 111-33 15 15 111-34 111-34 15 15 111-35 111-35 15 15 111-36 111-36 15 15 111-37 111-37 15 15 111-38 15 15 Revision 21 UFSAR Revision UFSAR 21 EP 3-1 3-1 October 2009

Nine Mile Point Unit 1 UFSAR LIST OF EFFECTIVE PAGES PAGES SECTION SECTION IV IV Page Page Revision Page Page Revision Number Number Number Number Number Number Number Number IV-1 IV-i 15 15 F IV-4 IV-4 15 15 IV-2 IV-2 15 15 F IV-4a IV-4a 15 15 IV-3 IV-3 15 15 F IV-5 IV-5 14 14 IV-4 IV-4 15 15 F IV-6 IV-6 14 14 IV-5 IV- 5 17 17 F IV-7 IV-7 16 16 IV-6 IV- 6 15 15 F IV-7a IV-7a 16 16 IV-7 IV- 7 18 18 F IV-8 IV-8 14 14 IV-8 IV-8 15 15 F IV-9 IV-9 14 14 IV-9 IV- 9 15 15 IV-10 IV-10 15 15 IV-11 IV-1I 18 18 IV-12 17 17 IV-13 16 16 IV-14 18 18 IV-15 18 18 IV-16 15 15 IV-17 15 15 IV-18 IV-18 17 17 IV-18a IV-18a 17 17 IV-18b IV-18b 17 17 IV-19 21 21 IV-19a IV-19a 21 21 IV-19b IV-19b 16 16 IV-20 15 15 IV-21 IV-21 16 16 IV-22 15 15 IV-23 16 16 IV-24 17 17 IV-25 IV-25 20 20 IV-25a IV-25a 20 20 IV-25b IV-25b 20 IV-26 IV-26 16 16 IV-27 IV-27 16 16 IV-28 IV-28 16 16 IV-29 IV-29 16 16 IV-30 IV-30 16 16 IV-31 IV-31 16 16 IV-32 IV-32 21 21 IV-33 IV-33 20 F IV-1 IV-I 14 14 IV-2 F IV-2 14 14 IV-3 F IV-3 14 14 UFSAR Revision 21 UFSAR 21 EP 4-1 4-1 October 2009 2009

Nine Mile Point Point Unit 1 UFSAR LIST OF EFFECTIVE EFFECTIVE PAGES PAGES SECTION V SECTION V Page Page Revision Page Page Revision Revision Number Number Number Number Number Number Number Number V-I V-i 17 17 V-2 V-2 IS 15 V-3 V-3 21 21 V-3a V-3a 21 21 V-3b V-3b 21 21 V-4 V-4 17 17 V-S V-5 19 19 V-6 V-6 20 20 V-6a V-6a 20 20 V-6b V-6b 17 17 V-7 V-7 16 16 V-8 V-8 17 17 V-9 V-9 IS 15 V-I0 V-10 IS 15 V-II V-Il IS 15 V-12 V-12 19 19 V-13 19 19 V-14 V-14 17 17 V-IS V-15 16 16 V-16 V-16 19 19 V-17 V-17 17 17 V-18 17 17 V-19 V-19 17 17 V-20 V-20 17 17 V-21 V-21 21 21 T V-I V-1 Sh 1 17 17 T V-I V-i Sh 2 17 17 T V-I V-1 Sh 3 16 16 T V-2 V-2 Sh 1 21 21 T V-2 V-2 Sh 2 21 21 T V-3 V-3 21 21 T V-4 V-4 19 19 T V-5 V-5 IS 15 F V-I V-1 19 19 F V-2 V-2 16 16 F V-3 V-3 14 14 F V-4 V-4 16 16 F V-5 V-5 16 16 F V-6 V-6 16 16 F V-7 V-7 16 16 F V-8 V-8 16 16 UFSAR Revision 21 21 EP S-1 5-1 October 2009 2009

Nine Mile Mile Point Unit 1 UFSAR LIST OF EFFECTIVE EFFECTIVE PAGES PAGES SECTION VIVI Page Page Revision Revision Page Page Revision Number Number Number Number Number Number Number Number VI-1 VI-I 17 17 T VI-1 VI-i Sh 2 15 15 VI-2 VI-2 15 15 T VI-2 VI-2 15 15 VI-3 VI-3 15 15 T VI-3a Sh 1 18 18 VI-4 VI-4 15 15 T VI-3a Sh 2 18 18 VI-5 VI-5 15 15 T VI-3a Sh 3 18 18 VI-6 VI-6 15 15 T VI-3a Sh 4 21 21 VI-7 VI-7 17 17 T VI-3b Sh 1 17 17 VI-8 VI-8 17 17 T VI-3b Sh 2 18 18 VI-9 VI-9 17 17 T VI-3b Sh 3 18 18 VI-10 VI-10 17 17 T VI-3b Sh 4 18 18 VI-10a VI-lOa 17 17 T VI-4 VI-4 17 17 VI-lOb VI-10b 17 17 T VI-5 Sh 1 15 15 VI-11 VI-li 15 15 T VI-5 Sh 2 15 15 VI-12 15 15 F VI-1 VI-l 14 14 VI-13 16 16 F VI-2 VI-2 14 14 VI-14 16 16 F VI-3 VI-3 14 14 VI-15 17 17 F VI-4 VI-4 14 14 VI-16 17 17 F VI-4a VI-4a 14 14 VI-17 19 19 F VI-5 VI-5 14 14 VI-18 19 19 F VI-6 VI-6 14 14 VI-19 VI-19 15 15 F VI-7 VI-7 14 14 VI-20 VI-20 21 21 F VI-8 VI-8 14 14 VI-21 VI-21 21 21 F VI-9 VI-9 14 14 VI-21a VI-21a 20 20 F VI-10 14 14 VI-21b VI-21b 16 16 F VI-11 VI-lI 14 14 VI-22 VI-22 21 21 F VI-12 14 14 VI-23 21 21 F VI-13 14 14 VI-23a VI-23a 21 21 F VI-14 14 14 VI-23b VI-23b 17 17 F VI-15 14 14 VI-24 VI-24 21 21 F VI-16 14 14 VI-25 VI-25 17 17 F VI-17 14 14 VI-26 VI-26 17 17 F VI-18 14 14 VI-27 VI-27 19 19 F VI-19 14 14 VI-27a VI-27a 17 17 F VI-20 17 17 VI-27b VI-27b 17 17 F VI-21 VI-21 14 14 VI-28 VI-28 15 15 F VI-22 17 17 VI-29 VI-29 21 21 F VI-23 16 16 VI-30 VI-30 15 15 F VI-24 18 18 VI-31 VI-31 15 15 VI-32 VI-32 15 15 VI-33 VI-33 16 16 T VI-1 VI-i Sh 1 15 15 UFSAR Revision 21 UFSAR 21 6-1 EP 6-1 October 2009 October

Nine Mile Point Unit Unit 1 UFSAR LIST OF EFFECTIVE EFFECTIVE PAGES PAGES SECTION VII SECTION VII Page Page Revision Page Page Revision Number Number Number Number Number Number Number Number VII-l VII-1 15 15 VII-2s VII-25 15 15 VII-2 VII-2 17 17 VII-26 16 16 VII-2a VII-2a 17 17 VII-27 16 16 VII-2b VII-2b 17 17 VII-28 16 16 VII-3 VII-3 17 17 VII-29 15 15 VII-4 VII-4 17 17 VII-30 15 15 VII-s VII-5 19 19 VII-31 VII-31 15 15 VII-6 VII-6 19 19 VII-32 15 15 VII-7 VII-7 20 VII-33 15 15 VII-7a VII-7a 20 20 VII-34 18 18 VII-7b 20 20 VII-35 19 19 VII-8 VII-8 21 21 VII-36 17 17 VII-9 VII-9 21 21 VII-37 18 18 VII-I0 VII-10 21 21 VII-38 18 18 VII-II VII-II 21 21 VII-39 18 18 VII-12 21 21 VII-40 18 18 VII-12a VII-12a 21 21 VII-41 20 20 VII-12b VII-12b 21 21 VII-42 20 20 VII-12c VII-12c 21 21 VII-42a VII-42a 20 20 VII-12d VII-12d 21 21 VII-42b VII-42b 17 17 VII-13 20 VII-43 15 15 VII-14 20 20 VII-44 21 21 VII-14a VII-14a 20 T VII-l VII-1 15 15 VII-14b VII-14b 20 F VII-l VII-1 17 17 VII-14c VII-14c 20 F VII-2 VII-2 17 17 VII-14d VII-14d 20 20 F VII-3 VII-3 18 18 VII-14e VII-14e 21 21 F VII-4 VII-4 14 14 VII-14f VII-14f 21 21 F VII-s VII-5 14 14 VII-14g VII-14g 21 21 F VII-6 VII-6 16 16 VII-14h VII-14h 21 21 F VII-7 VII-7 14 14 VII-15 VII-IS 21 21 F VII-8 VII-8 14 14 VII-16 21 21 F VII-9 VII-9 14 14 VII-17 21 21 F VII-I0 VII-10 16 16 VII-18 21 21 F VII-II VII-II 14 14 VII-19 21 21 F VII-12 17 17 VII-20 21 21 F VII-13 17 17 VII-21 21 21 F VII-14 19 19 VII-21a VII-21a 21 21 F VII-15 VII-IS 19 19 VII-21b 21 21 F VII-16 19 19 VII-22 15 15 F VII-17 14 14 VII-23 17 17 VII-24 16 16 UFSAR Revision Revision 21 21 7-1 EP 7-1 October 2009

Nine Mile Mile Point Unit 1 UFSAR UFSAR LIST OF EFFECTIVE PAGES PAGES SECTION VIII SECTION VIII Page Page Revision Revision Page Page Revision Number Number Number Number Number Number Number Number VIII-1 VIII-I 16 16 VIII-35 18 18 VIII-2 VIII-2 16 16 VIII-36 17 17 VIII-3 VIII-3 19 19 VIII-37 16 16 VIII-3a VIII-3a 19 19 VIII-38 18 18 VIII-3b VIII-3b 19 19 VIII-38a VIII-38a 21 21 VIII-4 VIII-4 19 19 VIII-38b VIII-38b 21 21 VIII-5 VIII-5 16 16 VIII-39 18 18 VIII-6 VIII-6 16 16 VIII-40 15 15 VIII-7 VIII-7 16 16 VIII-41 15 15 VIII-8 VIII-8 17 17 VIII-42 18 18 VIII-9 VIII-9 18 18 VIII-42a VIII-42a 18 18 VIII-9a VIII-9a 16 16 VIII-42b 18 18 VIII-9b 16 16 VIII-43 18 18 VIII-10 17 17 VIII-44 18 18 VIII-II VIII-lI 15 15 VIII-45 18 18 VIII-12 16 16 VIII-46 18 18 VIII-13 17 17 VIII-47 16 16 VIII-14 18 18 VIII-48 16 16 VIII-15 15 15 VIII-49 18 18 VIII-16 17 17 VIII-50 16 16 VIII-17 20 VIII-51 16 16 VIII-18 20 VIII-52 17 17 VIII-19 20 20 VIII-53 15 15 VIII-20 20 20 VIII-54 15 15 VIII-20a VIII-20a 20 VIII-55 20 20 VIII-20b VIII-20b 20 20 VIII-56 20 20 VIII-21 15 15 VIII-1 T VIII-1 16 16 VIII-22 15 15 VIII-2 T VIII-2 15 15 VIII-23 15 15 VIII-3 T VIII-3 Sh 1 20 20 VIII-24 16 16 T VIII-3 VIII-3 Sh 2 21 21 VIII-25 17 17 T VIII-3 Sh 3 20 20 VIII-26 17 17 T VIII-3 VIII-3 Sh 4 20 20 VIII-27 17 17 T VIII-3 VIII-3 Sh 55 20 20 VIII-28 15 15 T VIII-3 VIII-3 Sh 6 15 15 VIII-29 16 16 T VIII-3 VIII-3 Sh 7 19 19 VIII-30 19 19 T VIII-3 VIII-3 Sh 8 19 19 VIII-31 15 15 T VIII-3 VIII-3 Sh 9 21 21 VIII-32 20 20 T VIII-3 VIII-3 Sh 1010 21 21 VIII-33 20 20 T VIII-3 VIII-3 Sh 10a lOa 21 21 VIII-34 20 20 T VIII-3 VIII-3 Sh lob lOb 21 21 VIII-34a VIII-34a 20 20 VIII-4 T VIII-4 Sh 11 17 17 VIII-34b VIII-34b 20 20 T VIII-4 VIII-4 Sh 2 21 21 UFSAR Revision 21 Revision 21 EP 8-1 8-1 October 2009

Nine Mile Nine Mile Point Point Unit Unit 11 UFSAR UFSAR LIST OF EFFECTIVE PAGES OF EFFECTIVE PAGES SECTION VIII SECTION VIII (Cont I d. ) (Cont'd.) Page Page Revision Revision Page Page Revision Revision Number Number Number Number Number Number Number Number VIII-4 Sh 33 T VIII-4 T 17 17 VIII-S Sh T VIII-5 T Sh 11 19 19 VIII-S Sh 22 T VIII-5 Sh T 19 19 VIII-1 F VIII-1 F 16 16 VIII-2 F VIII-2 F 20 20 VIII-3 F VIII-3 F 17 17 VIII-4 F VIII-4 F 17 17 VIII-S F VIII-5 F 14 14 VIII-6 F VIII-6 20 20 VIII-7 F VIII-7 F 14 14 VIII-8 F VIII-8 F 20 20 VIII-9 F VIII-9 F 14 14 VIII-10 F VIII-10 F 14 14 VIII-11 F VIII-II F 14 14 VIII-12 F VIII-12 F 20 20 VIII-13 F VIII-13 F 17 17 VIII-14 F VIII-14 17 17 VIII-1S F VIII-15 14 14 VIII-16 F VIII-16 14 14 F VIII-17 VIII-17 14 14 F VIII-18 VIII-18 16 16 F VIII-19 14 14 F VIII-20 VIII-20 14 14 F VIII-21 16 16 F VIII-22 FVIII-22 17 17 F VIII-23 14 14 F VIII-24 17 17 F VIII-2S VIII-25 14 14 F VIII-26 17 17 F VIII-26a VIII-26a 20 20 F VIII-26b 17 17 F VIII-27 14 14 F VIII-28 18 18 F VIII-29 17 17

• UFSAR Revision 21 21 EP 8-2 8-2 October 2009

Nine Mile Nine Mile Point Point Unit 1 UFSAR LIST OF EFFECTIVE PAGES PAGES SECTION IX SECTION IX Page Page Revision Revision Page Page Revision Revision Number Number Number Number Number Number Number Number IX-l IX-1 15 15 F IX-4 IX-4 14 14 IX-2 IX-2 20 20 F IX-5 IX-5 14 14 IX-2a IX-2a 20 20 F IX-6 IX-6 21 21 IX-2b IX-2b 20 20 F IX-7 IX-7 14 14 IX-3 IX-3 18 18 F IX-8 IX-8 17 17 IX-4 IX-4 17 17 IX-5 IX-5 15 15 IX-6 IX-6 15 15 IX-7 IX-7 15 15 IX-8 IX-8 19 19 IX-9 IX-9 19 19 IX-I0 IX-10 19 19 IX-II IX-II 19 19 IX-lla IX-lla 19 19 IX-lIb IX-lib 19 19 IX-12 15 15 IX-13 16 16 IX-14 16 16 IX-IS IX-15 17 17 IX-16 15 15 IX-17 17 17 IX-17a IX-17a 18 18 IX-17b IX-17b 17 17 IX-18 17 17 IX-19 15 15 IX-20 IX-20 15 15 IX-21 IX-21 15 15 IX-22 IX-22 17 17 IX-23 IX-23 17 17 IX-24 IX-24 17 17 IX-25 IX-25 18 18 IX-26 IX-26 20 IX-27 IX-27 20 20 IX-28 IX-28 20 IX-28a IX-28a 20 20 IX-28b IX-28b 20 IX-29 IX-29 15 15 T IX-1 IX-1 Sh 1 20 IX-1 Sh 22 T IX-l 17 17 F IX-l IX-1 20 20 F IX-2 IX-2 14 14 F IX-3 IX-3 14 14 UFSAR Revision 21 Revision 21 EP 9-1 9-1 October 2009 October 2009

Nine Mile Point Unit 1 UFSAR LIST OF EFFECTIVE EFFECTIVE PAGES PAGES SECTION SECTION X Page Revision Page Page Revision Revision Number Number Number Number Number Number Number Number X-I X-1 17 17 X-33 X-33 21 21 X-la X-1a 17 17 X-34 X-34 21 21 X-lb X-1b 17 17 X-35 X-35 20 20 X-2 X-2 16 16 X-36 X-36 20 20 X-3 X-3 19 19 X-37 X-37 20 20 X-4 X-4 17 17 X-38 X-38 20 20 X-5 X-5 17 17 X-39 X-39 19 19 X-Sa X-5a 17 17 X-40 X-40 19 19 X-5b X-5b 17 17 X-40a X-40a 19 19 X-6 X-6 18 18 X-40b X-40b 17 17 X-7 X-7 19 19 X-41 X-41 17 17 X-8 X-8 19 19 X-42 X-42 21 21 X-9 X-9 19 19 X-43 21 21 X-10 21 21 X-43a X-43a 21 21 X-lOa X-10a 21 21 X-43b X-43b 21 21 X-lOb X-10b 21 21 X-43c X-43c 21 21 X-II X-11 17 17 X-43d X-43d 21 21 X-12 17 17 X-43e X-43e 21 21 X-13 17 17 X-43f X-43f 21 21 X-14 17 17 X-44 X-44 21 21 X-IS X-15 17 17 X-45 X-45 16 16 X-16 16 16 X-46 X-46 16 16 X-17 20 20 X-47 X-47 20 20 X-18 20 20 X-48 X-48 20 20 X-19 20 20 X-49 X-49 16 16 X-20 20 20 X-50 X-50 16 16 X-20a X-20a 20 20 X-51 X-51 20 20 X-20b X-20b 20 20 X-52 X-52 20 20 X-21 X-21 20 X-52a X-52a 20 20 X-22 X-22 20 20 X-52b X-52b 20 20 X-23 20 20 X-53 X-53 17 17 X-23a X-23a 20 20 X-54 X-54 17 17 X-23b X-23b 20 20 X-55 X-55 17 17 X-24 19 19 X-56 X-56 21 21 X-25 20 20 X-57 X-57 21 21 X-26 20 20 X-58 X-58 21 21 X-27 X-27 16 16 F X-I X-1 14 14 X-28 20 20 X-2 F X-2 17 17 X-29 X-29 20 X-3 F X-3 21 21 X-30 X-30 21 21 X-4 F X-4 17 17 X-31 X-31 21 21 X-5 F X-5 20 20 X-32 X-32 21 21 F X-6 X-6 14 14 UFSAR Revision UFSAR Revision 21 21 10-1 EP 10-1 October 2009 October 2009

Nine Mile Point Point Unit 1 UFSAR PAGES LIST OF EFFECTIVE PAGES SECTION X (Cont SECTION X d. ) (Cont'd.) I Page Page Revision Page Page Revision Number Number Number Number Number Number Number Number F X-7 X-7 17 17 10A-18c 1OA-18c 20 20 F X-8 X-8 17 17 10A-18d 1OA-18d 20 20 F X-9 X-9 15 15 10A-18e 1OA-18e 20 20 F X-10 14 14 10A-18f 1OA-18f 20 20 F X-11 X-1l 14 14 10A-19 1OA-19 20 20 1OA-20 10A-20 16 16 lOA 10A 16 16 10A-21 1OA-21 18 18 10A-i 10A-i 20 20 10A-22 1OA-22 21 21 10A-ii 1OA-ii 21 21 10A-23 1OA-23 21 21 lOA-iii 10A-iii 21 21 1OA-24 10A-24 21 21 10A-iv 1OA-iv 20 20 10A-25 1OA-25 21 21 1OA-v 10A-v 20 20 10A-25a 1OA-25a 21 21 lOA-vi 1OA-vi 21 21 1OA-25b 10A-25b 21 21 lOA-via 1OA-via 20 20 1OA-26 10A-26 16 16 10A-vib 1OA-vib 20 20 10A-27 1OA-27 19 19 lOA-vii 1OA-vii 16 16 1OA-28 10A-28 16 16 lOA-viii 1OA-viii 16 16 1OA-29 10A-29 16 16 1OA-ix 10A-ix 16 16 10A-30 1OA-30 18 18 10A-1 10A-I 18 18 10A-31 1OA-31 18 18 1OA-2 10A-2 16 16 1OA-32 10A-32 T 2.5.1.1-1 16 16 1OA-3 10A-3 16 16 1OA-33 10A-33 T 2.5.1.1-2 16 16 1OA-4 10A-4 16 16 10A-34 1OA-34 T 2.5.1.1-3 16 16 10A-5 10A-5 16 16 1OA-35 10A-35 T 2.5.1.1-4 16 16 1OA-6 10A-6 21 21 1OA-36 10A-36 T 2.5.1.1-5 19 19 1OA-7 10A-7 21 21 1OA-37 10A-37 T 2.5.1.1-6 16 16 1OA-8 10A-8 21 21 1OA-38 10A-38 T 2.5.1.1-7 16 16 1OA-8a 10A-8a 21 21 1OA-39 10A-39 16 16 1OA-8b 10A-8b 21 21 1OA-40 10A-40 17 17 10A-9 1OA-9 21 21 1OA-41 10A-41 17 17 1OA-10 10A-10 19 19 1OA-42 10A-42 20 20 1OA-1I 10A-11 21 21 1OA-43 10A-43 20 20 1OA-12 10A-12 21 21 1OA-44 10A-44 20 20 1OA-13 10A-13 21 21 1OA-44a 10A-44a 20 20 1OA-14 10A-14 21 21 1OA-44b 10A-44b 17 17 1OA-14a 10A-14a 21 21 1OA-45 10A-45 19 19 1OA-14b 10A-14b 21 21 1OA-46 10A-46 17 17 1OA-15 10A-15 16 16 1OA-47 10A-47 17 17 1OA-16 10A-16 16 16 1OA-48 10A-48 2.5.3.4-1 T 2.5.3.4-1 16 16 1OA-17 10A-17 19 19 1OA-49 10A-49 2.5.3.4-1 T 2.5.3.4-1 16 16 1OA-18 10A-18 21 21 10A-50 lOA-50 19 19 1OA-18a 10A-18a 20 20 1OA-51 lOA-51 21 21 1OA-18b 10A-18b 20 20 1OA-52 lOA-52 20 UFSAR Revision 21 Revision 21 10-2 EP 10-2 October 2009 October 2009

Nine Mile Nine Mile Point Unit 1 UFSAR UFSAR LIST OF EFFECTIVE EFFECTIVE PAGES PAGES SECTION X SECTION (Cont d. ) X (Cont'd.) I Page Revision Page Revision Number Number Number Number Number Number Number Number 10A-52a 1OA-52a 20 20 1OA-85 10A-85 T 1. 2.2 1.2.2 18 18 10A-52b 1OA-52b 20 20 1OA-86 10A-86 T 1. 2.2 1.2.2 18 18 lOA-53 1OA-53 20 20 10A-87 1OA-87 T 1. 2.2 1.2.2 20 20 lOA-54 1OA-54 17 17 1OA-87a 10A-87a T 1. 2.2 1.2.2 20 20 lOA-55 1OA-55 16 16 10A-88 1OA-88 T 1. 2.2 1.2.2 18 18 lOA-56 1OA-56 16 16 10A-89 1OA-89 T 1. 2.2 1.2.2 18 18 lOA-57 1OA-57 20 20 1OA-89a 10A-89a T 1. 2.2 1.2.2 18 18 lOA-58 1OA-58 20 20 1OA-90 10A-90 T 3.2-1 3.2-1 17 17 lOA-59 1OA-59 20 20 1OA-91 10A-9l T 3.3-1 3.3-1 21 21 1OA-60 10A-60 20 20 1OA-92 10A-92 T 3.3-1 3.3-1 21 21 10A-60a 1OA-60a 20 20 10A-93 1OA-93 T 3.4-1 3.4-1 16 16 10A-60b 10A-60b 20 20 10A-94 1OA-.94 T 3.5-1 3.5-1 16 16 10A-61 1OA-61 19 19 10A-95 1OA-95 T 3.6-1 16 16 10A-62 1OA-62 16 16 1OA-96 10A-96 T 3.7-1 16 16 1OA-63 10A-63 18 18 1OA-97 10A-97 T 3.8-1 3.8-1 16 16 1OA-64 10A-64 19 19 1OA-98 10A-98 T 3.9-1 3.9-1 16 16 10A-65 1OA-65 16 16 1OA-99 10A-99 T 3.10-1 16 16 10A-66 1OA-66 21 21 1OA-100 10A-lOO T 3.10-1 16 16 1OA-67 10A-67 21 21 1OA-101 10A-10l T 3.1-1 16 16 1OA-67a 10A-67a 21 21 10A-102 10A-102 T 3.1-1 16 16 1OA-67b 10A-67b 21 21 10A-103 10A-103 T 3.1-1 31212 16 16 1OA-68 10A-68 16 16 10A-104 10A-l04 T 3.1-1 18 18 10A-69 1OA-69 17 17 10A-105 10A-l05 T 3.1.1-1 18 18 1OA-70 10A-70 20 20 10A-106 T 3.1.1-1 18 18 1OA-71 10A-7l 20 10A-107 10A-107 T 3.1.1-1 18 18 1OA-72 10A-72 20 20 10A-108 10A-l08 T 3.1.1-1 3.1.1-1 18 18 1OA-73 10A-73 20 1OA-108a 10A-108a T 3.1.1-1 3.1.1-1 19 19 1OA-74 10A-74 20 1OA-108b 10A-108b T 3.1.1-1 3.1.1-1 18 18 1OA-74a 10A-74a 21 21 10A-109 10A-109 T 3.1.1-2 3.1.1-1 18 18 10A-74b 1OA-74b 20 20 10A-l10 10A-110 T 3.1.1-2 3.1.1-1 19 19 10A-75 1OA-75 16 16 lOA-Ill 10A-Ill T 3.1.1-2 3.1.1-1 18 18 IOA-76 10A-76 16 16 10A-112 T 3.1.1-2 3.1.1-2 18 18 1OA-77 10A-77 21 21 10A-1l3 10A-113 T 3.1.1-2 3.1.1-2 21 21 1OA-78 10A-78 21 21 10A-114 T 3.1.1-2 3.1.1-2 18 18 1OA-79 10A-79 21 21 10A-114a 1OA-114a T 3.1.1-2 3.1.1-2 18 18 1OA-80 10A-80 21 21 10A-114b 1OA-114b T 3.1.1-2 18 18 1OA-80a 10A-80a 21 21 10A-115 T 3.1.1-3 3.1.1-2 18 18 1OA-80b 10A-80b 21 21 10A-1l6 10A-116 T 3.1.1-4 3.1.1-2 18 18 1OA-81 10A-8l 19 19 10A-1l7 10A-117 T 3.1.1-4 18 18 1OA-82 10A-82 21 21 10A-1l8 10A-118 T 3.1.1-5 18 18 1OA-83 10A-83 16 16 10A-119 10A-119 T 3.1.1-6 19 19 3.1.1-6 1OA-84 10A-84 20 20 10A-120 10A-120 T 3.1.1-6 18 18 Revision 21 UFSAR Revision 21 10-3 EP 10-3 October 2009 October 2009

Nine Mile Nine Mile Point Unit Unit 1 UFSAR LIST OF EFFECTIVE EFFECTIVE PAGES PAGES SECTION X SECTION X (Cont' d.) (Cont'd.) Page Page Revision Page Page Revision Number Number Number Number Number Number Number Number 10A-121 10A-121 T 3.1.1-7 3.1.1-7 18 18 F 1OA-8B 10A-8B 16 16 10A-122 1OA-122 T 3.1.1-8 3.1.1-8 18 18 F 1OA-8C 10A-8C 16 16 10A-123 1OA-123 T 3.1.1-8 3.1.1-8 18 18 F 1OA-8D 10A-8D 16 16 10A-123a 10A-123a T 3.1.1-8 3.1.1-8 18 18 F 1OA-9 10A-9 16 16 10A-124 1OA-124 T 3.1.1-9 3.1.1-9 18 18 F 1OA-9A 10A-9A 16 16 10A-125 IOA-125 T 3.1.1-9 3.1.1-9 18 18 F 1OA-9B 10A-9B 16 16 10A-126 1OA-126 T 3.1.1-9 3.1.1-9 19 19 F 1OA-9C 10A-9C 16 16 1OA-127 10A-127 T 3.1.1-9 3.1.1-9 19 19 F 10A-9D 1OA-9D 16 16 10A-127a 1OA-127a T 3.1.1-9 3.1.1-9 19 19 F 10A-1 1OA-I 16 16 10B lOB 16 16 F 10A-2 1OA-2 16 16 1OB-i 10B-i 18 18 F 10A-2A 1OA-2A 16 16 1OB-ii 10B-ii 18 18 F 10A-2B 1OA-2B 16 16 1OB-iii lOB-iii 20 20 F 10A-2C 1OA-2C 16 16 1OB-I 10B-1 18 18 1OA-2D F 10A-2D 16 16 10B-2 1OB-2 16 16 F 10A-3 1OA-3 18 18 10B-3 1OB-3 16 16 F 10A-3A 1OA-3A 18 18 10B-4 1OB-4 16 16 F 10A-3B 1OA-3B 18 18 10B-5 1OB-5 16 16 1OA-3C F 10A-3C 18 18 10B-6 1OB-6 16 16 F lOA-3D 10A-3D 18 18 10B-7 1OB-7 16 16 10A-4 F 10A-4 19 19 10B-8 1OB-8 16 16 1OA-4A F 10A-4A 17 17 10B-9 1OB-9 20 20 1OA-4B F 10A-4B 17 17 1OB-9a 10B-9a 20 20 1OA-4C F 10A-4C 17 17 1OB-9b 10B-9b 20 1OA-4D F 10A-4D 17 17 1OB-10 10B-10 18 18 F 10A-5 1OA-5 17 17 1OB-1l 10B-11 16 16 1OA-5A F 10A-5A 17 17 1OB-12 10B-12 18 18 1OA-5B F 10A-5B 17 17 1OB-13 10B-13 18 18 1OA-5C F 10A-5C 17 17 1OB-14 10B-14 18 18 1OA-5D F 10A-5D 17 17 1OB-14a 10B-14a 18 18 10A-6 F 10A-6 16 16 1OB-14b 10B-14b 17 17 1OA-6A F 10A-6A 16 16 1OB-15 10B-15 18 18 1OA-6B F 10A-6B 16 16 1OB-16 10B-16 16 16 1OA-6C F 10A-6C 16 16 1OB-17 10B-17 18 18 F 10A-6D 16 16 1OB-18 10B-18 18 18 F 10A-7 1OA-7 17 17 1OB-18a 10B-18a 18 18 1OA-7A F 10A-7A 17 17 1OB-18b 10B-18b 18 18 1OA-7B F 10A-7B 17 17 1OB-19 10B-19 16 16 1OA-7C F 10A-7C 17 17 1OB-20 10B-20 16 16 F 10A-7D 1OA-7D 17 17 1OB-21 10B-21 16 16 F 10A-8 1OA-8 16 16 lOB-22 10B-22 16 16 1OA-8A F 10A-8A 16 16 1OB-23 10B-23 16 16 UFSAR Revision 21 Revision 21 10-4 EP 10-4 October 2009 October

Nine Mile Mile Point Unit Unit 1 UFSAR LIST OF EFFECTIVE EFFECTIVE PAGES PAGES SECTION X SECTION (Cont' d. ) X (Cont'd.) Page Revision Revision Page Revision Number Number Number Number Number Number Number Number 10B-24 1OB-24 T 1 17 10B-60 1OB-60 17 17 10B-24a 1OB-24a T 1 17 10B-61 1OB-61 16 16 10B-25 1OB-25 T 1 17 10B-62 1OB-62 16 16 1OB-26 10B-26 T 1 17 10B-63 1OB-63 18 18 1OB-27 10B-27 T 1 16 1OB-64 10B-64 16 16 10B-28 1OB-28 T 1 17 10B-65 1OB-65 18 18 10B-29 1OB-29 T 1 16 10B-66 1OB-66 16 16 10B-30 1OB-30 T 1 17 10B-67 1OB-67 19 19 1OB-30a 10B-30a T 1 17 10B-68 1OB-68 16 16 1OB-31 10B-31 T 1 17 1OB-69 10B-69 19 19 1OB-31a 10B-31a T 1 17 10B-70 1OB-70 16 16 1OB-32 10B-32 T 1 16 1OB-71 10B-71 16 16 1OB-33 10B-33 T 1 20 10B-72 1OB-72 20 20 10B-33a 10B-33a T 1 20 1OB-73 10B-73 19 19 1OB-34 10B-34 T 1 16 1OB-74 10B-74 19 19 1OB-35 10B-35 T 1 16 1OB-75 10B-75 16 16 1OB-36 10B-36 16 1OB-76 10B-76 16 16 1OB-37 10B-37 16 1OB-77 10B-77 20 20 1OB-38 10B-38 19 1OB-78 10B-78 20 20 1OB-39 10B-39 16 1OB-79 10B-79 20 20 10B-40 16 1OB-79a 10B-79a 20 20 1OB-41 10B-41 T 2A 16 1OB-79b 10B-79b 20 20 1OB-42 10B-42 T T 2B 18 1OB-80 10B-80 16 16 1OB-43 10B-43 T 3 20 1OB-81 10B-81 20 20 1OB-44 10B-44 T 3 20 1OB-82 10B-82 16 16 1OB-45 10B-45 T 3 20 1OB-83 10B-83 16 16 1OB-46 10B-46 T 3 20 1OB-84 10B-84 16 16 1OB-47 10B-47 T 3 20 1OB-85 10B-85 16 16 1OB-48 10B-48 T 3 20 1OB-86 10B-86 16 16 10B-48a 10B-48a T 3 20 1OB-87 10B-87 20 20 1OB-49 10B-49 16 1OB-88 10B-88 19 19 1OB-50 lOB-50 16 1OB-89 10B-89 16 16 10B-51 lOB-51 20 1OB-90 10B-90 16 16 1OB-52 lOB-52 20 1OB-91 10B-91 16 16 1OB-53 lOB-53 20 1OB-92 10B-92 16 16 1OB-54 lOB-54 18 18 1OB-93 10B-93 20 20 1OB-55 lOB-55 18 18 1OB-94 10B-94 19 19 1OB-56 lOB-56 16 1OB-95 10B-95 16 16 1OB-57 lOB-57 16 1OB-96 10B-96 16 16 1OB-57a 10B-57a 18 18 1OB-97 10B-97 16 16 1OB-58 lOB-58 18 1OB-98 10B-98 16 16 1OB-59 lOB-59 19 1OB-99 10B-99 16 16 UFSAR Revision Revision 21 21 10-5 EP 10-5 October 2009 2009

Nine Mile Point Unit 1 UFSAR EFFECTIVE PAGES LIST OF EFFECTIVE PAGES X (Cont' SECTION X SECTION d.) (Cont'd.) Page Page Revision Page Page Revision Revision Number Number Number Number Number Number Number Number 10B-100 1OB-100 20 20 10B-140 1OB-140 16 16 10B-101 1OB-101 19 19 1OB-141 10B-141 16 16 10B-102 1OB-102 16 16 10B-142 lOB-142 16 16 10B-103 1OB-103 16 16 10B-143 lOB-143 20 20 10B-104 1OB-104 20 20 10B-144 lOB-144 19 19 10B-104a 1OB-104a 20 10B-14S lOB-145 16 16 10B-104b 1OB-104b 20 20 lOB-146 10B-146 16 16 lOB-lOS 1OB-105 16 16 lOB-147 10B-147 16 16 10B-106 1OB-106 20 20 10B-148 lOB-148 20 20 10B-107 1OB-107 19 19 10B-149 lOB-149 19 19 10B-108 1OB-108 16 16 lOB-ISO 1OB-150 16 16 10B-109 1OB-109 16 16 1OB-151 lOB-lSI 16 16 10B-110 1OB-l10 16 16 lOB-152 10B-1S2 16 16 lOB-Ill 1OB-ill 18 18 10B-1S3 lOB-153 20 20 10B-112 1OB-112 16 16 10B-1S4 lOB-154 19 19 10B-113 1OB-113 20 20 lOB-155 lOB-ISS 16 16 1OB-114 10B-114 16 16 10B-1S6 lOB-156 16 16 lOB-lIS 1OB-115 19 19 1OB-157 10B-1S7 16 16 10B-116 10B-116 16 16 lOB-158 10B-1S8 16 16 10B-117 10B-117 16 16 10B-1S9 1OB-159 16 16 10B-118 10B-118 20 1OB-160 10B-160 19 19 10B-119 10B-119 19 19 10B-161 1OB-161 16 16 10B-120 10B-120 16 16 lOB-162 10B-162 20 20 10B-121 10B-121 16 16 10B-162a 1OB-162a 20 1OB-122 10B-122 16 16 10B-162b 1OB-162b 20 20 1OB-123 10B-123 20 lOB-163 10B-163 16 16 1OB-124 10B-124 19 19 lOB-164 10B-164 20 20 1OB-125 10B-12S 16 16 lOB-165 10B-16S 19 19 1OB-126 10B-126 16 16 lOB-166 10B-166 16 16 1OB-127 10B-127 16 16 lOB-167 10B-167 16 16 1OB-128 10B-128 20 lOB-168 10B-168 16 16 1OB-129 10B-129 19 19 lOB-169 10B-169 20 1OB-130 10B-130 16 16 1OB-170 10B-170 19 19 10B-131 10B-131 16 16 1OB-171 10B-171 16 16 1OB-132 10B-132 16 16 10B-172 lOB-172 16 16 1OB-133 10B-133 20 20 lOB-173 10B-173 16 16 1OB-134 10B-134 19 19 lOB-174 10B-174 16 16 1OB-135 10B-13S 16 16 lOB-175 10B-17S 19 19 1OB-136 10B-136 16 16 lOB-176 10B-176 16 16 1OB-137 10B-137 16 16 lOB-177 10B-177 20 1OB-138 10B-138 20 20 lOB-178 10B-178 20 1OB-139 10B-139 19 19 1OB-179 10B-179 16 16 21 UFSAR Revision 21 EP 10-6 10-G October 2009

Nine Mile Nine Mile Point Point Unit unit 11 UFSAR LIST OF LIST OF EFFECTIVE EFFECTIVE PAGES PAGES SECTION X SECTION X (Cont'd.) (Cont' d.) Page Page Revision Revision Page Page Revision Number Number Number Number Number Number Number Number 10B-180 10B-180 16 10B-218 1OB-218 16 16 10B-181 1OB-181 19 10B-219 1OB-219 16 16 10B-182 1OB-182 16 10B-220 1OB-220 21 21 10B-183 1OB-183 20 20 10B-221 1OB-221 21 21 10B-184 1OB-184 20 20 10B-222 1OB-222 21 21 10B-185 1OB-185 16 10B-223 1OB-223 21 21 10B-186 1OB-186 16 10B-224 1OB-224 21 21 10B-187 1OB-187 19 19 10B-188 1OB-188 16 16 10B-189 1OB-189 20 20 10B-190 1OB-190 20 20 10B-191 1OB-191 16 16 10B-192 1OB-192 16 16 10B-193 1OB-193 19 19 10B-194 1OB-194 18 18 10B-195 1OB-195 18 18 10B-196 1OB-196 16 16 10B-197 1OB-197 16 16 10B-198 1OB-198 16 16 10B-199 1OB-199 20 20 10B-200 1OB-200 20 20 10B-201 1OB-201 20 20 10B-202 1OB-202 20 20 10B-202a 1OB-202a 20 20 10B-202b 1OB-202b 20 20 10B-203 1OB-203 19 19 10B-204 1OB-204 19 19 10B-205 1OB-205 21 21 10B-205a 10B-205a 21 21 10B-206 1OB-206 16 16 10B-207 1OB-207 20 20 10B-207a 10B-207a 20 20 10B-208 1OB-208 16 16 10B-209 1OB-209 16 16 10B-210 1OB-210 16 16 10B-211 10B-211 16 16 10B-212 1OB-212 16 16 10B-213 1OB-213 16 16 10B-214 1OB-214 16 16 10B-215 1OB-215 16 16 10B-216 1OB-216 18 18 10B-217 1OB-217 18 18 21 UFSAR Revision 21 10-7 EP 10-7 2009 October 2009

Nine Mile Nine Mile Point Point Unit 1 UFSAR LIST OF EFFECTIVE PAGES PAGES SECTION SECTION XIXI Page Revision Page Revision Number Number Number Number Number Number Number Number XI-1 15 15 XI-2 15 15 XI-3 15 15 XI-4 15 15 XI-S XI-5 19 19 XI-6 XI-6 21 21 XI-7 21 21 XI-8 21 21 XI-9 21 21 XI-9a 21 21 XI-9b 21 21 XI-9c XI-9c 21 21 XI-9d XI-9d 21 21 XI-10 20 20 XI-11 XI-II 18 18 XI-11a XI-lla 18 18 XI-11b XI-llb 17 17 XI-12 20 20 XI-13 17 17 XI-14 20 20 XI-15 20 XI-16 15 15 F XI-1 XI-1 17 17 F XI-2 XI-2 14 14 F XI-3 XI-3 21 21 F XI-4 XI-4 14 14 F XI-5 XI-5 21 21 F XI-6 XI-G 16 16 F XI-7 XI-7 18 18

• UFSAR Revision 21 Revision 21 EP 11-1 11-1 2009 October 2009

Nine Mile Point Unit 1 UFSAR LIST OF EFFECTIVE EFFECTIVE PAGES PAGES SECTION XII XII Page Revision Page Revision Number Number Number Number Number Number XlI-l XII-1 18 T XII-4 XII-4 17 17 XII-2 XII-2 20 T XII-S XII-5 15 15 XII-3 XII-3 20 T XII-6 XII-6 15 15 XII-4 XII-4 20 T XII-7 XII-7 15 15 XII-4a XII-4a 20 T XII-8 Sh 1 20 20 XII-4b 20 T XII-8 Sh 2 XII-8 20 20 XII-S XII-5 20 T XII-8 Sh 3 20 20 XII-6 17 F XlI-l XII-1 20 20 XII-7 17 17 XII-8 17 17 XII-9 XII-9 20 20 XII-9a XII-9a 20 20 XII-9b 18 18 XII-I0 XII-10 20 20 XII-II XII-11 17 17 XII-12 15 15 XII-13 15 15 XII-14 XII-14 17 17 XII-14a XII-14a 17 17 XII-14b 17 17 XII-IS XII-15 21 21 XII-16 21 21 XII-16a XII-16a 21 21 XII-16b 21 21 XII-17 18 18 XII-18 18 18 XII-19 18 18 XII-20 17 17 XII-21 21 21 XII-21a 21 21 XII-21b 21 21 XII-22 XII-22 15 15 XII-23 XII-23 17 17 XII-24 XII-24 19 19 XII-2S XII-25 19 19 XII-26 XII-26 19 19 XII-27 XII-27 20. 20. XII-28 XII-28 20 20 T XlI-l XII-1 15 15 T XII-2 Sh 1 17 17 T XII-2 Sh 22 20 20 T XII-3 T 20 20 UFSAR Revision UFSAR Revision 21 21 EP 12-1 12-1 October 2009

Nine Mile Point Point Unit 1 UFSAR LIST OF EFFECTIVE PAGES PAGES SECTION SECTION XIII XIII Page Revision Page Page Revision Number Number Number Number Number Number XllI-l XIII-1 21 21 XIII-2 21 21 XIII-3 21 21 XIII-4 21 21 XIII-S XIII-5 21 21 XIII-6 21 21 XIII-7 21 21 XIII-8 21 21 XIII-9 21 21 XIII-I0 XIII-10 21 21 XIII-II XIII-11 21 21 XIII-12 XIII-12 21 21 XIII-13 21 21 XIII-14 21 21 XIII-IS XIII-15 21 21 XIII-16 21 21 XIII-17 21 21 XIII-18 21 21 XIII-19 21 21 XIII-20 21 21 XIII-21 21 21 XIII-22 21 21 XIII-23 21 21 XIII-24 21 21 XIII-2S XIII-25 21 21 T XIII-1 T XllI-l Sh 1 21 21 T XIII-1 T XllI-l Sh 22 21 21 T XIII-2 T XIII-2 21 21 XllI-l F XIII-1 21 21 XIII-2 F XIII-2 21 21 XIII-3 F XIII-3 21 21 XIII-3a F XIII-3a 21 21 XIII-4 F XIII-4 21 21 XIII-4a Sh 1 F XIII-4a 21 21 XIII-4a Sh 2 F XIII-4a 21 21 XIII-4b F XIII-4b 21 21 XIII-4c F XIII-4c 21 21 XIII-S F XIII-5 21 21

• UFSAR Revision 21 Revision 21 13-1 EP 13-1 October 2009

Nine Mile Point Point Unit Unit 1 UFSAR LIST OF EFFECTIVE PAGES PAGES SECTION XIV SECTION Page Page Revision Revision Page Page Revision Number Number Number Number Number Number Number Number XIV-l XIV-1 16 16 XIV-2 XIV-2 15 15 XIV-3 XIV-3 15 15 XIV-4 XIV-4 15 15 XIV-5 15 15 XIV-6 XIV-6 15 15 XIV-77 XIV- 15 15 XIV-88 XIV- 15 15 XIV-9 XIV-9 15 15 XIV-I0 XIV-10 16 16 XIV-II XIV-11 16 16 XIV-12 XIV- 12 16 16 XIV-13 XIV-13 18 18 XIV-14 XIV-14 18 18

• UFSAR Revision UFSAR Revision 21 21 EP 14-1 14-1 October 2009 2009

Nine Mile Point Point Unit 1 UFSAR LIST OF EFFECTIVE PAGES PAGES SECTION XVXV Page Page Revision Page Page Revision Number Number Number Number Number Number Number Number XV-l XV-1 16 16 XV-40 XV-40 16 16 XV-2 XV-2 17 17 XV-41 XV-41 16 16 XV-3 XV-3 18 18 XV-42 XV-42 16 16 XV-4 XV-4 16 16 XV-43 XV-43 16 16 XV-5 XV-5 18 18 XV-44 XV-44 21 21 XV-6 XV-6 18 18 XV-45 XV-45 21 21 XV-7 XV-7 18 18 XV-46 XV-46 21 21 XV-8 XV-8 17 17 XV-47 XV-47 21 21 XV-8a XV-8a 17 17 XV-48 XV-48 21 21 XV-8b XV-8b 17 17 XV-48a XV-48a 21 21 XV-9 XV-9 17 17 XV-48b XV-48b 21 21 XV-I0 XV-10 16 16 XV-49 XV-49 21 21 XV-II XV-11 16 16 XV-50 XV-50 21 21

 .XV-12 XV-12                   16 16                XV-51 XV-51          16 16 XV-13 XV-  13                 16 16                XV-52 XV-52          16 16 XV-14 XV-  14                 16 16                XV-53 XV-53          16 16 XV-15 XV-15                   18 18                XV-54 XV-54          16 16 XV-16 XV-  16                 16 16                XV-55 XV-55          16 16 XV-17 XV-17                   16 16                XV-56 XV-56          21 21 XV-18 XV-18                   16 16                XV-57 XV-57          21 21 XV-19 XV-19                   17 17                XV-58 XV-58          21 21 XV-20 XV-20                   16 16                XV-59 XV-59          21 21 XV-21 XV-21                   16 16                XV-60 XV-60          21 21 XV-22 XV-22                   17 17                XV- 61 XV-61          21 21 XV-23 XV-23                   16 16                XV-61a XV-61a         21 21 XV-24 XV-24                   16 16                XV-61b XV-61b         21 21 XV-25 XV-25                   16 16                XV-62 XV-62          20 20 XV-26 XV-26                   16 16                XV-63 XV-63          20 20 XV-27 XV-27                   16 16                XV- 64 XV-64          20 20 XV-28 XV-28                   16 16                XV-65 XV-65          20 20 XV-29 XV- 29                  16 16                XV-65a XV-65a         20 20 XV-30 XV-30                   16 16                XV-65b XV-65b         20 20 XV-31 XV-31                   21 21                XV-66 XV-66          20 20 XV-32 XV-32                   21 21                XV-67 XV-67          17 17 XV-33 XV-33                   21 21                XV-68 XV-68          21 21 XV-34 XV-34                   21 21                XV-69 XV-69          21 21 XV-35 XV-35                   18 18                XV-70 XV-70          21 21 XV-36 XV-36                   16 16               XV-70a XV-70a          21 21 XV-37 XV-37                   16 16                XV-70b XV-70b         21 21 XV-38 XV-38                   16 16               XV-71 XV-71           16 16 XV-39 XV-39                   16 16               XV-XV-7272         16 16 XV-73 XV-73                   16 16               XV-74 XV-74           16 16 Revision 21 UFSAR Revision UFSAR            21            EP 15-1 15-1            October 2009

Nine Mile Point Unit 1 UFSAR LIST OF EFFECTIVE EFFECTIVE PAGES PAGES SECTION XV (Cont' d.) (Cont'd.) Page Page Revision Revision Page Revision Number Number Number Number Number Number Number Number XV-7S XV-75 16 T XV-29 21 21 XV-76 XV-76 21 T XV-29a 16 16 XV-77 21 T XV-29b 21 21 XV-78 XV-78 21 T XV-29c 21 21 XV-79 XV-79 21 T XV-29d 21 21 XV-80 XV-80 21 T XV-30 21 21 XV-81 21 T XV-3l XV-31 Sh 1 21 21 T XV-1 XV-l 17 T XV-3l XV-31 Sh 2 21 21 T XV-2 17 T XV-3l XV-31 Sh 3 21 21 T XV-3 16 T XV-32 21 21 T XV-4 16 T XV-32a Sh 1 20 20 T XV-5 XV-S 16 T XV-32a Sh 2 20 20 T XV-6 21 T XV-33 21 21 T XV-7 21 T XV-34 21 21 T XV-7a 21 T XV-34a 21 21 T XV-7b 21 T XV-34b 21 21 T XV-8 21 T XV-35 XV-3S 21 21 T XV-9 Sh 1 16 T XV-3Sa XV-35a 21 21 T XV-9 Sh 22 16 T T XV-3Sb XV-35b 21 21 T XV-9a 17 T T XV-3Sc XV-35c 21 21 XV-10 T XV-I0 17 T T XV-35d XV-3Sd 21 21 T XV-II XV-11 16 T T XV-36 XV-36 16 16 T XV-12 16 F XV-1 XV-l 17 17 T XV-13 16 F XV-2 14 14 T XV-14 16 F XV-3 XV-3 14 14 T XV-15 XV-IS 16 F XV-4 XV-4 14 14 T XV-16 16 F XV-S XV-5 14 14 T XV-17 16 F XV-6 XV-6 14 14 T XV-18 16 F XV-7 XV-7 14 14 T XV-19 16 F XV-8 XV-8 14 14 T XV-20 16 F XV-9 XV-9 14 14 T XV-21 16 F XV-I0 XV-10 14 14 T XV-21a 16 F XV-II XV-11 14 14 T XV-22 21 F XV-12 XV-12 14 14 T XV-23 21 F XV-13 XV-13 14 14 T XV-24 21 F XV-14 XV-14 14 14 T XV-2S XV-25 Sh 1 21 F XV-15 XV-IS 14 14 XV-25 Sh 22 T XV-2S 21 F XV-16 XV-16 14 14 T XV-26 Sh 1 21 F XV-17 XV-17 14 14 T XV-26 Sh 2 21 F XV-18 XV-18 14 14 T XV-27 21 F XV-19 14 14 T XV-28 21 F XV-20 14 14 UFSAR Revision 21 21 EP 15-2 15-2 October 2009 October 2009

Nine Mile Nine Mile Point Point Unit unit 1 UFSAR UFSAR LIST OF LIST OF EFFECTIVE EFFECTIVE PAGES PAGES SECTION XV SECTION XV (Cont' d. ) (Cont'd.) Page Page Revision Page Page Revision Number Number Number Number Number Number Number Number F XV-21 XV-21 14 14 F XV-56g XV-56g 14 14 F XV-22 XV-22 14 14 F XV-57 XV-57 14 14 F XV-23 XV-23 14 14 F XV-58 XV-58 14 14 F XV-24 XV-24 14 14 F XV-59 XV-59 14 14 F XV-25 XV-25 14 14 F XV-60 XV-60 14 14 F XV-26 XV-26 14 14 F XV-60a XV-60a 20 20 F XV-27 XV-27 14 14 F XV-60b XV-60b 20 20 F XV-28 XV-28 14 14 F F XV-61 XV-61 14 14 F XV-29 XV-29 14 14 F F XV-62 XV-62 14 14 F XV-30 XV-30 14 14 F F XV-63 XV-63 14 14 F XV-31 XV-31 14 14 F F XV-64 XV-64 14 14 F XV-32 XV-32 14 14 F F XV-65 XV-65 14 14 F XV-33 XV-33 14 14 F F XV-66 XV-66 14 14 F XV-34 XV-34 14 14 F F XV-67 XV-67 14 14 F XV-35 XV-35 14 14 F F XV-68 XV-68 14 14 F XV-36 XV-36 14 14 F F XV-69 XV-69 14 14 F XV-37 XV-37 14 14 F F XV-70 XV-70 14 14 F XV-38 XV-38 14 14 F F XV-71 XV-71 14 14 F XV-39 XV-39 14 14 F F XV-72 XV-72 14 14 F XV-40 XV-40 14 14 F F XV-73 XV-73 21 21 F XV-41 XV-41 14 14 F XV-74 XV-74 21 21 F XV-42 XV-42 14 14 F XV-43 XV-43 14 14 F XV-44 XV-44 14 14 F XV-45 XV-45 14 14 F XV-46 XV-46 14 14 F XV-47 XV-47 14 14 F XV-48 XV-48 14 14 F XV-49 XV-49 14 14 F XV-50 XV-50 14 14 F XV-51 XV-51 14 14 F XV-52 XV-52 14 14 F XV-53 XV-53 14 14 F XV-54 XV-54 14 14 F XV-55 XV-55 14 14 F XV-56 XV-56 14 14 F XV-56a XV-56a 14 14 F XV-56b XV-56b 14 14 F XV-56c XV-56c 14 14 F XV-56d XV-56d 14 14 F XV-56e XV-56e 14 14 F XV-56£ XV-56f 14 14 UFSAR Revision 21 21 15-3 EP 15-3 October 2009

Mile Point Unit 1 UFSAR Nine Mile PAGES LIST OF EFFECTIVE PAGES SECTION XVI XVI Page Page Revision Page Revision Number Number Number Number Number Number Number Number XVI-1 XVI-1 16 16 XVI-34 XVI-34 16 16 XVI-2 XVI-2 16 16 XVI-35 XVI-35 16 16 XVI-3 XVI-3 16 16 XVI-36 XVI-36 16 16 XVI-4 XVI-4 16 16 XVI-37 XVI-37 16 16 XVI-5 XVI-5 16 16 XVI-38 XVI-38 16 16 XVI-6 XVI-6 16 16 XVI-39 XVI-39 16 16 XVI-7 XVI-7 16 16 XVI-40 XVI-40 16 16 XVI-8 XVI-8 16 16 XVI-41 XVI-41 16 16 XVI-9 XVI-9 16 16 XVI-42 XVI-42 16 16 XVI-10 XVI-10 16 16 XVI-43 16 16 XVI-11 XVI-11 16 16 XVI-44 16 16 XVI-12 XVI-12 16 16 XVI-45 XVI-45 20 20 XVI-13 19 19 XVI-45a XVI-45a 20 20 XVI-13a XVI-13a 19 19 XVI-45b XVI-45b 20 20 XVI-13b XVI-13b 19 19 XVI-46 XVI-46 16 16 XVI-14 16 16 XVI-47 XVI-47 16 16 XVI-15 XVI-15 16 16 XVI-48 XVI-48 16 16 XVI-16 XVI-16 16 16 XVI-49 XVI-49 16 16 XVI-17 XVI- 17 16 16 XVI-50 XVI-50 17 17 XVI-18 16 16 XVI-51 XVI-51 20 20 XVI-19 XVI- 19 18 18 XVI-52 XVI-52 20 20 XVI-20 XVI-20 19 19 XVI-52a XVI-52a 20 20 XVI-21 XVI-21 19 19 XVI-52b XVI-52b 16 16 XVI-21a XVI-21a 20 20 XVI-53 XVI-53 16 16 XVI-21b XVI-21b 16 16 XVI-54 XVI-54 17 17 XVI-22 XVI-22 16 16 XVI-55 XVI-55 17 17 XVI-23 XVI-23 16 16 XVI-55a XVI-55a 20 20 XVI-24 XVI-24 21 21 XVI-55b XVI-55b 17 17 XVI-25 XVI-25 21 21 XVI-56 XVI-56 20 20 XVI-25a XVI-25a 21 21 XVI-57 XVI-57 16 16 XVI-25b XVI-25b 21 21 XVI-58 XVI-58 16 16 XVI-26 XVI-26 16 16 XVI-59 XVI-59 16 16 XVI-27 XVI-27 16 16 XVI-60 XVI-60 16 16 XVI-28 16 16 XVI-60a XVI-60a 17 17 XVI-29 XVI-29 16 16 XVI-60b XVI-60b 17 17 XVI-30 XVI-30 16 16 XVI-61 XVI-61 16 16 XVI-31 XVI-31 16 16 XVI-62 XVI-62 16 16 XVI-32 XVI-32 17 17 XVI-63 XVI-63 16 16 XVI-32a XVI-32a 17 17 XVI-64 XVI-64 21 21 XVI-32b XVI-32b 17 17 XVI-65 XVI-65 21 21 XVI-33 XVI-33 16 16 XVI-65a XVI-65a 21 21 XVI-65b XVI-65b 21 21 XVI-66 XVI-66 21 21 UFSAR Revision Revision 21 21 16-1 EP 16-1 October 2009 October 2009

Nine Mile Point Unit 1 UFSAR LIST OF EFFECTIVE EFFECTIVE PAGES PAGES SECTION XVI (Cont' SECTION d.) (Cont'd.) Page Revision Page Revision Number Number Number Number Number Number Number Number XVI-67 XVI-67 21 21 XVI-I07 XVI-107 16 16 XVI-68 XVI-68 21 21 XVI-I08 XVI-108 16 16 XVI-69 XVI-69 21 21 XVI-I09 XVI-109 16 16 XVI-70 XVI-70 16 16 XVI-II0 XVI-110 16 16 XVI-71 XVI-71 17 17 XVI-Ill XVI-111 16 16 XVI-72 XVI-72 16 16 XVI-112 16 16 XVI-73 16 16 XVI-113 16 16 XVI-74 16 16 XVI-114 16 16 XVI-75 XVI-75 16 16 XVI-115 16 16 XVI-76 XVI-76 16 16 XVI-116 16 16 XVI-77 XVI-77 16 16 XVI-117 16 16 XVI-78 XVI-78 16 16 XVI-118 16 16 XVI-79 XVI-79 16 16 XVI-119 16 16 XVI-80 XVI-80 16 16 XVI-120 XVI-120 16 16 XVI-81 XVI-81 16 16 XVI-121 XVI-121 16 16 XVI-82 XVI-82 16 16 XVI-122 XVI-122 16 16 XVI-83 XVI-83 16 16 XVI-123 XVI-123 16 16 XVI-84 XVI-84 16 16 XVI-124 XVI-124 19 19 XVI-85 XVI-85 16 16 T XVI-1 XVI-l 16 16 XVI-86 XVI-86 16 16 XVI-2 Sh 1 T XVI-2 18 18 XVI-87 XVI-87 16 16 T XVI-2 XVI-2 Sh 2 17 17 XVI-88 XVI-88 16 16 T XVI-3 XVI-3 17 17 XVI-89 XVI-89 16 16 T XVI-4 XVI-4 16 16 XVI-90 XVI-90 16 16 T XVI-5 XVI-5 16 16 XVI-91 XVI-91 16 16 T XVI-6 XVI-6 16 16 XVI-92 XVI-92 16 16 T XVI-7 XVI-7 18 18 XVI-93 XVI-93 16 16 T XVI-8 XVI-8 16 16 XVI-94 16 16 T XVI-9 XVI-9 19 19 XVI-95 XVI-95 16 16 T XVI-9a XVI-9a Sh 1 16 16 XVI-96 XVI-96 16 16 T , XVI-9a XVI-9a Sh 2 21 21 XVI-97 XVI-97 16 16 T XVI-I0 XVI-10 16 16 XVI-98 16 16 T XVI-II XVI-11 16 16 XVI-99 XVI-99 16 16 T XVI-12 XVI-12 16 16 XVI-I00 XVI-100 16 16 T XVI-13 16 16 XVI-I0l XVI-101 16 16 T XVI-14 16 16 XVI-I02 XVI-102 16 16 T XVI-15 16 16 XVI-I03 XVI-103 16 16 T XVI-16 16 16 XVI-I04 XVI-104 16 16 T XVI-17 16 16 XVI-I05 XVI-105 16 16 T XVI-18 16 16 XVI-106 XVI-I06 16 16 T XVI-19 16 16 XVI-20 T XVI-20 16 16 T XVI-22 XVI-22 16 16 XVI-21 T XVI-21 16 16 T XVI-23 T XVI-23 16 16 UFSAR Revision 21 21 EP 16-2 16-2 October 2009

Nine Mile Point Unit 1 UFSAR EFFECTIVE PAGES LIST OF EFFECTIVE PAGES (Cont'd. ) SECTION XVI (Cont'd.) Page Page Revision Page Revision Number Number Number Number Number Number Number Number T XVI-24 XVI-24 16 16 F XVI-28 XVI-28 14 14 T XVI-25 XVI-25 16 16 F XVI-29 XVI-29 14 14 T XVI-26 Sh XVI-26 1 20 20 F XVI-30 XVI-30 14 14 T XVI-26 Sh XVI-26 2 20 20 F XVI-31 XVI-31 14 14 T XVI-27 Sh XVI-27 1 16 16 F XVI-32 XVI-32 14 14 T XVI-27 Sh XVI-27 2 16 16 F XVI-33 XVI-33 14 14 T XVI-28 XVI-28 16 16 F XVI-34 XVI-34 14 14 T XVI-29 XVI-29 16 16 F XVI-35 XVI-35 14 14 T XVI-30 XVI-30 16 16 F XVI-36 XVI-36 14 14 T XVI-31 Sh XVI-31 1 16 16 F XVI-37 XVI-37 14 14 T XVI-31 Sh XVI-31 2 16 16 F XVI-38 XVI-38 14 14 F XVI-l XVI-1 14 14 F XVI-39 XVI-39 14 14 F XVI-2 XVI-2 14 14 F XVI-40 XVI-40 14 14 F XVI-3 XVI-3 14 14 F XVI-41 XVI-41 14 14 F XVI-4 XVI-4 14 14 F XVI-42 XVI-42 14 14 F XVI-5 XVI-5 14 14 F XVI-43 XVI-43 14 14 F XVI-6 XVI-6 14 14 F XVI-44 XVI-44 14 14 F XVI-7 XVI-7 14 14 F XVI-45 XVI-45 14 14 F XVI-8 XVI-8 14 14 F F XVI-46 XVI-46 14 14 F XVI-9 XVI-9 14 14 F F XVI-47 XVI-47 14 14 F XVI-I0 XVI-10 14 14 F XVI-48 XVI-48 14 14 F XVI-II XVI-11 14 14 F XVI-49 XVI-49 14 14 F XVI-12 XVI-12 14 14 F XVI-50 XVI-50 14 14 F XVI-12a XVI-12a 14 14 F XVI-51 XVI-51 14 14 F XVI-12b XVI-12b 14 14 F XVI-52 XVI-52 14 14 F XVI-12c XVI-12c 16 16 F XVI-53 14 14 F XVI-12d XVI-12d 16 16 F XVI-54 XVI-54 14 14 F XVI-13 14 14 F XVI-55 XVI-55 14 14 F XVI-14 XVI- 14 14 14 F XVI-56 XVI-56 14 14 F XVI-15 XVI-15 14 14 F XVI-57 XVI-57 14 14 F XVI-16 XVI-16 14 14 F XVI-58 14 14 F XVI-17 16 16 F XVI-59 XVI-59 14 14 F XVI-18 14 14 F XVI-60 XVI-60 14 14 F XVI-19 XVI-19 14 14 F XVI-61 XVI-61 14 14 F XVI-20 XVI-20 14 14 F XVI-21 XVI-21 14 14 F XVI-22 XVI-22 14 14 F XVI-23 XVI-23 14 14 F XVI-24 XVI-24 14 14 F XVI-25 XVI-25 14 14 F XVI-26 XVI-26 14 14 F XVI-27 XVI-27 14 14 UFSAR UFSAR Revision 21 Revision 21 EP 16-3 16-3 October 2009 October 2009

Nine Mile Nine Mile Point Point Unit 1 UFSAR UFSAR OF EFFECTIVE LIST OF LIST EFFECTIVE PAGES PAGES SECTION XVII SECTION XVII Page Page Revision Page Page Revision Revision Number Number Number Number Number Number Number Number XVII-l XVII-I 15 15 T T XVII-7 XVII-7 15 15 XVII-2 XVII-2 15 15 T T XVII-8 XVII-8 15 15 XVII-3 XVII-3 15 15 T XVII-9 XVII-9 15 15 XVII-4 XVII-4 15 15 T XVII-I0 XVII-10 15 15 XVII-5 XVII-5 15 15 T XVII-II XVII-11 15 15 XVII-6 XVII-6 15 15 T XVII-12 XVII-12 15 15 XVII-7 XVII-7 15 15 T XVII-13 15 15 XVII-8 XVII-8 15 15 T XVII-14 XVII-14 15 15 XVII-9 XVII-9 15 15 T XVII-15 XVII-IS 15 15 XVII-I0 XVII-10 15 15 T XVII-16 XVII-16 15 15 XVII-II XVII-11 15 15 T XVII-17 XVII-17 15 15 XVII-12 XVII-12 15 15 T XVII-18 15 15 XVII-13 XVII-13 15 15 T XVII-19 XVII-19 15 15 XVII-14 XVII-14 15 15 T XVII-20 XVII-20 15 15 XVII-IS XVII-15 15 15 T XVII-21 XVII-21 15 15 XVII-16 XVII-16 15 15 T XVII-22 XVII-22 15 15 XVII-17 XVII-17 15 15 T XVII-23 XVII-23 15 15 XVII-18 XVII-18 15 15 T XVII-24 15 15 XVII-19 XVII-19 15 15 T XVII-25 15 15 XVII-20 XVII-20 15 15 T XVII-26 15 15 XVII-21 XVII-21 15 15 T XVII-27 15 15 XVII-22 XVII-22 15 15 T XVII-28 XVII-28 15 15 XVII-23 XVII-23 15 15 T XVII-29 XVII-29 15 15 XVII-24 XVII-24 15 15 T XVII-30 XVII-30 15 15 XVII-25 XVII-25 15 15 F XVII-1 XVII-l 15 15 XVII-26 XVII-26 15 15 F XVII-2 XVII-2 15 15 XVII-27 XVII-27 15 15 F XVII-3 XVII-3 15 15 XVII-28 XVII-28 15 15 F XVII-4 XVII-4 15 15 XVII-29 XVII-29 15 15 F XVII-5 XVII-5 15 15 XVII-30 XVII-30 15 15 F XVII-6 XVII-6 15 15 XVII-31 XVII-31 15 15 F XVII-7 XVII-7 15 15 XVII-32 XVII-32 15 15 F XVII-8 XVII-8 15 15 XVII-33 XVII-33 18 18 F XVII-9 XVII-9 15 15 XVII-34 XVII-34 15 15 F F XVII-10 XVII-I0 15 15 XVII-35 XVII-35 15 15 F XVII-11 XVII-II 15 15 T XVII-l XVII-1 15 15 F XVII-12 XVII-12 15 15 T XVII-2 XVII-2 15 15 F XVII-13 15 15 T XVII-3 XVII-3 15 15 F XVII-14 15 15 T XVII-4 XVII-4 15 15 F XVII-15 XVII-IS 15 15 T XVII-5 XVII-5 15 15 F XVII-16 XVII-16 15 15 T XVII-6 XVII-6 15 15 F XVII-17 15 15 F XVII-18 15 15 F XVII-19 15 15 UFSAR Revision 21 21 17-1 EP 17-1 October 2009

Nine Mile Point Point Unit 1 UFSAR LIST OF EFFECTIVE PAGES PAGES SECTION XVII (Cont' d.) (Cont'd.) Page Page Revision Revision Page Page Revision Number Number Number Number Number Number Number Number F XVII-20 15 15 F XVII-62 XVII-62 15 15 15 XVII-63 15 15 F XVII-21 XVII-21 15 F XVII-63 F XVII-22 15 15 F XVII-64 XVII-64 15 15 F XVII-23 15 15 F XVII-65 XVII-65 15 15 F XVII-24 15 15 F XVII-25 15 15 F XVII-26 15 15 F XVII-27 15 15 F XVII-28 15 15 F XVII-29 XVII-29 15 15 F XVII-30 XVII-30 15 15 F XVII-31 XVII-31 15 15 F XVII-32 XVII-32 15 15 F XVII-33 15 15 F XVII-34 15 15 F XVII-35 XVII-35 15 15 F XVII-36 XVII-36 15 15 F XVII-37 XVII-37 15 15 F XVII-38 15 15 F XVII-39 XVII-39 15 15 F XVII-40 XVII-40 15 15 F XVII-4l XVII-41 15 15 F XVII-42 XVII-42 15 15 F XVII-43 15 15 F XVII-44 XVII-44 15 15 F XVII-45 15 15 F XVII-46 XVII-46 15 15 F XVII-47 XVII-47 15 15 F XVII-48 15 15 F XVII-49 XVII-49 15 15 F XVII-50 15 15 F XVII-51 XVII-51 15 15 F XVII-52 15 15 F XVII-53 15 15 F XVII-54 15 15 F XVII-55 15 15 F XVII-56 15 15 F XVII-57 15 15 F XVII-58 15 15 F XVII-59 15 15 F XVII-60 XVII-60 15 15 F XVII-61 XVII-61 15 15 Revision 21 UFSAR Revision UFSAR 21 17-2 EP 17-2 October 2009 2009

Nine Mile Point Unit 1 UFSAR LIST OF EFFECTIVE EFFECTIVE PAGES PAGES SECTION XVIII SECTION XVIII Page Revision Revision Page Revision Number Number Number Number Number Number Number XVIII-l XVIII-1 18 18 XVIII-2 XVIII-2 15 15 XVIII-3 XVIII-3 16 16 XVIII-4 XVIII-4 15 15 XVIII-5 XVIII-5 15 15 XVIII-6 XVIII-6 18 18 XVIII-7 XVIII-7 15 15 XVIII-8 XVIII-8 15 15 XVIII-9 XVIII-9 15 15 XVIII-I0 XVIII-10 18 18 XVIII-II XVIII-11 18 18 XVIII-12 XVIII-12 15 15 XVIII-13 XVIII-13 15 15 XVIII-14 XVIII-14 15 15 XVIII-15 XVIII-15 15 15 XVIII-16 XVIII-16 15 15 T XVIII-l XVIII-1 18 18

• UFSAR Revision Revision 21 21 18-1 EP 18-1 October 2009

Nine Mile Point Unit 1 UFSAR LIST OF EFFECTIVE PAGES PAGES APPENDIX APPENDIX A Page Page Revision Page Page Revision Number Number Number Number Number Number Number Number A-1 A-I 11 Ii

• Revision 21 UFSAR Revision 21 EP A-1 A-1 October 2009

Nine Mile Point Unit 1 UFSAR EFFECTIVE PAGES LIST OF EFFECTIVE PAGES APPENDIX B Page Page Revision Page Revision Number Number Number Number Number Number Number Number B-I B-1 20 20

• UFSAR Revision Revision 21 21 EP B-1 B-1 October 2009

Nine Mile Point Unit 1 UFSAR Nine EFFECTIVE PAGES LIST OF EFFECTIVE PAGES APPENDIX C Page Revision Page Page Revision Number Number Number Number Number Number Number Number c-i C-i 20 20 C.2-4 C.2-4 20 20 C-ii C-ii 20 20 C.2-S C.2-5 20 20 C-iii C-iii 20 20 C.2-6 C.2-6 20 20 C-iv 20 20 C.2-7 C.2-7 20 20 C.2-S C. 2-8 20 20 C.O-1 C. 0-1 20 20 C.2-9 C.2-9 20 20 C.0-2 C. 0-2 20 C.2-10 C. 2-10 20 20 C.2-11 C. 2-11 20 20 C.1-1 C. 1-i 20 20 C.2-12 C. 2-12 20 20 C.1-2 C. 1-2 20 20 C.2-13 C. 2-13 20 20 C.1-3 C. 1-3 20 20 C.1-4 C. 1-4 20 20 C.3-1 C.3-1 20 20 C.1-S C. 1-5 20 20 C.1-6 C. 1-6 20 20 C.4-1 C. 4-1 20 20 C.1-7 C. 1-7 21 21 C.4-2 C.4-2 20 20 C. 1-7a C.1-7a 21 21 C.1-7b C. 1-7b 21 21 T C-1 Sh 1 C-I 21 21 C.1-S C. 1-8 21 21 T C-1 Sh C-I Sh 2 21 21 C.1-9 C. 1-9 21 21 T C-1 Sh C-I Sh 3 21 21 C.1-10 C. 1-10 20 20 T C-I Sh C-1 Sh 4 21 21 C.1-11 C. 1-11 20 20 T C-I Sh C-1 Sh 5 21 21 C.1-12 C. 1-12 21 21 T T Sh C-1 Sh 6 C-I 21 21 C.1-13 C. 1-13 20 20 T T C-1 C-I Sh Sh 7 21 21 C.1-14 C. 1-14 20 20 T T C-I Sh C-1 Sh S 8 21 21 C.1-1S C. 1-15 20 T T C-I Sh C-1 Sh 9 21 21 C. 1-16 C.1-16 20 20 T T C-i Sh C-1 Sh 10 21 21 C.1-17 C. 1-17 20 20 C.1-1S C. 1-18 20 20 C.1-19 C. 1-19 20 20 C. 1-20 C.1-20 20 20 C. 1-21 C.1-21 20 20 C.1-22 C. 1-22 20 20 C.1-23 C. 1-23 20 C.1-24 C. 1-24 20 C. 1-25 C.1-2S 20 20 C.1-26 C. 1-26 20 C.1-27 C. 1-27 20 C.1-2S C.1-28 20 C.2-1 C.2-1 20 20 C.2-2 C.2-2 20 20 C. 2-3 C.2-3 20 UFSAR Revision UFSAR Revision 21 21 EP C-1 C-1 October 2009

U.S. NUCLEAR REGULATORY NUCLEAR REGULATORY

• COMMISSION COMMISSION DOCKET 50-220 50-220 LICENSE DPR-63 DPR-63 NINE MILE POINT NINE NUCLEAR STATION NUCLEAR STATION UNIT 1 FINAL FINAL SAFETY ANALYSIS REPORT ANALYSIS (UPDATED)

VOLUME 1 VOLUME OCTOBER 2009 OCTOBER REVISION 21 REVISION

Nine Mile Point Unit 1 UFSAR

• Section Section Title TABLE OF CONTENTS TABLE OF CONTENTS CONTENTS Page i

LIST OF TABLES xxxiv LIST OF FIGURES xxxix SECTION I SECTION INTRODUCTION AND

SUMMARY

INTRODUCTION

SUMMARY

I-I I-1 A. A. PRINCIPAL PRINCIPAL DESIGN CRITERIA DESIGN CRITERIA 1-2 1.0 General General 1-7 2.0 Buildings and Structures Structures 1-8 3.0 Reactor 1-8 4.0 Reactor Vessel Reactor 1-10 5.0 Containment Containment 1-10 6.0 Control Control and Instrumentation Instrumentation 1-12 7.0 Electrical Electrical Power 1-14 8.0 Radioactive Waste Disposal Radioactive Disposal 1-14 9.0 Shielding and Access Control Shielding Control 1-14 10.0 10.0 Fuel Handling Handling and Storage I-14a I-14a B. B. CHARACTERISTICS CHARACTERISTICS 1-15 I-IS 1.0 Site 1-15 I-IS 2.0 Reactor 1-15 I-IS 3.0 Core 1-15 I-IS 4.0 Fuel Assembly Assembly 1-15 I-IS 5.0 Control System 1-15 I-IS 6.0 Core Design and Operating Conditions 1-16 1-16 7.0 Design Design Power Peaking Peaking Factor 1-16 1-16 8.0 Nuclear Nuclear Design Data 1-16 1-16 9.0 Reactor Reactor Vessel Vessel 1-17 10.0 10.0 Coolant Recirculation Recirculation Loops 1-17 11.00

11. Containment Primary Containment 1-17 1-17 12.0 12.0 Secondary Containment Secondary Containment 1-17 1-17 13.0 13.0 Structural Structural Design 1-18 14.0 14.0 Electrical System Station Electrical 1-18 15.0 15.0 Reactor Instrumentation System Reactor Instrumentation 1-18 16.0 16.0 Reactor Protection Reactor Protection System System I-18a I-18a C.

C. IDENTIFICATION OF CONTRACTORS IDENTIFICATION CONTRACTORS 1-19

• D.

D. UFSAR Revision GENERAL CONCLUSIONS GENERAL CONCLUSIONS Revision 21 21 i 1-20 1-20 October 2009

Nine Mile Point Unit 1 UFSAR Section Section E. E. Title TABLE OF CONTENTS REFERENCES REFERENCES (Cont'd.) CONTENTS (Cont'd.) Page 1-21 1-21 SECTION II SECTION II STATION STATION SITE AND ENVIRONMENT ENVIRONMENT II-I 11-1 A. A. SITE DESCRIPTION DESCRIPTION II-i 11-1 1.0 General General II-1 11-1 2.0 Physical Features Physical Features II-i 11-1 3.0 3.0 Property Use and Development Property 11-2 11-2 B. B. ADJACENT TO DESCRIPTION OF AREA ADJACENT THE SITE 11-3 11-3 1.0 General 11-3 11-3 1.2 1.1 Population Population 11-3 11-3 2.0 Agriculture, Industrial and Agriculture, Industrial Recreational Use Recreational 11-3 11-3 2.1 Agricultural Agricultural Use 11-3 11-3 2.2 Industrial Industrial Use 11-4 11-4 2.2.1 2.2.1 Toxic Chemicals Toxic Chemicals 11-4 11-4 2.3 Recreational Recreational Use II-4c II-4c C. C. METEOROLOGY METEOROLOGY 11-5 11-5 D. D. LIMNOLOGY LIMNOLOGY 11-6 11-6 E. E. EARTH EARTH SCIENCES SCIENCES 11-7 11-7 F. F. ENVIRONMENTAL ENVIRONMENTAL RADIOLOGY 11-8 11-8 G. G. REFERENCES REFERENCES 11-9 11-9 III SECTION III BUILDINGS AND STRUCTURES BUILDINGS III-1 111-1 A. A. TURBINE BUILDING BUILDING 111-3 1.0 Design Bases 111-3 1.1 1.1 Wind and Snow Loadings Loadings 111-3 1.2 Pressure Pressure Relief Relief Design 111-3 111-3 1.3 Seismic Design and Internal Seismic Internal Loadings 111-3 111-3 1.4 Heating Heating and Ventilation Ventilation 111-4 111-4 1.5 Shielding Shielding and Access Access Control Control III-4a III-4a 2.0 Structure Design Structure 111-5 111-5 2.1 General Structural General Structural Features Features 111-5 111-5 21 UFSAR Revision 21 ii ii October 2009

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• Section Section Title (Cont'd.)

TABLE OF CONTENTS (Cont'd.) Page 2.2 Heating and Ventilation Ventilation System 111-6 111-6 2.3 Smoke and Heat Removal Removal 111-7 111-7 2.4 Shielding Control Shielding and Access Control 111-7 2.5 Additional Building Additional Building Cooling 111-7 3.0 Safety Analysis Analysis 111-8 B. B. CONTROL ROOM CONTROL 111-9 111-9 1.0 Design Bases 111-9 1.1 Wind and Snow Snow Loadings 111-9 111-9 1.2 Pressure Relief Design Pressure Relief 111-9 111-9 1.3 Seismic Seismic Design and Internal Internal Loadings 111-9 1.4 1.4 Heating and Ventilation Heating Ventilation 111-9 111-9 1.5 1.5 Shielding and Access Shielding Access Control Control 111-9 111-9 2.0 Structure Design Structure III-10 111-10 2.1 General General Structural Structural Features Features III-10 111-10 2.2 Heating, Ventilation and Air Heating, Ventilation Air Conditioning Conditioning System III-11 III-II 2.3 Smoke and Heat Removal Smoke 111-12 2.4 Shielding and Access Shielding Access Control Control III-12a III-12a 3.0 3.0 Safety Analysis Safety Analysis III-12a III-12a C. C. WASTE DISPOSAL BUILDING BUILDING 111-13 1.0 Design Bases 111-13 1.1 Wind and Snow Snow Loadings 111-13 1.2 Pressure Relief Design Pressure Relief 111-13 1.3 Seismic Design and Internal Seismic Internal Loadings 111-13 1.4 Heating and Ventilation Heating Ventilation 111-14 1.5 Access Control Shielding and Access Shielding Control 111-14 2.0 Structure Structure Design 111-14 2.1 General Structural General Features Structural Features 111-14 2.2 Heating Ventilation System Heating and Ventilation System 111-15 2.3 Shielding Access Control Shielding and Access 111-17 3.0 Safety Analysis Safety Analysis 111-17 D. D. OFFGAS BUILDING 111-19 1.0 Design Bases 111-19 1.1 Wind and Snow Snow Loadings 111-19 1.2 Pressure Relief Design Pressure Relief 111-19 1.3 Seismic Design and Internal Seismic Internal Loadings 111-19 1.4 Heating and Ventilation Heating Ventilation 111-19

• Shielding and Access Access Control 1.5 Shielding 111-19 UFSAR Revision 21 21 iii iii October 2009

Nine Mile Nine Unit 11 UFSAR Point Unit Mile Point UFSAR Section Section Title Title TABLE OF TABLE CONTENTS (Cont'd.) OF CONTENTS (Cont'd.) Page 111-19 2.0 2.0 Structure Design Structure Design 111-19 2.1 2.1 General Structural Features General Structural Features 111-19 111-19 2.2 2.2 Heating Ventilation System and Ventilation Heating and System 111-20 111-20 2.3 2.3 Shielding Access Control and Access Shielding and Control 111-20 111-20 3.0 3.0 Safety Analysis Safety Analysis 111-20 111-20 E. E. NONCONTROLLED BUILDINGS NONCONTROLLED BUILDINGS 111-22 111-22 1.0 1.0 Administration Building Administration Building 111-22 111-22 1.1 1.1 Design Bases Design Bases 111-22 111-22 1.1.1 1.1.1 Wind Snow Loadings and Snow Wind and Loadings 111-22 111-22 1.1.2 1.1. 2 Relief Design Pressure Relief Pressure Design 111-22 111-22 1.1.3 1.1. 3 Seismic Design and Internal Loadings Seismic Design and Internal Loadings 111-22 111-22 1.1.4 1.1.4 Heating, Cooling and Ventilation Heating, Cooling and Ventilation 111-23 111-23 1.1.5 1.1. 5 Shielding Access Control and Access Shielding and Control 111-23 111-23 1.2 1.2 Structure Design Structure Design 111-23 111-23 1.2.1

1. 2.1 Structural Features General Structural General Features 111-23 111-23 1.2.2

1. 2.2 Heating, Ventilation and Heating, Ventilation and Air Air Conditioning Conditioning 111-24 111-24 1.2.3

1. 2.3 Access Control Access Control 111-24 111-24 1.3 1.3 Safety Analysis Safety Analysis 111-24 111-24 2.0 2.0 Treatment Building Sewage Treatment Sewage Building 111-25 111-25 2.1 2.1 Design Bases Design Bases 111-25 111-25 2.1.1 2.1.1 Wind Snow Loadings and Snow Wind and Loadings 111-25 111-25 2.1.2 2.1.2 Relief Design Pressure Relief Pressure Design 111-25 111-25 2.1.3 2.1. 3 Seismic Internal Loadings Seismic Design and Internal Design and Loadings 111-25 111-25 2.1.4 2.1.4 Electrical Design Electrical Design 111-25 111-25 2.1.5 2.1. 5 Fire Gas Detection Explosive Gas and Explosive Fire and Detection 111-25 111-25 2.1.6 2.1. 6 Heating and Ventilation Heating and Ventilation 111-26 111-26 2.1.7 2.1. 7 Access Control Shielding and Access Shielding Control 111-26 111-26 2.2 2.2 Structure Design Structure Design 111-26 111-26 2.2.1 2.2.1 Structural Features General Structural General Features 111-26 111-26 2.2.2 2.2.2 Ventilation System Ventilation System 111-28 111-28 2.2.3 2.2.3 Access Control Access Control 111-28 111-28 3.0 3.0 Information Center Energy Information Energy Center 111-28 111-28 3.1 3.1 Design Bases Design Bases 111-28 111-28 3.1.1 3.1.1 Wind Snow Loadings and Snow Wind and Loadings 111-28 111-28 3.1.2 3.1. 2 Relief Design Pressure Relief Pressure Design 111-28 111-28 3.1.3 3.1. 3 Seismic Design and Internal Seismic Design and Internal Loadings Loadings 111-28 111-28 3.1.4 3.1.4 and Ventilation Heating and Heating Ventilation 111-29 111-29 3.1.5 3.1. 5 Shielding and Access Control Shielding and Access Control 111-29 111-29 3.2 Structure Design Structure Design 111-29 111-29 UFSAR 21 UFSAR Revision 21 iv iv October Octob~_r 2009 2009

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• Section TABLE OF CONTENTS (Cont'd.)

TABLE Title (Cont'd.) Page 3.2.1. 3.2.1 General Structural Features General Structural 111-29 111-29 3.2.2 Heating Ventilation System Heating and Ventilation 111-30 111-30 3.2.3 Access Control 111-30 111-30 F. F. SCREENHOUSE, INTAKE AND DISCHARGE SCREENHOUSE, TUNNELS TUNNELS 111-31 111-31 1.0 1.0 Screenhouse Screenhouse 111-31 111-31 1.1 Design Basis Basis 111-31 111-31 1.1.1 Wind and Snow Loadings 111-31 111-31 1.1.22 1.1. Pressure Relief Design 111-31 111-31 1.1.33 1.1. Seismic Design and Internal Internal Loadings Loadings 111-31 111-31 1.1. 1.1.44 Heating Ventilation Heating and Ventilation 111-31 111-31 1.1.55 1.1. Shielding and Access Control Shielding 111-31 111-31 1.2 Structure Design Structure 111-31 111-31 2.0 Intake Discharge Tunnels Intake and Discharge 111-33 111-33 2.1 Design Bases Bases 111-33 111-33 2.2 Structure Design Structure 111-33 111-33 3.0 Safety Analysis 111-34 111-34 G. G. STACK STACK 111-35 111-35 1.0 Design Bases Bases 111-35 111-35 1.1 General General 111-35 111-35 1.2 Wind Loading 111-35 111-35 1.3 Seismic Seismic Design 111-35 111-35 1.4 Shielding Shielding and Access Control Control 111-35 111-35 2.0 Structure Structure Design Design 111-35 3.0 Safety Safety Analysis 111-36 3.1 Radiology Radiology 111-36 3.2 Stack Stack Failure Failure Analysis 111-37 3.2.1 Reactor Building Reactor Building 111-37 3.2.2 Generator Building Diesel Generator Building 111-38 111-38 3.2.3 Screen Screen and Pump House House 111-38 111-38 H. H. SECURITY BUILDING SECURITY BUILDING WEST AND SECURITY BUILDING BUILDING ANNEX 111-39 I.

1. SOLIDIFICATION AND RADWASTE SOLIDIFICATION STORAGE BUILDING STORAGE BUILDING 111-40 111-40 1.0 Design Bases 111-40 1.1 Wind and Snow Loadings 111-40 111-40 UFSAR Revision 21 21 vV October 2009

Nine Mile Point Unit 1 UFSAR Section 1. 1.2-2' Title TABLE OF CONTENTS Pressure Relief Pressure CONTENTS (Cont'd.) Relief Design (Cont'd.) Page 111-40 111-40 1.3 1.3 Seismic Design and Internal Seismic Internal Loadings Loadings 111-40 111-40 1.4, 1.4. Heating, ventilation Heating, Ventilation and AirAir Conditioning 111-40 111-40 1.5 1.5 Shielding Access Control Shielding and Access Control 111-40 111-40 2.0 2.0 Structure and Design Structure 111-41 111-41 2.1 2.1 Structural Features General Structural Features 111-41 111-41 2.2 2.2 Heating, Ventilation Heating, Ventilation andAir and ,Air Conditioning 111-41 111-41 2.3 2.3 Shielding and Access Shielding Access Control Control 111-43 3.0 3.0 Use Use 111-43 J. J. REFERENCES REFERENCES 111-45 111-45 SECTION IV REACTOR IV-l IV-l A. DESIGN BASES BASES IV-l IV-1 1.0 2.0 General General IV-l IV-1 '2.0 Performance Objectives Performance Objectives IV-l IV-1 3.0 3.0 Design Limits and Targets Targets IV-2 IV-2 B. B. REACTOR DESIGN REACTOR IV-3 IV-3 1.0 General General IV-3 IV-3 2.0 2.0 Nuclear Nuclear Design Technique Technique IV-4 IV-4 2.1 2.1 Reference Reference Loading Loading Pattern IV-5 IV-5 2.2 2.2 Final Loading Loading Pattern IV-6 IV-6 2.2.1 2.2.1 Acceptable Deviation Acceptable Deviation From Reference Reference Loading Pattern IV-6 IV- 6 2.2.2 2.2.2 Reexamination of Licensing Basis Reexamination Basis IV-6 IV-6 2.3 2.3 Refueling Cycle Refueling Cycle. Reactivity Reactivity Balance Balance IV-7 IV-7 3.0 3.0 Thermal Thermal and Hydraulic Hydraulic Characteristics Characteristics IV-7 IV-7 3.1 3.1 Thermal and Hydraulic Thermal Hydraulic Design IV-7 IV-7 3.1.1 3.1.1 Recirculation Recirculation Flow Control Control IV-77 IV-3.1. 3.1.22 Core Thermal Thermal Limits Limits IV-7 IV-7 3.1.2.1 3.1.2.1 Excessive Clad Temperature Excessive Temperature IV-IV-88 3.1.2.2 3.1.2.2 Cladding Strain Cladding IV-9 IV-9 3.1.2.3 3.1.2.3 Flow Coolant Flow IV-9 IV-9 3.2 3.2 Hydraulic Analyses Thermal and Hydraulic Analyses IV-9 IV-9 3.2.1 3.2.1 Hydraulic Analysis Hydraulic Analysis IV- 9 IV-9 3.2.2 3.2.2 Thermal Analysis Thermal Analysis IV-1I IV-II 21 UFSAR Revision 21 vi vi October 2009 2009

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• Section Section Title Title CONTENTS (Cont'd.)

OF CONTENTS TABLE OF TABLE (Cont'd.) Page 3.2.2.1 3.2.2.1 Fuel Integrity Safety Cladding Integrity Fuel Cladding Safety Limit Analysis Limit Analysis IV-il IV-11 3.2.2.2 3.2.2.2 MCPR MCPR Operating Limit Analysis Operating Limit Analysis IV-12 IV-12 3.3 3.3 Reactor Transients Reactor Transients IV-13 IV-13 4.0 4.0 Stability Analysis Stability Analysis IV-14 IV-14 4.1 4.1 Design Bases Design Bases IV-14 IV-14 4.2 4.2 Analysis Method Stability Analysis Stability Method IV-14 IV-14 5.0 5.0 Mechanical Design and Evaluation Mechanical Design and Evaluation IV-15 IV-15 5.1 5.1 Mechanical Design Fuel Mechanical Fuel Design IV-15 IV-IS 5.1.1 5.1.1 Design Bases Design Bases IV-15 IV-IS 5.1.2 5.1. 2 Fuel Rods Fuel Rods IV-15 IV-15 5.1.3 5.1. 3 Water Rods Water Rods IV-16 IV-16 5.1.4 5.1. 4 Fuel Assemblies Fuel Assemblies IV-16 IV-16 5.1.5 5.1. 5 Mechanical Limits and Design Limits Mechanical Design and Stress Stress Analysis Analysis IV-16 IV-16 5.1.6 5.1. 6 Relationship Fuel Design Between Fuel Relationship Between Design Limits and Limits Damage Limits Fuel Damage and Fuel Limits IV-16 IV-16 5.1.7 5.1. 7 Surveillance and Surveillance and Testing Testing IV-16 IV-16 6.0 6.0 Control Rod Mechanical Design Control Rod Mechanical Design and and Evaluation Evaluation IV-16 IV-16 6.1 6.1 Design Design IV-16 IV-16 6.1.1 6.1.1 Control Rods and Control Rods and Drives Drives IV-16 IV-16 6.1.2 6.1. 2 Standby Liquid Poison System Standby Liquid Poison System IV-19 IV-19 6.2 6.2 System Evaluation Control System Control Evaluation IV-19 IV-19 6.2.1 6.2.1 Rod Rod Withdrawal Errors Evaluation Withdrawal Errors Evaluation IV-19 IV-19 6.2.2 6.2.2 Overall Control System Evaluation Overall Control System Evaluation IV-21 IV-2I 6.3 6.3 Limiting Conditions for Limiting Conditions for Operation Operation and Surveillance and Surveillance IV-23 IV-23 6.4 6.4 Rod Lifetime Control Rod Control Lifetime IV-23 IV-23 7.0 7.0 Reactor Internal Structure Vessel Internal Reactor Vessel Structure IV-24 IV-24 7.1 7.1 Design Bases Design Bases IV-24 IV-24 7.1.1 7.1.1 Core Shroud Core Shroud IV-25 IV-25 7.1.2 7.1. 2 Core Support Core Support IV-27 IV-27 7.1.3 7.1. 3 Top Top Grid Grid IV-27 IV-27 7.1.4 7.1. 4 Control Guide Tubes Rod Guide Control Rod Tubes IV-27 IV-27 7.1.5 7.1. 5 Feedwater Sparger Feedwater Sparger IV-27 IV-27 7.1.6 7.1. 6 Spray Spargers Core Spray Core Spargers IV-27 IV-27 7.1.7 7.1. 7 Poison Sparger Liquid Poison Liquid Sparger IV-28 IV-28 7.1.8 7.1. 8 Steam Separator and Steam Separator and Dryer Dryer IV-28 IV-28 7.1.9 7.1. 9 Core Shroud Stabilizers Core Shroud Stabilizers IV-28 IV-28 7.1.10 7.1.10 Core Weld Repair Vertical Weld Shroud Vertical Core Shroud Repair IV-30 IV-30 7.2 7.2 Design Evaluation Design Evaluation IV-30 IV-30 UFSAR UFSAR Revision Revision 21 21 vii vii October 2009 October 2009

Nine Mile Point Unit 1 UFSARUFSAR Section Section 7.3 Title Surveillance and Testing Surveillance Testing (Cont'd.) TABLE OF CONTENTS (Cont'd.) IV-31 IV-31 C. C. REFERENCES REFERENCES IV-32 IV-32 SECTION V SECTION V REACTOR REACTOR COOLANT COOLANT SYSTEM V-i V-I A. A. DESIGN DESIGN BASES V-I V-I 1.0 l.0 General General V-I V-I 2.0 Performance Objectives Performance Objectives V-i V-I 3.0 Design Design Pressure Pressure V-2 V-2 4.0 Cyclic Loads (Mechanical (Mechanical and Thermal) Thermal) V-3 V-3 5.0 Codes V-3a V-3a B. B. SYSTEM DESIGN AND OPERATION OPERATION V-4 V-4 1.0 l.0 General General V-4 V-4 1.1 l.1 Drawings V-4 V-4 1.2 l.2 Materials of Construction Materials Construction V-4 V-4 1.3 l.3 Thermal Thermal Stresses V-4 V-4 1.4 l.4 Primary Leakage Primary Coolant Leakage V-5 V-5 1.5 l.5 Coolant Coolant Chemistry V-6 V-6 2.0 Reactor Vessel Reactor V-6 V-6 3.0 Reactor Recirculation Reactor Recirculation Loops V-7 V-7 4.0 Reactor Steam and Auxiliary Systems Reactor Systems Piping Piping V-8 V-8 5.0 Relief Devices V-8 V-8 C. C. SYSTEM DESIGN EVALUATION SYSTEM EVALUATION V-10 V-10 1.0 l.0 General V-10 V-10 2.0 Pressure Pressure V-10 V-10 3.0 Design Design Heatup and Cooldown Cooldown Rates V-lI V-II 4.0 Materials Radiation Exposure Materials Radiation V-12 4.1 Pressure-Temperature Limit Curves Pressure-Temperature Curves V-12 4.2 Temperature Temperature Limits for Boltup V-12 4.3 Temperature Temperature Limits Limits for In-Service Pressure Tests System Pressure V-13 4.4 Operating Operating Limits During Heatup, Heatup, Cooldown, Operation Cooldown, and Core Operation V-13 4.5 Predicted Predicted Shift in RT~T in RTNDT V-13 4.6 Fluence Calculations Neutron Fluence Calculations V-13 5.0 Mechanical Considerations Mechanical Considerations V-14 5.1 5.1 Jet Reaction Forces V-14 UFSAR Revision Revision 21 21 viii viii October 2009

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• Section Section 5.2 Title TABLE OF CONTENTS Seismic Seismic Forces CONTENTS (Cont'd.)

(Cont'd.) Page V-14 6.0 G.0 Safety Limits, Limiting Safety Limits, Limiting Safety Safety Settings Settings and Minimum Conditions for for Operation Operation V-14 D. D. TESTS AND INSPECTIONS INSPECTIONS V-l6 V-16 1.0 Prestartup Testing Prestartup V-16 V-16 2.0 Inspection Inspection and Testing Testing Following Startup Startup V-16 V-16 2.1 Pressure Test Pressure V-16 V-16 2.2 Pressure Vessel Vessel Irradiation Irradiation V-16 V-16 E. E. EMERGENCY COOLING EMERGENCY COOLING SYSTEM SYSTEM V-17 1.0 Design Bases Bases V-17 2.0 Design and Operation System Design Operation V-17 3.0 Design Evaluation V-19 3.1 Redundancy Redundancy V-19 3.2 Makeup Makeup Water V-19 3.3 System System Leaks V-19 V-19 3.4 Containment Isolation Containment V-19 V-19 4.0 Tests Tests and Inspections Inspections V-20 V-20 4.1 Prestartup Prestartup Test V-20 V-20 4.2 Inspections and Tests Subsequent Inspections Tests V-20 V-20 F. F. REFERENCES REFERENCES V-21 V-21 SECTION VI CONTAINMENT SYSTEM CONTAINMENT SYSTEM VI-1 VI-l A. A. PRIMARY CONTAINMENT CONTAINMENT - MARK I CONTAINMENT PROGRAM CONTAINMENT PROGRAM VI-2 VI-2 1.0 General Structure General Structure VI-2 VI-2 2.0 Pressure Suppression Hydrodynamic Suppression Hydrodynamic Loads VI-2 VI-2 2.1 Safety/Relief Valve Discharge Safety/Relief Discharge VI-2 VI-2 2.2 Loss-of-Coolant Loss-of-Coolant Accident VI-3 VI-3 2.3 Summary Phenomena Summary of Loading Phenomena VI-4 VI-4 3.0 Plant-Unique Modifications Plant-Unique Modifications VI-5 VI-5 B. B. PRIMARY CONTAINMENT - PRESSURE PRIMARY CONTAINMENT PRESSURE SUPPRESSION SUPPRESSION SYSTEM VI-6 VI-6 1.0 Design Bases Bases VI-6 VI-6 1.1 General General VI-6 VI-6 21 UFSAR Revision 21 ix lX October 2009

Nine Mile Nine Unit 11 UFSAR Point Unit Mile Point UFSAR Section Section 1.2 1.2 Title Title TABLE OF CONTENTS TABLE CONTENTS (Cont'd.) Design Basis Accident Design (Cont'd.) Accident (DBA) (DBA) Page Page VI-6 VI-6 1.3 1.3 Containment Heat Removal Containment Removal VI-8 VI-8 1.4 1.4 Isolation Criteria Isolation Criteria VI-8 VI-8 1.5 1.5 Vacuum Relief Vacuum Relief Criteria Criteria VI-9 VI-9 1.6 1.6 Flooding Criteria Flooding Criteria VI-9 VI-9 1.7 1.7 Shielding Shielding VI-9 VI-9 2.0 Structure Design Structure Design VI-9 VI-9 2.1 General General VI-9 VI-9 2.2 Penetrations and Access Penetrations Access Openings VI-II VI-11 2.3 Jet and Missile Protection Protection VI-12 2.4 Materials Materials VI-13 2.5 Shielding Shielding VI-14 2.6 Vacuum Relief Vacuum Relief VI-14 2.7 Containment Flooding Containment Flooding VI-14 C. C. SECONDARY CONTAINMENT SECONDARY CONTAINMENT -- REACTOR REACTOR BUILDING BUILDING VI-16 VI-16 1.0 Design Bases Design VI-16 1.1 Snow Loadings Wind and Snow VI-16 1.2 Pressure Relief Pressure Relief Design VI-16 1.3 Seismic Design Seismic VI-17 1.4 Shielding Shielding VI-17 2.0 Structure Structure Design VI-17 2.1 Features Structural Features General Structural VI-17 D. D. CONTAINMENT ISOLATION CONTAINMENT ISOLATION SYSTEM VI-20 1.0 Design Bases VI-20 1.1 Containment Containment Spray Appendix J WaterWater Seal Requirements Requirements VI-23a VI-23a 2.0 System Design VI-24 3.0 Tests and Inspections VI-26 E. E. CONTAINMENT VENTILATION SYSTEM CONTAINMENT VENTILATION VI-27 1.0 Containment Primary Containment VI-27 1.1 Design Bases VI-27 1.2 System Design System VI-27 2.0 Containment Secondary Containment Secondary VI-28 2.1 Design Bases VI-28 2.2 System Design VI-28 F. F. TEST AND INSPECTIONS VI-30 1.0 Drywell and Suppression Chamber VI-30 UFSAR Revision 21 .UFSAR 21 xX October 2009

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• Section Title TABLE OF CONTENTS CONTENTS (Cont'd.)

Preoperational Testing Preoperational (Cont'd.) Page VI-30 VI-30 1.1 1.2 1.2 Postoperational Postoperational Testing VI-30 VI-30 2.0 2.0 Containment Penetrations and Containment Penetrations Isolation Valves Valves VI-31 VI-31 2.1 2.1 Penetration and Valve Penetration Leakage Valve Leakage VI-31 VI-31 2.2 2.2 Valve Operability Test Operability Test VI-31 VI-31 3.0 3.0 Containment Ventilation System Containment Ventilation System VI-32 4.0 4.0 Other Containment Other Tests Containment Tests VI-32 5.0 S.O Reactor Building Reactor VI-32 VI-32 5.1 S.l Reactor Building Reactor Building Normal Ventilation System System VI-32 VI-32 5.2 S.2 Reactor Building Reactor Isolation Valves Building Isolation Valves VI-32 VI-32 S.3 5.3 Emergency Ventilation System Emergency Ventilation System VI-33 G. G. REFERENCES REFERENCES VI-33 VII SECTION VII SAFEGUARDS ENGINEERED SAFEGUARDS VII-1 VII-2

• A.

A. 1.0 1.0 2.0 2.0 2.1 2.1 2.2 2.2 CORE SPRAY Design Bases General General SYSTEM SPRAY SYSTEM Bases

System Design

Operator Assessment Operator Assessment VII-2 VII-2 VII-2 VII-2 VII-2 VII-2 VII-2 VII-2 VII-S VII-5 3.0 3.0 Design Evaluation VII-6 VII-6 4.0 4.0 Tests and Inspections Tests Inspections VII-7 B. B. CONTAINMENT SPRAY CONTAINMENT SPRAY SYSTEM VII-8 VII-8 1.0 1.0 Licensing Basis Requirements Licensing Requirements VII-8 VII-8 1.1 IOCFR50.49 - Environmental 10CFRSO.49 Environmental Qualification Electric Equipment Qualification of Electric Equipment Important to Safety Important Nuclear Safety for Nuclear Power Plants Plants VII-8 VII-8 1.2 1.2 Appendix A 10CFR50 Appendix 10CFRSO General Design A - General Criteria Criteria for Nuclear Power Plants VII-8 VII-8 2.0 2.0 Design Bases Bases VII-ll VII-11 2.1 2.1 Design Basis Functional Requirements Functional Requirements VII-ll VII-11 2.2 2.2 Parameters Controlling Parameters VII-12 3.0 3.0 System Design VII-12a VII-12a 3.1 3.1 System Function VII-12a VII-12a 3.2 3.2 System Design Description VII-12b VII-12b 3.3 3.3 System Design VII-12c VII-12c 3.4 3.4 Standards Codes and Standards VII-14b VII-14b UFSAR Revision 21 Revision 21 xa xa October 2009

Nine Mile Nine Unit 11 UFSAR Point Unit Mile Point UFSAR Section Section 3.5 3.5 Title Title TABLE OF TABLE CONTENTS (Cont'd.) OF CONTENTS System Instrumentation System Instrumentation (Cont'd.) Page Page VII-14b VII-14b 3.6 3.6 Design Features

System Design

System Features VII-14c VII-14c 4.0 4.0 Design Performance Evaluation Design Performance Evaluation VII-14d VII-14d 4.1 4.1 Performance Analyses System Performance System Analyses VII-14d VII-14d 4.2 4.2 System Response System Response VII-14e VII-14e 4.3 4.3 With Other Interdependency With Interdependency Other Engineered Safeguards Systems Engineered Safeguards Systems VII-14e VII-14e 5.0 5.0 System Operation System Operation VII-14f VII-14f 5.1 5.1 Limiting for Operation Conditions for Limiting Conditions Operation VII-14f VII-14f 6.0 6.0 and Inspection Tests and Tests Inspection VII-14g VII-14g C. C. LIQUID POISON LIQUID INJECTION SYSTEM POISON INJECTION SYSTEM VII-15 VII-15 1.0 1.0 Design Bases Design Bases VII-15 VlI-1s 2.0 2.0 System Design System Design VII-16 VII-16 2.1 2.1 Operator Assessment Operator Assessment VII-19 VII-19 3.0 3.0 Design Evaluation Design Evaluation VII-20 VII-20 4.0 4.0 and Inspections Tests and Tests Inspections VII-21 VII-21 5.0 5.0 Alternate Boron Injection Alternate Boron Injection VII-21a VII-21a D. D. CONTROL ROD CONTROL VELOCITY LIMITER ROD VELOCITY LIMITER VII-22 VII-22 1.0 1.0 Design Bases Design Bases VII-22 VII-:-22 2.0 2.0 System Design System Design VII-22 VII-22 3.0 3.0 Design Evaluation Design Evaluation VII-24 VII-24 3.1 3.1 General General VII-24 VII-24 3.2 3.2 Design Sensitivity Design Sensitivity VII-24 VII-24 3.3 3.3 Normal Operation Normal Operation VII-25 VII-2s 4.0 4.0 and Inspections Tests and Tests Inspections VII-25 VII-2s E. E. CONTROL ROD CONTROL HOUSING SUPPORT ROD HOUSING SUPPORT VII-26 VII-26 1.0 1.0 Design Bases Design Bases VII-26 VII-26 2.0 2.0 System Design System Design VII-26 VII-26 2.1 2.1 Loads and Deflections Loads and Deflections VII-28 VII-28 3.0 3.0 Design Evaluation Design Evaluation VII-28 VII-28 4.0 4.0 Tests and Tests and Inspections Inspections VII-29 VII-29 F. F. FLOW RESTRICTORS FLOW RESTRICTORS VII-30 VII-30 1.0 1.0 Design Bases Design Bases VII-30 VII-30 2.0 2.0 System Design System Design VII-30 VII-30 3.0 3.0 Design Evaluation Design Evaluation VII-30 VII-30 4.0 4.0 Tests and Inspections Tests and Inspections VII-31 VII-31 UFSAR Revision 21 UFSAR Revision 21 xb xb October 2009 October 2009 0

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• Section Section G.

G. Title TABLE OF CONTENTS (Cont'd.) COMBUSTIBLE GAS CONTROL COMBUSTIBLE (Cont'd.) CONTROL SYSTEM Page VII-32 1.0 Design Design Bases VII-32 2.0 Containment Inerting System Containment Inerting VII-32 2.1 System Design VII-32 2.2 Design Design Evaluation VII-33 VII-33 3.0 Containment Containment Atmospheric Atmospheric Dilution System VII-33 3.1 System Design VII-33 3.2 Design Evaluation Design Evaluation VII-35 4.0 Inspections Tests and Inspections VII-35 H. H. EMERGENCY VENTILATION EMERGENCY SYSTEM VENTILATION SYSTEM VII-36 VII-36 1.0 Design Bases Bases VII-36 VII-36 2.0 System Design VII-36 VII-36 2.1 Operator Operator Assessment Assessment VII-38 VII-38 3.0 Design Evaluation Design Evaluation VII-39 VII-39 4.0 Inspections Tests and Inspections VII-39

• II..

1.0 2.0 3.0 4.0 HIGH-PRESSURE HIGH-PRESSURE COOLANT INJECTION Design Bases

System Design

Design Evaluation Evaluation Tests and Inspections Inspections INJECTION VII-41 VII-41 VII-41 VII-41 VII-41 VII-41 VII-42 VII-42 VII-43

• UFSAR Revision 21 21 xc XC October 2009 October 2009

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• Section Section Title Title CONTENTS (Cont'd.)

OF CONTENTS TABLE OF TABLE (Cont'd.) Page J. J. FUEL AND FUEL AND REACTOR REACTOR COMPONENTS COMPONENTS HANDLING HANDLING SYSTEM SYSTEM X-39 X-39 1.0 1.0 Design Design Bases Bases X-39 X-39 2.0 2.0 System

System Design

Design X-40 X-40 2.1 2.1 Description of Description of Facility Facility X-40 X-40 2.1.1 2.1.1 Cask Cask Drop Protection System Drop Protection System X-42 X-42 2.2 2.2 Operation of Operation of the the Facility Facility X-43 X-43 2.3 2.3 Control Control ofof Heavy Loads Program Heavy Loads Program X-43a X-43a 2.3.1 2.3.1 Introduction/Licensing Introduction/Licensing Background Background X-43a X-43a 2.3.2 2.3.2 Safety Safety Basis Basis X-43b X-43b 2.3.3 2.3.3 Scope Scope of of Heavy Heavy Load Load Handling Handling Systems Systems X-43b X-43b 2.3.4 2.3.4 Control of Control of Heavy Heavy Loads Loads Program Program X-43c X-43c 2.3.4.1 2.3.4.1 NMPNS Commitments in NMPNS Commitments Response to in Response to NUREG-0612, Phase II Elements NUREG-0612, Phase Elements X-43c X-43c 2.3.4.2 2.3.4.2 Reactor Pressure Reactor Pressure Vessel Head and Vessel Head and Spent Fuel Spent Fuel Cask Cask Lifts Lifts X-43d X-43d 2.3.5 Safety Safety Evaluation X-43d 2.3.5 Evaluation X-43d 3.0 3.0 Design Evaluation Design Evaluation X-43d X-43d 4.0 4.0 Tests Tests and and Inspections Inspections X-44 X-44 K. K. FIRE PROTECTION FIRE PROTECTION PROGRAM PROGRAM X-45 X-45 1.0 1.0 Program Program Bases Bases X-45 X-45 1.1 1.1 Nuclear Nuclear Division Directive -- Fire Division Directive Fire Protection Program Protection Program X-45 X-45 1.2 1.2 Nuclear Division Interface Nuclear Division Interface Procedure Fire Protection Procedure -- Fire Protection Program Program X-45 X-45 1.3 1.3 Fire Fire Hazards Hazards Analysis Analysis X-45 X-45 1.4 1.4 Appendix Review Safe Appendix RR Review Safe Shutdown Shutdown Analysis Analysis X-46 X-46 1.5 1.5 Protection and Fire Protection Fire Appendix RR and Appendix Related Portions of Operations Related Portions of Operations (OPs, SOPs, Procedures (OPs, Procedures SOPs, and and EOPs) EOPs) and Damage and Damage Repair Repair Procedures Procedures X-46 X-46 1.6 1.6 Fire Fire Protection Protection Portions Portions ofof the the Emergency Emergency Plan Plan X-46 X-46 2.0 2.0 Program Implementation and Program Implementation and Design Design Aspects Aspects X-46 X-46 2.1 2.1 Fire Protection Implementing Fire Protection Implementing Procedures Procedures X-46 X-46 2.2 Fire Fire Protection Protection Administrative 2.2 Administrative Controls Controls X-47 X-47 UFSAR UFSAR Revision Revision 21 21 xvii XVll October 2009 October 2009

Nine Mile Point Unit 1 UFSAR Section Section Title TABLE (Cont'd.) TABLE OF CONTENTS (Cont'd.) Page 2.3 Fire Protection Protection System Drawings and Calculations Calculations X-47 X-47 2.4 Protection Engineering Fire Protection Evaluations (FPEEs) Evaluations (FPEEs) X-47 X-47 3.0 Monitoring and Evaluating Monitoring Evaluating Program Program Implementation Implementation X-47 X-47 3.1 Quality Assurance Program Quality Assurance X-47 X-47 3.2 Brigade Manning, Fire Brigade Manning, Training, Training, Drills and Responsibilities Responsibilities X-47 X-47 4.0 Surveillance and Tests Surveillance X-48 X-48 L. L. REMOTE SYSTEM REMOTE SHUTDOWN SYSTEM X-49 X-49 1.0 1.0 Design Bases Bases X-49 X-49 2.0 System Design X-49 X-49 3.0 System Evaluation Evaluation X-49 X-49 4.0 Tests Tests and Inspections Inspections X-50 X-SO M. 1.0 1.0 1.1 HYDROGEN HYDROGEN WATER CHEMISTRY AND NOBLE METAL CHEMICAL SYSTEMS SYSTEMS CHEMICAL ADDITION (NOBLECHEM) Design Basis Noble Metal (NOBLECHEM) Chemical Addition System Metal Chemical X-51 X-S1 X-51 X-S1 X-52 X-S2 1.2 Hydrogen Water Chemistry Hydrogen System Chemistry System X-52a X-S2a 2.0 System System Design X-53 X-S3 2.1 Noble Metal Chemical Addition Metal Chemical Addition X-53 X-S3 2.2 Hydrogen Water Chemistry Hydrogen Chemistry System X-53 X-S3 2.2.1 HWC Feedwater Hydrogen Injection Feedwater Hydrogen Injection X-54 X-S4 2.2.2 HWC Offgas Oxygen Injection Oxygen Injection X-54 X-S4 2.2.3 HWC Offgas Offgas Sample X-54 X-S4 3.0 System Evaluation System Evaluation X-54 X-S4 4.0 Inspections Tests and Inspections X-55 X-SS N. N. REFERENCES REFERENCES X-56 X-S6 10A APPENDIX lOA HAZARDS ANALYSIS FIRE HAZARDS ANALYSIS APPENDIX APPENDIX 10B lOB SHUTDOWN ANALYSIS SAFE SHUTDOWN ANALYSIS SECTION XI STEAM-TO-POWER CONVERSION SYSTEM STEAM-TO-POWER CONVERSION XI-I XI-1 A. A. UFSAR Revision UFSAR DESIGN DESIGN BASES 21 Revision 21 xviia XI-I XI-1 October 2009 2009

Nine Mile Point Unit 1 UFSAR

• Section Section TABLE OF CONTENTS Title Title CONTENTS (Cont'd.)

(Cont'd.) Page B. B. SYSTEM DESIGN AND OPERATION OPERATION XI-2 XI-2 1.0 Turbine Generator Generator XI-2 XI-2 2.0 Turbine Condenser Condenser XI-4 XI-4 3.0 Condenser Condenser Air Removal Removal and Offgas Offgas System XI-5 XI-5 4.0 Circulating Water System Circulating Water XI-9a 5.0 Condensate Pumps Condensate XI-9a XI-9a 6.0 Condensate Filtration System Condensate Filtration XI-9b 6.0A 6.OA Condensate Demineralizer Condensate Demineralizer System XI-9c XI-9c 7.0 Transfer System Condensate Transfer Condensate System XI-II XI-11 8.0 Feedwater Booster Feedwater Booster Pumps XI-II XI-11 9.0 Feedwater Pumps Feedwater XI-11a 10.0 Feedwater Heaters Feedwater Heaters XI-12 C. C. SYSTEM ANALYSIS ANALYSIS XI-13 D. D. TESTS AND INSPECTIONS INSPECTIONS XI-16

• SECTION XII A.

A. 1.0 1.0 RADIOLOGICAL CONTROLS RADIOLOGICAL Design Design Bases CONTROLS RADIOACTIVE WASTES RADIOACTIVE WASTES XII-l XlI-1 XII-l XlI-1 XII-1 XlI-1 XII-1 1.1 Objectives Objectives XlI-1 1.2 Types of Radioactive Radioactive Wastes XII-l XlI-1

• 21 UFSAR Revision 21 xviib 2009 October 2009

Nine Mile Point Unit 1 UFSAR Section Section TABLE TABLE OF CONTENTS Title Title (Cont'd.) CONTENTS (Cont'd.) Page .1.2.1 1.2.1 Gaseous Waste Waste XII-I XlI-l

1. 2.2 1.2.2 Liquid Wastes Wastes XII-1 XlI-l

1. 2.3 1.2.3 Solid Wastes Solid XII-2 XII-2 2.0 System System Design and Evaluation Evaluation XII-2 XII-2 2.1 Gaseous Waste Waste System XII-2 XII-2 2.1.1 Offgas System System XII-3 XII-3 2.1.

2.1.22 Steam-Packing Exhausting Steam-Packing System Exhausting System XII-3 XII-3 2.1. 2.1.33 Building Ventilation Building Systems Ventilation Systems XII-3 XII-3 2.1.4 Stack XII-3 XII-3 2.2 Liquid Waste Liquid Waste System XII-4 XII-4 2.2.1 Liquid Waste Waste Handling Processes Handling Processes XII-4 XII-4 2.2.2 Sampling and Monitoring Monitoring Liquid Wastes XII-6 XII-6 2.2.3 Liquid Waste Liquid Waste Equipment Arrangement Arrangement XII-7 XII-7 2.2.4 Radioactive Waste System Liquid Radioactive System Control Control XII-7 XII-7 2.3 Solid Waste System Solid XII-8 XII-8 2.3.1 Solid Waste Handling Solid Handling Processes Processes XII-8 XII-8 2.3.2 Solid Waste System Equipment System Equipment XII-9 XII-9 2.3.3 Process Control Program Process Program XII-9a XII-9a 3.0 Safety Limits Safety XII-9a XII-9a 4.0 Tests and Inspections Inspections XII-10 XII-I0 4.1 Waste Process Process Systems XII-10 XII-I0 4.2 Filters Filters XII-I0 XII-10 4.3 Effluent Monitors Effluent Monitors XII-IO XII-10 4.3.1 Offgas and Stack Monitors Monitors XII-IO XII-10 4.3.2 Liquid Waste Waste Effluent Effluent Monitor XII-IO XII-10 B. B. RADIATION RADIATION PROTECTION PROTECTION XII-11 XII-II 1.0 Primary and Secondary Secondary Shielding Shielding XII-11 XII-II 1.1 Bases Design Bases XII-1I XII-II 1.2 Design Design XII-12

1. 2.1 1.2.1 Reactor Reactor Shield Wall XII-12

1. 2.2 1.2.2 Biological Shield Biological Shield XII-12

1. 2.3 1.2.3 Miscellaneous Miscellaneous XII-12 1.3 Evaluation Evaluation XII-13 2.0 Radioactivity Monitoring Area Radioactivity Systems Systems XII-13 2.1 Area Area Radiation Monitoring System Radiation Monitoring XII-13 2.1.1 Design Bases Bases XII-13 2.1.22 2.1. Design XII-14 2.1.33 2.1. Evaluation Evaluation XII-15 XII-IS UFSAR Revision 20 Revision 20 xviii xviiL October October 2007 2007

Nine Mile Point Unit 1 UFSAR

• Section Section Title Area Air Contamihation (Cont'd.)

TABLE OF CONTENTS (Cont'd.) Contamination Monitoring Page 2.2 System XII-16 2.2.1 Bases Design Bases XII-16 2.2.2 Design XII-16 2.2.3 Evaluation Evaluation XII-16a XII-16a 3.0 Radiation Protection Radiation Protection XII-17 3.1 Facilities Facilities XII-17 3.1.1 Laboratory, Counting Room Laboratory, Room and Calibration Calibration Facilities Facilities XII-17 3.1.22 3.1. Shower Facilities Change Room and Shower Facilities XII-18 3.1.33 3.1. Personnel Decontamination Facility Personnel Decontamination Facility XII-19 3.1.4 Tool and Equipment Tool Equipment Decontamination Decontamination Facility Facility XII-19 3.2 Radiation Control Radiation Control XII-20 3.2.1 Shielding Shielding XII-20 3.2.2 Control Access Control XII-20 3.3 Contamination Control Contamination XII-21a XII-21a 3.3.1 Facility Contamination Facility Contamination Control XII-21a XII-21a 3.3.2 Personnel Contamination Control Personnel Contamination Control XII-22 3.3.3 Airborne Contamination Control Airborne Contamination XII-23 3.4 Personnel Dose Determinations Personnel Determinations XII-24 3.4.1 3.4.1 Radiation Radiation Dose XII-24 3.5 Radiation Protection Radiation Instrumentation Instrumentation XII-24 3.5.1 Instrumentation Counting Room Instrumentation XII-24 3.5.2 Radiation Instrumentation Portable. Radiation Instrumentation Portable XII-25 3.5.3 Air Sampling Instrumentation Sampling Instrumentation XII-25 3.5.4 Personnel Monitoring Personnel Instruments Monitoring Instruments XII-25 3.5.5 Emergency Instrumentation Emergency Instrumentation XII-25 4.0 Tests and Inspections Tests Inspections XII-26 4.1 Shielding XII-26 4.2 Radiation Monitors Area Radiation Monitors XII-27 4.3 Contamination Monitors Area Air Contamination Monitors XII-27 4.4 Radiation Protection Facilities Radiation Protection Facilities XII-27 4.4.1 Ventilation Air Flows Ventilation XII-27 4.4.2 Instrument Calibration Well Instrument Calibration Well Shielding XII-28 4.5 Radiation Protection Radiation Instrumentation Instrumentation XII-28

• UFSAR Revision UFSAR 21 Revision 21 xix October 2009

Nine Mile Nine Mile Point Point Unit 1 UFSAR TABLE OF CONTENTS TABLE CONTENTS (Cont'd.) (Cont'd.) Section 0 Section Title SECTION XIII SECTION XIII CONDUCT OF OPERATIONS OPERATIONS XllI-1 XIII-I A. ORGANIZATION ORGANIZATION AND RESPONSIBILITY XllI-1 XIII-l 1.0 1.0 Management and Technical Management Technical Support Support Organization Organization XllI-1 XIII-l 1.1 Station Organization XIII-l XllI-1 1.1.1 1.1.1 Vice President President Nine Nine Mile PointPoint XllI-1 XIII-l 1.1.22 1.1. Matrixed Matrixed Reporting XllI-1 XIII-l 1.1.33 1.1. Qualifications of Support Personnel Qualifications Personnel XIII-2 XIII-2 2.0 2.0 Nine Mile Mile Point Nuclear Station, LLC, Nuclear Station, LLC, Organization XIII-2 XIII-2 2.1 2.1 General Manager Plant General Manager XIII-3 XIII-3 2.2 2.2 Other Functions Reporting to the Functions Reporting Vice President President Nine Mile Point Point XIII-6 XIII-6 2.3 2.3 Supervisor Engineering-Nuclear Fuels Supervisor Engineering-Nuclear Fuels XIII-8 XIII-8 3.0 3.0 Assurance Quality Assurance XIII-9 XIII-9 4.0 4.0 Operating Shift Crews Operating Crews XIII-9 XIII-9 5.0 5.0 Qualifications Qualifications of Staff Personnel Staff Personnel XIII-10 B. B. 1.0 1.0 2.0 3.0 QUALIFICATIONS AND TRAINING QUALIFICATIONS PERSONNEL PERSONNEL section deleted This section section deleted This section section deleted This section TRAINING OF XIII-II XIII-ll XIII-II XIII-II XIII-ll XIII-II XIII-ll XIII-II 4.0 Training of Personnel Training Personnel XIII-ll XIII-II 4.1 General Responsibility XIII-ll XIII-II 4.2 Implementation Implementation XIII-II XIII-II 4.3 Quality XIII-II XIII-II 4.3.1 4.3.1 Operator Training For Operator XIII-ll XIII-II 4.3.2 4.3.2 Maintenance For Maintenance XIII-12 4.3.3 4.3.3 For Technicians Technicians XIII-12 4.3.4 4.3.4 For General Employee Employee Training/Radiation Protection and Training/Radiation Protection Emergency Emergency Plan XIII-12 4.3.5 4.3.5 For Industrial Industrial Safety XIII-12 4.3.6 4.3.6 For Nuclear Quality Assurance Nuclear Quality Assurance XIII-12 4.3.7 4.3.7 For Fire Brigade Brigade XIII-12 4.3.8 4.3.8 For Manager Operations and General Manager Operations General Supervisor Operations Supervisor Operations XIII-12 UFSAR Revision 21 21 xx XX October 2009 2009

Mile Point Unit 11 UFSAR Nine Mile (Cont'd.) TABLE OF CONTENTS (Cont'd.) Section Section Title Title Page 4.4 Training of Licensed Operator Training Operator Candidates/Licensed NRC Operator Candidates/Licensed Operator Retraining Retraining XIII-12 5.0 Local, Cooperative Training With Local, State State and Federal Officials Federal Officials XIII-14 C. C. OPERATING PROCEDURES XIII-15 XIII-IS D. D. EMERGENCY PLAN AND PROCEDURES EMERGENCY XIII-17 E. E. SECURITY SECURITY XIII-19 F. F. RECORDS XIII-20 1.0 1.0 Operations Operations XIII-20 1.1 1.1 Control Room Log XIII-20 1.2 1.2 Shift Manager's Manager's Log XIII-20 1.3 1.3 Radwaste Radwaste Log XIII-20 1.4 Waste Quantity Level Level Shipped XIII-20 1.4 Shipped 2.0 2.0 Maintenance Maintenance XIII-21 3.0 Radiation Protection Radiation Protection XIII-21 XIII-21 3.1 Personnel Personnel Exposure Exposure XIII-21 3.2 3.2 By-Product By-Product Material Material as Required Required byby 10CFR30 10CFR30 XIII-21 3.3 Meter Calibrations Meter Calibrations XIII-21 3.4 Station Radiological Radiological Conditions Conditions in in Accessible Areas Accessible XIII-21

• UFSAR UFSAR Revision Revision 21 21 xxa xxa 2009 October 2009

Nine Mile Point Unit 1 UFSAR Point Unit UFSAR THIS PAGE INTENTIONALLY THIS INTENTIONALLY BLANK 17 UFSAR Revision 17 xxb xxb October 2001

Mile Point Nine Mile Nine Unit 1 UFSAR Point Unit TABLE OF TABLE CONTENTS (Cont'd.) OF CONTENTS (Cont'd.) Section Title Page 3.5 3.5 Administration of Administration of the Radiation Radiation Program and Procedures Protection Program Protection Procedures XIII-21 XIII-21 4.0 4.0 Chemistry and Radiochemistry Chemistry Radiochemistry XIII-21 XIII-21 5.0 5.0 Special Nuclear Materials Special Nuclear Materials XIII-22 6.0 6.0 Calibration of Instruments Calibration Instruments XIII-22 7.0 7.0 Administrative and Reports Records and Administrative Records Reports XIII-22 G. G. REVIEW AUDIT OF OPERATIONS REVIEW AND AUDIT OPERATIONS XIII-23 XIII-23 1.0 1.0 Operations Review Plant Operations Committee Review Committee XIII-23 1.1

1. 1 Function Function XIII-23 2.0 2.0 Nuclear Review Board Nuclear Safety Review Board XIII-23 2.1 2.1 Function Function XIII-24 3.0 3.0 Operating Experience Review of Operating Experience XIII-24 SECTION SECTION XIV INITIAL TESTING AND OPERATIONS INITIAL OPERATIONS XIV-I XIV-1 A.

A. INITIAL REACTOR TESTS PRIOR TO INITIAL REACTOR FUELING XIV-1 XIV-1

• B.

B. 1.0 1.0 1.1 INITIAL CRITICALITY AND INITIAL POSTCRITICALITY TESTS POSTCRITICALITY Initial Fuel Loading and Near-Zero Initial Power General Requirements Requirements Near-Zero Atmospheric Pressure Power Tests at Atmospheric XIV-5 XIV-S XIV-XIV-S5 XIV-5 XIV-S 1.2 1.2 Procedures General Procedures XIV-5 XIV-S 1.3 1.3 Core Loading and Critical Test Test Program Program XIV-7 XIV-7 2.0 2.0 Heatup from Ambient Rated Ambient to Rated Temperature Temperature XIV- 9 XIV-9 2.1 2.1 General General XIV-9 XIV-9 2.2 2.2 Conducted Tests Conducted XIV-9 XIV-9 3.0 3.0 From Zero to 100 Initial 100 Percent Initial Reactor Rating XIV-10 XIV-10 4.0 4.0 Full-Power Demonstration Run Full-Power Demonstration Run XIV-12 XIV-12 5.0 5.0 Comparison of Base Conditions Comparison XIV-12 XIV-12 6.0 6.0 Additional Tests at Design Rating Additional XIV- 13 XIV-13 7.0 7.0 Report Startup Report XIV- 13 XIV-13 SECTION XV XV ANALYSIS SAFETY ANALYSIS XV- 1 XV-1 A. XV- 1 A. INTRODUCTION XV-1 21 UFSAR Revision 21 xxi XXl October 2009

Nine Mile Point Unitunit 1 UFSAR TABLE OF CONTENTS (Cont'd.) (Cont'd.) Section Section Title Page B. B. BOUNDARY PROTECTION SYSTEMS BOUNDARY PROTECTION XV-2 XV-2 1.0 Transients Considered Transients Considered XV-2 XV-2 2.0 Assumptions Methods and Assumptions XV-3 XV-3 3.0 Transient Analysis Transient Analysis XV-3 XV-3 3.1 Turbine Trip Without Bypass Turbine XV-3 XV-3 3.1.1 Objectives Objectives XV-3 XV-3 3.1. 3.1.22 Assumptions and Initial Assumptions Initial Conditions Conditions XV-3 XV-3 3.1. 3.1.33 Comments Comments XV-3 XV-3 3.1.4 Results XV-3 XV-3 3.2 100OF Feedwater Loss of 100°F Feedwater Heating XV-4 XV-4 3.2.1 Objective Objective XV-4 XV-4 3.2.2 Assumptions and Initial Assumptions Initial Conditions Conditions XV-4 XV-4 3.2.3 Results XV-4 XV-4 3.3 Feedwater Controller Failure-Feedwater Controller Failure-Maximum Demand Maximum XV-5 XV-5 3.3.1 Objective Objective XV-5 XV-5 3.3.2 Assumptions and Initial Assumptions Conditions Initial Conditions XV-5 XV-5 3.3.3 Comments XV-5 XV-5 3.3.4 3.4 3.4.1 3.4.2 3.4.3 3.4.4 Results Control Rod Withdrawal Error Control Objective Objective Assumptions Comments Results Initial Assumptions and Initial Conditions XV-5 XV-5 XV-5 XV-5 XV-5 XV-5 XV-5 XV-5 XV-6 XV-6 XV-6 XV-6 3.5 Main Steam Line Isolation Main Isolation Valve Valve Closure (With Scram) Scram) XV-6 XV-6 3.5.1 Objective Objective XV-6 XV-6 3.5.2 Assumptions Initial Assumptions and Initial Conditions XV-6 XV-6 3.5.3 Comments Comments XV-7 XV-7 3.5.4 Results XV-7 XV-7 3.6 Inadvertent Inadvertent Startup of Cold Recirculation Loop Recirculation XV-7 XV-7 3.6.1 Objective Objective XV-7 XV-7 3.6.2 Assumptions Initial Assumptions and Initial Conditions Conditions XV-7 XV-7 3.6.3 Comment XV-8 XV-8 3.6.4 Results Results XV-9 XV-9 3.7 Recirculation Pump Trips Recirculation Trips XV-9 XV-9 3.7.1 Objectives Objectives XV-9 XV-9 3.7.2 Assumptions and Initial Assumptions Initial Conditions Conditions XV-9 XV-9 3.7.3 Comments XV-9 XV-9 3.7.4 Results Results XV-10 3.8 UFSAR Revision Recirculation Recirculation Pump Stall 16 Revision 16 xxi i xxii Stall XV-10 XV-10 November 1999 1999

Nine Mile Point Unit 1 UFSAR

• Section Section 1.2.5 Title Title TABLE OF CONTENTS (Cont'd.)

(Cont'd.) Page 1.2.5 Subcooled Subcooled Liquid XV-30 XV-30 1.2.6

1. 2.6 System System Pressure and Steam-Water Steam-Water Mass XV-31 XV-31 1.2.7

1. 2.7 Mixture Impact Forces XV-31 XV-31 1.2.8

1. 2.8 Core Internal Internal Forces Forces XV-31 XV-31 1.3 Radiological Effects Radiological Effects XV-32 XV-32 1.3.1

1. 3.1 Radioactivity Releases Radioactivity Releases XV-32 XV-32 1.3.2

1. 3.2 Meteorology and Dose Rates Meteorology XV-33 XV-33 2.0 Loss-of-Coolant Accident Loss-of-Coolant XV-33 XV-33 2.1 Introduction Introduction XV-33 XV-33 2.2 Input Input to Analysis XV-35 XV-35 2.2.1 Operational and ECCS Input Operational Input Parameters Parameters XV-35 XV-35 2.2.2 Single Failure Failure Study on ECCS ECCS Manually-Controlled Manually-Controlled Electrically-Operated Valves Electrically-Operated Valves XV-35 XV-35 2.2.3 Single Failure Basis XV-35 XV-35 2.2.4 Pipe Whip Basis XV-36 XV-36 2.3 This section deleted XV-36 XV-36 2.4 Appendix K Appendix K LOCA Performance Performance Analysis XV-36 XV-36 2.4.1 Computer Codes XV-36 XV-36 2.4.2 Description Description of Model Model Changes Changes XV-37 XV-37 2.4.3 Analysis Procedure Procedure XV-37 XV-37 2.4.3.1 BWR/2 BWR/2 Generic Generic Analysis XV-37 XV-37 2.4.3.2 Unit I-Specific 1-Specific Analysis Analysis Break Spectrum Spectrum Evaluation XV-38 XV-38 2.4.4 Analysis Results Results XV-38 XV-38 3.0 Refueling Accident Refueling Accident XV-40 XV-40 3.1 Identification of Causes Identification XV-40 XV-40 3.2 Accident Accident Analysis XV-41 XV-41 3.3 Radiological Effects Radiological XV-44 XV-44 3.3.1 Fission Fission Product Product Releases Releases XV-44 XV-44 3.3.2 Meteorology Meteorology and Dose Rates Rates XV-45 XV-4S 4.0 4.0 Control Rod Drop Accident Accident XV-46 XV-46 4.1 Identification Identification of Causes XV-46 XV-46 4.2 Accident Accident Analysis Analysis XV-46 XV-46 4.3 Designed Safeguards Designed Safeguards XV-47 XV-47 4.4 Procedural Procedural Safeguards Safeguards XV-48 XV-48 4.5 Radiological Radiological Effects XV-48 XV-48

• Fission Product Releases 4.5.1 4.5.1 Releases XV-48a XV-48a 4.5.2 Meteorology Meteorology and Dose Rates XV-49 XV-49 UFSAR Revision 21 Revision 21 xxv XXV October 2009

Nine Mile Point Nine Mile Unit 1 UFSAR Point Unit Section Section 5.0 5.0 Title TABLE Containment CONTENTS (Cont'd.) OF CONTENTS TABLE OF (Cont'd.) Design Basis Accident Containment Design Accident Page XV-50 XV-50 0 5.1 5.1 Original Recirculation Line Rupture Original Recirculation Analysis -- With Analysis With Core Spray XV-50 XV-50 5.1.1 5.1.1 Purpose Purpose XV-50 XV-50 5.1. 2 5.1.2 Analysis Method and Assumptions Analysis Method Assumptions XV-51 XV-51 5.1. 3 5.1.3 Core Core Heat Buildup XV-51 XV-51 5.1.4 5.1.4 Spray System Core Spray Core System XV-52 XV-52 5.1. 5 5.1.5 Pressure Immediately Containment Pressure Containment Immediately Following Blowdown Following Blowdown XV-53 XV-53 5.1. 6 5.1.6 Containment Spray Containment Spray XV-54 XV-54 5.1. 7 5.1.7 Blowdown Effects Blowdown Core Components Effects on Core Components XV-55 XV-55 5.1. 8 5.1.8 Radiological Effects Radiological Effects XV-56 XV-56 5.1.8.1 5.1.8.1 Fission Product Releases Fission Releases XV-56 XV-56 5.1.8.2 5.1.8.2 Meteorology and Dose Meteorology Dose Rates Rates XV-58 XV-58 5.2 Original Containment Design Original Containment Basis Design Basis Accident Analysis - Without Accident Analysis Without Core Core Spray XV-59 XV-59 5.2.1 5.2.1 Purpose Purpose XV-59 XV-59 5.2.2 5.2.2 5.2.3 5.2.3 Core Heatup Containment Response Containment Response XV-59 XV-60 0 5.3 5.3 Design Basis Reconstitution Suppression Suppression Chamber Analysis Chamber Heatup Analysis XV-61 XV-61 5.3.1 5.3.1 Introduction XV-61 XV-61 5.3.2 5.3.2 Input to Analysis Analysis XV-61a XV-61a 5.3.3 5.3.3 DBR Suppression Chamber Heatup Suppression Chamber Analysis Analysis XV-61a XV-61a 5.3.3.1 5.3 .3.1 Codes Computer Codes XV-62 XV-62 5.3 .3.2 5.3.3.2 Methods Analysis Methods XV-62 5.3 .3.3 5.3.3.3 Results for Containment Analysis Results Containment Spray Design Basis Assumptions XV-64 XV- 64 5.3.3.4 5.3.3.4 Analysis Results EOP Results for EOP Operation Assumptions Operation Assumptions XV-65 XV-65 5.3.4 5.3.4 Conclusions Conclusions XV-65a XV-65a 6.0 6.0 New Fuel Bundle Error Bundle Loading Error Analysis Analysis XV-66 XV-66 6.1 6.1 Identification Causes Identification of Causes XV-66 6.2 6.2 Analysis Accident Analysis XV-67 XV-67 6.3 6.3 Requirements Safety Requirements XV-67 XV-67 7.0 7.0 Meteorological Models Used in Meteorological in Analyses Accident Analyses XV-68 7.1 7.1 Introduction XV -68 XV-68 0 UFSAR Revision Revision 21 21 xxvi XXVl October October 2009

Mile Point Unit 1 UFSAR Nine Mile

• Section Section Title Title TABLE OF CONTENTS (Cont'd.)

Atmospheric Dispersion Factor (Cont'd.) Page 7.2 Atmospheric Factor Calculations Calculations XV-68 XV-68 7.2.1 Offsite Offsite - EAB and LPZ XV-69 XV-69 7.2.2 Control Control Room and Technical Technical Support Support Center Center (Excluding MSLB)MSLB) XV-69 XV-69 7.2.3 Control Control Room - MSLB Puff Release Puff Release XV-70 XV-70 7.3 Summary of Results Summary XV-70 XV-70 7.4 7.4 Exfiltration Exfiltration XV-70 XV-70 7.5 Secondary Containment Secondary Containment Drawdown Drawdown XV-76 XV-76 7.5.1 Introduction Introduction XV-76 XV-76 7.5.2 Analysis Analysis XV-76 XV-76 7.5.3 Results Results XV-77 XV-77

• UFSAR Revision 21 21 xxvia xxvia October 2009

Nine Mile Mile Point Unit unit 1 UFSAR 0 THIS PAGE THIS PAGE INTENTIONALLY BLANK INTENTIONALLY BLANK UFSAR Revision UFSAR 18 Revision 18 xxvib xxvib October 2003

Nine Nine Mile Point Unit 1I UFSAR

• Section Title TABLE OF CONTENTS (Cont'd.)

CONTENTS (Cont'd.) Page 5.2 5.2 NUREG-0737, Supplement NUREG-0737, Supplement 1, 1, Section 4.1.b 4.1.b XVIII-13 5.2.1 5.2.1 Convenient Location Location XVIII-13 XVIII- 13 5.2.2 5.2.2 Continuous Display Continuous XVIII-13 XVIII- 13 5.3 5.3 NUREG-0737, Supplement NUREG-0737, Supplement 1, 1, Section 4.l.c c 4.1. XVIII-13 5.3.1 5.3.1 Procedures Procedures and Training XVIII-13 5.3.2 5.3.2 Isolation of SPDS from from Safety-Related Safety-Related SystemsSystems XVIII-13 5.4 5.4 NUREG-0737, Supplement 1, NUREG-0737, Supplement 1, Section 4.l.e e 4.1. XVIII-14 5.4.1 5.4.1 Incorporation Incorporation of Accepted Human Accepted Human Factors Engineering Factors Engineering Principles Principles XVIII-14 5.4.2 5.4.2 Information Information Can Be Readily Readily Perceived and Perceived Comprehended and Comprehended XVIII-14 5.5 5.5 NUREG-0737, Supplement 1, NUREG-0737, Supplement 1, Section 4.l.f, 4.1.f, Sufficient Information Sufficient XVIII-15 XVIII-1S

• 6.0 6.0 Procedures Procedures XVIII-15 6.1 6.1 Operating Operating Procedures Procedures XVIII-15 XVIII-1S 6.2 6.2 Surveillance Procedures Surveillance Procedures XVIII-1S XVIII-15 7.0 7.0 References References XVIII- 16 XVIII-16 APPENDIX A Unused Unused APPENDIX B NINE MILE MILE POINT NUCLEAR STATION, LLC, NUCLEAR STATION, LLC, QUALITY ASSURANCE PROGRAM PROGRAM TOPICAL REPORT, REPORT, NINE MILE POINT NUCLEAR STATION NUCLEAR STATION UNITS 1 AND 22 OPERATIONS OPERATIONS PHASE PHASE APPENDIX C LICENSE RENEWAL SUPPLEMENT LICENSE RENEWAL SUPPLEMENT - AGING MANAGEMENT AGING MANAGEMENT PROGRAMS AND TIME-LIMITED TIME-LIMITED AGING ANALYSES ANALYSES
• UFSAR Revision 20 20 xxxiii xxxiii October 2007 2007

Nine Mile Nine Mile Point Point Unit Unit 1 UFSAR UFSAR Table Table Number Number Title Title LIST OF LIST OF TABLES TABLES 1-1 I-1 COMPARISON TO STANDARDS COMPARISON STANDARDS -- HISTORICAL HISTORICAL (PROVIDED (PROVIDED WITH APPLICATION WITH APPLICATION TO CONVERT CONVERT TO FULL-TERM FULL-TERM OPERATING LICENSE) OPERATING LICENSE) 1-2 ACRONYMS USED ABBREVIATIONS AND ACRONYMS ABBREVIATIONS USED ININ UFSAR 11-1 1980 POPULATION 1980 POPULATION AND POPULATION POPULATION DENSITY DENSITY FOR FOR TOWNS TOWNS AND CITIES WITHIN 12 MILES MILES OF NINE MILE NINE MILE POINT - UNIT 1 11-2 CITIES WITHIN CITIES WITHIN AA 50-MILE 50-MILE RADIUS OF THE THE STATION WITH POPULATIONS WITH POPULATIONS OVER 10,000 10,000 11-3 REGIONAL AGRICULTURAL AGRICULTURAL USEUSE 11-4 REGIONAL AGRICULTURAL AGRICULTURAL STATISTICS STATISTICS -- CATTLE AND MILK PRODUCTION 11-5 INDUSTRIAL FIRMS WITHIN INDUSTRIAL WITHIN 88 KM (5 (5 MI) UNIT 1 MI) OF UNIT 11-6 PUBLIC UTILITIES IN IN OSWEGO COUNTY 11-7 PUBLIC WATER SUPPLY LOCATIONS WITHIN SUPPLY DATA FOR LOCATIONS WITHIN AN APPROXIMATE APPROXIMATE 30-MILE 30-MILE RADIUS RADIUS .11-8 11-8 RECREATIONAL AREAS IN THE REGION RECREATIONAL 11-9 11-9 SOURCES OF TOXIC CHEMICALS WITHIN WITHIN 8 KM (5 (5 MI) OF OF UNIT 1 SITE SITE 11-10 II-10 PREDICTED VAPOR CONCENTRATION CONCENTRATION IN IN THE UNIT 1 CONTROL ROOM V-1 V-i REACTOR COOLANT SYSTEM DATA V-2 OPERATING CYCLES AND TRANSIENT ANALYSIS RESULTS RESULTS V-3 FATIGUE RESISTANCE RESISTANCE ANALYSIS ANALYSIS V-4 CODES FOR SYSTEMS FROM REACTOR VESSEL CONNECTION TO SECOND ISOLATION ISOLATION VALVE UFSAR Revision 21 21 xxxiv xxxiv October October 2009

Nine Mile Nine Mile Point Unit 1 UFSAR

• Table Table Number Title Title LIST OF TABLES (Cont'd.)

(Cont'd.) V-5 TIME TO AUTOMATIC AUTOMATIC BLOWDOWN BLOWDOWN VI-1 VI-l DRYWELL PENETRATIONS PENETRATIONS VI-2 SUPPRESSION SUPPRESSION CHAMBER CHAMBER PENETRATIONS PENETRATIONS VI-3a VI-3a REACTOR COOLANT COOLANT SYSTEM ISOLATION ISOLATION VALVES VALVES VI-3b VI-3b PRIMARY CONTAINMENT PRIMARY CONTAINMENT ISOLATION ISOLATION VALVES VALVES - LINES LINES ENTERING FREE SPACE OF THE CONTAINMENT ENTERING FREE CONTAINMENT VI-4 VI-4 SEISMIC SEISMIC DESIGN CRITERIA CRITERIA FOR ISOLATION VALVES VALVES VI-5 INITIAL TESTS PRIOR INITIAL PRIOR TO STATION STATION OPERATION VII-1 VII-l PERFORMANCE PERFORMANCE TESTS

• VIII-l VIII-1 VIII-2 VIII-2 ASSOCIATION ASSOCIATION BETWEEN EMERGENCY BETWEEN PRIMARY OPERATING PROCEDURES EMERGENCY OPERATING SAFETY FUNCTIONS AND PRIMARY SAFETY PROCEDURES LIST OF EOP KEY PARAMETERS PARAMETERS VIII-3 VIII-3 TYPE AND INSTRUMENT TYPE INSTRUMENT CATEGORY CATEGORY FOR UNIT 1 RG 1.97 VARIABLES VARIABLES VIII-4 VIII-4 PROTECTIVE PROTECTIVE SYSTEM FUNCTION SYSTEM FUNCTION VIII-5 VIII-5 NON-TECHNICAL NON-TECHNICAL SPECIFICATION INSTRUMENTATION THAT SPECIFICATION INSTRUMENTATION INITIATES CONTROL INITIATES CONTROL ROD WITHDRAWAL WITHDRAWAL BLOCK IX-l IX-1 MAGNITUDE MAGNITUDE AND DUTYDUTY CYCLE CYCLE OF MAJOR MAJOR STATION STATION BATTERY LOADS LOADS XII-1 XII-1 FLOWS AND ACTIVITIES ACTIVITIES OF MAJOR MAJOR SOURCES SOURCES OF GASEOUS GASEOUS ACTIVITY XII-2 XII-2 QUANTITIES QUANTITIES AND ACTIVITIES OF LIQUID RADIOACTIVE RADIOACTIVE WASTES WASTES XII-3 XII-3 ANNUAL SOLID WASTE ACCUMULATION ACCUMULATION AND ACTIVITY Revision 21 UFSAR Revision 21 xxxv XXXV 2009 October 2009

Nine Mile Point Unit Unit 1 UFSAR Table Table Number Number Title Title LIST OF TABLES TABLES (Cont'd.) (Cont'd.) XII-4 XII-4 LIQUID WASTE WASTE DISPOSAL DISPOSAL SYSTEM SYSTEM MAJOR COMPONENTS COMPONENTS XII-5 XII-5 SOLID WASTE DISPOSAL DISPOSAL SYSTEM SYSTEM MAJOR COMPONENTS COMPONENTS XII-6 XII-6 OCCUPANCY OCCUPANCY TIMES TIMES XII-7 GAMMA ENERGY GROUPS GROUPS XII-8 XII-8 AREA RADIATION AREA RADIATION MONITOR DETECTOR DETECTOR LOCATIONS XIII-1 XIII-l ANSI STANDARD CROSS-REFERENCE CROSS-REFERENCE UNIT 1 XIII-2 XIII-2 MINIMUM SHIFT CREW COMPOSITION MINIMUM XV-1 XV-1 TABLE DELETED XV-2 XV-2 TRIP POINTS FOR PROTECTIVE PROTECTIVE FUNCTIONS FUNCTIONS XV-3 thru thru TABLES DELETED XV-4 XV-4 XV-5 XV- 5 BLOWDOWN RATES BLOWDOWN RATES XV-6 XV-6 REACTOR COOLANT CONCENTRATIONS REACTOR CONCENTRATIONS (pCi/gm) XV- 7 XV-7 TABLE DELETED TABLE XV-7a XV-7a MSLB ACCIDENT ACCIDENT ANALYSIS INPUTS INPUTS AND ASSUMPTIONS ASSUMPTIONS XV-7b XV-7b MSLB ACCIDENT ACCIDENT RELEASE RATES RELEASE RATES XV-XV-88 MAIN STEAM LINE BREAK BREAK ACCIDENT DOSES DOSES XV-9 XV-9 SIGNIFICANT INPUT SIGNIFICANT INPUT PARAMETERS PARAMETERS TO THE THE LOSS-OF-COOLANT ACCIDENT ANALYSIS LOSS-OF-COOLANT ANALYSIS XV-XV-9a9a CORE SPRAY SYSTEM FLOW PERFORMANCE CORE PERFORMANCE ASSUMED ININ ANALYSIS LOCA ANALYSIS XV-10 XV-10 ECCS SINGLE VALVE FAILURE FAILURE ANALYSIS ANALYSIS Revision 21 UFSAR Revision 21 xxxvi xxxvi October 2009

Nine Mile Point Unit 1 UFSAR Nine LIST OF TABLES TABLES (Cont'd.) (Cont'd.) Table Table Number Number Title Title XV-II XV-11 SINGLE FAILURES FAILURES CONSIDERED IN LOCA ANALYSIS ANALYSIS XV-12 thru XV-12 thru TABLES DELETED XV-21 XV-21 XV-21a XV-21a ANALYSIS ASSUMPTIONS FOR NINE MILE ANALYSIS ASSUMPTIONS MILE POINT 1 CALCULATIONS CALCULATIONS XV-22 XV-22 ACTIVITY RELEASED TO THE REACTOR ACTIVITY REACTOR BUILDING FOLLOWING THE FHA (CURIES) FOLLOWING (CURIES) XV-23 XV-23 UNIFORM UNFILTERED STACK DISCHARGE UNFILTERED STACK DISCHARGE RATES FROM 0 TO 2 HRHR AFTER AFTER THE FHA (CURIES/SEC) (CURIES/SEC) XV-24 XV-24 HANDLING ACCIDENT DOSES FUEL HANDLING DOSES XV-25 XV-25 FHA ANALYSIS FHA ANALYSIS INPUTS AND ASSUMPTIONS ASSUMPTIONS

• XV-26 XV-26 XV-27 XV-27 XV-28 XV-28 ACCIDENT ANALYSIS CRD ACCIDENT CRDA NOBLE ANALYSIS INPUTS NOBLE GAS RELEASE RELEASE HALOGEN RELEASE CRDA HALOGEN RELEASE INPUTS AND ASSUMPTIONS ASSUMPTIONS XV-29 XV-29 CONTROL ROD DROP ACCIDENT DOSES CONTROL DOSES XV-29a XV-29a WETTING OF FUEL CLADDING WETTING CLADDING BY CORE SPRAY XV-29b XV-29b POST-LOCA AIRBORNE DRYWELLFISSION PRODUCT DRYWELL FISSION PRODUCT INVENTORY (CURIES)

INVENTORY (CURIES) XV-29c XV-29c POST-LOCA REACTOR BUILDING POST-LOCA BUILDING FISSION FISSION PRODUCT PRODUCT INVENTORY (CURIES) INVENTORY (CURIES) XV-29d XV-29d POST-LOCA POST-LOCA DISCHARGE RATES (CURIES/SEC) (CURIES/SEC) XV-30 XV-30 CORE FISSION PRODUCT INVENTORY INVENTORY XV-31 XV-31 LOCA ANALYSIS INPUTS AND ASSUMPTIONS LOCA ASSUMPTIONS

• XV-32 XV-32 UFSAR Revision LOSS-OF-COOLANT ACCIDENT LOSS-OF-COOLANT 21 Revision 21 ACCIDENT DOSES xxxvi i xxxvii DOSES 2009 October 2009

Nine Mile Nine Mile Point Unit 1 UFSAR Table Table Number Title Title LIST OF TABLES (Cont'd.) (Cont'd.) XV-32a XV-32a SIGNIFICANT INPUT PARAMETERS SIGNIFICANT INPUT PARAMETERS TO THE DBR DBR CONTAINMENT CONTAINMENT SUPPRESSION CHAMBER HEATUP ANALYSIS SUPPRESSION CHAMBER ANALYSIS XV-33 TABLE DELETED XV-34 XV-34 TABLE DELETED XV-34a XV-34a RELEASE/INTAKE ELEVATIONS RELEASE/INTAKE ELEVATIONS XV-34b XV-34b RELEASE/INTAKE DISTANCE AND DIRECTIONS RELEASE/INTAKE DIRECTIONS XV-35 XV-35 TABLE DELETED TABLE XV-35a VALUES FOR THE CONTROL X/Q VALUES CONTROL ROOM XV-35b XV-35b VALUES FOR THE TECHNICAL X/Q VALUES TECHNICAL SUPPORT SUPPORT CENTER CENTER XV-35c XV-35c XV-35d XV-35d XV-36 OFFSITE OFFSITE X/Q OFFSITE VALUES FOR GROUND-LEVEL X/QVALUES GROUND-LEVEL RELEASES OFFSITE X/Q VALUES FOR ELEVATED REACTOR BUILDING RELEASES ELEVATED RELEASES BUILDING LEAKAGE PATHS PATHS RELEASES

• XVI-1 XVI-I CODE CALCULATION CALCULATION

SUMMARY

XVI-2 XVI-2 STEADY-STATE - (100% STEADY-STATE (100% FULL POWER OPERATION) POWER NORMAL OPERATION) PERTINENT STRESSES OR STRESS INTENSITIES PERTINENT INTENSITIES XVI-3 LIST OF REACTIONS FOR REACTOR REACTOR VESSEL NOZZLES NOZZLES XVI-4 XVI-4 EFFECT OF VALUE OF INITIAL INITIAL FAILURE FAILURE PROBABILITY XVI-5 XVI-5 TRANSIENT EVENT FOR REACTOR PRESSURE SINGLE TRANSIENT VESSEL VESSEL XVI-6 POSTULATED EVENTS POSTULATED EVENTS XVI-7 MAXIMUM STRAINS FROM POSTULATED MAXIMUM STRAINS EVENTS POSTULATED EVENTS XVI-8 STRUCTURE ANALYSIS CORE STRUCTURE ANALYSIS RECIRCULATION RECIRCULATION LINE BREAK UFSAR Revision 21 21 xxxviia xxxviia October 2009 October 2009

Nine Mile Nine Mile Point Point Unit Unit 1 UFSAR

• Table Table Number Number Title Title LIST OF TABLES LIST TABLES (Cont'd.)

(Cont'd.) XVI-9 XVI-9 CORE STRUCTURE CORE STRUCTURE ANALYSIS ANALYSIS STEAM LINE BREAK BREAK XVI-9a XVI-9a CORE SHROUD CORE SHROUD REPAIR REPAIR DESIGN SUPPORTING DOCUMENTATION DOCUMENTATION XVI-I0 XVI-10 DRYWELL JET DRYWELL JET AND AND MISSILE MISSILE HAZARD HAZARD ANALYSIS ANALYSIS DATA DATA XVI-II XVI-1 DRYWELL MISSILE HAZARD DRYWELL JET AND MISSILE HAZARD ANALYSIS ANALYSIS RESULTS RESULTS XVI-12 STRESS DUE TO DRYWELL STRESS DRYWELL FLOODING XVI-13 ALLOWABLE ALLOWABLE WELD SHEARSHEAR STRESS STRESS XVI-14 LEAK RATE TEST LEAK TEST RESULTS RESULTS XVI-IS XVI-15 OVERPRESSURE TEST--PLATE OVERPRESSURE TEST--PLATE STRESSES STRESSES

• XVI-16 STRESS

SUMMARY

STRESS

SUMMARY

• UFSAR Revision 21 21 xxxviib xxxviib October 2009

Unit 1 UFSAR Nine Mile Point unit LIST OF TABLES (Cont'd.) (Cont'd.) Table Number Nimber Tjtle Title XVI-17 XVI-17 HEAT TRANSFER HEAT TRANSFER COEFFICIENTS COEFFICIENTS AS A A FUNCTION FUNCTION OF DROP DROP DIAMETER DIAMETER XVI-18 HEAT TRANSFER HEAT TRANSFER COEFFICIENT COEFFICIENT AS A FUNCTION FUNCTION OF OF PRESSURE PRESSURE XVI-19 XVI-19 RELATIONSHIP RELATIONSHIP BETWEEN PARTICLE SIZE SIZE AND TYPE OF OF SPRAY PATTERN SPRAY XVI-20 ALLOWABLE STRESSES FOR FLOOR SLABS, BEAMS, XVI-20 ALLOWABLE STRESSES FOR FLOOR SLABS, BEAMS, COLUMNS, WALLS, FOUNDATIONS, COLUMNS, WALLS, FOUNDATIONS, ETC. ETC. XVI-21 ALLOWABLE STRESSES FOR STRUCTURAL STRUCTURAL STEEL XVI-22 XVI-22 ALLOWABLE STRESSES - REACTOR VESSELVESSEL CONCRETE CONCRETE PEDESTAL PEDESTAL XVI-23 XVI-23 DRYWELL - ANALYZED DRYWELL ANALYZED DESIGN LOAD COMBINATIONS COMBINATIONS XVI-24 XVI-24 SUPPRESSION CHAMBER - ANALYZED SUPPRESSION CHAMBER ANALYZED DESIGN LOAD COMBINATIONS COMBINATIONS XVI-25 ACI CODE 505 ALLOWABLE STRESSES AND ACTUAL ACTUAL VENTILATION STACK CONCRETE VENTILATION STRESSES FOR CONCRETE XVI-26 ALLOWABLE ALLOWABLE STRESSES FOR CONCRETE SLABS, SLABS, WALLS, WALLS, BEAMS, STRUCTURAL BEAMS, WALLS STEEL, AND CONCRETE BLOCK WALLS STRUCTURAL STEEL, XVI-27 XVI-27 SYSTEM LOAD COMBINATIONS COMBINATIONS XVI-28 XVI-28 HIGH-ENERGY SYSTEMS - HIGH-ENERGY SYSTEMS INSIDE CONTAINMENT CONTAINMENT XVI-29 HIGH-ENERGY SYSTEMS HIGH-ENERGY SYSTEMS - OUTSIDE OUTSIDE CONTAINMENT CONTAINMENT XVI-30 SYSTEMS WHICH WHICH MAY BE AFFECTED BY PIPE WHIPWHIP XVI-31 CAPABILITY TO RESIST WIND PRESSURE AND WIND VELOCITY XVII-l XVII-I DISPERSION AND ASSOCIATED ASSOCIATED METEOROLOGICAL METEOROLOGICAL PARAMETERS PARAMETERS XVII-2 XVII-2 RELATION OF SATELLITE AND NINE MILE POINT WINDS WINDS XVII-J XVII-3 FREQUENCY OF OCCURRENCE FREQUENCY OCCURRENCE OF LAPSE LAPSE RATES - 1963 AND 1964 1964 16 UFSAR Revision 16 xxxviii xxxviii November 1999

Nine Mile Point Unit Unit 1 UFSAR

• Figure Number Number Title Title LIST OF FIGURES FIGURES (Cont'd.)

(Cont'd.) X-2 REACTOR CLEANUP CLEANUP SYSTEM SYSTEM X-3 CONTROL CONTROL ROD DRIVE DRIVE HYDRAULIC HYDRAULIC SYSTEM SYSTEM X-4 REACTOR BUILDING BUILDING CLOSED LOOP COOLING SYSTEM X-5 TURBINE BUILDING BUILDING CLOSED LOOP COOLING SYSTEM X-6 SERVICE WATER SYSTEM SYSTEM X-7 FIGURE DELETED FIGURE X-8 SPENT FUEL STORAGE FILTERING AND COOLING STORAGE POOL FILTERING SYSTEM SYSTEM X-9 BREATHING, INSTRUMENT, AND SERVICE BREATHING, INSTRUMENT, SERVICE AIR

• X-10 X-10 X-11 XI-1 XI-l REACTOR REFUELING STEAM REFUELING SYSTEM PICTORIAL CASK DROP PROTECTION PROTECTION SYSTEM FLOW AND REHEATER STEAM FLOW SYSTEM REHEATER VENTILATION VENTILATION SYSTEM SYSTEM XI-2 EXTRACTION EXTRACTION STEAM FLOWFLOW XI-3 MAIN CONDENSER CONDENSER AIR REMOVAL AND OFFGAS SYSTEM XI-4 CIRCULATING WATER SYSTEM CIRCULATING WATER SYSTEM XI-5 CONDENSATE FLOW CONDENSATE FLOW XI-6 CONDENSATE TRANSFER CONDENSATE SYSTEM TRANSFER SYSTEM XI-7 FEEDWATER FLOW SYSTEM FEEDWATER SYSTEM XII-1 RADIOACTIVE DISPOSAL SYSTEM RADIOACTIVE WASTE DISPOSAL SYSTEM XIII-1 SENIOR LEVEL STATION SENIOR STATION MANAGEMENT MANAGEMENT ORGANIZATION CHART CHART
• XIII-2 XIII-2 ENGINEERING ENGINEERING SERVICES Revision 21 UFSAR Revision 21 SERVICES ORGANIZATION xlvii xlvii CHART ORGANIZATION CHART October 2009

Mile Point Nine Mile Point Unit Unit 1 UFSAR UFSAR FIGURES LIST OF FIGURES (Cont'd.) (Cont'd.) Figure Figure Number Number Title Title XIII-3 XIII-3 QUALITY ASSURANCE QUALITY ORGANIZATION ASSURANCE ORGANIZATION XIII-3a XIII-3a NUCLEAR SAFETY NUCLEAR SAFETY & & SECURITY SECURITY ORGANIZATION ORGANIZATION XIII-4 XIII-4 NINE MILE NINE MILE POINT POINT NUCLEAR NUCLEAR STATION STATION ORGANIZATION ORGANIZATION CHART CHART XIII-4a XIII-4a NINE MILE POINT NUCLEAR NINE NUCLEAR STATION STATION ORGANIZATION ORGANIZATION CHART CHART XIII-4b XIII-4b NINE MILE NINE MILE POINT NUCLEAR NUCLEAR STATION STATION ORGANIZATION ORGANIZATION CHART CHART XIII-4c XIII-4c NINE MILE POINT NUCLEAR NINE NUCLEAR STATION ORGANIZATION STATION ORGANIZATION CHART CHART XIII-5 XIII-5 SAFETY ORGANIZATION ORGANIZATION XV-1 XV-I XV-2 XV-2 XV-3 XV-3 STATION STATION TRANSIENT TRANSIENT DIAGRAM FIGURE DELETED PLANT RESPONSE DIAGRAM RESPONSE TO LOSS OF 100°F FEEDWATER HEATING 100OF FEEDWATER UFSAR Revision 21 21 xlviia xlviia October 2009 October

Nine Mile Nine Mile Point Point Unit Unit 1 UFSAR UFSAR

• Figure Figure Number Number Title Title LIST OF LIST OF FIGURES FIGURES (Cont'd.)

(Cont'd.) XV-56E XV-56E LOSS-OF-COOLANT ACCIDENT LOSS-OF-COOLANT ACCIDENT DRYWELL DRYWELL PRESSURE PRESSURE XV-56F XV-56F LOSS-OF-COOLANT ACCIDENT LOSS-OF-COOLANT ACCIDENT SUPPRESSION SUPPRESSION CHAMBER CHAMBER PRESSURE PRESSURE XV-56G XV-56G LOSS-OF-COOLANT ACCIDENT LOSS-OF-COOLANT CONTAINMENT TEMPERATURE ACCIDENT CONTAINMENT TEMPERATURE

                - WITH CORE CORE SPRAY XV-57 XV-57          CONTAINMENT DESIGN CONTAINMENT    DESIGN BASIS BASIS CLAD TEMPERATURE TEMPERATURE RESPONSE - WITHOUT RESPONSE      WITHOUT CORE SPRAY XV-58 XV-58          CONTAINMENT DESIGN CONTAINMENT    DESIGN BASIS BASIS METAL-WATER METAL-WATER REACTION XV-59 XV-59          CONTAINMENT DESIGN CONTAINMENT    DESIGN BASIS BASIS CLAD PERFORATION    WITHOUT PERFORATION WITHOUT CORE SPRAY CORE XV-60 XV-60          CONTAINMENT CONTAINMENT DESIGN BASISBASIS CONTAINMENT   TEMPERATURE CONTAINMENT TEMPERATURE
                - WITHOUT CORE CORE SPRAY XV-60A '

XV-60A DBR ANALYSIS SUPPRESSION POOL AND WETWELL SUPPRESSION POOL WETWELL AIRSPACE TEMPERATURE RESPONSE - CONTAINMENT TEMPERATURE RESPONSE CONTAINMENT SPRAY DESIGN BASIS ASSUMPTION DESIGN XV-60B XV-60B DBR ANALYSIS SUPPRESSION POOL AND WETWELL SUPPRESSION POOL WETWELL AIRSPACE TEMPERATURE RESPONSE - EOP OPERATION TEMPERATURE RESPONSE ASSUMPTIONS XV-61 REACTOR BUILDING MODEL REACTOR MODEL XV-62 EXFILTRATION EXFILTRATION VS.VS. WIND SPEED - NORTHERLY WIND XV-63 REACTOR BUILDING DIFFERENTIAL PRESSURE REACTOR PRESSURE XV-64 EXFILTRATION EXFILTRATION VS.VS. WIND SPEED - SOUTHERLY SOUTHERLY WIND XV-65 REACTOR BUILDING - ISOMETRIC ISOMETRIC XV-66 REACTOR BUILDING - CORNER SECTIONS XV-67 REACTOR BUILDING - ROOF SECTIONS SECTIONS UFSAR UFSAR Revision 21 21 xlix xlix October 2009 2009

Nine Mile Point Point Unit unit 1 UFSAR Figure Number Number Title Title LIST OF FIGURES (Cont'd.) (Cont'd.) 0 XV-68 XV-68 REACTOR BUILDING BUILDING - PANEL PANEL TO CONCRETE SECTIONS CONCRETE SECTIONS XV-69 XV-69 REACTOR BUILDING BUILDING - EXPANSION EXPANSION JOINT SECTIONS SECTIONS XV-70 XV-70 REACTOR BUILDING EXFILTRATION EXFILTRATION - NORTHERLY NORTHERLY WIND XV-71 XV-71 REACTOR BUILDING EXFILTRATION EXFILTRATION - SOUTHERLY SOUTHERLY WIND XV-72 XV-72 REACTOR BUILDING DIFFERENTIAL PRESSURE BUILDING DIFFERENTIAL PRESSURE XV-73 XV-73 REACTOR BUILDING PRESSURE REACTOR VS. PRESSURE VS. TIME BY REACTOR TIME REACTOR BUILDING ELEVATION BUILDING XV-74 XV-74 REACTOR BUILDING PRESSURE REACTOR VS. TIME PRESSURE VS. TIME BY REACTOR REACTOR BUILDING ELEVATION BUILDING ELEVATION (FOCUSED (FOCUSED ON THE INITIAL 2.S2.5 HR) HR) XVI-1 XVI-' XVI-2 XVI-2 REACTOR SUPPORT DYNAMIC REACTOR MOMENT MOMENT REACTOR VESSEL SEISMIC ANALYSIS OF REACTOR LUMPED MASS REPRESENTATION LUMPED REPRESENTATION VESSEL GEOMETRIC AND DYNAMIC ANALYSIS ANALYSIS -- ELEVATION ELEVATION VS.VS. XVI-3 XVI-3 REACTOR SUPPORT DYNAMIC DYNAMIC ANALYSIS ANALYSIS -- ELEVATION ELEVATION VS.VS. SHEAR SHEAR XVI-4 XVI-4 REACTOR SUPPORT DYNAMIC REACTOR DYNAMIC ANALYSIS ANALYSIS - ELEVATION VS. ELEVATION VS. DEFLECTION XVI-5 XVI-S REACTOR SUPPORT DYNAMIC DYNAMIC ANALYSIS ANALYSIS - ELEVATION VS. ELEVATION VS. ACCELERATION ACCELERATION XVI-6 thru thru FIGURES DELETED FIGURES XVI-8 XVI-8 XVI-9 XVI-9 SUPPORT STRUCTURE STRESS REACTOR VESSEL SUPPORT STRESS

SUMMARY

XVI-10 XVI-10 THERMAL ANALYSIS THERMAL ANALYSIS XVI-11 XVI-II FAILURE PROBABILITY DENSITY FUNCTION PROBABILITY DENSITY UFSAR Revision 21 21 1 October 2009

Mile Point Unit 1 UFSAR Nine Mile

• Figure Figure Number Title Title LIST OF FIGURES FIGURES (Cont'd.)

(Cont'd.) XVI-12 ADDITION STRAINS STRAINS PAST 4% REQUIRED TO EXCEED EXCEED DEFINED SAFETY MARGIN DEFINED MARGIN XVI-12a SHROUD WELDS WELDS XVI-12b CORE SHROUD CORE SHROUD STABILIZERS STABILIZERS XVI-12c CORE SHROUD CORE SHROUD WELDS WELDS XVI-12d XVI-12d V9/VI0 V9/V10 VERTICAL CLAMP ASSEMBLY VERTICAL WELD CLAMP ASSEMBLY XVI-13 LOSS-OF-COOLANT ACCIDENT - CONTAINMENT LOSS-OF-COOLANT CONTAINMENT PRESSURE NO CORE OR CONTAINMENT CONTAINMENT SPRAYS XVI-14 FIGURE DELETED FIGURE XVI-IS XVI-15 DRYWELL TO CONCRETE GAP CONCRETE AIR GAP XVI-16 TYPICAL PENETRATIONS TYPICAL PENETRATIONS

• Revision 21 UFSAR Revision 21 la la October 2009

Nine Nine Mile Unit 11 UFSAR Mile Point Unit INTENTIONALLY BLANK THIS PAGE INTENTIONALLY 21 Revision 21 UFSAR Revision lb lb October 2009 October 2009

Nine Mile Mile Point Unit 1 UFSAR

• normal auxiliary refueling Section auxiliary cooling refueling is Section X-A..

described described in heat in main in the event condenser while main condenser cooling means during is the shutdown V-E, is event the reactor reactor is still while still during shutdown shutdown cooling system described X-A. A redundant redundant emergency in Section V-E, is isolated under under pressure. shutdown and described in emergency cooling system, remove decay is provided to remove isolated from the decay Additional pressure. Additional in cooling capability cooling capability is available from the is also available high-pressure injection (HPCI) high-pressure coolant injection (HPCI) system system and the fire fire protection system. protection system. Redundant Redundant and independent independent core spray systems are provided to cool the core in in the event of a accident (LOCA). loss-of-coolant accident loss-of-coolant (LOCA). Automatic Automatic depressurization is depressurization included to rapidly reduce is included reduce assist with core pressure to assist pressure core spray operation (see operation (see Section VII-A). Section VII-A) . Operation Operation of the core core spray spray system system assures assures that any any metal-water metal-water reaction following a postulated postulated LOCA will will be limited limited to less than 1 percent percent of the Zircaloy Zircaloy clad. clad.

77. . Reactivity Reactivity shutdown capability capability is is provided provided to make and adequately subcritical, hold the core adequately control rod subcritical, by control action, from any point in action, in the operating cycle and at operating cycle at temperature down to room temperature, any temperature temperature, assuming that anyone any one control rod is is fully withdrawn and unavailable for use.

unavailable use. capability is This capability is demonstrated demonstrated in in Section IV-B. Section IV-B. A description of the movable physical description movable control control rods is is given given in Section IV-B. in Section IV-B. The control rod drive (CRD) (CRD) hydraulic system is hydraulic system is described described in in Section X-C.X-c. available to scram The force available control rod is scram a control is approximately 3000 lb at the beginning approximately beginning of a scramscram stroke. stroke. is This is well in excess in excess of the 440-lb force required in required in the event of fuel channel pinching of the channel pinching control rod blade during a LOCA, control LOCA, as discussed in in Section XV. Section XV. Even Even with scram accumulator failure, scram accumulator failure, a force of at least 1100 lb from reactor reactor pressure pressure acting alone is available with reactor pressures in is available in excess excess of of 800 psig. psig. 8.

8. reactivity shutdown capability Redundant reactivity capability is is provided reactivity control independent of normal reactivity independent provisions.

provisions. capability, as shown This system has the capability, in Section in UFSAR Revision 21 21 1-9 I-9 October 2009 2009

Nine Mile Nine Unit 11 UFSAR Point Unit Mile Point UFSAR 9.9. VII-C, to VII-C, condition at condition the bring the to bring at any control rod the control AA flow rod system reactor to the reactor time in any time contained in are contained results are results restrictor in flow restrictor in the in the the SRLR( in the cold shutdown to aa cold 2 . SRLR(2). steam line main steam the main shutdown independent of life, independent core life, the core capabilities. Cycle-specific system capabilities. Cycle-specific (MSL) limits line (MSL) limits of loss from coolant loss coolant vessel in reactor vessel the reactor from the the event in the of aa event of MSL MSL break (Section VII-F). break (Section VII-F) . UFSAR Revision 21 UFSAR 21 I-ga I-9a 2009 October 2009 October

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• UFSAR Revision 21 21 I-9b I-gb 2009 October 2009

Nine Mile Point unit Unit 1 UFSAR UFSAR 4.0 Reactor Vessel Reactor Vessel 1.

1. The reactor core vessel are designed core and vessel accommodate designed to accommodate tripping of the turbine turbine generator, generator, loss of power to the reactor recirculation system reactor recirculation system and other transients, transients, and maneuvers which maneuvers which can be expected expected without compromising safety and without safety damage.

without fuel damage. A bypass system having capacity of having aa design capacity of approximately approximately 40 percent percent of turbine steam flow for the throttle valves wide open (VWO) condition partially (VWO) condition mitigates the effects mitigates rejection. effects of sudden load rejection. An An bypass system test was performed actual bypass performed and the results indicated aa system bypass indicated capacity of about 2,500,000 bypass capacity lb/hr. lbJhr. This and other transients and maneuvers which been analyzed are detailed in have been Section xv. in section XV. 2.

2. Separate systems to prevent serious reactor coolant Separate coolant system (RCS) incorporated in overpressure are incorporated (RCS) overpressure the in the design.

design. include an overpressure These include scram, overpressure scram, solenoid-actuated relief solenoid-actuated valves, safety valves and the relief valves, turbine bypass system. bypass system. An analysis adequacy of analysis of the adequacy of RCS pressure relief devices devices is included in section V-C. is included in Section v-c. 3.

3. excursions which could Power excursions could result from any credible reactivity reactivity addition accident accident will not cause damage,damage, either by motion or rupture, either rupture, to the pressure vessel, or pressure vessel, or impair operation of required impair operation required safeguards systems.

safeguards systems. The magnitude of credible reactivity reactivity addition accidents addition accidents curtailed by control is curtailed is control rod velocity limiters (Section VII-D), by a control VII-D), control rod housing support structure procedural controls VII-E), and by procedural (Section VII-E), supplemented by a rod supplemented worth minimizer minimizer (RWM)(RWM) (Section VIII-C). VIII-C). Power excursion excursion analyses analyses for control rod included in accidents are included dropout accidents Section xv. in Section XV. 4.

4. reactor vessel will not be substantially The reactor pressurized until the vessel pressurized temperature is vessel wall temperature in is in excess of the nil ductility excess reference temperature ductility reference temperature 60 0 F.

(RTNDT) + 60°F. (RTNDT) initial RTNDT of the reactor The initial reactor vessel vessel 0 F. The change of RTNDT material is material greater than 40 is no greater 40°F. evaluated in exposure has been evaluated radiation exposure with radiation in accordance with Regulatory accordance Regulatory Guide (RG) Revision 2 (RG) 1.99 Revision to determine an adjusted reference temperature (ART) reference temperature (ART) for the most limiting vessel material. vessel material. Vessel material material surveillance samples

          . surveillance                         located within the reactor samples are located                        reactor vessel to permit periodic vessel                    periodic verification              material verification of material properties properties with exposure.exposure.

5.0 Containment containment 1.

1. including the drywell, containment, including The primary containment, drywell, suppression chamber, pressure suppression associated access chamber, and associated access UFSAR Revision 17 17 I-10 1-10 october 2001 October

Nine Mile Point Point Unit 11 UFSAR

• openings and penetrations, erected penetrations, is accommodate, without erected to accommodate, and temperatures is designed, designed, fabricated and without failure, the pressures temperatures resulting from or subsequent to the double-ended double-ended rupture rupture (DER) any coolant pipe within pressures equivalent failure of (DER) or equivalent within the drywell.

drywell. the of The primary primary containment containment is is designed accommodate the designed to accommodate pressures pressures following a LOCA, LOCA, including the generation generation of hydrogen from a metal-watermetal-water reaction. reaction. Pressure transients, including hydrogen transients, effects, are presented hydrogen effects, presented in Section xv. in Section XV. initial The initial NDTT for the primary containment system primary containment system is is about -20°F about -20°F and is is not expected increase during the expected to increase lifetime lifetime of the Station. These described in structures are described These structures in Sections VI-A, BBand Sections VI-A, and C. C. Additional details, related to particularly those related details, particularly to design and fabrication, are included design included inin Section XVI.XVI. 2.

2. Provisions are made for the removal of heat from Provisions from containment, for reasonable within the primary containment, within reasonable containment from fluid jets protection of the containment protection jets or or missiles, and such other missiles, other measures as may be necessary to maintain the integrity containment system as integrity of the containment as necessary following a LOCA.

long as necessary LOCA. Redundant containment Redundant systems, described in containment spray systems, in Section VII, Section VII, pump water from the suppression chamber suppression chamber independent heat exchangers through independent through nozzles, exchangers to spray nozzles, which discharge into the drywell and suppression which suppression chamber. chamber. sprayed into the drywell Water sprayed drywell isis returned gravity to the suppression chamber to complete by gravity complete the cooling cycle. cycle. Studies verify the performed to verify Studies performed containment system to withstand capability of the containment capability potential fluid jets potential summarized in jets and missiles are summarized in Section Section XVI. XVI. 3.

3. Provision Provision is periodic integrated is made for periodic leakage rate integrated leakage tests tests (ILRT)(ILRT) to be performed performed inin accordance accordance with 10CFR50 Appendix 10CFR50 Appendix J.J. Provision is Provision is also made for leak testing penetrations and access testing penetrations openings and for access opening~ for periodically demonstrating the integrity periodically demonstrating integrity of thethe reactor building.

reactor building. These These provisions are all all described in in Section VI-F. VI-F.

• Revision 21 UFSAR Revision 21 I-11 I-II October 2009 October 2009

Nine Mile Point Unit unit 1 UFSAR 4.

4. The containment containment system and all engineered .safeguards engineered maintained maintained such postulated accidents postulated 10CFRIO0.

10CFRI00. accidents are below The offsite offsite accordance with RG 1.183 accordance all safeguards were originally other necessary originally designed and such that offsite doses resulting resulting from below the values stated in doses have been re-analyzed 1.183 and 10CFR50.67. 10CFR50.67. from re-analyzed in in The analysis analysis in results are detailed detailed in in Section XV. XV. 5.

5. Double isolation valves are provided Double isolation provided on most lines directly entering directly entering the primary primary containment containment freespace, freespace, or penetrating penetrating the primaryprimary containment containment and connected to the RCS.

RCS. Lines which are equipped with double are not equipped isolation valves isolation determined to be acceptable valves have been determined acceptable based based upon the fact that the system reliability is reliability is notnot compromised, compromised, the systemsystem is is closed outside containment,containment, and a single active failure can be accommodated accommodated with isolation valve in only one isolation in the line. Periodic these valves will assure their testing of these their capability to isolate isolate at all times. all times. The isolation valve system valve system is is discussed discussed in in detail in in Section VI-D. VI-D. 6.

6. reactor building provides The reactor secondary containment provides secondary containment pressure suppression when the pressure suppression system is is inin service and containment barrier serves as the primary containment barrier during refueling and otherother periods when the pressure suppression system is suppression system is open or not required.

required. This This structure is is described described in in Section Section VI-C. VI-C. An emergency emergency ventilation system (Section VII-H) provides ventilation provides a means means controlled release for controlled release of halogens particulates halogens and particulates filters from the reactor via filters building to the stack reactor building accident conditions. under accident conditions. 6.0 Control and Instrumentation Instrumentation 1.

1. The Station Station is provided with a control room (Section is provided III-B) which has adequate shielding and other adequate shielding other emergency features to permit occupancy emergency features occupancy during all all credible accident credible accident situations.

situations. 2.

2. Interlocks Interlocks or other protectiveprotective features are provided provided reliability of procedural to augment the reliability procedural controls controls in in preventing serious accidents.

preventing accidents. Interlock systems are provided which block or prevent prevent rod withdrawal withdrawal from a multitude of abnormal abnormal conditions. The control conditions. control rod block logic logic is is shown onon

• Figures Figures VIII-6 VIII-8, respectively, VIII-6 and VIII-8, respectively, for the UFSAR Revision 21 UFSAR 1-12 October 2009 October 2009

Mile Point Unit Nine Mile Nine Unit 1 UFSAR

• source range source range, range monitor monitor (IRM) monitor monitor (SRM)

(IRM) neutron range, average average power instrumentation provides instrumentation recirculation flow recirculation power range flow control (SRM) and and intermediate neutron instrumentation. range monitor provides both intermediate range instrumentation. monitor (APRM) both control control blocks, (APRM) control rod and blocks, as shown and range In the power In power shown on Figure Figure VIII-14. VIII-14. excursions involving Reactivity excursions Reactivity control rods are involving the control either prevented or either prevented or their consequences substantially their consequences substantially mitigated by a control mitigated control RWM VIII-C.4.0) which RWM (Section VIII-C.4.0) which supplements procedural supplements procedural controls controls in avoiding patterns in avoiding patterns worths, a local power range of high rod worths, range monitor monitor (LPRM) (LPRM) neutron monitoring neutron monitoring and alarm alarm system system (Section VIII-C.1.1.3), and a control VIII-C.l.l.3), control rod position position indicating system IV-B.6.0), both of which system (Section IV-B.6.0), which enable enable the Operator Operator to observe observe rod movement, movement, thus verifying his verifying his actions. actions. A control A overtravel position light control rod overtravel light verifies verifies that the blade is the blade coupled to a withdrawn is coupled withdrawn CRD. CRD. A refueling platform A refueling operation interlock platform operation interlock is is discussed in Section XV, in Accident, which, along Refueling Accident, XV, Refueling which, along with other procedures other supplemented by automatic procedures and supplemented automatic interlocks, interlocks, serves to prevent criticality accidents in criticality accidents in the refueling mode. mode. A cold water addition reactivity A excursion is reactivity excursion is prevented prevented by the procedures interlocks described procedures and interlocks Section XV, in Section in XV, Startup Recirculation Loop Startup of Cold Recirculation Loop Analysis). . (Transient Analysis) Containment integrity Containment integrity is is maintained through the use of maintained through of strict strict procedural procedural controls and is is enforced enforced by interlocking mechanisms interlocking mechanisms at the airlock airlock doors to the drywell and a local local alarm system at the access access openings of the reactor building. reactor building. 3.

3. A reliable, reliable, dual-logic channel channel reactor reactor protection (RPS), , described in system (RPS) in Section VIII-A, VIII-A, is is provided provided to automatically initiate automatically initiate appropriate action whenever appropriate whenever parameters exceed preset limits.

various parameters limits. Each logic contains two subchannels channel contains channel completely subchannels with completely sensors, each capable of tripping the independent sensors, independent the channel. logic channel. A trip A trip of one-of-two subchannels in one-of-two subchannels in each logic channel results in in aa reactor scram. reactor scram. The trip trip in each logic channel may occur from unrelated in parameters, i.e., parameters, i.e., high neutron flux in in one logic channel coupled with high pressure in in the other logic Revision 21 UFSAR Revision 1-13 1-13 October 2009200O9

Nine Nine Mile Mile Point Unit 1 UFSAR channel will result channel fails fails in scram calibration result in in a direction solenoid valves. scram solenoid calibration of individual assure system valves. system reliability. in a scram. direction to cause event of loss of power reliability. scram. The RPS circuitry cause a reactor scram power or loss of air air Periodic testing and Periodic individual subchannels subchannels is The ability circuitry scram in supply to the is performed in the performed to ability of the RPS to to to safely terminate a variety safely terminate Variety of Station Station malfunctions malfunctions is is demonstrated demonstrated in Section xv. in Section XV. 4.

4. Redundant sensors and circuitry Redundant circuitry are provided provided for the actuation actuation of equipmentequipment required to function under under post-accident post-accident conditions.

conditions. This redundancy redundancy is is described described in various sections in the various sections of the text text discussing system design. design. 7.0 Electrical Power Electrical Power Sufficient Sufficient normal and standby standby auxiliary auxiliary sources sour~es of electrical electrical power power are provided provided to assure a capability for prompt shutdown capability continued maintenance and continued maintenance of the Station Station inin a safe condition all under all credible circumstances. credible circumstances. These features are discussed These features discussed in Section IX. in IX. 8.0 Radioactive Radioactive Waste Disposal 1.

1. Gaseous, are designed Disposal Gaseous, liquid and solid waste designed so that discharge accordance with 10CFR20 accordance waste disposal disposal facilities discharge of effluents 10CFR20 and 10CFRSO effluents is 10CFRSO Appendix facilities Appendix I.

is in I. in The facility facility descriptions are given descriptions in Section given in Section XII-A while while development of appropriate the development appropriate limits is is covered covered in in Section II. Section II. 2.

2. Gaseous discharge from the Station is Gaseous discharge is appropriately appropriately monitored, as discussed monitored, discussed in in Section VIII, automatic VIII, and automatic isolation isolation featuresfeatures are incorporated incorporated to maintain releases releases below below the limits of 10CFR20 10CFR20 and 10CFRSO 10CFR50 Appendix I. I.

9.0 Shielding and Access Control Access Control Radiation Radiation shielding and access access control control patterns patterns are such that that doses will be less than those specified specified in 10CFR20. in 10CFR20. These These features are described described in in Section XII-B. XII-B. UFSAR Revision 21 21 1-14 October 2009 2009

Nine Mile Point Unit 11 UFSAR

• 10.0 10.0 Fuel Fuel Handling and Storage Appropriate accidental accidental criticality criticality fuel are described in Storage Appropriate fuel handling and storage and provide facilities storage facilities provide adequate cooling Section x.

in Section X. which preclude cooling for spent spent

• UFSAR Revision Revision 21 1-14a I-14a October 20C 2009 39 9

Nine Mile Point Point Unit 1 UFSAR THIS PAGE INTENTIONALLY INTENTIONALLY BLANK UFSAR UFSAR Revision 21 21 1-14b I-14b October 2009

Nine Mile Nine Mile Point Unit Unit 1 UFSAR Standby Liquid Standby Liquid Control Control System System Capability: Capability: Shutdown Margin Shutdown Margin (Ak) (~k) PPm (20C, Xenon Free) (20C, Xenon Free) SRLR SRLR(2) (2) SRLR (2) SRLR(2) 9.0 9.0 Reactor Vessel Reactor Vessel Inside Diameter Inside Diameter 17 ftft -- 99 inin Internal Height Internal Height ft - 63 ft - 10 in in Design Pressure Design 1250 psig at 575°F 1250 575°F 10.0 Coolant Recirculation Coolant Loops Recirculation Loops Location of Recirculation Location Recirculation Containment Drywell Containment Drywe1l Loops Loops Number of Recirculation Recirculation 5 Loops and Pumps Pumps Pipe Size Pipe Size 28 in in 11.0 Primary Containment Primary Containment

• Type Type Design Pressure Pressure of Drywell Vessel Vessel Pressure of Design Pressure of of Suppression Chamber Suppression Chamber Pressure Pressure Suppression 62 psig 35 psig Vessel Vessel Design Leakage Design Leakage Rate Rate percent per day at 35 0.5 weight percent 35 psig 12.0 Secondary Containment Secondary Containment Type Type Reinforced concrete and steel Reinforced steel superstructure with metal siding superstructure siding 2

Internal Design Pressure Internal lb/ft 40 lb/fe Leakage Rate Design Leakage Rate 100% free volume per day 100% day discharged via stack while maintaining maintaining 0.25-in water negative pressure in pressure in the reactor building reactor building atmosphere relative to atmosphere

• UFSAR Revision 2121 1-17 1-17 October 2009 2009

Nine Mile Point unit Unit 11 UFSAR 13.0 13.0 Structural Structural Design Ground Seismic Ground Acceleration Acceleration Sustained Wind Loading Sustained Control Room Shielding O.llg

0. 11g 125 125 mph, mph, 30 ft Normal Operation Normal ft above ground ground level Operation - Dose not to level to 0

exceed exceed hourly equivalent equivalent (based on on 40-hr week) week) of maximum permissible permissible quarterly dosedose specified in specified in 10CFR20. 10CFR20. Accident Conditions - Meets the Accident Conditions design total total effective dose effective dose equivalent (TEDE) equivalent (TEDE) dose for for personnel personnel in in the control roomroom such such that the exposure exposure limits of of 10CFR50.67 will not be exceeded 10CFR50.67 exceeded in in the course course of the LOCA.LOCA. In In addition, the cumulative addition, cumulative dose fromfrom any design basis accident accident (DBA) (DBA) I would also meet 10CFR50.67 10CFR50.67 limits limits.. 14.0 Station Station Electrical Incoming Power Outgoing Outgoing Power Onsite Electrical System Sources Power Sources Lines Power Lines Sources Onsite Power Sources System Two 115-kV transmission transmission lines 345-kV transmission Two 345-kV transmission lines Two diesel generators generators Provided Two safety-related safety-related Station batteries Q-related 125-V dc battery One Q-related system system 15.0 Reactor Reactor Instrumentation Instrumentation System System Location of Neutron Location Neutron In-core Sensors Monitor Sensors Ranges of Nuclear Nuclear Instrumentation: Instrumentation: Range Four Startup Range 0.01% rated power Source to 0.01% power and to to Monitors Monitors 8.3% 8.3% with chamber chamber retraction Intermediate Range Eight Intermediate Range 0.0003% to 40% 0.0003% 40% rated power power Monitors Monitors 120 Power Power Range Monitors Monitors 5% to 125% 125% rated power power UFSAR Revision 2121 1-18 October October 2009 2009

Nine Mile Point Unit 11 UFSAR

• 16.0 16.0 Number Reactor Protection Reactor Reactor Reactor System System Number Protection System Number of Channels in Protection in System 22 Number of Channels 22 Required Required to Scram or or Effect Effect Other Protective Functions Functions Number of Sensors Number Sensors per 22 Monitored Variable Monitored Variable inin each Channel (Minimum each (Minimum scram function) for scram
• UFSAR Revision 21 UFSAR 21 1-18a I-18a October 2009

Nine Mile Nine Mile Point Point Unit Unit 1 UFSAR THIS PAGE THIS PAGE INTENTIONALLY INTENTIONALLY BLANK BLANK UFSAR Revision 21 I-18b U-18b October 2009

Nine Mile Nine Unit 11 UFSAR Point Unit Mile Point UFSAR

• E.

E. 1. 1. REFERENCES REFERENCES USAEC Press USAEC Nuclear 1965. 1965. Release H-252, Press Release Power Plant Nuclear Power "General Design H-252, "General Design Construction Permits," Plant Construction Permits," Criteria Criteria November November for for 22, 22, 2.

2. 0000-0084-3226-SRLR, "Supplemental Reload Revision 0,0, "Supplemental 0000-0084-3226-SRLR, Revision Reload Licensing Report for NMP1, Reload 20, Cycle 19," December Licensing Report for NMPI, Reload 20, Cycle 19," December 2008.

2008. 3.

3. GE GE Fuel Designs, General Bundle Designs, Fuel Bundle Electric Company General Electric Company Proprietary, NEDE-31152P, Proprietary, June 1996.

Revision 5,5, June NEDE-31152P, Revision 1996.

• 21 UFSAR Revision 21 1-21 2009 October 2009

Mile Point Unit 1 UFSAR Nine Mile Nine

• ABBREVIATIONS TABLE 1-2 ABBREVIATIONS AND ACRONYMS ACRONYMS USED IN UFSAR ACI American Concrete American Concrete Institute Institute ADS Automatic depressurization depressurization system system AISC American American Institute Institute of Steel Construction Construction ALARA As low as reasonably reasonably achievable achievable ALRA Amended license license renewal application AMP Aging Management Management Program Program ANS American Nuclear Society American Nuclear ANSI American National National Standards Standards Institute Institute AOV Air-operated Air-operated valve APRM APRM Average power range monitor monitor ARI Alternate Alternate rod injection ARMS ARMS Area radiation radiation monitoring monitoring system system ART Adjusted reference reference temperaturetemperature AST Alternative source term Alternative term ASTM American American Society Society for Testing Testing and Materials Materials ATWS ATWS Anticipated transient without Anticipated scram without scram
• BOC BOP BPWS BTP BWR Beginning of cycle Beginning Balance of plant Balance plant position withdrawal Banked position Banked Branch technical Branch Boiling water Boiling sequence withdrawal sequence technical position reactor water reactor Owners' Group BWROG BWROG Water Reactor Boiling Water Boiling Reactor Owners' Group BWRT Backwash receiving Backwash receiving tank tank BWRVIP BWRVIP Boiling Reactor Vessel Water Reactor Boiling Water Vessel and Internals Internals Program Program CAD Containment atmosphere dilution (device)

Containment atmosphere CCCWS Closed-cycle cooling water Closed-cycle system water system CEO Chief Executive Officer Officer CFR Code of Federal Federal Regulations Regulations CFS Condensate filtration Condensate filtration system system CGCS Combustible gas control system Combustible system CHF Critical heat flux Critical flux CIV intermediate valve Combined intermediate CND Condensate demineralizer Condensate demineralizer CO 22 Carbon dioxide COLR Operating Limits Report Core Operating Report CPR Critical power ratio Critical ratio Control rod drive CRD CRDA Control rod drop accident accident CRDRL Control rod drive return return lineline 21 UFSAR Revision 21 11 of 7 October 2009 2009

Nine Mile Nine Mile Point Point Unit Unit 1 UFSAR CRPI CRPI CRS CRS Control TABLE 1-2 TABLE (Cont'd.) 1-2 (Cont'd.) Control rod position position indication Control Room Supervisor Supervisor CRT Cathode Cathode ray tube CSO Chief Shift Shift Operator Operator CST storage tank Condensate storage Condensate CUF Cumulative usage factor Cumulative usage factor CWT Concentrated waste tank Concentrated waste DAC Dominant Dominant area of concern DBA Design basis accident accident DBE Design basis earthquake basis earthquake DCRDR DCRDR Detailed Detailed control room design review control review DE Dose equivalent equivalent DEC Department Department of Environmental Environmental Conservation Conservation DER Deviation/Event Deviation/Event ReportReport DER Double-ended Double-ended rupture DG Diesel generator Diesel generator DOP Dioctylphthalate Dioctylphthalate DOT Department of Transportation Department ECCS Emergency Emergency core cooling system core cooling system ECP Electrochemical corrosion potential Electrochemical corrosion potential EDG Emergency generator Emergency diesel generator EFPY Effective full-power Effective full-power years years EIC Energy Information Center Energy Information Center EOC End of cycle EOF Emergency Operations Facility Emergency Operations Facility EOL End of life life EOP Emergency operating Emergency operating procedure EPA Environmental Environmental Protection Agency Agency EPDM Ethylene-propylene-diene-monomer Ethylene-propylene-diene-monomer EPG Emergency Emergency procedure procedure guideline EPRI Electric Electric Power Research Institute Research Institute EQ Environmental Environmental qualification ESF Engineered safety Engineered safety feature ESW Emergency service water Emergency service water FA Fire area FAC Flow-accelerated corrosion Flow-accelerated FCV Flow control valve FHA Hazards Analysis Fire Hazards Analysis FMEA Failure modes and effects analysis analysis FMP Program Fatigue Monitoring Program FRC Franklin Research Research Center Center 21 UFSAR Revision 21 22 of 7 October 2009 2009

Nine Mile PointPoint Unit 1 UFSAR

• FSA Fire subarea subarea (Cont'd.)

TABLE 1-2 (Cont'd.) FSAR Final Safety Final Safety Analysis ReportReport FZ Fire zone Fire GALL Generic Generic aging lessons lessons learned learned GDC General General Design Criterion GE General General Electric Company Electric Company GL Generic Generic Letter Letter GSI Generic Generic Safety Safety Issue HAZ Heat-affected zone Heat-affected zone HCU Hydraulic control unit Hydraulic control unit HEM Homogeneous equilibrium Homogeneous model equilibrium model HEO Human engineering engineering observation HEPA HEPA High-efficiency particulate High-efficiency particulate air/absolute (filter) air/absolute (filter) HPCI High-pressure coolant High-pressure coolant injection HVAC HVAC Heating, ventilating, Heating, ventilating, and air air conditioning HWC Hydrogen water chemistry Hydrogen HX Heat exchanger Heat exchanger I&C Instrumentation & control Instrumentation control ID diameter Inner diameter IGSCC Intergranular Intergranular stress corrosion corrosion cracking ILRT ILRT Integrated leakage rate test Integrated leakage test INPO Institute Institute Nuclear Power of Nuclear Power Operations Operations ISEG Independent Engineering Group Independent Safety Engineering ISI lSI Inservice inspection Inservice ISP Integrated Surveillance Integrated Program Surveillance Program IST 1ST Inservice testing Inservice testing LCO Limiting condition Limiting condition of operation LHGR Linear generation rate Linear heat generation LLD Lower Lower limit of detection LLL Low-low Low-low limitlimit LOCA Loss-of-coolant Loss-of-coolant accident accident LOFW Loss of feedwater feedwater LOOP offsite power Loss of offsite power LPCS Low-pressure Low-pressure core spray LPRM Local power Local monitor power range monitor LPSP setpoint Low power setpoint LPZ population zone Low population zone

• LRA LSSS License renewal License Limiting UFSAR Revision 21 renewal application safety system Limiting safety system setting 3 of 7 setting October 2009 October 2009

Nine Mile Nine Mile Point Point UnitUnit 11 UFSAR UFSAR LTC Load Load tap TABLE 1-2 (Cont'd.) TABLE tap changer changer (Cont'd.) M&TE Measuring Measuring and and testing equipment testing equipment MAPLHGR MAPLHGR Maximum Maximum average average planar planar linear linear heat heat generation generation rate MCC control center Motor control center MCPR Minimum Minimum critical critical power ratio power ratio MG Motor Motor generator generator MLHGR MLHGR Maximum Maximum linear linear heat generation generation rate MOV Motor-operated valve Motor-operated MSIV Main steam steam isolation valve MSL Main steam steam lineline MSLB MSLB Main steam steam line break NDT ductility Nil ductility transition transition NDT Nondestructive Nondestructive testing testing NDTT Nil ductility ductility transition transition temperature temperature NEI Nuclear Energy Institute Nuclear Institute NEIL Nuclear Electric Insurance Insurance Limited NFPA Protection Association National Fire Protection NMPC Niagara Mohawk Mohawk Power Corporation NPSH Net positive positive suction head head NRC Nuclear Regulatory Regulatory Commission Commission NRV Nonreturn valve Nonreturn NSRB Nuclear Safety Review Review Board NSSS Nuclear steam supply system system NVLAP NVLAP National Voluntary Laboratory Accreditation Voluntary Laboratory Accreditation Program Program NYPA New York Power Authority NYPP New York Power Pool Pool OBE Operating basis earthquake Operating OCCWS Open-cycle cooling water system Open-cycle system OEA Operating experience Operating experience assessment assessment OL Operating license Operating OLNC NobleChem On-Line NobleChem OOS Out of service OSC Operational Support Center Operational Center OT Operational transient Operational transient PA Public address (system) PASS Post-accident sampling system Post-accident system pcr PCI Pellet-cladding Pellet-cladding interaction UFSAR Revision Revision 21 21 44 of of 7 October 2009 2009

Nine Mile Point Unit 1 UFSAR

• PCT p.f.

TABLE 1-2 (Cont'd.) Peak cladding (Cont'd.) cladding temperature factor temperature p.f. Power factor P&ID Piping and instrumentation instrumentation diagram diagram PM Preventive Preventive maintenance maintenance PORC Plant Operations Operations Review Committee Committee PP/PA Page party/public address (system) party/public address PSAR Preliminary Safety Preliminary Safety Analysis ReportReport PSTG Plant-specific technical guideline Plant-specific technical P-T Pressure-temperature Pressure-temperature PVC Polyvinyl Polyvinyl chloride QA Quality assurance Quality assurance QATR Quality Topical Report Assurance Topical Quality Assurance Report RBCLCW Reactor closed loop cooling Reactor building closed cooling water water RBM Rod block monitor monitor RCA Radiologically-controlled area Radiologically-controlled RCPB Reactor coolant pressure boundary Reactor coolant boundary RCS Reactor coolant system Reactor coolant system RG Regulatory Guide Regulatory Guide RIP Reactor Reactor internals protection RIS Regulatory Summary Regulatory Issue Summary RMS Radiation monitoring Radiation monitoring systemsystem RO Reactor Operator Reactor Operator RPIS information system Rod position information system RPS Reactor protection Reactor (trip) protection (trip) system system RPT Recirculation Recirculation pump trip trip RPV Reactor vessel Reactor pressure vessel RPVH Reactor pressure vessel head Reactor RSP Remote shutdown panel Remote shutdown panel RSS Remote system shutdown system Remote shutdown RTD Resistance temperature Resistance temperature detectordetector RTNDT temperature nil Reference temperature Reference ductility transition nil ductility transition RWCU Reactor Reactor water cleanup RWE Rod withdrawal error withdrawal error RWM Rod worth minimizer minimizer RWP Radiation work permit Radiation permit SAG accident guideline Severe accident SAP accident procedure Severe accident procedure SAR report Safety analysis report SAS Secondary system Secondary alarm system SBO Station blackout blackout SCBA Self-contained breathing Self-contained apparatus breathing apparatus UFSAR Revision UFSAR 21 Revision 21 5 of 7 October 2009 2009

Nine Mile Point Point Unit 1 UFSAR UFSAR SCC Stress TABLE 1-2 (Cont'd.) (Cont'd.) corrosion cracking Stress corrosion Shutdown SDM Shutdown margin margin SDV Scram Scram discharge discharge volume volume SER Safety Evaluation Safety Report Evaluation Report SFC Spent fuel pool Spent pool cooling cooling and cleanup SIL Service Information Service Information Letter Letter SJAE Steam Steam jet air ejector jet air ejector SM Shift Manager Shift Manager SOE Sequence Sequence of events events SOP Special operating Special operating procedure procedure SORC Station Operations Station Operations Review Committee Review Committee SOV SOV Solenoid-operated Solenoid-operated valve SPDS Safety parameter Safety parameter display system system SR Surveillance Surveillance requirement requirement SRAB Safety Safety Review Review and Audit Board Board SRLR SRLR Supplemental Supplemental Reload Licensing Licensing Report Report SRM Source monitor Source range monitor SRO SRO Senior Reactor Senior Reactor Operator Operator SRP Standard Standard Review Review Plan SRV Safety/relief valve Safety/relief SRVDL Safety/relief valve discharge Safety/relief discharge line line SSA Safe Shutdown Analysis Analysis SSC Structures, systems Structures, systems and components components SWEC Stone & Webster Engineering Engineering Corporation SWP Service water system Service system TAF Top of active fuel fuel TBCLCW Turbine building closed loop water loop cooling water TCV Turbine control control valve TDH developed head Total developed TEDE effective dose equivalent Total effective equivalent TER Technical Evaluation Report Technical Evaluation Report TIP Traversing in-core Traversing in-core probe TLAA Time-Limited Aging Time-Limited Aging Analyses Analyses TLD Thermoluminescence dosimeter Thermoluminescence dosimeter TMI Three Mile Mile Island TSC Technical Support Technical Center Support Center TSVC Turbine stop valve valve closure TVD Test, vent and drain Test, UBC Uniform Building Code UFSAR Updated Updated Final Safety Analysis Report Analysis Report UHS Ultimate heat sink Ultimate UFSAR Revision 21 Revision 21 6 of 7 October 2009 2009

Nine Mile Point Unit unit 1 UFSAR

• UL TABLE 1-2 Underwriters' Nine Mile Point I-2 (Cont'd.)

(Cont'd.) Underwriters' Laboratories Laboratories Inc. Inc. Unit 1. I' Point Nuclear Nuclear Station - Unit 1 Unit 2 Nine Mile Point Point Nuclear Nuclear Station - Unit 2 UPS Uninterruptible Uninterruptible power supply URC Ultrasonic Ultrasonic resin cleaning U.S. U.S. United States States USBM U.S. Bureau U.S. Bureau of MinesMines USE Upper-shelf upper-shelf energy USLS U.S. U.s. Land Survey Survey UT Ultrasonic testing Ultrasonic testing VWO Valve Valve wide open open WNT Waste Waste neutralizer neutralizer tank WSLR Within Within scope of license renewal renewal

• UFSAR Revision Revision 21 7 of 7 October October 2009 2009

Nine Mile Nine Unit 11 UFSAR Point Unit Mile Point UFSAR

• B.

B. 1.0 1.0 The Scriba DESCRIPTION OF DESCRIPTION General General The Station Station isis located in the located on north-central ADJACENT TO AREA ADJACENT OF AREA on the the Lake THE SITE TO THE Lake Ontario portion SITE Ontario coast of coast in Oswego in the County, the town town of of Scriba in the north-central portion of Oswego County, approximately 55 mi approximately north-northeast of mi north-northeast of the the nearest boundary of nearest boundary of the city of the city of Oswego. Oswego. 1.1 1.1 Population Population Population Population growth growth in in the the vicinity vicinity of of the the Station Station has has been been very very slow, with the city of Oswego showing a decrease slow, with the city of Oswego showing a decrease in population. in population. The The 19601960 census census enumerated enumerated 22,155 22,155 residents residents compared compared to to approximately approximately 19,793 19,793 people people in in 1980. 1980. However, However, county county population population increased increased from 86,118 in 1960 to 113,901 in 1980. The total from 86,118 in 1960 to 113,901 in 1980. The total 1980 1980 population population within within 12 12 mi mi of of the the Station Station is is estimated estimated to to be be 46,349 46,349 (see Figure 11-4). This area contains all or portions of (see Figure 11-4). This area contains all or portions of one one citycity andand tenten towns. towns. Population Population and and population population density density forfor the the ten ten towns towns and and one one city city within within this this area area are are shown shown in in Table Table II-1. Counties and towns within this area II-I. Counties and towns within this area are shown on Figure are shown on Figure II-5. 11-5.

• Transient Transient population to to thethe rural, however, population within however, aa number facilities changes in number of area within 12 of school, area 12 mi undeveloped character rural, undeveloped mi of of the character of school, industrial, that populations.

changes in area populations. the Station industrial, and facilities in the area that create small daily and in the create small Station is the area. of the is limited area. There and recreational daily limited due There are, recreational and seasonal are, seasonal due The The population population withinwithin aa 50-mi 50-mi area area surrounding surrounding the the Station Station waswas approximately 914,193 in 1980 (see Figure approximately 914,193 in 1980 (see Figure 11-6). The city of 11-6) . The city of Syracuse Syracuse is is the the largest largest population population center center within within this this area, area, with with aa population population of of 170,105 170,105 in in 1980. 1980. Table 11-2 lists Table 11-2 lists cities cities within within this this 50-mi50-mi radius radius with with populations populations over over 10,000. 10,000. The The 50-mi radius contains 50-mi radius contains portions portions of of three three Canadian Canadian Census Census Divisions located in the province of Ontario: Prince Divisions located in the province of Ontario: Prince Edward, Edward, Frontenac, and Frontenac, Addington/Lennox. The and Addington/Lennox. The 1976 1976 population population countscounts totaled 22,559, 108,052, and 32,633, totaled 22,559, 108,052, and 32,633, respectively. respectively. 2.0 2.0 Agriculture, Agriculture, Industrial Industrial and and Recreational Recreational Use Use 2.1 2.1 Agricultural Agricultural Use Use The The areaarea within within aa 50-mi 50-mi radius radius ofof the the site site encompasses encompasses all all or or portions portions of ten New York counties: of ten New York counties: Cayuga, Cayuga, Jefferson, Jefferson, Lewis,Lewis, Madison, Madison, Oneida, Oneida, Onondaga, Onondaga, Ontario, Ontario, Oswego, Oswego, Seneca, and Seneca, and Wayne. Wayne. UFSAR UFSAR Revision Revision 21 21 11-3 11-3 October October 2009 2009

Nine Mile Mile Point Unit 1 UFSAR Approximately 37 percent Approximately percent of the land within this ten-county ten-county

• region is is used for agricultural production.

agricultural production. Tables 11-3 and Tables 11-4 present present agricultural agricultural statistics statistics for this ten-county ten-county region. 2.2 Industrial Use Industrial Use Several Several industrial establishments establishments are located in in Oswego County, County, with the Novelis Corporation with Corporation and the Independence Independence Generation operated by Sithe Energies Plant operated Energies USA being being located located nearest to to the Station. Station. The lakeshore lakeshore east of Oswego Oswego is is the most most industrially developed industrially developed area near the site. site. The cities cities of Fulton and Mexico Mexico are the only other industrial industrial sites sites within 15 mi of of the site. site. Two natural natural gas pipelines pipelines lie lie within 88 km of the plant; one pipeline plant; supplies the Independence pipeline supplies Independence Plant and the other supplies Indeck Indeck Energy. Energy. Both pipelines pipelines are located located on the north-south and east-west transmission line corridors. east-west transmission corridors. The The major industrial establishments establishments in in Oswego County, their County, their locations, and their principal locations, principal products are listed listed in Tables in Tables II-5 and 11-6. 11-5 nearest public water supply The nearest supply intake in in Lake Ontario is is located approximately 8 mi southwest of the Station location. located approximately location. This intake intake supplies the city of Oswego Onondaga County. Oswego and Onondaga County. Data on these and other other vicinity vicinity public water water supplies supplies are listed in listed in Table 11-7. Figure 11-2 shows the locations locations of the communities communities listed. listed. 2.2.1 Toxic Toxic Chemicals Chemicals Potential Sources of Toxic Chemicals Chemicals According Regulatory Guide (RG) According to Regulatory (RG) 1.78, 1.78, both onsite and offsite offsite potential toxic gas hazards potential considered. hazards must be considered. Any toxic substance substance stored onsite in in a quantity greater greater than 45 kg (100 (100 lb) must be evaluated. lb) evaluated. sources to be evaluated Offsite sources evaluated include stationary facilities stationary facilities transportation of toxic and frequent transportation substances (truck, rail, substances rail, and barge) within 8 km (5 (5 mi) mi) of the site. site. shipments are defined Frequent shipments defined as exceeding exceeding 10/yr 10/yr for for truck shipments, shipments, 30/yr for rail rail shipments, and 50/yr shipments, 50/yr for barge shipments. shipments. For the NMPNS site, NMPNS site, sources potential toxic chemical sources of potential chemical hazards hazards include chemicals include chemicals stored onsite, onsite, as well as stationary stationary and transportation sources transportation sources within 88 km of the site. site. Table 11-9 11-9 lists the chemicals lists chemicals associated associated with each source source along with their their quantities and distances from the Unit 1 control room air air

• intake.

intake. The stationary sources include the James A. A. FitzPatrick FitzPatrick 0 UFSAR Revision 21 21 11-4 11-4 October 2009

Nine Mile Point Unit 1 UFSAR

• plant, hazardous Novelis Corporation, plant, Novelis Station, and Unit 2.

passes within passes Corporation, Oswego Wire hazardous materials is

2. One transportation traffic within 6.2 km (3.9 mi) of the site.

transportation transportation source is Wire Inc., transportation source is truck traffic Inc., Sithe along Route Sithe Independence Route 104, site. Another is the railroad line between Another Independence source of possible which 104, which between Oswego and Mexico, NY. Mexico, NY. Discussions with Conrail indicate that on average, Conrail indicate average, only one hazardous chemical shipment during an 18-mo period hazardous chemical passes through the Oswego terminal. passes terminal. Traffic on a spur Traffic spur to the site site is not frequent is frequent enough enough <<30/yr) (<30/yr) to warrant warrant consideration. consideration. Only those chemicals that have the potential potential to form a toxic cloud or plume after vapor cloud vapor after release environment need release to the environment need to be be evaluated. evaluated. criterion is This criterion is met by all all chemicals listed chemicals listed in in Table 11-9. Table Control Room Habitability Control Habitability Determination Determination The effect effect of an accidental accidental release of each each of the chemicals chemicals described in previous section on control in the previous control room habitability habitability is evaluated is evaluated by calculating concentrations inside the calculating vapor concentrations control control room as a function of time following the accident. accident. This This calculation is calculation performed using the conservative is performed methodology conservative methodology outlined in in NUREG-0570 utilizing NUREG-0570 and utilizing the assumptions assumptions described described in RG 1.78. in 1.78. In postulated accident, In a postulated accident, the entire content content of the largest largest container is storage container single storage released, resulting is released, resulting in in a toxic vapor vapor cloud and/or plume cloud plume that is conservatively assumed to be is conservatively be transported by the wind directly toward transported toward the control control room intake. intake. The formation of the toxic cloud cloud and/or plume is is dependent dependent on on characteristics of the chemical the characteristics environment. chemical and the environment. The The chemical stored as a gas is amount of a chemical entire amount entire is treated treated as a puff puff or cloud that has a finite determined from the quantity finite volume determined and density density of the stored chemical. chemical. A substance substance stored as a boiling point below the ambient temperature liquid with a boiling temperature forms instantaneous puff due to flashing (rapid gas formation) of an instantaneous of some fraction of the stored stored quantity. The remaining liquid forms a puddle puddle that quickly spreadsspreads into a thin layer on the ground, subsequently ground, vaporizing and forming a ground-level subsequently vaporizing ground-level vapor vapor plume. plume. A high boiling boiling point liquid liquid (above ambient temperature) temperature) evaporates by forced convection forms a puddle that evaporates convection with no involved. flashing involved. calculations are done by a computer The calculations computer program program (VAPOR) (VAPOR) or a spreadsheet, spreadsheet, both based on NUREG-0570 methodology that requires NUREG-0570 methodology requires

• the following input information:

UFSAR Revision information: parameters, meteorology, control room parameters, Revision 21 21 II-4a ll-4a chemical physical properties, chemical physical meteorology, distance from the spill properties, spill October 2009 to to 2009

Mile Point Unit Nine Mile Nine Unit 1 UFSAR UFSAR the control the control room room intake, intake, and and the quantity of the quantity chemical released. of chemical released.

• The following The following Unit 1 control control parameters parameters are used: used: ventilation rate of rate 2530 ftft3/min~

of 2530 3 /min, and net free and free volume volume of 130,600 130,600 ft ft 33. The The most conservative most conservative meteorological conditions are meteorological conditions assumed for the are assumed calculations, consisting calculations, consisting of Pasquill Pasquill Class Class AA stability, stability, 0.5 0.5 m/sec wind speed m/sec speed and an ambient ambient temperature temperature of of 33°C 330C for sodiumsodium . bisulfite bisulfite solution stored onsite, solution stored onsite, and Class F stability, stability, a wind wind speed of 1.0 speed 1.0 m/sec, m/sec, and an ambient ambient temperature temperature of 90OF 90°F for all all other chemical other chemical releases. releases. The criteria The criteria for determining chemical toxicity and determining chemical and setting setting limits for habitability limits habitability determinations determinations are are taken from regulatory regulatory guidance documents. According guidance documents. According to RG 1.78, 1.78, the toxicity toxicity limit of of a chemical chemical is is the maximum maximum concentration concentration that can be tolerated tolerated by by an average average human human for 2 min without without physical physical incapacitation incapacitation (severe coughing, (severe coughing, eye burn, burn, severe severe skin irritation). irritation). Standard Standard Review Plan (SRP) (SRP) Section Section 6.4 states states that acute effectseffects shouldshould be reversible be reversible within a short period period of time (several (several minutes) without the benefit without medication other benefit of medication other than the use of of self-contained breathing self-contained apparatus (SCBA). breathing apparatus acute toxicity (SCBA). The acute toxicity limits listed limits listed in in RG 1.78 are used used in study except in this study except that, where more appropriate, where appropriate, documented documented sources available(2 -s. sources are available(2-S). Nonguideline toxicity Nonguideline toxicity limits are based concentrations that based on concentrations that

• produce no effects or minor irritation produce irritation affecting mental affecting mental alertness and physical coordination, assuming a 15-min exposure physical coordination, exposure time.

time. cases where appropriate In cases In appropriate human data are are not available, available, data are used by applying a conservative conservative factor of 10 to lower lower acute exposure the acute exposure limit. limit. effect of the continuous outside The effect venting of the onsite outside venting bisulfite sodium bisulfite storage tank on control habitability is control room habitability is evaluated evaluated by calculatingcalculating the maximum sulfur dioxide vapor vapor concentration concentration at the control control room intake.intake. evaluation is The evaluation is performed using the guidance described performed described in in RG 1.78. 1.78. The toxicity limit is is set at the TLV-TWA limit established established in NUREG/CR-5669(S) in NUREG/CR-5669(S) for sulfur dioxide. dioxide. Results and Conclusions The results of the analysis are summarized summarized in II-10, which in Table II-10~ indicates that none of the toxic chemicals indicates evaluated have the chemicals evaluated incapacitate the control room operators. potential to incapacitate potential operators. UFSAR Revision 21 21 ll-4b II-4b October October 2009 2009

Nine Mile Nine Mile Point Point Unit Unit 1 UFSAR UFSAR

• 2.3 2.3 Recreational Use Recreational Seventeen state Seventeen located within a located within a 50-mi identifies the identifies the state visitor counts.

visitor counts. Use state parks parks and 50-mi and one one national radius radius state parks parks and national wildlife of of the Montezuma National The Montezuma wildlife refuge Station. Station. and their facilities, refuge are Table Table 11-8 11-8 facilities, capacities, National Wildlife capacities, and Refuge is Wildlife Refuge is and located north located north of Cayuga Cayuga Lake Lake in in Seneca approximately 44 County, approximately Seneca County, 44 southwest of mi southwest of the Station. Station.

• UFSAR Revision 21 21 II-4c 11-4c October 2009

Nine Mile Nine Mile Point Point Unit 1 UFSAR THIS PAGE PAGE INTENTIONALLY INTENTIONALLY BLANK UFSAR Revision 21 21 II-4d October 2009

Nine Mile Point Unit 1 UFSAR

• G.

G. 1. 1. 2. 2. REFERENCES REFERENCES Nine Mile Point Manual." Manual." Sax,N. E. Sax,-N. E. Nuclear Station "Offsite Dose Calculation Point Nuclear Dangerous Properties Dangerous Properties of Industrial Calculation Materials, Industrial Materials, 3rd Edition, 3rd York, NY, Nostrand Reinhold, New York, Edition, Van Nostrand NY, 1968. 1968. 3.

3. NUREG/CR-5669, NUREG/CR-5669, "Evaluation of Exposure Exposure Limits to Toxic Gases for Nuclear Reactor Control Control Room Operators,"

Operators," July 1991. 1991. 4.

4. Physics, 76th Edition, David CRC Handbook of Chemistry and Physics, David R.

R. Lide, Lide, Editor-in-Chief. Editor-in-Chief. 5.

5. Air Contaminants Contaminants - Permissible Limits, Title Permissible Exposure Limits, Title 29 29 Code of Federal Federal Regulations Regulations Part 1910-1000, 3112, 1910-1000, OSHA 3112, 1989.

1989.

• Revision 21 UFSAR Revision 21 11-9 11-9 October 2009 2009

Nine Mile Point Unit 11 UFSAR

• INDUSTRIAL FIRMS WITHIN INDUSTRIAL TABLE 11-5 11-5 WITHIN 8 KM (5 MI) OF UNIT 11 (5 MI)

Distance/ Direction from Site Firm Firm (km) (km) Products Products Employment Employment Novelis Corporation Novelis 4.5/SW 4.S/SW Aluminum Aluminum 1,000 I sheet and plate James A. A. FitzPatrick <1/E

                               <l/E             Electrical Electrical          500 500 Nuclear Power Plant Nuclear         Plant                         generation Nine Nine Mile  Point Mile Point             Adjacent Adjacent         Electrical Electrical        1,100 Unit 2                       to Unit 1        generation Sithe Energies Energies USA USA         3.s/SW 3.5/SW           Electrical Electrical           75 75 Independence Independence                                  generation

• Generation Plant Generation Oswego Wire Wire Incorporated Incorporated Plant 7.0/SW 7.0/SW Copper wire 40 40 I

NOTE: For complete listing listing of major industries in in Oswego Oswego NOTE: reference Oswego County County, reference County, County Industrial Industrial Directory. Directory. Revision 21 UFSAR Revision 21 1 of 11 October 2009 October 2009

Point Unit 11 UFSAR Mile Point Nine Mile Nine UFSAR TABLE 11-6 TABLE 11-6 UTILITIES IN OSWEGO COUNTY PUBLIC UTILITIES PUBLIC Location Tn sotiten Servjce I Power Niagara Mohawk Power sites Many sites Gas Gas Corporation Corporation New york Telephone York Telephone Many sites Communications communications Company Company Penn Central Railroad Shipping Telephone Oswego County Telephone oswego Oswego oswego Communications communications Company Alltel All tel New York, Inc... York, Inc Fulton Communications communications New York Power Authority Many sites Gas and Electric 0 UFSAR Revision 18 UFSAR Revision 18 11 of of I1 October 2003 October 2003

Nine Mile Point Unit 1 UFSAR

• TABLE 11-9 TABLE 11-9 SOURCES OF TOXIC CHEMICALS 8 KM (5 CHEMICALS WITHIN MI) OF UNIT 1 SITE (5 MI)

Chemical Chemical Quantity Quantity Distance Distance Location Location Chemical Chemical Stored to Intake James A. A. N2 N2 40,432 lb 0.497 mimi FitzPatrick FitzPatrick Plant C0 CO 22 26,000 lb Propane Propane <10,000 lb Halon 1301 1301 6,000 6,000 lb NaOCI 4,537 lb NaOH 1,694 1,694 lb Gasoline 6,005 lb H2 H2 1,150 1,150 lb lb Freon R-12 Freon R-12 1,695 1,695 lb Freon Freon R-22 R-22 6,010 lb Novelis Corporation Novelis Corporation C1 CI 22 1,500 1,500 lb 3.0 mi mi HCI HCl 5,000 lb CO 22 118,000 118,000 lb lb Propane 80,000 lb N N22 50,000 50,000 lb lb NaOH 4,200 4,200 gal gal H22 SO 44 5,000 gal gal Route 104 HCl 12,000 lb mi 3.4 mi CO 22 6,000 lb lb N N22 40,000 lb 40,000 lb Nine Nine Mile Point H22 SO S0 44 165 gal gal Unit 1 H2 H2 12,000 12,000 ft 3 ft 112 m CO C0 22 20,000 20,000 lb 100 m N N22 15,300 15,300 gal 140 m Halon Halon 1301 500 lb NaOH 165 gal gal NaOCI 1,200 gal gal

• UFSAR UFSAR Revision Revision 21 21 1 of 2 2009 October 2009

Nine Mile Point Unit 1 UFSAR Chemical Chemical Location Location TABLE 11-9 Chemical Chemical (Cont'd.) 11-9 (Cont'd.) Quantity Quantity Stored Distance Distance to Intake Nine Mile Point Nine H22 SO 44 11,925 gal 11,925 gal Unit 2 NaHSO NaHS0 3 25,548 gal 100 m H H22 353.1 lb 353.1 103m 103 m CO 22 52,000 lb 52,000 167 m 167 Propane Propane 250 gal 640 m N N22 2,052,000 scf 2,052,000 229 m Halon 1301 Halon 250 lb NaOH 10,160 gal 10,160 gal NaOCI gal 3,930 gal Ethylene Glycol Ethylene gal 2,400 gal Oswego Oswego Wire H22 SO 44 70 gal 4.39 mi mi Incorporated Incorporated HCI HCl 15 gal gal Isopropyl Isopropyl Alcohol 65 gal gal Propane Propane 500 gal gal N2 N2 2,500 gal gal Energies, Sithe Energies, NaOH Ammonia 60,280 67 gal gal lb 2.17 mi mi 0 Inc. Inc. H2 H2 531 lb CO 22 C0 64,000 64,000 lb N N22 2,691,731 lb NaOH NaOH 6,747 6,747 gal gal NaOCI NaOCI 9,112 9,112 lb H22SO 4 11,997 11,997 gal gal UFSAR UFSAR Revision 2121 2 of 2 October 2009

• po~nit Nine Mile Po*Unit TABLE II-10 TABLE II-10 1 UFSAR UFSAR PREDICTED VAPOR CONCENTRATION CONCENTRATION ININ THE UNIT 1 CONTROL ROOM Criteria Exclusion Criteria Comparisons Comparisons Control Room Weight Weight Max. Single Max. Single Distance to Unit 1 Distance Toxic Limit Limit Limit Container Storage Storage Control Room Control Room 3

Chemical Chemical Location Chemical (mg/m)) (mg/m ) (lb) (lb) (ib) (mi) J. A. J. FitzPatrick A. FitzPatrick CO 2 CO, 17,998.15 17,998.15 37,200 28,000 0.497 H,2 H 7,495.04 15,500 2,084 N,2 N 104,147.59 104,147.59 215,500 35,983 35,983 Propane 163,991. 163,991.17 17 339,300 4,880 Corporation Novelis Corporation C0 CO,2 17,998.15 17,998.15 4,840,900 4,840,900 114,000 114,000 3.0 3.0 CI, C1 2 43.5 11,700 4,000 HCI HCl 52.19 52.19 14,000 11,891 N N,2 104,147.59 104,147.59 28,012,100 28,012,100 50,000 50,000 Propane 163,991.17 163,991.17 44,108,000 44,108,000 390,400 Oswego Wire Wire Incorporated Incorporated HC1 HCl 52.19 52.19 64,800 149 4.39 N N,2 104,147.59 104,147.59 129,286,700 129,286,700 17,992 Propane 163,991.17 163,991.17 203,575,200 2,440 2,440 Isopropyl Alcohol 1,229.12 1,229.12 1,525,800 1,525,800 435 435 Sithe Sithe Energy CO2 CO, 17,998.15 17,998.15 4,840,900 4,840,900 64,000 2.17 H H,2 7,495.04 2,015,900 2,015,900 531 531 N N,2 104,147.59 104,147.59 28,012,100 28,0l2,100 2,691,731 2,691,731 Revision 21 UFSAR Revision 21 11 of of 22 October 2009 October 2009

• Nine po~nit Nine Mile Po0Unit TABLE II-10 TABLE (Cont'd.)

(Cont'd.) 1 UFSAR UFSAR 0 Quantitative Results Comparisons Quantitative Comparisons Distance to Unit 1 Distance Toxic Limit Limit Control Room Max. Max. Conc. Conc. Control Room Control Location Chemical Location Chemical Chemical (ppm) (ppm) (ppm) (m) Nine Mile Point Point Unit 1 CO CO,2 10,000 10,000 6,160 100 H2 H, 90,909 90,909 2,760 2,760 112 N2 N, 90,909 90,909 11,760 140 140 2 3 Nine Mile Point Unit 22 NaHSO3asSO 2 NaHSO,asSO, 0.262 0.262 g/mg/m] g/m]** 2.67E-2 g/m 100 2 5.24E-03 g/m' g/m] 1.69E-3 g/m 1.69E-3 ** g/m'** H H,2 90,909 4,090 4,090 103 C0 CO,2 10,000 10,000 6,400 167 167 Propane Propane 1,000 1,000 142 640 640 N2 N, 90,909 14,720 14,720 229 229 Sithe Energy Sithe Energy Ammonia Ammonia 300 207 3,500 3,500 Notes: Notes:

• • For aa release For release of sodium bisulfite of sodium bisulfite solution to the solution to the containment containment berm.

berm.

**     For   continuous outside For continuous  outside venting venting of of the   sodium bisulfite the sodium  bisulfite      storage tank.

storage tank. UFSAR Revision 21 UFSAR Revision 21 22 of of 2 2 October 2009 October 2009

Nine Nine Mile Mile PointPoint Unit Unit 11 UFSAR UFSAR

• B.

B. The CONTROL The control building ROOM CONTROL ROOM control room turbine turbine building room is building at building offices on west, west, andand the is located at el on the the control located in el 277. 277. in the south and the south control room It It is bounded by is bounded and east, break area, room break corner of southeast corner the southeast the turbine east, the the of the administration the administration by the room on turbine room instrumentation and area, instrumentation and on the the control control (I&C) (I&C) office area,area, and diesel building on and diesel on the north. the north. 1.0 1.0 Design Bases Design Bases 1.1 1.1 Wind Loadings Snow Loadings Wind and Snow The wind loadings for the wind and snow loadings the control room are the same control room same as as for the turbine building. for 1.2 1.2 Pressure Relief Design There are no special pressure relief relief requirements for control for the control room. room. 1.3 Seismic Design and Internal Loadings Seismic Loadings The structural design for the control structural auxiliary control room, as well as the control room below at el 261, is Class I seismic based auxiliary control room below at el 261, is Class I seismic based on the maximum credible earthquake earthquake motion outlined outlined in in the introduction III. Section III. introduction to Section Components whose Components whose functional failure functional failure cause significant could cause release of radioactivity, significant release radioactivity, or which are vital vital to safe shutdown and isolation isolation of the reactor, reactor, are also designed as Class I. I. The seismic analysis The seismic resulted in analysis resulted in the application acceleration factors of 20.0 application of acceleration percent gravity 20.0 percent horizontal and 10.0 horizontal percent gravity vertical. 10.0 percent vertical. TheseThese acceleration factors were factors calculated from the dynamic were calculated analysis of the dynamic analysis the turbine building. building. Although control room Although the control is structurally a part is structurally part ofof the turbine building, functional load building, functional load stresses when stresses combined with stresses when combined stresses earthquake loading due to earthquake loading maintained within are maintained within thethe established established working stresses* for the working the structural material structural material involved. involved. 1.4 1.4 and Ventilation Heating and Heating ventilation Heating and air Heating conditioning are air conditioning provided for are provided personnel comfort for personnel comfort and instrument and protection. The instrument protection. ventilating system The ventilating also provides system also provides air to clean air clean to the control room the control following an room following an accident. accident. 1.5 1.5 and-Access Control Shielding andAccess Shielding Control access to Normal access Normal control room the control to the room is is provided from the provided from administration building administration building through through security-controlled security-controlled doors. doors . UFSAR Also Section XVI, see Section Also see Revision 19 UFSAR Revision 19 Subsection G. XVI, Subsection I!I-9 111-9 G. October 2005 October 2005

Nine Mile Point Unit 1 UFSAR Shielding Shielding is is supplied to allow continuous occupancy occupancy during any reactor accident. accident. The most limiting accidentsaccidents are the controlcontrol

• rod drop accident accident (CRDA)

(CRDA) and the loss-of-coolant loss-of-coolant accidentaccident (LOCA) (LOCA) without without core core spray, described in spray, which are described in Section xv. Section XV. shielding also meets the design TEDE dose rate for personnel The shielding personnel in in the control room such such that the exposure exposure limits of IOCFR50.67 10CFR50.67 will not be exceededexceeded inin the course course of the LOCA. LOCA. In addition, In the cumulative cumulative dose from any design design basis accident accident (DBA) (DBA) would would also meet 10CFR50.67 limits. IOCFR50.67 limits. Credit is is taken for automatic automatic initiation initiation of the control room air air treatment treatment system for the LOCA. LOCA. If air outside the building is If air is contaminated, contaminated, the ventilating ventilating system system will be controlled to assure that that contamination contamination within the control control room is is minimized and kept kept within the above limits, limits, as shown in in Section Section 3.0,3.0, following. following. 2.0 Structure Structure Design Plans showing location and principal principal dimensions are shown on on Figures 111-4, 111-4, 111-5, and 111-6. 2.1 General Structural General Structural Features Features The structural structural steel enclosing the control room and the steel enclosing

• auxiliary auxiliary control room below is is supported supported on concreteconcrete walls and concrete concrete foundations foundations bearing bearing on and keyed keyed into soundsound rock.

rock. Actual rock bearing bearing pressures pressures are less than one-third one-third of the allowable allowable working bearing pressure. pressure. Lateral Lateral earthquake earthquake forces forces or wind loads loads are transmitted transmitted to the concrete concrete foundations foundations by the combination of structural combination structural steel bracingbracing and concreteconcrete walls. walls. The control control room walls, walls, roof and floors are framed with structural steel. structural steel. The west and north interior interior walls are 12-in solid reinforced concrete. reinforced concrete. The east wall is is enclosed with insulated metal insulated metal wall panels mademade up of FK-16 x 16 16 metallic-coated interior metallic-coated elements, 1 1/2-in insulation interior liner elements, insulation and 16 B B & S gage F-2 porcelainized porcelainized aluminum exterior face sheets, exterior sheets, manufactured by H. H. Robertson Company. as manufactured Company. The wall panel panel joints joints are sealed with a synthetic elastomer caulking synthetic elastomer caulking material. material. This wall is separated from the administration is separated administration building extension by a 3-in rattle extension space. rattle space. The south interior wall wall consists of 8-in concrete consists concrete blocks laid with steel-reinforced steel-reinforced joints. mortar joints. An interior interior metal partition partition wall parallelparallel to the south wall forms a 6'-6" corridor corridor and is is provided provided with windows windows for observing observing the control room operations operations from the corridor. corridor. UFSAR Revision Revision 21 21 III-i0 111-10 October 2009 2009

Nine Mile Point Unit 1 UFSAR

• The slab immediately 1/2-in immediately above to the walls and provides structural steel beam structural a roof above decking, above the control provides radiation 1/2-in thick poured-in-place poured-in-place reinforced framing.

beam framing. above at el 333 which is control room at el 300 is shielding, and consists of 8 radiation shielding, reinforced concreteconcrete supported on Two-thirds of this slab area Two-thirds is made up of 3-in 3-in deep metal is pinned metal pinned on area has has decking, 2 in in of insulation insulation and a 5-ply roof with slag surface. surface. The remaining remaining third of the slab area provides provides a shielding shielding roof roof over over the control room and consists consists of the 8 8 1/2-in thick poured-in-place reinforced concrete poured-in-place reinforced concrete slab to which is is applied 1 1/2 in of rigid insulation and a 5-ply roof with slag surface. 1/2 in surface. The control control room floor is is poured-in-place poured-in-place reinforced reinforced concrete concrete on on 14-gauge 14-gauge metal decking. decking. The gross depth depth of the floor slab is is 8 in average depth of concrete in and the average concrete is is 5 3/4 in. in. 2.2 Heating, Heating, Ventilation and Air Conditioning Conditioning System System The ventilation ventilation system shown on Figure Figure 111-14 is is designed to to provide provide outside and recirculated recirculated air air to the control control room and auxiliary control room areas during normal auxiliary emergency normal and emergency conditions. conditions. In In the normal ventilation ventilation mode, mode, outside air air enters the system system through through a louvered louvered intake after which which it it passes through a 15-kW 15-kW duct heater heater and normal normal supply isolation isolation dampers, dampers, which which are interlocked interlocked with the emergency ventilation inlet dampers. emergency ventilation dampers. Outside air air is is needed needed to recoup air air from leakage and losses and to maintain a habitable environment for personnel. habitable environment personnel. The outside air then flows through through an outside air air mix damper damper and is is then mixed with recirculated recirculated control control room return return airair from the the recirculation damper, recirculation damper, which is is set to maintain maintain a positive pressure pressure inin the control room. room. The total total amount of air air (14,500 (14,500 cfm minimum) then passes through a two-element two-element dust filter filter and redundant redundant cooling coils where it it will be cooled,cooled, if if necessary, necessary, to ensure ensure the control room temperature temperature does not exceed exceed the maximum calculated temperature maximum calculated temperature of 80.5°F. 80.5 0 F. The cooled cooled air air enters enters the control room circulation circulation fan for distribution distribution to various areas through ducts. through ducts. Air will circulate circulate through through the control control room to the return ductwork ductwork for recirculation recirculation and mixing with additional outside air. outside air. In In order to prevent prevent infiltration infiltration of of potentially contaminated air, potentially contaminated air, doors are weather-stripped weather-stripped and penetrations penetrations are sealed sealed to maintain maintain a positive pressure pressure to the turbine turbine building of 1/16 1/16 inin of water.water. The emergency ventilation system is emergency ventilation is automatically automatically initiatedinitiated on on high high radiation radiation signal signal from the intake radiation radiation monitors, monitors, LOCA UFSAR Revision Revision 2121 III-II III-ii October 2009

Nine Mile Point Unit 1 UFSAR and/or MSLB signal and/or manually initiated manually emergency ventilation emergency air full full capacity signal from the reactor protection initiated supply isolation dampers when when required ventilation inlet inlet air will then flow through a 15-kW duct heater control room emergency capacity control protection system (RPS) required by procedures. dampers will be automatically procedures. dampers will be opened. dampers emergency fans. opened. (RPS), , or normal The normal automatically closed, and the The outside heater and one of the fans . . The design flow rate the or for the control room emergency ventilation system is emergency ventilation is 2250 cfmcfm +10%. +/-10%. Air then passes through a manual throttling throttling damper,damper, a high efficiency efficiency particulate particulate filter, filter, and an activated charcoal charcoal filter filter unit. filtered This filtered air air will then join the normal supply supply ductwork and mix with control control room return air return air to be circulated circulated by the normal control room circulation circulation fan. fan. The design flow design flow rate for the emergency emergency ventilation ventilation system outside outside air is air is determined as that necessary determined necessary to maintain maintain a positive positive pressure of of 1/16 in 1/16 in of water to the turbine building, building, administration

building, building, and outside atmosphere, atmosphere, and is is a function of control control room boundary leakage. leakage. The design flow rate of 2250 cfm +/-10% +10% isis within the required range'of range of 1000 to 3750 cfm which is is based on on minimum required required fresh air air for personnel personnel and maximum filter filter capability.

capability. emergency ventilation The emergency ventilation fans may be manually manually started started for for periodic periodic testing. Heating is provided by thermostatically-controlled is provided thermostatically-controlled ventilation heaters. duct heaters. Cooling is provided by two chiller is provided chiller units. units. Both Both the temperature temperature control control valve and/or the bypass bypass valvevalve for the chilled water chilled water system may be open without overcooling the control without overcooling control room. room. Tests and inspections inspections on the control room emergency emergency ventilation filters filters are are done in in accordance accordance with Technical Technical Specifications. Specifications. 2.3 Smoke and Heat Removal Removal To assist assist in maintaining a habitable in atmosphere in habitable atmosphere in the control control room and auxiliary auxiliary control room, room, a smoke purge capability capability is is provided from two independent provided independent fans, one 6000-cfm 6000-cfm makeupmakeup fan and one 8000-cfm 8000-cfm exhaust fan (Figure III-14). 111-14). Qualitative smoke Qualitative evaluations have been performed smoke evaluations performed for NMP1. NMPl. The The evaluations assessed the effects evaluations effects of both external external and internal internal fire/smoke events on the capability capability to maintain maintain reactor reactor control control either the control from either control room or remote shutdown panels. panels. The The evaluations considered evaluations considered various plant design design and procedural procedural criteria criteria in in accordance accordance with RG 1.196, 1.196, "Control Room Habitability Habitability at Light Water Nuclear Power Water Nuclear Power Plants," Plants," and NEI 99-03, 99-03, "Control "Control

• UFSAR Revision 21 21 II1-12 111-12 October 2009

Mile Point Nine Mile Point Unit 11 UFSAR

• Room Habitability Guidance,"

Room Habitability confirmed that egress confirmed shutdown shutdown the control panels panels preclude use of both preclude are are control room and Guidance," Revision egress pathways served served and that both the by by the control Revision 1. pathways to to and and including ventilation ventilation that no single single smoke/fire control room evaluations

1. The evaluations including the systems systems the remote smoke/fire event room and remote remote independent independent event could remote shutdown could shutdown of of panels.

panels. 2.4 Shielding and Access Shielding and Access Control Control Normal personnel Normal personnel access control room access to the control room is provided by three is provided controlled access controlled access doors all all located located on el el 277. 277. north door The north door opens into into the control room break area, the south door break area, door opensopens into the administration into administration building, building, andand the the west west door opens opens into a corridor, access to the administration corridor, giving access building at el administration building el 277 277 and also also making available available the the stairway stairway to el 261 of the administration building. administration building. In addition In stair is above, a stair addition to the above, is provided within the control provided within control room (northwest corner) down to the auxiliary auxiliary control room on on the ground floor, shownshown on Figure 111-4. In In case of a reactor reactor accident, access to or from the control personnel access accident, personnel room would be control room extreme of southerly extreme from the southerly all of all buildings and approximately buildings approximately ft from the center of the reactor. 400 ft reactor. The walls, walls, roof and floors are are designed concrete designed to have concrete thicknesses which provide thicknesses shielding during the design basis provide shielding basis accident (DBA). accident (DBA). 3.0 3.0 Safety Analysis Safety Analysis The control control room is is designed occupancy by continuous occupancy designed for continuous by operating personnel operating personnel during normal operatingoperating or accident accident conditions. conditions. Concrete shielding provided Concrete shielding in the roof and floors provided in above and in in the walls facing the reactor reactor building building isis more than sufficient to ensure sufficient ensure the exposure I0CFR50.67 will not exposure limits of 10CFR50.67 not exceeded in be exceeded course of aa LOCA. in the course LOCA. Maintaining positive Maintaining pressure inside the control room and regulating filtered regulating the filtered concentration of radioactive outside air supply prevents the concentration cumulative dose from the LOCA ensures that the cumulative materials and ensures accident will be within IOCFR50.67. within the exposure limits of 10CFR50.67. In In addition, supplied air air available in respirators are available respirators in thethe control room for use if if necessary. Tracer gas testing is is performed periodically periodically using the constantconstant injection method of ASTM E741-00, injection "Standard Test Method for E741-00, "Standard for Determining Air Change in in a Single Zone by Means of a Tracer Gas Gas UFSAR UFSAR Revision 21 21 lll-12a III-12a October 20092009

Nine Mile Point Unit 1 UFSAR Dilution." Dilution." For the constant of tracer gas is This occurs when constant injection is injected until the resulting injection method, injected into the control resulting concentration when the amount of tracer the same as the amount leaving gas in in the outside leaving the eRE. method, a constant control room envelope reaches a steady state concentration reaches tracer gas entering the CRE CRE. outside airflow used for pressurization constant flow envelope (eRE) flow (CRE) state value. By injecting the tracer pressurization of the value. eRE isis tracer envelope, an estimate envelope, estimate of the filtered filtered and unfiltered unfiltered airflow airflow that provides provides this pressurization pressurization can be made by measuring measuring the concentration concentration of tracer gas in in the outside outside airflow while while at the same time measuring measuring the steady state concentration concentration in in the eRE.CRE. A A eRE habitability program CRE habitability program has been established established to ensure that that eRE CRE occupants occupants can control control the reactor reactor safelysafely under normalnormal conditions and maintain it in maintain it in a safe condition condition following a radiological event, hazardous radiological event, hazardous chemical chemical release, release, or a smoke smoke challenge. challenge. Both normal and emergency emergency lighting-fdre lighting':~~are provided provided in in the control control room together together with communications, air conditioning, communications, air conditioning, ventilation, heating heating and sanitary plumbing facilities. plumbing facilities. If If normal electric electric power service service is is not available, available, provision has has been made to power the cooling, cooling, ventilating ventilating and heating units from the emergency emergency diesel generators. generators. Building components and finish materials Building components materials are noncombustible noncombustible and* and combustible combustible materials are not stored stored in in the control control room.room. distance of the control The minimum distance control room to the centerline centerline of of the reactor is is 330 ftft and there there are no direct connections direct connections from from passageways, ventilating passageways, ventilating ducts or tube connectionsconnections between between the reactor building and the control room. reactor building room. The floor of the control room is is 1616 ftft above above yard grade and 28 28 ft ft above maximum lake level (el 249). 249). Therefore, the Therefore, possibility possibility of flooding or inundation inundation is incredible. is incredible. UFSAR UFSAR Revision Revision 2121 IlI-12b III-12b October 2009

Nine Mile Point Unit 1 UFSAR UFSAR

• H.

The security located on the southwest corner perimeter. perimeter. III-i. See Figure III-I. SECURITY BUILDING SECURITY BUILDING WEST AND SECURITY security building west and security BUILDING ANNEX security building building annex annex are corner of the Station security Administrative Administrative offices offices are contained contained within these buildings for buildings for support of the duties duties associated associated with Station Station security. security. Because of the nature of this subject, Because subject, a detailed detailed description of description of these these buildings will not be discussed in in this document. document. For For additional information regarding this subject, additional information subject, refer to the Station Station security security plan. plan.

• UFSAR Revision 21 111-39 October 2009 October 2009

Nine Mile Mile Point Unitunit 1 UFSAR I. I. SOLIDIFICATION AND STORAGE BUILDING RADWASTE SOLIDIFICATION 1.0 1.0 Design Bases Bases 1.1 Loadings Wind and Snow Loadings Wind and snow loadings for the radwaste radwaste solidification solidification and building (RSSB) storage building designed to meet or exceed (RSSB) are designed exceed those of of building. disposal building. the waste disposal 1.2 Pressure Relief Design Pressure Relief pressure relief special pressure There are no special requirements for this relief requirements this building. building. 1.3 1.3 Seismic Design and Internal Loadings(I) Loadings(l) The foundation foundation mat, structural walls, mat, structural columns, floors and roof of walls, columns, of classified as primary the RSSB are classified primary structural elements. structural elements. All All primary primary structural seismically designed to withstand structural elements are seismically effects of an operating basis earthquake the effects (OBE) in earthquake (OBE) in accordance accordance with Regulatory Regulatory Guide (RG) 1.143.. (RG) 1.143 Secondary structure Secondary structure elements, including platforms, elements, including catwalks, pipe platforms, catwalks, supports, equipment and vessel supports, equipment supports, and internal masonry vessel supports, walls, are classified walls, nonseismic-resistant items and are classified as nonseismic-resistant 1.4 Heating, conventional method. designed by conventional Heating, ventilation method. Air Conditioning(2) Ventilation and Air,Conditioning(2) e ventilation and air conditioning heating, ventilation The heating, conditioning (HVAC) chilled (HVAC) and chilled designed for the following primary functional water systems are designed functional requirements: requirements: heat, ventilate and air condition heat, condition the RSSBj remove RSSB; remove particulates from the RSSB atmosphere; airborne particulates airborne prevent atmosphere; prevent unfiltered radioactivity from the exfiltration of airborne radioactivity unfiltered exfiltration infiltration of airborne building; prevent infiltration radioactivity into the airborne radioactivity RSSB control electrical room; control and provide control room and electrical provide a means for monitoring (via the main stack) the release release of airborne radioactivity via the ventilation radioactivity exhaust system; ventilation exhaust minimize the system; minimize effects on the facility and its occupants effects occupants from releases of of radioactivity into the RSSB atmosphere; radioactivity atmosphere; collect filter collect and filter air air displaced via the vents from all RSSB tanks containing displaced radioactive fluids; continuously radioactive continuously purge the RSSB of truck exhaust exhaust fumes and other other hazardous occupancy at all gases to ensure safe occupancy hazardous gases all times. times. 1.5 1.5 Shielding Control(3) Access control(3) Shielding and Access Shielding is Shielding designed to limit is designed radiation levels on the building radiation .exterior, exterior, in in the control room, the control room, in in the electrical room, the electrical room, stairwells, and stairwells, the passageway passageway to the truck bays.bays. Access to the exterior of the RSSB is Access protected area, which protected area, which isis controlled controlled by controlled by access to the is controlled Security. Nuclear Security. Normal Normal e/ UFSAR Revision Revision 1515 111-40 111-40 November 1997 1997

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 ,I               ~I   II                                           ~I                ~I !I                                                            NINE MILE POINT

• 11 l\ 1\ 1\ 1\ NUCLEAR STATION - UNIT SCALE: 1'=300' SCRIBA, N. Y.

                          -- ---- . ---                                                                                             UPDATED SAFETY ANALYSIS REPORT UFSAR Rev.21 (October 2009)

Nine Mile Nine Mile Point Unit Unit 11 UFSAR

• can be can be removed removed from the established to established from the from the CRD.

The drives CRD. to prevent drives position position the reactor. These the reactor. prevent accidental the control control rods These procedures accidental separation rods in procedures are separation of of the the control 6-in increments in 6-in control rod increments of of stroke and stroke and hold themthem in in these these discrete discrete latchlatch positions positions until until actuated for movement actuated movement by the the hydraulic hydraulic system to a new new position. position. Visible indication Visible indication of the position position of each each drive is displayed in is displayed in the control the control room room by means of of illuminated illuminated numerals numerals which which correspond with the respective correspond respective latched positions. latched positions. In addition, In indication indication is is provided provided thatthat shows shows insert insert and and withdraw withdraw travel travel limits of the drive limits drive andand an overtravel overtravel withdraw withdraw limit limit on on the drive have been reached. reached. Control rod seating seating at the lower lower end end of the stroke prevents prevents the overtravel withdraw limit from being overtravel withdraw unless the control reached unless reached control rod is is uncoupled uncoupled from the drive. drive. coupling to be checked. This allows the coupling checked. These These indicators indicators and those for the in-core those in-core monitors monitors are grouped together together and and displayed on the control displayed control panel and arrangedarranged on the board board to to correspond relative rod and in-core correspond to relative positions in in-core monitor positions in the core. core. During During reactor shutdown, the SDM can be verified. reactor shutdown, verified. The SDMSDM demonstration is demonstration is performed performed as described described in in the Technical Technical Specifications. Specifications. 6.1.2 6.1.2 Standby Standby Liquid Poison System System This system is is described described in in detail in in Section VII-C. Section VII-C. The The standby liquid poison system is standby designed to provide is designed provide the capability capability of bringing bringing the reactor, reactor, at any time in in a cycle, cycle, fromfrom a full full power and minimum minimum control inventory (defined to be at control rod inventory at peak xenon) to a subcritical subcritical condition with the reactor in in the the reactive xenon-free most reactive state. xenon-free state. The liquid poison solution is is sodium pentaborate enriched in pentaborate enriched in boron-10 isotope. isotope. The calculated liquid poison system SDM for the cold (20 (20oC), 0 C), xenon-free core xenon-free condition is is provided in in the SRLR(l). SRLR('). This SDM corresponds corresponds to a boron (B-10 isotope) concentration boron concentration of 109.8 109.8 ppm in in the reactor reactor core. core. 6.2 Control System Evaluation 6.2.1 Withdrawal Errors Evaluation Rod Withdrawal Design features provided to minimize the possibility of of inadvertent inadvertent continuous control rod withdrawal, withdrawal, and to limit limit

• potential power transients in include the following:

UFSAR Revision 21 21 in the event they should occur, IV-19 IV_-19 occur, October 2009

Nine Mile Mile Point Point Unit Unit 11 UFSAR 1. 1. 2. 2. The The control control system can be can system is be withdrawn Normal rod operation Normal control is designed operation is designed so that withdrawn at a time. time. that only one one rod step (notch) at a time. is a step operated at the same switches must be operated control switches time. Two Two same time to to withdraw withdraw a rod continuously. continuously. 0 21 UFSAR Revision 21 IV-19a IV-19a October 2009 2009

Nine Nine Mile Mile Point Point Unit Unit 1I UFSAR UFSAR

• structural components The structural examined to examined (including are designed are will not will not prevent components which to determine, determine. the (including a steam steam line designed so that which guide the loadings line break).

prevent insertion guide the loadings which break). The that deformations insertion of control the control The core control rods which would would occur core structural deformations produced produced by control rods. rods. by accident rods have occur in have been in a LOCA structural components been components accident loadings loadings Considerable effort Considerable effort waswas expended expended to eliminate possible failures eliminate possible failures or control or instability due to the vibration control instability vibration of reactor internal of reactor internal components. The components. The reactor reactor system was analyzed analyzed as a multidegree-of-freedom system. multidegree-of-freedom system. This determined the analysis determined This analysis system's natural frequencies, system's natural frequencies, the resultant vibration mode shapes resultant vibration shapes and the and the relationship relationship betweenbetween the the vibration vibration amplitudes amplitudes and the critical stresses critical stresses in in the the system, system, to show show that that system integrity integrity would be maintained. would maintained. 7.3 Surveillance Surveillance and Testing Rigid quality Rigid quality control requirements assured that the design requirements assured specifications specifications of the vessel components were met. vessel internal components met. These These quality control quality control methods fabrication of utilized during the fabrication methods were utilized of the individual components as well individual components assembly process. during the assembly well as during process. Preoperational performance tests Preoperational performance tests and the startup program startup program demonstrated the design demonstrated design adequacy adequacy of reactor reactor vessel vessel internals internals and and operability of the core spray operability spray spargers spargers. .

• Periodic testing of the control Periodic margin - core loading margin times and reactivity Specification Specification..

loading and stuck reactivity anomalies, anomalies, rod system, i.e., control control is rods; described is described i.e., in in rod reactivity reactivity scram insertion the Technical Technical

• UFSAR Revision 16 16 IV-3 1 IV-31 November November 19991999

Nine Mile Nine Mile Point Point Unit Unit 11 UFSAR UFSAR C. C. 1. REFERENCES REFERENCES 0000-0084-3226-SRLR, Revision 0000-0084-3226-SRLR, Licensing Report for Licensing Report for NMP1, 2008. 2008. Revision 0, NMP1, Reload Reload "Supplemental Reload 0, "Supplemental 20, 20, Cycle Cycle 19," 19," Reload December December 2.

2. Randall and Randall and St.

St. John, John, Nucleonics Nucleonics 16(3), 16(3), 82-86, 129 (1958). 82-86, 129 (1958). 3.

3. General Electric General Electric Standard Standard Application Application for Reactor Reactor Fuel, Fuel, NEDE-24011-P-A-16-US, October NEDE-24011-P-A-16-US, October 2007.

2007. 4.

4. J. A.

J. A. Wooley, Wooley, "Three-Dimensional "Three-Dimensional BWR BWR Core Simulator," Simulator," NEDO-20953A, January NEDO-20953A, January 1977. 1977. 5.

5. "Qualification of of the the One-Dimensional Transient Model One-Dimensional Core Transient Model BWRls," NEDO-24154, for BWR's," NEDO-24154, Vol.Vol. 1 and 2, 2, and and NEDE-24154-P-A, NEDE-24154-P-A, Vol. 3, Vol. 3, February February 1, 1, 1986.

1986. 6.

6. R. B.

R. B. Linford, Methods of Plant Transient Linford, "Analytical Methods Transient Evaluations for the General Evaluations General Electric Boiling Water Electric Boiling Water Reactor," NEDO-10802, Reactor," NEDO-I0802, December December 1986, 1986, and Amendments. Amendments. 7.

7. GE Fuel Bundle Designs, Designs, General General Electric Company Company Proprietary, NEDE-31152P, Revi 9 ion 8, NEDE-31152P, Revision 2001.

8, April 2001. 8.

8. GENE-770-31-1292, Rev.

GENE-770-31-1292, 2, "Engineering Rev. 2, Report for "Engineering Report for Application Application of GEl1GEl1 to NMP1 NMPI Reload 12," 12," General Electric Co. Proprietary Document, Co. Proprietary Document, April 1993. 1993. 9.

9. Safety Evaluation General Electric Advanced Long Evaluation of the General Long Assembly, NEDE-22290, Life Control Rod Assembly, NEDE-22290, January 1985. 1985.

10. Letter from C. C. o.

0. Thomas (NRC)

(NRC) to J.J. F.F. Klapproth (GE), (GE), July 1,1, 1985. 1985.

11. Safety Safety Evaluation Evaluation of the General Electric Duralife 230 230 NEDE-22290-P-A Supplement 3, Assembly, NEDE-22290-P-ASupplement Control Rod Assembly, 3, May May 1988.

1988. 12.

12. GENE-523-113-0894,Rev.

GENE-523-113-0894, Rev. 1, 1, "BWR "BWR Core Shroud Inspection and Core Shroud Evaluation Evaluation Guidelines," Guidelines," MarchMarch 1995. 1995. 13.

13. BWRVIP-01, BWRVIP-01, Rev.

Rev. 2,2, "BWR Inspection and "BWR Core Shroud Inspection Flaw and Flaw Guideline," October Evaluation Guideline," 1996. October 1996. 14.

14. BWRVIP-07, BWRVIP-07, "Guidelines Reinspection of "Guidelines for Reinspection Core of BWR Core Shrouds," February Shrouds," February 1996.1996.

UFSAR UFSAR Revision Revision 21 21 IV-32 IV-32 October October 2009 2009

Nine Mile Point Unit 11 UFSAR Nine Mile ~ 4.0 4.0 Cyclic Loads Cyclic Loads (Mechanical (Mechanical and and Thermal) Thermal) Fatigue resistance Fatigue resistance of of the reactor vessel the reactor vessel was originally analyzed was originally analyzed based on based on the expected expected number number of operatingoperating cyclescycles over the 40-yr 40-yr life of life of the the vessel. vessel. Table Table V-2 V-2 lists lists operating cycles the operating the cycles evaluated and evaluated and their their expected expected number of cycles. cycles. Using the Using operating cycles operating cycles in in Table Table V-2, V-2, stress analyses analyses were performed on were performed on feedwater nozzles, the feedwater nozzles, control control rod rod drive (CRD) (CRD) penetrations, penetrations, lower vessel lower head, vessel support vessel head, support skirt, skirt, core support cone, core support cone, vessel wall, vessel wall, other other nozzles nozzles in in thethe vessel, closure studs and the vessel, closure studs basin seal basin seal skirt skirt weld. weld. Fatigue usage factors, Fatigue factors, utilizing utilizing expected number the expected number ofof operating operating cycles cycles in in Table Table V-2, V-2, were calculated follows: calculated as follows: ni U-- u - N, Where: Where: uu = fatigue usage usage factorfactor ~ ni = expected number of cycles of a given expected amplitude amplitude given stress Ni allowable number maximum allowable number of cyclescycles at the same stress amplitude calculated usage factors (Table V-3) The calculated were all V-3) were all within the allowable allowable design limits (General Electric Company (GE) (GE) - 0.8, 0.8, ASME Section 111-1965 - 1.0). 1.0). recirculation'nozzles, Except for the reactor recirculation "nozzles, stress analysis on on other nozzles in in the reactor vessel concluded concluded that they were subjected to significantly subjected significantly less severe transients than the feedwater nozzles and, therefore, feedwater nozzles therefore, their fatigue usage factors were allall negligible. negligible. Stress analyses recirculation nozzle analyses of recirculation thermal transients conclude conclude that the nozzle fatigue usages are negligible. negligible. The vessel wall was also concluded to have a negligible fatigue usage factor. analyses were based on the expected The above analyses expected number of of operating cycles using assumed parameters parameters for each transient.transient. In In addition, NMP has implemented implemented a a Fatigue Monitoring Program Program ~ (FMP) that is (FMP) is used to manage the fatigue life life analyzed of analyzed components. components. utilizes This program utilizes the FatiguePro software to to UFSAR Revision 21 21 V-3 V-3 October October 2009

Nine Mile Point Unit 11 UFSAR maintain all all actual calculated below their corresponding based on the actual, experienced experienced by the plant cumulative usage factors (CUF) calculated cumulative actual, rather allowables. The calculations corresponding allowables. rather than expected, calculations are expected, number of cycles plant for each transient and, in the actual rather than assumed (CUF) cycles in some cases, parameters experienced assumed parameters cases, each experienced by each cycle. cycle. 5.0 Codes Codes Applicable Applicable- codes for the RCS are included in in Table Table V-4. V-4. Discussion of calculations Discussion demonstrating Code adherence calculations demonstrating adherence are given given in Section xv. in Section XV. summaries are provided Further summaries in Section provided in XVI. XVI. applicable up to the outside of the second Codes applicable second isolation isolation valve on all all emergency systems are also given auxiliary and emergency given inin Table Table V-4. V-4. UFSAR Revision 21 21 V-3a V-3a October 2009

Nine Mile Point Unit 1 UFSAR INTENTIONALLY BLANK THIS PAGE INTENTIONALLY BLANK

• UFSAR Revision 21 UFSAR 21 V-3b V-3b October 2009 October

Nine Mile Mile Point unit Unit 1 UFSAR B. B. SYSTEM DESIGN AND OPERATION SYSTEM 1.0 General 0 1.1 Drawings Drawings A flow diagram A diagram of the RCS is shown on Figure is shown Figure V-1.V-i. This system is is defined defined as encompassing encompassing the reactor primary system, system, the solenoid-actuated solenoid-actuated relief valves, primary system safety valves, primary safety valves and the emergency cooling system. system. The reactor primary primary system, system, in in turn, includes the reactorreactor vessel vessel and itsits internals, internals, recirculation recirculation piping, valves, valves, pumps and all all connected connected piping to to the external external containment containment isolation isolation valve. valve. Many other other systems systems connect directly to the RCS besides besides those shown on Figure Figure V-1. V-i. These are included included in in separate flow separate flow diagrams as given below: diagrams below: System SystemigurpiNo. Figure No. Feedwater Feedwater XI-7 XI-7 Shutdown Cooling Shutdown X-1 Cleanup X-2 X-2 Core Spray VII-1 VlI-1 Liquid Poison VII-6 VII-6 1.2 1.2 metal Materials Materials of Construction Insulation throughout the RCS within the drywell consists of Insulation metal reflective reflective inSUlation insulation and blanket insulation. inSUlation. In leakage leading to wetting of the inSUlation of small coolant leakage contact with the outer contact material of the loop, outer surface material of In the event event insulation in adverse loop, no adverse in chemical reaction leading to excessive electrochemical or chemical electrochemical corrosion is anticipated. is anticipated. 1.3 Thermal Stresses Thermal Heatup and cooldown rates for water Heatup water inin the reactor reactor system during normal operation operation will be limited to 100 100 0o F/hr by procedural procedural control. control. Holding this limit will assure that stresses are well well within Code limits as discussed in within in Section V-A.4.0 above. above. In short-term, more rapid blowdown greater In the event of a short-term, greater than 100 0 o F/hr, 100 F/hr, the vessel would be held for an equivalent equivalent amount of of time at constant temperature and pressure constant temperature pressure before heating or or cooling cooling is 100 0 F/hr. is resumed at 100°F/hr. example, the design For example, calculations specifically calculations specifically considered operation of a inadvertent operation considered inadvertent single bypass or solenoid-actuated solenoid-actuated .relief leading to a relief valve leading 17.50F/min blowdown

• 17.5°F/min blowdown forfor aa period period of 10 370 0 F.

10 min to 370°F. Following vessel would be held at a constant temperature this, the vessel temperature of 370°F370OF heatup would be resumed at cooldown or heatup hr, then cooldown for 1 3/4 hr,then at 100 0 F/hr. 100°F/hr. UFSAR Revision Revision 17 17 V-4 V-4 october October 2001

Nine Mile Nine Mile Point unit 1 UFSAR Point Unit

• F.

F. REFERENCES REFERENCES

1. NRC Standard NRC Standard Review Plan, Section Review Plan, Section 5.2.2, 5.2.2, Overpressurization Overpressurization Protection - NUREG Protection NUREG 75-087.

75-087. 2.

2. 0000-0084-3226-SRLR, Revision 0, 0000-0084-3226-SRLR, Revision 0, IISupplemental Reload "Supplemental Reload Licensing Report for NMPl, Licensing NMP1, Reload Reload 20, 20, Cycle December 19," December Cycle 19,11 2008.

2008. 3.

3. Shure and D.

K. Shure D. J. Dudziak, "Calculating J. Dudziak, IICalculating Energy Energy Release Release byby Fission Products," Fission Products,1I AEC Report WAPD-T-1309, March 1961. WAPD-T-1309, March 1961. 4.

4. K. Shure, IIFission Product Decay Shure, "Fission Heat,1I AEC Report Decay Heat," Report WAPD-BT-24, WAPD-BT-24, December 1961.

December 1961. 5.

5. Letter to NMPNS NRC Letter NMPNS dated dated November November 8,8, 2004, 2004, "Nine IINine Mile Mile Point Nuclear Nuclear Station Unit Unit Nos.

Nos. 1 and 22 - Issuance Issuance ofof Amendments RE: Implementation Amendments Reactor Pressure Implementation of the Reactor Vessel Integrated Program (TAC Surveillance Program Integrated Surveillance (TAC Nos. Nos. MC1758 andand MC1759) . II MC1759)." 6.

6. NRC Letter to NMPNS dated dated October 27, 2003, October 27, 2003, "Nine Mile IINine Mile

• Point Nuclear Station, Unit No.

Point Nuclear No. 1 - Issuance Pressure-Temperature Limit Curves Re: Pressure-Temperature MB6687) MB6687) .. II Issuance of Amendment Amendment Curves and Tables (TAC No. (TAC No.

• UFSAR Revision 21 21 V-21 V-21 October 2009

Nine Nine Mile Mile Point Point Unit unit 11 UFSAR UFSAR TABLE TABLE V-i V-l (Cont'd.) (Cont'd.) Safety Safety ValvesValves Number Number 99 Capacity Capacity (each) (each) 633,000 633,000 to to 651,000 651,000 lb/hrIbjhr Pressure Pressure setting setting (nominal) (nominal) 1218 1218 to to 1254 1254 psig psig Capacity Capacity (minimum(minimum each, each, 644,543 644,543 lb/hr lbjhr at at 1278 1278 psig psig certified) certified) Design Design code code ASME ASME Sec Sec 1-1962 1-1962 EmergencY*Coo]jng V-mn-v-rfi=nr-iT' OrNe-0 i nrT System qx7czi-on Condensers: Condensers: Design Design pressure pressure -- shell shell 15 15 psig psig at at 300OF 300°F

                                 --  tubing tubing                   1250    psig 1250.psig at   at 575*F 575°F Design Design code  code             --  shell shell                    ASME ASME Sec Sec VIII VIII (nuclear (nuclear cases) cases) and and ASME ASME Sec Sec III, III, Subsection        ND, Subsection ND, 1986     1986 edition edition        -
                                 -- tubing tUbing                   ASME ASME SecSec III, III, Subsection Subsection NC, NC, Class Class 2,2, 1986 1986 edition edition Number Number of    of tube tube bundles bundles                          44 Capacity Capacity (rated  (rated capacity capacity of                             10 7 Btu/hr of              3838 xx 107    Btu/hr atat 11351135 I

four four units) units) psig psig andand 5621F 562°P on on tube tube side; 5 psig and side; 5 psig and 228°F 228°F onon shell shell side side Operating operating time time withwith gravity gravity 88 hrhr makeup makeup Isolation Isolation valves valves inin inlet inlet line line 22 normally normally open open motor motor operated operated Isolation Isolation valves valves in in outlet outlet line line 11 normally normally closed closed air air operated and operated and 11 check check valve valve System* Pre-surps S;y-teT Pressures Design 1250 psig 0F Design 1250 psig at at 575 575°F Initial Initial vesselvessel hydrostatic hydrostatic test test 1875 1875 psig psig pressure pressure Maximum Maximum safety safety valve valve setting setting 1254 1254 psig psig Minimum safety valve Minimum safety valve setting setting 1218 1218 psig psig Solenoid-actuated Solenoid-actuated relief relief valve valve 22 @@ 1090 1090 psig psig settings settings 22 @

                                                                    @ 1095 1095 psig psig 22 @  1100 psig
                                                                    @ 1100    psig Emergency Emergency coolingcooling system  system                      >1080   psig for 12
                                                                  ~1080 psig for         12 sec sec pressure        actuation pressure actuation Reactor Reactor scram scram                                          :1080
                                                                  ~1080 psig psig Normal Normal operating operating pressurepressure                      1030 1030 psig psig at at 550°F 550°F UFSAR This This value lifetime UFSAR Revision value represents lifetime neutron Revision 16 represents the neutron fluence 16 the original fluence for original design for the 33 of design estimate the reactor of 33 estimate of reactor vessel.

vessel. of November 1999 November 1.999

Nine Mile Point Unit 1 UFSAR

• OPERATING TABLE V-2 V-2 OPERATING CYCLES AND TRANSIENT ANALYSIS RESULTS**

ANALYSIS TRANSIENT RESULTS** I OPERATING CYCLE OPERATING NO. OF CYCLES EXPECTED NO. CYCLES Removal Vessel Head Removal 50 50 Vessel Head Reinstallation 50 50 100 o F/hr Heatup 100OF/hr 240 240 100 o F/hr Cooldown* 1000F/hr 229 229 3000F/hr 300 0 F/hr Emergency Cooldown Emergency Cooldown 10 Blowdown Blowdown 1 Cycles Scram Cycles 280

• Emergency Emergency Condenser Condenser Initiation Loop into Isolated Loop Unisolation unisolation of an Isolated Initiation Isolated Loop Loop 30 30 30 30 Emergency Emergency Condenser Condenser Initiation Initiation into Idle Loop Loop 30 30 Shutdown Shutdown Cooling Initiation Initiation into Isolated Loop Isolated Loop 240 240 Inadvertent Start of Cold Loop Inadvertent Loop 20 20 Emergency Emergency Condenser Condenser into Pumped Pumped Loop Loop 500 500 Recirculation Pump Hot Loop Recirculation Loop Startup 300 300 The number The of 100 number of 1000F/hr cooldowns was o F/hr cooldowns was determined determined by by subtracting subtracting the emergency cooldowns emergency cooldowns and blowdown from the
• number number of 100 o 100OF/hr heatups.

F/hr heatups. UFSAR Revision 2121 1 of 2 October 2009

Nine Mile Point Unit 1 UFSAR

**   This table was used in TABLE V-2 TABLE       (Cont'd.)

(Cont'd.) in the original original fatigue evaluations evaluations ofof the reactor reactor vessel vessel as detailed detailed inin Section V-A.4.0. V-A.4.0. The The NMP Fatigue Fatigue Monitoring Monitoring Program is is now used to manage the the fatigue evaluations evaluations at NMP. NMP. This program program uses the the software to maintain all FatiguePro software all actual calculated cumulative usage factors below cumulative below their corresponding allowables. The calculations allowables. calculations are based on the actual actual number number of cycles experienced cycles experienced by the plant for each each transient transient and,and, in in some cases, cases, the actual parameters actual parameters experienced experienced by each cycle rather than the number of cycles cycles listed listed in this table. in table. 0 UFSAR Revision Revision 21 21 2 of 2 October 2009

Nine Mile Point Unit 1 UFSAR

• TABLE V-3 FATIGUE RESISTANCE FATIGUE V-3 ANALYSIS RESISTANCE ANALYSIS Region of Vessel Vessel Usage Factor*

Usage I Closure Studs 0.205 Basin Seal Skirt WeldWeld 0.782 Feedwater Nozzles Feedwater Nozzles Cavities With Repair Cavities 0.489 Without Repair Cavities Repair Cavities 0.163 Control Rod Drive Penetrations Penetrations 0.060 Lower Lower Vessel Head, Vessel Support Vessel Head, Support 0.0833 0.0833 Skirt and Core Support Cone Cone

• Reactor Recirculation Reactor Nozzles Recirculation Nozzles 0.006
• Listed usage factor values values are based on original original fatigue evaluations evaluations of the reactor vessel detailed in vessel as detailed in Section V-A.4.0.

V-A.4.0. Values Values are managed managed and maintained maintained below their below their corresponding allowables corresponding allowables through the NMP Fatigue Monitoring Monitoring Program. Program. UFSAR Revision 21 Revision 21 1 of of 1 October 2009 2009

Nine Mile Mile Point Unit 11 UFSAR TABLE V-4V-4 CODES FOR SYSTEMS FROM REACTOR VESSELVESSEL CONNECTION TO SECOND ISOLATION ISOLATION VALVE VALVE Piping Vessel Vessel Nozzle to Second Nozzle Second Isolation Valve Isolation Valve Isolation Valves Valves Shutdown Cooling B31.1-1955; ASA B31.1-1955; ASME Sec 1-1962 Sec 1-1962 Cleanup Sec 1-1962 and ASME Sec and Articles N324 and Articles N460 to N469 of of 111-1965; ASME Sec 111-1965; ASME Sec III, III, F, 1986 Appendix F, Appendix 1986 Edition* ASA B31.1-1955; B31.1-1955; B31.1-1955, ASA B31.1-1955, I 1-1962 and ASME Sec 1-1962 certain requirements certain requirements Articles Articles N324 and Sec of ASME Sec N460 to N469 of of IIIA-1965, and ASME IIIA-1965, ASME Sec 111-1965; 111-1965; 111-1986 Sec 111-1986 ASME Sec III, III, (IV-38-13) (IV-38-13) Appendix F, 1986 F, 1986 Edition* Feedwater Feedwater ASA B31.1-1955; B31.1-1955; ASME Sec 1-1962 Core Spray Core 1-1962 and ASME Sec 1-1962 Articles N324 and N460 to N469 of of 111-1965 ASME Sec 111-1965 ASA B31.1-1955; B31.1-1955; B31.1-1955 and ASA B31.1-1955 1-1962 and ASME Sec 1-1962 certain requirements certain requirements Articles N324 and of ASME Sec Sec IIIA-1965 IIIA-1965 N460 to N469 of of 111-1965; ASME Sec 111-1965; III, ASME Sec III, Appendix F, 1986 Appendix F, 1986 Edition* Liquid Poison ASA B31.1-1955, B31.1-1955, ASME Sec 1-1962 1-1962 and ASME Sec 1-1962 and Articles N324 and Articles N460 to N469 of of ASME Sec Sec 111-1965

• For analyzing analyzing thermally-induced conditions between isolation valves.

conditions 19 Revision 19 UFSAR Revision 1 of of 1 overpressurization thermally-induced overpressurization valves. October 2005

u.s. U.S. NUCLEAR REGULATORY NUCLEAR REGULATORY COMMISSION COMMISSION DOCKET 50-220 50-220 LICENSE DPR-63 NINE MILE POINT NINE NUCLEAR STATION NUCLEAR UNIT 1 UNIT FINAL SAFETY FINAL ANALYSIS ANALYSIS REPORT (UPDATED) VOLUME 2 VOLUME OCTOBER OCTOBER 2009 REVISION 21 REVISION

Nine Mile Point Unit 1 UFSAR CONTENTS TABLE OF CONTENTS Section Section Title Page TABLE OF CONTENTS CONTENTS i LIST OF TABLES xxxiv xxxiv LIST OF FIGURES FIGURES xxxix xxxix SECTION I SECTION INTRODUCTION AND

SUMMARY

INTRODUCTION I-I I-1 A. A. PRINCIPAL DESIGN PRINCIPAL DESIGN CRITERIA 1-2 1.0 General General 1-7 2.0 Buildings Buildings and Structures Structures I-B 1-8 3.0 Reactor Reactor 1-8 4.0 Reactor Vessel Reactor 1-10 5.0 5.0 Containment Containment 1-10 6.0 Control and Instrumentation Control Instrumentation 1-12 7.0 Electrical Power Electrical 1-14 8.0 B.O Radioactive Waste Radioactive Waste Disposal 1-14 9.0 Shielding Shielding and Access Control 1-14

• 10.0 B.

B. 1.0 2.0 Fuel Handling Site Handling and Storage CHARACTERISTICS CHARACTERISTICS Reactor Reactor Core I-14a I-14a I-IS 1-15 I-IS 1-15 I-IS 1-15 I-IS 1-15 3.0 4.0 Fuel Assembly Assembly I-IS 1-15 5.0 Control System I-IS 1-15 6.0 Core Design and Operating Conditions Conditions 1-16 7.0 Design Power Power Peaking Factor 1-16 8.0 B.O Nuclear Design Data Nuclear 1-16 9.0 Reactor Reactor Vessel 1-17 10.0 Coolant Recirculation Coolant Recirculation Loops Loops 1-17 11.00

11. Primary Containment Primary Containment 1-17 12.0 Secondary Secondary Containment Containment 1-17 1-17 13.0 Structural Structural Design 1-18 14.0 Station Electrical System Station Electrical 1-18 15.0 15.0 Reactor Instrumentation Instrumentation System 1-18 16.0 16.0 Reactor Reactor Protection Protection System I-IBa I-18a C.

C. IDENTIFICATION OF CONTRACTORS IDENTIFICATION CONTRACTORS 1-19

• D.

D. UFSAR Revision GENERAL Revision 21 21 CONCLUSIONS GENERAL CONCLUSIONS ii 1-20 1-20 October 2009

Nine Mile Nine Unit 11 UFSAR Point Unit Mile Point UFSAR CONTENTS (Cont'd.) 0 TABLE OF CONTENTS TABLE OF (Cont'd.) Section Section Title Title Page E. E. REFERENCES REFERENCES 1-21 1-21 SECTION IIII SECTION STATION AND ENVIRONMENT SITE AND STATION SITE ENVIRONMENT II-I 11-1 A. A. SITE DESCRIPTION SITE DESCRIPTION II-I 11-1 1.0 1.0 General General II-i 11-1 2.0 2.0 Physical Features Physical Features II-i 11-1 3.0 3.0 Property and Development Property Use and Use Development 11-2 11-2 B. B. DESCRIPTION OF DESCRIPTION OF AREA ADJACENT TO AREA ADJACENT TO THE SITE THE SITE 11-3 11-3 1.0 1.0 General General 11-3 11-3 1.1 1.1 Population population 11-3 11-3 2.0 2.0 Industrial and Agriculture, Industrial Agriculture, and Recreational Use Recreational Use 11-3 11-3 2.1 2.1 Agricultural Use Agricultural Use 11-3 11-3 2.2 2.2 Industrial Use Industrial Use 11-4 11-4 2.2.1 2.2.1 Toxic Chemicals Toxic Chemicals 11-4 11-4 2.3 2.3 Recreational Use Recreational Use II-4c II-4c C. METEOROLOGY METEOROLOGY 11-5 11-5 C. D. D. LIMNOLOGY LIMNOLOGY 11-6 11-6 E. E. EARTH SCIENCES EARTH SCIENCES 11-7 11-7 F. F. ENVIRONMENTAL RADIOLOGY ENVIRONMENTAL RADIOLOGY 11-8 11-8 G. REFERENCES REFERENCES 11-9 11-9 G. SECTION III SECTION III AND STRUCTURES BUILDINGS AND BUILDINGS STRUCTURES III-1 111-1 A. A. TURBINE BUILDING TURBINE BUILDING 111-3 111-3 1.0 1.0 Design Bases Design Bases 111-3 111-3 1.1 Wind Snow Loadings and Snow Wind and Loadings 111-3 III-3 1.1 1.2 1.2 Pressure Relief Design Pressure Relief Design 111-3 111-3 1.3 1.3 Seismic and Internal Design and Seismic Design Internal Loadings Loadings 111-3 111-3 1.4 Heating and Ventilation Heating and Ventilation 111-4 111-4 1.4 1.5 Shielding Access Control Shielding and Access and Control III-4a III-4a 1.5 2.0 Structure Design Structure Design 111-5 111-5 2.0 2.1 2.1 Structural Features General Structural General Features 111-5 111-5 Revision 21 UFSAR Revision UFSAR 21 iiii October 2009 October 2009

Nine Mile Point Unit 11 UFSAR (Cont'd.) TABLE OF CONTENTS (Cont'd.) section Section Title Page 2.2 2.2 Heating and Ventilation Ventilation System 111-6 111-6 2.3 Smoke and Heat Heat Removal Removal 111-7 111-7 2.4 Shielding and Access Shielding Access Control Control 111-7 2.5 Additional Additional Building Cooling 111-7 3.0 Safety Analysis Safety 111-8 B. B. CONTROL ROOM CONTROL ROOM 111-9 111-9 1.0 Design Bases Bases 111-9 1.1 Wind and Snow Loadings 111-9 111-9 1.2 Pressure Relief Design 111-9 1.3 Seismic Design Seismic Design and Internal Internal Loadings Loadings 111-9 111-9 1.4 Heating Heating and Ventilation Ventilation 111-9 111-9 1.5 Shielding Shielding and Access Access Control Control 111-9 111-9 2.0 Structure Design Structure Design III-10 111-10 2.1 General Structural General Features Structural Features 111-10 III-10 2.2 Heating, Ventilation Heating, Ventilation and Air Air Conditioning System Conditioning III-11 III-II Smoke and Heat Heat Removal 111-12 2.3 Removal 2.4 Shielding Shielding and Access Control Control III-12a III-12a 3.0 Safety Analysis III-12a III-12a C. C. DISPOSAL BUILDING WASTE DISPOSAL BUILDING 111-13 1.0 Design Bases Design Bases 111-13 1.1 Wind and Snow Loadings Loadings 111-13 1.2 1.2 Pressure Relief Design Pressure 111-13 1.3 Seismic Seismic Design and Internal Internal Loadings Loadings 111-13 1.4 Heating and Ventilation Ventilation 111-14 1.5 Shielding Shielding and Access Control Control 111-14 2.0 Structure Design Structure 111-14 2.1 General Structural General Features Structural Features 111-14 2.2 Heating Heating and Ventilation System Ventilation System 111-15 2.3 Shielding and Access Control Shielding Control 111-17 3.0 Safety Analysis Safety Analysis 111-17 D. D. OFFGAS BUILDING 111-19 1.0 Design Bases 111-19 1.1 Wind and Snow Loadings Loadings 111-19 1.2 Pressure Relief Design Pressure 111-19 1.3 Seismic Design and Internal Loadings Seismic Loadings 111-19 111-19 1.4 Heating and Ventilation Heating Ventilation 111-19 1.5 Shielding and Access Control Shielding Control 111-19 UFSAR Revision 21 iii iii October 22009 October 009

Nine Mile Mile Point Unit 11 UFSAR TABLE OF CONTENTS (Cont'd.) CONTENTS (Cont'd.) Section Section Title Title Page 2.0 Structure Structure Design 111-19 2.1 General Structural General Structural Features 111-19 2.2 Heating and Ventilation Heating Ventilation System 111-20 111-20 2.3 Shielding Access Control Shielding and Access Control III-20 111-20 3.0 Safety Analysis Analysis III-20 111-20 E. E. NONCONTROLLED BUILDINGS NONCONTROLLED III-22 111-22 1.0 Administration Administration Building 111-22 III-22 1.1 Design Bases Design Bases III-22 111-22 1.1.1 1.1.1 Wind and Snow Loadings III-22 111-22 1.1..2 1.1.2 Pressure Pressure Relief Design Design III-22 111-22 1.1. 1.1.33 Seismic Design and Internal Loadings 111-22 III-22 1.1.4 Heating, Cooling Heating, Cooling and Ventilation Ventilation 111-23 1.1. 1.1.55 Shielding Shielding and Access Control Control III-23 111-23 1.2 Structure Structure Design 111-23 III-23

1. 2.1 1.2.1 General Structural General Features Structural Features 111-23 1.2.2 Heating, Ventilation Heating, Ventilation and Air Air Conditioning Conditioning 111-24 III-24 1.

1.3 2.0 2.1 2.3 1.2.3 2.1.1 2.1. 2.1.22 Access Control Safety Analysis Sewage Treatment Treatment Building Design Bases Design Bases Building Wind and Snow Loadings Loadings Pressure Relief Design Pressure III-24 111-24 III-24 111-24 III-25 111-25 111-25 111-25 III-25 111-25 III-25 2.1. 3 2.1.3 Seismic Design and Internal Internal Loadings 111-25 III-25 2.1.4 Electrical Electrical Design 111-25 III-25 2.1. 2.1.55 Fire and Explosive Gas Detection Detection 111-25 2.1. 2.1.66 Heating and Ventilation Ventilation 111-26 2.1. 2.1.77 shielding Shielding and Access Control 111-26 2.2 Structure Structure Design 111-26 2.2.1 General Structural General Structural Features Features 111-26 111-26 2.2.2 Ventilation System Ventilation 111-28 III-28 2.2.3 Access Control Access Control 111-28 111-28 3.0 Energy Information Information Center 111-28 111-28 3.1 Design Bases Design 111-28 III-28 3.1.1 Wind and Snow L6adings Loadings 111-28 3.1. 3.1.22 Pressure Pressure Relief Design 111-28 3.1. 3.1.33 Internal Seismic Design and Internal Loadings Loadings 111-28 111-28 3.1.4 Heating and Ventilation Heating Ventilation 111-29 111-29 Shielding and Access Control Control 111-29 111-29 3.1.55 3.1. 3.2 Structure Design Structure 111-29 III-29 UFSAR 21 UFSAR Revision 21 iv iv Octob~r October 2009

Nine Mile Point Unit 1 UFSAR CONTENTS (Cont'd.) TABLE OF CONTENTS (Cont'd.) Section Section Title Page 3.2.1 General Structural Structural Features Features 111-29 3.2.2 Heating and Ventilation Ventilation System 111-30 111-30 3.2.3 Access Control Access Control 111-30 111-30 F. F. SCREENHOUSE, SCREENHOUSE, INTAKE AND DISCHARGE TUNNELS 111-31 111-31 1.0 1.0 Screenhouse Screenhouse 111-31 111-31 1.1 Design Basis 111-31 111-31 1.1.1 Wind and Snow Loadings Loadings 111-31 111-31 1.1: 1.1*22 Pressure Relief Pressure Relief Design 111-31 111-31 1.1. 1.1.33 Seismic Seismic Design and Internal Internal Loadings Loadings 111-31 111-31 1.1.4 Heating Heating and Ventilation ventilation 111-31 111-31 1.1. 1.1.55 Shielding and Access Control Shielding 111-31 111-31 1.2 Structure Design Structure 111-31 111-31 2.0 Intake and Discharge Discharge Tunnels 111-33 2.1 Design Bases 111-33 2.2 Structure Design Structure 111-33 111-33 3.0 Safety Analysis Safety Analysis 111-34 G. STACK STACK 111-35 111-35 1.0 Design Bases 111-35 1.1 General General 111-35 1'.2 1.2 Wind Loading 111-35 111-35 1.3 Seismic Design III-35 111-35 1.4 Shielding and Access Access Control III-35 111-35 2.0 Structure Structure Design III-35 111-35 3.0 Safety Analysis III-36 111-36 3.1 Radiology Radiology III-36 111-36 3.2 Stack Failure Failure Analysis 111-37 III-37 3.2.1 Reactor Building Reactor Building 111-37 III-37 3.2.2 Diesel Generator Diesel Generator Building 111-38 III-38 3.2.3 Screen Screen and Pump House 111-38 III-38 H. H. SECURITY SECURITY BUILDING WEST AND SECURITY BUILDING BUILDING ANNEX 111-39 III-39 I.

1. RADWASTE RADWASTE SOLIDIFICATION AND SOLIDIFICATION AND STORAGE BUILDING 111-40 III-40 1.0 Design Bases Design III-40 111-40 1.1 Wind and Snow Loadings Loadings 111-40 III-40 UFSAR Revision UFSAR Revision 21 21 vV October 2009

Nine Mile Point Unit 1 UFSAR (Cont'd.) TABLE OF CONTENTS (Cont'd.) Section Section Title Page 1.2 1.2 Pressure Relief Pressure Relief Design Design 111-40 1.3 Seismic Design and Internal Seismic Internal Loadings Loadings 111-40 111-40 1.4 Heating, Heating, Ventilation Ventilation and AirAir Conditioning Conditioning 111-40 111-40 1.5 1.5 Shielding and Access Control 111-40 2.0 Structure Structure and Design 111-41 2.1 General Structural General Features Structural Features 111-41 111-41 2.2 Heating, Ventilation Heating, Air Ventilation and Air Conditioning Conditioning 111-41 111-41 2.3 Shielding Shielding and Access Control 111-43 3.0 Use 111-43 J. J. REFERENCES REFERENCES 111-45 SECTION IV REACTOR REACTOR IV-1 1V-l A. A. DESIGN BASES IV-l 1V-l 1.0 2.0 3.0 B. B. 1.0 General General Performance Objectives Performance Objectives Design Limits and Targets REACTOR DESIGN REACTOR General IV- 1 1V-l IV-l 1V-l IV-2 1V-2 IV-3 1V-3 IV-3 1V-3 2.0 Nuclear Design Technique Nuclear Technique IV-4 1V-4 2.1 Reference Loading Pattern Reference Loading Pattern IV-5 1V-5 2.2 Final Loading Pattern Pattern IV-6 1V-6 2.2.1 Acceptable Deviation From Reference Acceptable Deviation Reference Loading Pattern Loading Pattern IV-6 1V-6 2.2.2 2.2.2 Reexamination Licensing Basis Reexamination of Licensing IV-6 1V-6 2.3 Refueling Refueling Cycle Cycle Reactivity Reactivity Balance Balance IV-7 IV-7 3.0 Thermal and Hydraulic Thermal Hydraulic Characteristics Characteristics IV-7 1V-7 3.1 Hydraulic Design Thermal and Hydraulic Design IV-7 1V-7 3.1.1 Recirculation Flow Control Recirculation IV-7 1V-7 3.1. 3.1.22 Thermal Limits Core Thermal IV-7 1V-7 3.1.2.1 Excessive Temperature Excessive Clad Temperature IV-8 1V-8 3.1.2.2 Cladding Strain IV-9 1V-9 3.1.2.3 Coolant Flow IV-9 1V-9 3.2 Thermal Hydraulic Analyses Thermal and Hydraulic IV-9 1V-9 Hydraulic Hydraulic Analysis IV-9 1V-9 3.2.1 3.2.2 Thermal Analysis Thermal IV-ll IV-II 21 Revision 21 UFSAR Revision vi vi October 2009

Nine Mile Point Unit 1 1 UFSAR TABLE OF CONTENTS (Cont'd.) CONTENTS (Cont'd.) Section Section Title Page 3.2.2.1 3.2.2.1 Integrity Safety Fuel Cladding Integrity Safety Limit Analysis IV-11 IV-II 3.2.2.2 MCPR Operating Operating Limit Analysis IV-12 IV-12 3.3 Reactor Transients IV-13 IV-13 4.0 Stability Analysis Stability Analysis IV-14 4.1 Design Bases Bases IV-14 4.2 Stability Stability Analysis Method Method IV-14 5.0 Mechanical Mechanical Design and Evaluation Evaluation IV-15 IV-1S 5.1 Fuel Mechanical Mechanical Design IV-15 IV-1S 5.1.1 Bases Design Bases IV-IS IV-15 5.1. 5.1.22 Fuel Fuel Rods IV-IS IV-15 5.1. 5.1.33 Water Rods IV-16 IV-16 5.1.44 5.1. Fuel Assemblies Fuel Assemblies IV-16 IV-16 5.1.55 5.1. Mechanical Mechanical Design Limits and Stress Stress Analysis IV-16 IV-16 5.1.6 5.1. 6 Relationship Between Fuel Design Relationship Between Damage Limits Fuel Damage Limits and Fuel IV-16 IV-16 5.1.7 Surveillance and Testing Surveillance IV-16 IV-16 5.1. 7 6.0 Mechanical Design and Control Rod Mechanical Evaluation Evaluation IV-16 IV-16 6.1 Design IV-16 IV-16 6.1.1 Control Rods and Drives IV-16 6.1.22 6.1. Standby Liquid Poison System IV-19 6.2 Control System System Evaluation Evaluation IV-19 6.2.1 Withdrawal Errors Evaluation Rod Withdrawal Evaluation IV-19 6.2.2 6.2.2 Overall Overall Control Control System System Evaluation Evaluation IV-21 IV-21 6.3 Limiting Conditions Conditions for Operation Operation Surveillance and Surveillance IV-23 6.4 Control Rod Lifetime Control IV-23 7.0 Reactor Vessel Reactor Vessel Internal Internal Structure IV-24 7.1 Design Bases Design IV-24 7.1.1 Core Shroud IV-25 IV-2S 7.1. 7.1.22 Core Support IV-27 7.1.3 Top Grid IV-27 IV-27 7.1.4 Control Rod Guide Guide Tubes IV-27 IV-27 7.1.55 7.1. Feedwater Sparger Feedwater IV-27 7.1.6 7.1. 6 Core Spray Spargers IV-27 IV-27 7.1.77 7.1. Liquid Poison Sparger Sparger IV-28 7.1.8 7.1. 8 Stearn Steam Separator Separator and Dryer IV-28 IV-28 7.1.9 7.1. 9 Core Shroud Stabilizers Stabilizers IV-28 IV-28 7.1.10 Vertical Weld Repair Core Shroud Vertical Repair IV-30 IV-30 7.2 Design Evaluation Evaluation IV-30 IV-30 Revision 21 UFSAR Revision UFSAR 21 vii vii October 2009

Nine Mile Nine Unit 1 UFSAR Point Unit Mile Point UFSAR CONTENTS (Cont'd.) TABLE OF CONTENTS (Cont'd.) section Section Title Title Page 7.3 7.3 Surveillance and Testing Surveillance Testing IV-31 IV-31 C. C. REFERENCES REFERENCES IV-32 IV-32 V SECTION V SECTION REACTOR COOLANT REACTOR COOLANT SYSTEM SYSTEM V-I V-I A. A. DESIGN BASES DESIGN BASES V-I V-I 1.0 1.0 General General V-I V-1 2.0 Performance Objectives Performance Objectives V-I V-i 3.0 Design Pressure Design Pressure V-2 V-2 4.0 Cyclic Loads (Mechanical Cyclic (Mechanical and Thermal) Thermal) V-3 V-3 5.0 5.0 Codes V-3a V-3a B. B. SYSTEM DESIGN AND OPERATION SYSTEM OPERATION V-4 V-4 1.0 General General V-4 V-4 1.1 Drawings V-4 V-4 1.2 Materials of Construction Materials Construction V-4 V-4 1.3 1.4 1.5 2.0 3.0 Thermal Primary Coolant Reactor Reactor Stresses Thermal Stresses Primary Coolant Coolant Leakage Coolant Chemistry Chemistry Vessel Reactor Vessel Leakage Reactor Recirculation Reactor Recirculation Loops Steam and Auxiliary Reactor Steam Auxiliary Systems Systems V-4 V-4 V-S V-S V-6 V-G V-6 V-6 V-7 V-7 4.0 Piping V-8 V-8 5.0 Relief Devices Relief Devices V-8 V-8 C. C. SYSTEM DESIGN EVALUATION EVALUATION V-10 V-10 1.0 1.0 General V-10 V-IO 2.0 Pressure V-IO V-10 3.0 Design Heatup and Cooldown Cooldown Rates V-li V-II 4.0 Materials Radiation Exposure Materials Radiation Exposure V-12 4.1 Pressure-Temperature Pressure-Temperature Limit Curves V-12 4.2 Temperature Temperature Limits for Boltup V-12 4.3 Temperature Temperature Limits for In-Service In-Service System Pressure Pressure Tests V-13 V-13 4.4 Operating Limits During Heatup,Heatup, Cooldown, Operation Cooldown, and Core Operation V-13 4.5 Predicted Predicted Shift in in RT~T RTNDT V-13 4.6 Neutron Fluence Calculations Calculations V-13 Mechanical Mechanical Considerations 5.0 Considerations V-14 5.1 Jet Reaction Forces V-14 UFSAR 21 UFSAR Revision 21 viii viii October October 2009

Nine Mile Point Unit 11 UFSAR Nine TABLE OF CONTENTS (Cont'd.) CONTENTS (Cont'd.) Section Title Page 5.2 Seismic Seismic Forces V-14 6.0 6.0 Safety Limits, Limiting Safety Limits, Limiting Safety Settings and Minimum Conditions Settings for Conditions for Operation Operation V-14 D. D. TESTS AND TESTS AND INSPECTIONS INSPECTIONS V-16 V-16 1.0 Prestartup Prestartup Testing V-16 2.0 Inspection Inspection and Testing Following Startup Startup V-16 V-16 2.1 Pressure Pressure Test V-16 V-16 2.2 Pressure Vessel Irradiation Pressure Irradiation V-16 V-16 E. E. EMERGENCY COOLING SYSTEM EMERGENCY COOLING SYSTEM V-17 V-17 1.0 Design Bases Design V-17 V-17 2.0 System System Design and Operation Operation V-17 3.0 Design Evaluation Evaluation V-19 V-19 3.1 Redundancy Redundancy V-19 V-19 Makeup Water V-19 V-19 3.2 3.3 System System Leaks V-19 V-19 3.4 Containment Isolation Containment Isolation V-19 V-19 4.0 Tests and Inspections Inspections V-20 V-20 4.1 Prestartup Prestartup Test V-20 V-20 4.2 Subsequent Inspections and Tests Subsequent Inspections V-20 V-20 F. F. REFERENCES REFERENCES V-21 V-21 SECTION VI CONTAINMENT SYSTEM CONTAINMENT SYSTEM VI-i VI-1 A. A. PRIMARY CONTAINMENT PRIMARY CONTAINMENT - MARK I CONTAINMENT PROGRAM CONTAINMENT PROGRAM VI-2 VI-2 1.0 1.0 General Structure General Structure VI-2 VI-2 2.0 Pressure Suppression Pressure Suppression Hydrodynamic Hydrodynamic Loads VI-2 VI-2 2.1 Safety/Relief Valve Discharge Safety/Relief VI-2 VI-2 2.2 Loss-of-Coolant Accident Loss-of-Coolant VI-3 VI-3 2.3 Summary of Loading Phenomena Phenomena VI-4 VI-4 3.0 Plant-Unique Plant-Unique Modifications Modifications VI-5 VI-5 B. B. PRIMARY CONTAINMENT CONTAINMENT -- PRESSURE PRESSURE SUPPRESSION SYSTEM VI-6 VI-6 1.0 1.0 Design Bases VI-6 VI-6 1.1 General VI-6 VI-6 UFSAR Revision 21 21 ix lX October October 2009 2009

Nine Nine Mile Point Unit 1 UFSAR TABLE TABLE OF CONTENTS (Cont'd.) CONTENTS (Cont'd.) Section Section Title Page 1.2 Design Basis Accident Design Accident (DBA) VI-6 VI-6 1.3 Containment Heat Removal Containment Heat VI-8 VI-8 1.4 Isolation Isolation Criteria VI-8 VI-8 1.5 Vacuum Relief Criteria Criteria VI-9 VI-9 1.6 Flooding Criteria Criteria VI-9 VI-9 1.7 Shielding Shielding VI-9 VI-9 2.0 Structure Structure Design VI-9 VI-9 2.1 General General VI-9 VI-9 2.2 Penetrations Penetrations and Access Access Openings VI-II VI-ll 2.3 Jet and Missile Protection Protection VI-12 2.4 Materials Materials VI-13 2.5 Shielding Shielding VI-14 2.6 Vacuum Relief Vacuum Relief VI-14 2.7 Containment Containment Flooding VI-14 C. C. SECONDARY CONTAINMENT CONTAINMENT - REACTOR REACTOR BUILDING BUILDING VI-16 1.0 Design Bases Bases VI-16 1.1 1.2 1.3 1.4 1.4 2.0 2.1 Wind and Snow Loadings Pressure Pressure Relief Seismic Design Shielding Structure General Loadings Relief Design Design Structure Design General Structural Features Structural Features VI-16 VI-16 VI-17 VI-17 VI-17 VI-17 D. D. CONTAINMENT SYSTEM CONTAINMENT ISOLATION SYSTEM VI-20 VI-20 1.0 Design Bases VI-20 1.1 Containment Spray Appendix J Water Containment Water Requirements Seal Requirements VI-23a VI-23a 2.0 System Design VI-24 3.0 Inspections Tests and Inspections VI-26 VI-26 E. E. CONTAINMENT VENTILATION CONTAINMENT VENTILATION SYSTEM VI-27 1.0 Primary Containment Containment VI-27 1.1 Design Bases VI-27 VI-27 1.2 System Design System VI-27 VI-27 2.0 Secondary Containment Secondary Containment VI-28 2.1 Bases Design Bases VI-28 2.2 System Design System Design VI-28 F. TEST AND INSPECTIONS INSPECTIONS VI-30 F. VI-30 1.0 Drywell and Suppression Suppression Chamber VI-30 UFSAR Revision 21 21 xX October 2009

Nine Mile Point Unit 11 UFSAR TABLE OF CONTENTS (Cont'd.) CONTENTS (Cont'd.) section Section Title Page 1.1 Preoperational Testing Preoperational VI-30 VI-3~ 1.2 1.2 Postoperational Postoperational Testing VI-3D VI-30 2.0 2.0 Containment Penetrations and Containment Penetrations Isolation Valves Isolation Valves VI-31 2.1 2.1 Penetration Valve Leakage Penetration and Valve Leakage VI-31 VI-31 2.2 2.2 Test Operability Test Valve Operability VI-31 3.0 3.0 Ventilation System Containment Ventilation Containment System VI-32 4.0 4.0 Other Containment Tests Containment Tests VI-32 5.0 5.0 Reactor Building Reactor VI-32 5.1 5.1 Building Normal Reactor Building Normal Ventilation System System VI-32 5.2 5.2 Reactor Building Reactor Isolation Valves Building Isolation Valves VI-32 5.3 5.3 Emergency Ventilation Emergency System Ventilation System VI-33 G. REFERENCES REFERENCES VI-33 SECTION VII SECTION VII ENGINEERED SAFEGUARDS ENGINEERED SAFEGUARDS VII-1 VII-2

• A.

1.0 2.0 2.0 2.1 2.1 2.2 2.2 CORE SPRAY General General SYSTEM SPRAY SYSTEM Bases Design Bases

System Design

Operator Assessment Assessment VII-2 VII-2 VII-2 VII-2 VII-2 VII-2 VII-2 VII-2 VII-5 VII-5 3.0 3.0 Design Evaluation VII-6 VII-7 4.0 4.0 Inspections Tests and Inspections VII-7 B. B. CONTAINMENT SYSTEM CONTAINMENT SPRAY SYSTEM VII-8 VII-8 1.0 1.0 Basis Requirements Licensing Basis Requirements VII-8 VII-8 1.1 10CFR50.49 - Environmental IOCFR50.49 Environmental Qualification Electric Equipment Qualification of Electric Equipment Important to Safety for Nuclear Important Nuclear Power Plants Plants VII-8 VII-8 1.2 1.2 10CFR50 Appendix A - General Design 10CFR50 Appendix Nuclear Power Plants Criteria for Nuclear Criteria VII-8 VII-8 2.0 2.0 Bases Design Bases VII-ll VII-II 2.1 2.1 Functional Requirements Design Basis Functional Requirements VII-ll VII-II 2.2 2.2 Controlling Parameters Controlling Parameters VII-12 3.0 3.0 System Design System Design VII-12a VII-12a 3.1 3.1 System Function System Function VII-12a VII-12a 3.2 3.2 System Design System Design Description VII-12b VII-12b 3.3 3.3 System Design VII-12c VII-12c 3.4 3.4 Codes and Standards Standards VII-14b VII-14b Revision 21 UFSAR Revision 21 xa xa October 2009

Nine Mile Mile Point Unit 1 UFSAR (Cont'd.) CONTENTS (Cont'd.) TABLE OF CONTENTS Section Section Title Page Page 3.5 System Instrumentation Instrumentation VII-14b VII-14b 3.6 System Design Features Features VII-14c VII-14c 4.0 Design Performance Evaluation Design Performance VII-14d VII-14d 4.1 System Performance Performance Analyses Analyses VII-14d VII-14d 4.2 System System Response VII-14e VII-14e 4.3 Interdependency Other Interdependency With Other Engineered Engineered Safeguards Systems Safeguards Systems VII-14e VII-14e 5.0 System Operation System Operation VII-14f VII-14f 5.1 Limiting Conditions Operation Conditions for Operation VII-14f VII-14f 6.0 Tests and Inspection Inspection VII-14g VII-14g C. C. LIQUID POISON POISON INJECTION INJECTION SYSTEM SYSTEM VII-IS VII-15 1.0 Design Bases VII-IS VII-15 2.0 System Design System VII-16 2.1 Operator Assessment Operator VII-19 3.0 Design Evaluation Evaluation VII-20 4.0 Tests and Inspections Inspections VII-21 5.0 Alternate Boron Injection Alternate VII-2Ia VII-21a D. D. 1.0 2.0 3.0 3.1 CONTROL ROD VELOCITY LIMITER CONTROL Design Bases

System Design

Design Evaluation General General Evaluation LIMITER VII-22 VII-22 VII-22 VII-24 VII-24 3.2 Sensitivity Design Sensitivity VII-24 3.3 Normal Operation Operation VII-25 4.0 Tests and Inspections Inspections VII-25 E. E. CONTROL CONTROL ROD HOUSING SUPPORT SUPPORT VII-26 1.0 Design Bases VII-26 2.0 System Design VII-26 2.1 Loads and Deflections Deflections VII-28 3.0 Design Evaluation Design Evaluation VII-28 4.0 Tests and Inspections Inspections VII-29 F. F. FLOW RESTRICTORS VII-30 1.0 1.0 Design Bases Design VII-30 2.0 System Design VII-30 3.0 Design Evaluation Design Evaluation VII-30 VII-30 4.0 Tests and Inspections Inspections VII-31 VII-3I Revision 21 UFSAR Revision 21 xb xb October October 2009 2009

Nine Mile Point Unit 11 UFSAR (Cont'd.) TABLE OF CONTENTS (Cont'd.) Section Title Page G. G. COMBUSTIBLE CONTROL SYSTEM COMBUSTIBLE GAS CONTROL SYSTEM VII-32 VII-32 1.0 Design Bases Bases VII-32 2.0 2.0 Containment Inerting System Containment Inerting System VII-32 2.1 2.1 System Design VII-32 2.2 2.2 Design Evaluation Evaluation VII-33 3.0 3.0 Containment Atmospheric Dilution Containment Atmospheric System System VII-33 3.1 3.1 System Design VII-33 3.2 3.2 Design Evaluation Evaluation VII-35 4.0 4.0 Tests and Inspections Tests Inspections VII-35 VII-35 H. H. EMERGENCY VENTILATION EMERGENCY SYSTEM VENTILATION SYSTEM VII-36 1.0 Design Bases Bases VII-36 2.0 2.0 System Design VII-36 VII-36 2.1 2.1 Operator Assessment Operator Assessment VII-38 3.0 3.0 Design Evaluation Evaluation VII-39 VII-39 4.0 4.0 Tests and Inspections Inspections VII-39 VII-39

• I.

I. 1.0 2.0 3.0 4.0 Design Bases Bases

System Design

COOLANT INJECTION HIGH-PRESSURE COOLANT HIGH-PRESSURE Design Evaluation Design Evaluation Tests and Inspections Inspections VII-41 VII-41 VII-41 VII-41 VII-41 VII-42 VII-43

• UFSAR Revision Revision 21 xc October October 2009

Mile Point Unit 1 UFSAR Nine Mile INTENTIONALLY BLANK THIS PAGE INTENTIONALLY Revision 21 UFSAR Revision 21 xd xd October 2009 2009

Nine Mile Mile Point Unit 11 UFSAR

• section Section TABLE OF CONTENTS Title CONTENTS (Cont'd.)

(Cont'd.) Page J. J. REACTOR COMPONENTS FUEL AND REACTOR COMPONENTS HANDLING HANDLING SYSTEM SYSTEM X-39 X-39 1.0 Design Bases Bases X-39 X-39 2.0 System Design X-40 X-40 2.1 Description of Facility Description Facility X-40 X-40 2.1.1 Cask Drop Protection Protection System System X-42 X-42 2.2 Operation of the Facility Operation Facility X-43 X-43 2.3 Control of Heavy Loads Program X-43a X-43a 2.3.1 Introduction/Licensing Background Introduction/Licensing Background X-43a X-43a 2.3.2 Safety Basis Basis X-43b X-43b 2.3.3 Scope of Heavy Load Load Handling Handling Systems X-43b X-43b 2.3.4 Control of Heavy Loads Program X-43c X-43c 2.3.4.1 NMPNS Commitments in in Response toto NUREG-0612, Phase NUREG-0612, Phase I Elements X-43c X-43c 2.3.4.2 Reactor Pressure Reactor Vessel Head and Pressure Vessel Spent Fuel Cask Lifts X-43d X-43d 2.3.5 Safety Evaluation Evaluation X-43d X-43d 3.0 Design Evaluation Evaluation X-43d X-43d 4.0 Tests and Inspections Inspections X-44 X-44 K. FIRE PROTECTION PROTECTION PROGRAM X-45 X-45 1.0 1.0 Program Program Bases X-45 X-45 1.1 Nuclear Division Nuclear Division Directive Directive - Fire Protection Program Protection Program X-45 X-45 1.2 Nuclear Division Nuclear Division Interface Interface Procedure - Fire Procedure Protection Program Fire Protection Program X-45 X-45 1.3 Fire Hazards Hazards Analysis X-45 X-45 1.4 Appendix Shutdown Appendix R Review Safe Shutdown Analysis X-46 X-46 1.5 Fire Protection Protection and Appendix Appendix R Related Portions of Operations Related Operations Procedures (OPs, SOPs, and EOPs) Procedures (OPs, SOPs, EOPs) and Damage Damage Repair Procedures Repair Procedures X-46 X-46 1.6 1.6 Fire Protection Portions of the Protection Portions Emergency Plan Emergency X-46 X-46 2.0 Program Implementation Implementation and Design Aspects X-46 X-46 2.1 Protection Implementing Fire Protection Implementing Procedures Procedures X-46 X-46 2.2 Protection Administrative Fire Protection Administrative

• UFSAR Revision Controls 21 Revision 21 xvii xvii X-47 October October 2009

Mile Point Unit 11 UFSAR Nine Mile UFSAR Section Title Title TABLE OF CONTENTS (Cont'd.) (Cont'd.) Page 2.3 Fire Protection Protection System Drawings and Calculations Calculations X-47 X-47 2.4 Fire Protection Protection Engineering (FPEEs) Evaluations (FPEEs) Evaluations X-47 X-47 3.0 3.0 Monitoring Monitoring and Evaluating Evaluating Program Program Implementation Implementation X-47 X-47 3.1 Quality Assurance Quality Assurance Program Program X-47 X-47 3.2 Fire Brigade Brigade Manning, Training, Manning, Training, Drills and Responsibilities Responsibilities X-47 X-47 4.0 Surveillance Surveillance and Tests X-48 X-48 L. L. REMOTE SHUTDOWN REMOTE SHUTDOWN SYSTEM X-49 X-49 1.0 Design Bases Design Bases X-49 X-49 2.0 System Design X-49 X-49 3.0 3.0 System Evaluation Evaluation X-49 X-49 4.0 Tests and Inspections Inspections X-50 X-50 M. M. HYDROGEN WATER HYDROGEN WATER CHEMISTRY CHEMISTRY AND NOBLE NOBLE METAL CHEMICAL METAL (NOBLECHEM) CHEMICAL ADDITION (NOBLECHEM) SYSTEMS X-51 X-51 1.0 1.0 Design Basis X-51 X-51 1.1 Noble Chemical Addition Noble Metal Chemical Addition System X-52 X-52 1.2 Hydrogen Hydrogen Water Chemistry Chemistry System X-52a X-52a 2.0 System Design System X-53 2.1 Noble Metal Chemical Noble Chemical Addition Addition X-53 2.2 Hydrogen Hydrogen Water Chemistry System X-53 2.2.1 HWC Feedwater Feedwater Hydrogen Hydrogen Injection X-54 2.2.2 HWC Offgas Oxygen Injection Injection X-54 2.2.3 HWC Offgas Sample X-54 3.0 System Evaluation System Evaluation X-54 4.0 Tests and Inspections Inspections X-55 N. N. REFERENCES REFERENCES X-56 X-56 lOA APPENDIX 10A APPENDIX FIRE HAZARDS ANALYSIS FIRE ANALYSIS lOB APPENDIX 10B APPENDIX SAFE SHUTDOWN ANALYSIS SHUTDOWN ANALYSIS SECTION XI STEAM-TO-POWER CONVERSION SYSTEM STEAM-TO-POWER CONVERSION XI-I XI-1 A. A. DESIGN BASES BASES XI-I XI-1 Revision 21 UFSAR Revision 21 xviia October 2009

Nine Mile Point Unit Unit 1 UFSAR

• Section Section Title (Cont'd.)

TABLE OF CONTENTS (Cont'd.) Page B. B. SYSTEM DESIGN AND OPERATION OPERATION XI-2 XI-2 1.0 Turbine Generator Turbine Generator XI-2 XI-2 2.0 Turbine Condenser Turbine Condenser XI-4 XI-4 3.0 Removal and Offgas Condenser Air Removal Offgas System XI-5 XI-5 4.0 Circulating Water System Circulating Water XI-9a 5.0 Condensate Condensate Pumps XI-9a XI-9a 6.0 Condensate Filtration Condensate Filtration System XI-9b XI-9b 6.0A 6.OA Condensate Demineralizer System Demineralizer System XI-9c 7.0 Condensate Transfer Condensate Transfer System XI-ll XI-11 8.0 Feedwater Feedwater Booster Pumps XI-ll XI-l1 9.0 Feedwater Feedwater Pumps XI-lla 10.0 Feedwater Heaters Feedwater Heaters XI-12 XI-l2 C. C. SYSTEM ANALYSIS XI-13 XI-l3 D. D. TESTS AND INSPECTIONS INSPECTIONS XI-16

• SECTION XII SECTION A.

A. 1.0 1.1 RADIOLOGICAL RADIOLOGICAL CONTROLS RADIOACTIVE Design Bases Objectives Objectives CONTROLS WASTES RADIOACTIVE WASTES XII-1 XII-l XII-1 XII-l XII-1 XII-l XII-l XII-l 1.2 1.2 Types of Radioactive Radioactive Wastes XII-1 XII-l

• UFSAR 21 UFSAR Revision 21 xviib October 2009

Nine Mile Point Unit 1I UFSAR TABLE OF CONTENTS (Cont'd.) CONTENTS (Cont'd.) Section Section Title Page

1. 2.1 1.2.1 Gaseous Waste XII-1 XII-1

1. 2.2 1.2.2 Liquid Wastes XII-1 XII-I

1. 2.3 1.2.3 Solid Wastes Wastes XII-2 XII-2 2.0 System Design and Evaluation Evaluation XII-2 XII-2 2.1 Gaseous Waste System XII-2 XII-2 2.1.1 Offgas System System XII-3 XII-3 2.1.

2.1.2 2 Steam-Packing Steam-Packing Exhausting Exhausting System XII-3 XII-3 2.1. 2.1.33 Building Ventilation Ventilation Systems XII-3 XII-3 2.1.4 Stack XII-3 XII-3 2.2 Liquid Waste Waste System XII-4 XII-4 2.2.1 Liquid Waste Waste Handling Handling Processes Processes XII-4 XII-4 2.2.2 Sampling Sampling and Monitoring Monitoring Liquid Wastes XII-6 XII-6 2.2.3 2.2.3 Liquid Waste Waste Equipment Equipment Arrangement Arrangement XII-7 XII-7 2.2.4 Liquid Radioactive Radioactive Waste System System Control XII-7 XII-7 2.3 Solid Waste System System XII-8 XII-8 Solid Waste Handling Processes Processes 2.3.1 XII-8 XII-8 2.3.2 Solid Waste System Equipment Equipment XII-9 XII-9 2.3.3 Process Control Program Process XII-9a XII-9a 3.0 Safety Limits Safety Limits XII-9a XII-9a 4.0 Tests and Inspections Inspections XII-I0 XII-10 4.1 Waste Process Systems XII-10 4.2 Filters Filters XII-I0 XII-10 4.3 Effluent Monitors Effluent Monitors XII-I0 XII-10 4.3.1 Offgas and Stack Monitors Stack Monitors XII-10 4.3.2 Liquid Waste Effluent Effluent Monitor XII-10 B. B. RADIATION PROTECTION RADIATION PROTECTION XII-11 XII-II 1.0 Primary and Secondary Primary Secondary Shielding Shielding XII-l1 XII-II 1.1 Design Bases XII-II XII-!I 1.2 Design XII-12

1. 2.1 1.2.1 Reactor Shield Shield Wall XII-12

1. 2.2 1.2.2 Biological Biological Shield XII-12

1. 2.3 1.2.3 Miscellaneous Miscellaneous XII-12 1.3 Evaluation Evaluation XII-13 2.0 Area Radioactivity Radioactivity Monitoring Systems systems XII-13 2.1 Area Radiation Radiation Monitoring Monitoring System XII-13 2.1.1 2.1.1 Design Bases Bases XII-13 2.1.22 2.1. Design XII-14 2.1.

2.1.33 Evaluation Evaluation XII-15 UFSAR Revision 20 20 xviii xviii October October 2007

Nine Mile Nine Unit 11 UFSAR Point Unit Mile Point UFSAR TABLE OF TABLE CONTENTS (Cont'd.) OF CONTENTS (Cont'd.) Section Section Title Title Page 2.2 2.2 Area Air Area Contamination Monitoring Air Contamination Monitoring System System XII-16 XII-16 2.2.1 2.2.1 Design Bases Design Bases XII-16 XII-16 2.2.2 2.2.2 Design Design XII-16 XII-16 2.2.3 2.2.3 Evaluation Evaluation XII-16a XII-16a 3.0 3.0 Radiation Protection Radiation Protection XII-17 XII-17 3.1 3.1 Facilities Facilities XII-17 XII-17 3.1.1 3.1.1 Laboratory, Room and Counting Room Laboratory, Counting and Calibration Facilities Calibration Facilities XII-17 XII-17 3.1.2 3.1. 2 Change Room Change Room and Shower Facilities and Shower Facilities XII-18 XII-18 3.1.3 3.1. 3 Personnel Decontamination Facility Personnel Decontamination Facility XII-19 XII-19 3.1.4 3.1.4 Tool and Tool Equipment Decontamination and Equipment Decontamination Facility Facility XII-19 XII-19 3.2 3.2 Radiation Control Radiation Control XII-20 XII-20 3.2.1 3.2.1 Shielding Shielding XII-20 XII-20 3.2.2 3.2.2 Access Control Access Control XII-20 XII-20 3.3 3.3 Contamination Control Contamination Control XII-21a XII-21a 3.3.1 3.3.1 Facility Contamination Control Facility Contamination Control XII-21a XII-21a

• 3.3.2 3.3.2 3.3.3 3.3.3 3.4 3.4 3.4.1 3.4.1 3.5 3.5 Radiation Dose Radiation Contamination Control Personnel Contamination Personnel Airborne Contamination Airborne Personnel Dose Personnel Dose Radiation Protection Radiation Instrumentation Control Contamination Control Control Dose Determinations Determinations Protection XII-22 XII-22 XII-23 XII-23 XII-24 XII-24 XII-24 XII-24 XII-24 XII-24 Instrumentation 3.5.1 3.5.1 Counting Room Counting Room Instrumentation Instrumentation XII-24 XII-24 3.5.2 3.5.2 Portable Radiation Instrumentation Portable Radiation Instrumentation XII-25 XII-25 3.5.3 3.5.3 Sampling Instrumentation Air Sampling Air Instrumentation XII-25 XII-25 3.5.4 3.5.4 Personnel Monitoring Instruments Personnel Monitoring Instruments XII-25 XII-25 3.5.5 3.5.5 Emergency Instrumentation Emergency Instrumentation XII-25 XII-25 4.0 4.0 Tests and Tests and Inspections Inspections XII-26 XII-26 4.1 4.1 Shielding Shielding XII-26 XII-26 4.2 4.2 Area Radiation Monitors Area Radiation Monitors XII-27 XII-27 4.3 4.3 Area Air Contamination Monitors Area Air Contamination Monitors XII-27 XII-27 4.4 4.4 Radiation Protection Facilities Radiation Protection Facilities XII-27 XII-27 4.4.1 4.4.1 Ventilation Air Ventilation Air Flows Flows XII-27 XII-27 4.4.2 4.4.2 Instrument Calibration Well Instrument Calibration Well Shielding Shielding XII-28 XII-28 4.5 4.5 Radiation Protection Radiation Protection Instrumentation Instrumentation XII-28 XII-28
• Revision 21 UFSAR Revision UFSAR 21 xix xix October 2009 October 2009

Nine Mile Point Unit 1 UFSAR TABLE CONTENTS (Cont'd.) TABLE OF CONTENTS (Cont'd.) section Section SECTION XIII Title CONDUCT OF OPERATIONS CONDUCT XIII-I* XIII-l A. A. ORGANIZATION RESPONSIBILITY ORGANIZATION AND RESPONSIBILITY XllI-l XIII-l 1.0 Management Management and Technical Technical Support Support Organization Organization XllI-l XIII-l 1.1 Station Organization Station Organization XllI-l XIII-l 1.1.1 President Nine Mile Point Vice President XIII-l XllI-l 1.1.22 1.1. Matrixed Matrixed Reporting Reporting XllI-l XIII-l 1.1.33 1.1. Qualifications Qualifications of Support Personnel Personnel XIII-2 XIII-2 2.0 Nine Mile Mile ,point Point Nuclear Nuclear Station, Station, LLC, LLC, Organization Organization XIII-2 XIII-2 2.1 Plant General Manager Manager XIII-3 XIII-3 2.2 Other Functions Other Functions Reporting Reporting to the President Nine Mile Point Vice President XIII-6 XIII-6 2.3 Supervisor Engineering-Nuclear Fuels Supervisor Engineering-Nuclear XIII-8 XIII-8 3.0 Quality Quality Assurance XIII-9 XIII-9 4.0 Operating Shift Crews Operating Shift Crews XIII-9 XIII-9 5.0 Qualifications Qualifications of Staff Personnel Personnel XIII-I0 XIII-10 B. B. 1.0 1.0 2.0 3.0 QUALIFICATIONS QUALIFICATIONS AND TRAINING PERSONNEL PERSONNEL This section section deleted This section section deleted This section section deleted TRAINING OF XIII-ll XIII-II XIII-II XIII-II XIII-II XIII-11 XIII-II XIII-1l 4.0 Training of Personnel Training Personnel XIII-II XIII-ll 4.1 General Responsibility Responsibility XIII-II XIII-ll 4.2 Implementation Implementation XIII-II XIII-II 4.3 Quality XIII-II XIII-I1 4.3.1 For Operator Operator Training Training XIII-1I XIII-II 4.3.2 For Maintenance Maintenance XIII-12 4.3.3 For Technicians Technicians XIII-12 4.3.4 For General General Employee Employee Training/Radiation Protection-and Training/Radiation Protection-and Emergency Plan Emergency Plan XIII-12 4.3.5 For Industrial Industrial Safety XIII-12 4.3.6 For Nuclear Nuclear Quality Quality Assurance Assurance XIII-12 4.3.7 For Fire Brigade XIII-12 4.3.8 For Manager Manager Operations Operations and General General Operations Supervisor Operations XIII-12 21 Revision 21 UFSAR Revision xx xx October 2009

• Nine Mile Point Unit 1 UFSAR CONTENTS (Cont'd.)

TABLE OF CONTENTS (Cont'd.)

• Section 4.4 4.4 Title Training of Licensed Operator Operator Page Candidates/Licensed NRC Operator Candidates/Licensed Operator Retraining XIII-12 5.0 5.0 Cooperative Local, Cooperative Training With Local, Federal Officials State and Federal State Officials XIII-14 C.

C. OPERATING PROCEDURES OPERATING PROCEDURES XIII-15 D. EMERGENCY PROCEDURES EMERGENCY PLAN AND PROCEDURES XIII-17 E. E. SECURITY XIII-19 F. F. RECORDS RECORDS XIII-20 1.0 Operations Operations XIII-20 1. 1.1 Control Room Log Log XIII-20 1.2 1.2 Shift Manager's Log Manager's Log XIII-20 1.3 1.3 Radwaste Log Radwaste Log XIII-20 1.4 1.4 Waste Quantity Level Waste Quantity Shipped Level Shipped XIII-20 2.0 2.0 Maintenance Maintenance XIII-21 3.0 3.0 Radiation Protection XIII-21 3.1 3.1 Personnel Exposure Personnel Exposure XIII-21 3.2 3.2 By-Product Material Material as Required by 10CFR30 10CFR30 XIII-21 3.3 3.3 Meter Calibrations XIII-21 3.4 3.4 Station Radiological Conditions Station Radiological in Conditions in Accessible Areas Accessible Areas XIII-21

• UFSAR Revision 21 21 xxa xxa 2009 October 2009

Mile Point unit Nine Mile Unit 1 UFSAR 0 THIS PAGE THIS PAGE INTENTIONALLY INTENTIONALLY BLANK BLANK UFSAR Revision Revision 17 17 xxb xxb October 2001 October 2001

• Nine Mile Point Unit 11 UFSAR TABLE OF CONTENTS (Cont'd.)

(Cont'd.) Section Title 3.5 3.5 Administration of the Radiation Administration Radiation Protection Program and Procedures Protection Procedures XIII-21 4.0 Chemistry Chemistry and Radiochemistry Radiochemistry XIII-21 5.0 Special Special Nuclear Materials Materials XIII-22 6.0 Calibration Instruments Calibration of Instruments XIII-22 7.0 Administrative Records Administrative Records and Reports Reports XIII-22 G. G. REVIEW AND AUDIT OF OPERATIONS REVIEW OPERATIONS XIII-23 1.0 Plant Operations Review Committee Operations Review Committee XIII-23 1.1 1.1 Function Function XIII-23 2.0 Nuclear Safety Nuclear Review Board Safety Review XIII-23 2.1 Function Function XIII-24 3.0 Review of Operating Operating Experience Experience XIII-24 SECTION XIV SECTION INITIAL TESTING TESTING AND OPERATIONS OPERATIONS XIV-1 XIV-1 A. A. TESTS PRIOR INITIAL REACTOR PRIOR TO INITIAL FUELING FUELING XIV-1 XIV-1

• B.

B. 1.0 1.0 1.1 INITIAL CRITICALITY General CRITICALITY AND POSTCRITICALITY POSTCRITICALITY TESTS Initial Initial Fuel Loading and Near-Zero Fuel Atmospheric Pressure Power Tests at Atmospheric Requirements General Requirements Pressure XIV-5 XIV-5 XIV-5 XIV-5 XIV-5 XIV-5 1.2 General Procedures General Procedures XIV-5 XIV-5 1.3 Core Loading Loading and Critical Critical Test Test Program XIV-7 XIV-7 2.0 Heatup from Ambient to Rated Heatup Rated Temperature Temperature XIV-9 XIV-9 2.1 General XIV-9 XIV-9 2.2 Tests Conducted XIV-9 XIV-9 3.0 From ZeroZero to 100 Percent Initial Percent Initial Reactor Rating Rating XIV-10 XIV-10 4.0 Full-Power Demonstration Full-Power Demonstration Run XIV-12 XIV-12 5.0 Comparison of Base Conditions Comparison Conditions XIV-12 XIV-12 6.0 Additional Additional Tests at Design Rating Design XIV-13 XIV-13 7.0 Startup Report Startup Report XIV-13 XIV-13 SECTION XV SAFETY ANALYSIS SAFETY XV-l XV-1 A. A. INTRODUCTION INTRODUCTION XV-1 XV-1

• 21 UFSAR Revision 21 Xxi XXl 2009 October 2009

Nine Mile Point Unit 1 UFSAR 0 TABLE OF CONTENTS CONTENTS (Cont'd.) (Cont'd.) Section Title Title Page B. B. BOUNDARY PROTECTION BOUNDARY PROTECTION SYSTEMS XV-2 XV-2 1.0 Transients Considered Transients Considered XV-2 XV-2 2.0 Methods and Assumptions Assumptions XV-3 XV-3 3.0 Transient Transient Analysis XV-3 XV-3 3.1 Turbine Trip Without Bypass Turbine Bypass XV-3 XV-3 3.1.1 Objectives Objectives XV-3 XV-3 3.1.2 Assumptions and Initial Initial Conditions Conditions . XV-3 XV-3 3.1. 3.1.33 Comments XV-3 XV-3 3.1.4 Results Results XV-3 XV-3 3.2 Loss of 100°F Feedwater Heating 100OF Feedwater XV-4 XV-4 3.2.1 Objective Objective XV-4 XV-4 3.2.2 Assumptions and Initial Initial Conditions Conditions XV-4 XV-4 3.2.3 Results XV-4 XV-4 3.3 Feedwater Feedwater Controller Failure-Controller Failure-Maximum Demand XV-5 XV-5 3.3.1 Objective Objective XV-5 XV-5 3.3.2 Assumptions Assumptions and Initial Initial Conditions XV-5 XV-5 3.3.3 Comments XV-5 XV-5 3.3.4 Results Results XV-5 XV-5 3.4 Control Rod Withdrawal Withdrawal Error Error XV-5 XV-5 3.4.1 Objective Objective XV-5 XV-5 3.4.2 Assumptions Assumptions and Initial Initial Conditions XV-5 XV-5 3.4.3 Comments XV-6 XV-6 3.4.4 3.4.4 Results XV-6 XV-6 3.5 Steam Line Isolation Valve Main Steam Valve Closure (With Scram) XV-6 XV-6 3.5.1 3.5.1 Objective Objective XV-6 XV-6 3.5.2 3.5.2 Assumptions and Initial Assumptions Initial Conditions XV-6 XV-6 3.5.3 3.5.3 Comments XV-7 XV-7 3.5.4 3.5.4 Results Results XV-7 XV-7 3.6 3.6 Inadvertent Startup of Cold Inadvertent Cold Recirculatibn Loop Recirculatiion Loop XV-7 XV-7 3.6.1 Objective Objective XV-7 XV-7 3.6.2 Assumptions and Initial Assumptions Initial Conditions XV-7 XV-7 3.6.3 Comment XV-8 XV-8 3.6.4 Results Results XV-9 XV-9 3.7 Recirculation Recirculation Pump Trips XV-9 XV-9 3.7.1 Objectives Objectives XV-9 XV-9 3.7.2 Assumptions Initial Assumptions and Initial Conditions Conditions XV-9 XV-9 3.7.3 Comments XV-9 XV-9 3.7.4 Results Results XV-10 XV-I0 3.8 Recirculation Pump Recirculation Pump Stall XV-10 XV-I0 UFSAR Revision 16 Revision 16 xxii xxii November 1999 1999

Nine Mile Point Unit 1 UFSAR (Cont'd.) TABLE OF CONTENTS (Cont'd.) Section Title Page 2.5 1.2.5

1. Subcooled Subcooled Liquid XV-30 XV-30 2.6 1.2.6

1. System Pressure Pressure and Steam-Water Steam-Water Mass XV-31 XV-31 2.7 1.2.7

1. Mixture Impact Impact Forces XV-31 XV-31 2.8 1.2.8

1. Core Internal Internal Forces XV-31 XV-31 1.3 Radiological Radiological Effects Effects XV-32 XV-32 3.1 1.3.1

1. Radioactivity Radioactivity Releases Releases XV-32 XV-32 3.2 1.3.2

1. Meteorology Meteorology and Dose Rates XV-33 XV-33 2.0 Loss-of-Coolant Loss-of-Coolant Accident XV-33 XV-33 2.1 Introduction Introduction XV-33 XV-33 2.2 Input to Analysis Input Analysis XV-35 XV-35 2.2.1 Operational and ECCS Input Operational Input Parameters Parameters XV-35 XV-35 2.2.2 Single Failure Study Single Study on ECCS ECCS Manually-Controlled Manually-Controlled Electrically-Operated Valves Electrically-Operated Valves XV-35 XV-35 2.2.3 2.2.3 Single Failure Basis Single XV-35 XV-35 XV-36 2.2.4 Pipe Whip Basis Basis XV-36 2.3 This section section deleted deleted XV-36 XV-36 2.4 Appendix K LOCA Appendix LOCA Performance Performance Analysis XV-36 XV-36 2.4.1 Computer Codes XV-36 XV-36 2.4.2 Description of Model Changes Description Changes XV-37 XV-37 2.4.3 Analysis Procedure Procedure XV-37 XV-37 2.4.3.1 2.4.3.1 BWR/2 Generic BWR/2 Generic Analysis Analysis XV-37 XV-37 2.4.3.2 2.4.3.2 I-Specific Analysis Unit 1-Specific Analysis Break Evaluation Spectrum Evaluation XV-38 XV-38 2.4.4 Analysis Results XV-38 XV-38 3.0 Refueling Accident Refueling Accident XV-40 XV-40 3.1 Identification Identification of Causes XV-40 XV-40 3.2 Accident Analysis XV-41 XV-41 3.3 Radiological Effects Radiological XV-44 XV-44 3.3.1 Fission Product Releases Releases XV-44 XV-44 3.3.2 Meteorology and Dose Rates Meteorology Rates XV-45 XV-45 4.0 Control Rod Drop Accident Accident XV-46 XV-46 4.1 Identification Identification of Causes Causes XV-46 XV-46 4.2 Accident Accident Analysis XV-46 XV-46 4.3 Designed Safeguards Safeguards XV-47 XV-47 4.4 Procedural Safeguards Procedural Safeguards XV-48 XV-48 4.5 Radiological Effects Radiological Effects XV-48 XV-48 4.5.1 Fission Product Releases Fission Product Releases XV-48a XV-48a 4.5.2 Meteorology and Dose Rates Meteorology XV-49 XV-49 UFSAR Revision 21 21 xxv

                                  *xYv              October 2009 October

Nine Mile Point Unit 11 UFSAR TABLE OF CONTENTS (Cont'd.) (Cont'd.) Section Title Title Page Page 5.0 Containment Containment Design Basis Accident Accident XV-50 XV-50 5.1 Original Recirculation Original Recirculation Line Rupture With. COre Analysis - With Spray Core Spray XV-50 XV-50 5.l.1 5.1.1 Purpose XV-50 XV-50 5.1.2 5.l.2 Analysis Method Method and Assumptions Assumptions XV-51 XV-51 5.1.33 5.l. Core Heat Buildup Buildup XV-51 XV-51 5.1.4 5.l.4 Core Spray Spray System XV-52 XV-52 5.1.55 5.1. Containment Pressure Immediately Containment Pressure Immediately Following Blowdown Following Blowdown XV-53 XV-53 5.1. 5.1.66 Containment Containment Spray XV-54 XV-54 5.1. 7 5.1.7 Blowdown Effects Blowdown Effects on Core Components XV-55 XV-55 5.1. 8 5.1.8 Radiological Effects Radiological XV-56 XV-56 5.1.8.1 5.1.8.1 Fission Product Releases Releases XV-56 XV-56 5.1.8.2 5.1.8.2 Meteorology Meteorology and Dose Rates Rates XV-58 XV-58 5.2 Original Containment Containment Design Design Basis Basis Accident Accident Analysis - Without Core Without Core Spray XV-59 XV-59 5.2.1 Purpose XV-59 XV-59 5.2.2 5.2.3 5.3 5.3.1 5.3.1 5.3.2 Core Heatup Containment Response Containment Response Design Basis Reconstitution Reconstitution Suppression Chamber Heatup Suppression Introduction Introduction Input to Analysis Heatup Analysis XV-59 XV-59 XV-60 XV-60 XV-61 XV-61 XV-61 XV-61 XV-61a XV-61a 5.3.3 5.3.3 DBR Suppression Chamber Heatup Suppression Chamber Analysis XV-61a XV-61a 5.3.3.1 5.3.3.1 Computer Codes XV-62 XV-62 5.3.3.2 5.3.3.2 Analysis Methods XV-62 XV-62 5.3.3.3 Analysis Results for Containment Containment Spray Design Design Basis Basis Assumptions XV-64 XV-64 5.3.3.4 Analysis Results for EOPEOP Operation Assumptions Operation Assumptions XV-65 XV-65 5.3.4 Conclusions XV-65a XV-65a 6.0 New Fuel Bundle Loading Loading Error Error Analysis XV-66 XV-66 6.1 Identification Identification of Causes XV-66 XV-66 6.2 Accident Accident Analysis XV-67 XV-67 6.3 Safety Requirements Safety Requirements XV-67 XV-67 7.0 Meteorological Models Used in Meteorological in Analyses Accident Analyses XV-68 XV-68 Introduction Introduction XV-68 XV-68 7.1 Revision 21 UFSAR Revision 21 xxvi xxvi 2009 October 2009

Nine Mile Nine Unit 11 UFSAR Point Unit Mile Point UFSAR TABLE OF TABLE CONTENTS (Cont'd.) OF CONTENTS (Cont'd.) Section Section Title Title Page 7.2 7.2 Dispersion Factor Atmospheric Dispersion Atmospheric Fact6r Calculations Calculations XV-68 XV-68 7.2.1 7.2.1 Offsite -- EAB Offsite EAB and and LPZ LPZ XV-69 XV-69 7.2.2 7.2.2 Control Room Control Room and Technical Support and Technical Support Center (Excluding Center (Excluding MSLB) MSLB) XV-69 XV-69 7.2.3 7.2.3 Control Room Control Puff Release MSLB Puff Room -- MSLB Release XV-70 XV-70 7.3 7.3 Summary of Summary of Results Results XV-70 XV-70 7.4 7.4 Exfiltration Exfiltration XV-70 XV-70 7.5 7.5 Containment Drawdown Secondary Containment Secondary Drawdown XV-76 XV-76 7.5.1 7.5.1 Introduction Introduction XV-76 XV-76 7.5.2 7.5.2 Analysis Analysis XV-76 XV-76 7.5.3 7.5.3 Results Results XV-77 XV-77

• UFSAR Revision UFSAR 21 Revision 21 xxvia xxvia 2009 October 2009 October

Nine Mile Point Unit unit 1 UFSAR UFSAR 0 THIS PAGE THIS BLANK INTENTIONALLY BLANK PAGE INTENTIONALLY Revision 18 UFSAR Revision 18 xxvib xxvib October October 2003 2003

• Nine Mile Point Unit 11 UFSAR UFSAR
• Section 5.2 5.2 Title CONTENTS (Cont'd.)

TABLE OF CONTENTS (Cont'd.) Supplement 1, NUREG-0737, Supplement NUREG-0737, I, Section Page 4.1.b 4.1.b XVIII-13 5.2.1 5.2.1 Convenient Location Convenient Location XVIII-13 5.2.2 5.2.2 continuous Display Continuous XVIII-13 XVIII-13 5.3 5.3 NUREG-0737, Supplement NUREG-0737, Supplement I,1, Section 4.1.c 4.1. c XVIII-13 5.3.1 5.3.1 Procedures Procedures and Training XVIII-13 5.3.2 5.3.2 Isolation Isolation of SPDS from from Safety-Related Safety-Related Systems Systems XVIII-13 5.4 5.4 NUREG-0737, Supplement NUREG-0737, Supplement 1,I, Section 4.l.e 4.1. e XVIII-14 5.4.1 5.4.1 Incorporation Accepted Human Incorporation of Accepted Human Factors Engineering Engineering Principles Principles XVIII-14 5.4.2 5.4.2 Information Readily Information Can Be Readily Perceived Perceived and Comprehended Comprehended XVIII-14 XVIII-14 5.5 5.5 NUREG-0737, Supplement NUREG-0737, Supplement 1,I, Section 4.l.f, 4.1.f, Information Sufficient Information Sufficient XVIII-15 XVIII-IS 6.0 6.0 Procedures Procedures XVIII-15 XVIII-IS 6.1 6.1 Operating Procedures Operating Procedures XVIII-15 6.2 6.2 Surveillance Procedures Surveillance Procedures XVIII-15 XVIII-IS 7.0 7.0 References References XVIII-16 XVIII-16 APPENDIX A APPENDIX Unused Unused APPENDIX B APPENDIX NINE MILE POINT NUCLEAR NUCLEAR STATION, LLC, QUALITY STATION, LLC, QUALITY PROGRAM ASSURANCE PROGRAM TOPICAL REPORT, REPORT, NINE MILE POINT STATION UNITS 1 AND 2 OPERATIONS NUCLEAR STATION OPERATIONS PHASE PHASE APPENDIX C APPENDIX LICENSE RENEWAL SUPPLEMENT - AGING MANAGEMENT RENEWAL SUPPLEMENT MANAGEMENT TIME-LIMITED AGING ANALYSES PROGRAMS AND TIME-LIMITED ANALYSES

• UFSAR Revision UFSAR Revision 20 xxxiii October 200 7 October 2007

Nine Mile Point Unit 1 UFSAR LIST OF TABLES TABLES Table Table Number Title I-i I-I COMPARISON TO STANDARDS COMPARISON STANDARDS -- HISTORICAL (PROVIDED HISTORICAL (PROVIDED WITH APPLICATION APPLICATION TO CONVERT CONVERT TO FULL-TERM FULL-TERM OPERATING LICENSE) OPERATING LICENSE) 1-2 ABBREVIATIONS AND ACRONYMS USED IN UFSAR ABBREVIATIONS 11-1 11-1 1980 POPULATION POPULATION AND POPULATION POPULATION DENSITY DENSITY FOR TOWNS TOWNS CITIES WITHIN 12 MILES OF NINE MILE POINT - AND CITIES UNIT 1 11-2 CITIES WITHIN WITHIN A 50-MILE 50-MILE RADIUS OF THE STATION WITH POPULATIONS 10,000 POPULATIONS OVER 10,000 11-3 REGIONAL AGRICULTURAL REGIONAL USE AGRICULTURAL USE 11-4 REGIONAL AGRICULTURAL REGIONAL AGRICULTURAL STATISTICS - CATTLE CATTLE AND MILK PRODUCTION PRODUCTION 11-5 11-6 INDUSTRIAL FIRMS WITHIN PUBLIC UTILITIES WITHIN 8 KM (5(5 MI) UTILITIES IN OSWEGO COUNTY MI) OF UNIT 1 11-7 PUBLIC WATER SUPPLY DATA FOR LOCATIONS LOCATIONS WITHIN AN 30-MILE RADIUS APPROXIMATE 30-MILE APPROXIMATE 11-8 RECREATIONAL AREAS IN THE REGION RECREATIONAL 11-9 SOURCES OF TOXIC CHEMICALS WITHIN 88 KM (5 (5 MI) OF MI) OF UNIT 1 SITE SITE II-10 11-10 PREDICTED VAPOR PREDICTED VAPOR CONCENTRATION CONCENTRATION IN THE UNIT 1 CONTROL ROOM CONTROL V-I V-1 REACTOR COOLANT COOLANT SYSTEM DATA V-2 OPERATING CYCLES AND TRANSIENT OPERATING ANALYSIS RESULTS TRANSIENT ANALYSIS V-3 FATIGUE RESISTANCE RESISTANCE ANALYSIS ANALYSIS V-4 CODES FOR SYSTEMS FROM REACTOR REACTOR VESSEL CONNECTION VALVE ISOLATION VALVE TO SECOND ISOLATION UFSAR Revision Revision 21 21 xxxiv xxxiv October 2009 2009

Nine Mile Point Unit 11 UFSAR Nine TABLES LIST OF TABLES (Cont'd.) (Cont'd.) Table Table Number Title Title V-5 TIME TO AUTOMATIC BLOWDOWN AUTOMATIC BLOWDOWN VI-1 VI-l DRYWELL PENETRATIONS DRYWELL PENETRATIONS VI-2 SUPPRESSION CHAMBER PENETRATIONS SUPPRESSION CHAMBER PENETRATIONS VI-3a REACTOR COOLANT SYSTEM ISOLATION COOLANT SYSTEM ISOLATION VALVES VALVES VI-3b PRIMARY CONTAINMENT PRIMARY ISOLATION VALVES - LINES CONTAINMENT ISOLATION LINES ENTERING FREE SPACE ENTERING OF THE CONTAINMENT CONTAINMENT VI-4 SEISMIC DESIGN SEISMIC DESIGN CRITERIA FOR ISOLATION ISOLATION VALVES VALVES VI-5 VI-5 INITIAL INITIAL TESTS PRIOR TO STATION STATION OPERATION OPERATION VII-1 VII-l PERFORMANCE PERFORMANCE TESTS

• VIII-1 VIII-l VIII-2 VIII-2 ASSOCIATION BETWEEN PRIMARY SAFETY FUNCTIONS ASSOCIATION EMERGENCY OPERATING EMERGENCY OPERATING PROCEDURES PROCEDURES LIST OF EOP KEY PARAMETERS PARAMETERS FUNCTIONS AND VIII-3 VIII-3 TYPE AND INSTRUMENT TYPE INSTRUMENT CATEGORY CATEGORY FOR UNIT 1 RG 1.97 VARIABLES VARIABLES VIII-4 PROTECTIVE SYSTEM FUNCTION PROTECTIVE SYSTEM VIII-5 VIII-5 NON-TECHNICAL SPECIFICATION INSTRUMENTATION NON-TECHNICAL THAT INSTRUMENTATION THAT INITIATES CONTROL INITIATES WITHDRAWAL BLOCK CONTROL ROD WITHDRAWAL IX-1 IX-l MAGNITUDE MAGNITUDE AND DUTY CYCLE OF MAJOR MAJOR STATION BATTERY LOADS LOADS XII-1 XII-l FLOWS AND ACTIVITIES FLOWS ACTIVITIES OF MAJOR SOURCES SOURCES OF GASEOUS ACTIVITY XII-2 QUANTITIES AND ACTIVITIES QUANTITIES ACTIVITIES OF LIQUID RADIOACTIVE RADIOACTIVE WASTES WASTES XII-3 ANNUAL SOLID WASTE ACCUMULATION ACCUMULATION AND ACTIVITY 21 UFSAR Revision 21 UFSAR xxxv XXXV October 2009

Nine Nine Mile Point Unit 1 UFSAR LIST OF TABLES (Cont'd.) (Cont'd.) Table Table Number Number Title Title XII-4 XII-4 LIQUID WASTE DISPOSAL SYSTEM MAJOR COMPONENTS SYSTEM MAJOR COMPONENTS XII-5 XII-5 WASTE DISPOSAL SOLID WASTE DISPOSAL SYSTEM MAJOR COMPONENTS COMPONENTS XII-6 XII-6 OCCUPANCY TIMES XII-7 XII-7 GAMMA ENERGY GROUPS GROUPS XII-8 XII-8 AREA RADIATION RADIATION MONITOR DETECTOR LOCATIONS MONITOR DETECTOR XIII-l XIII-I CROSS-REFERENCE UNIT ANSI STANDARD CROSS-REFERENCE UNIT 1 XIII-2 XIII-2 MINIMUM SHIFT SHIFT CREW CREW COMPOSITION XV-l XV- 1 TABLE DELETED TABLE XV-2 XV-2 TRIP POINTS FOR PROTECTIVE PROTECTIVE FUNCTIONS FUNCTIONS XV-4 XV-5 XV-5 thru XV-3 thru XV-4 TABLES DELETED BLOWDOWN RATES BLOWDOWN RATES XV-6 XV- 6 REACTOR COOLANT REACTOR COOLANT CONCENTRATIONS CONCENTRATIONS (pCi/gm) XV-7 XV- 7 TABLE DELETED TABLE XV-7a XV-7a MSLB ACCIDENT ANALYSIS ASSUMPTIONS ANALYSIS INPUTS AND ASSUMPTIONS XV-7b XV-7b MSLB ACCIDENT RELEASE RATESRATES XV-88 XV- MAIN STEAM LINE BREAK ACCIDENT DOSES DOSES XV- 9 XV-9 SIGNIFICANT INPUT PARAMETERS SIGNIFICANT PARAMETERS TO THETHE LOSS-OF-COOLANT ACCIDENT LOSS-OF-COOLANT ANALYSIS ACCIDENT ANALYSIS XV-9a XV- 9a CORE SPRAY CORE SPRAY SYSTEM SYSTEM FLOW PERFORMANCE ASSUMED IN FLOW PERFORMANCE IN LOCA ANALYSIS ANALYSIS XV-10 XV-I0 SINGLE VALVE FAILURE ECCS SINGLE ANALYSIS FAILURE ANALYSIS UFSAR Revision 21 Revision 21 xxxvi xxxvi October 2009

Nine Mile Point Unit 1 UFSARUFSAR LIST OF TABLES (Cont'd.) (Cont'd.)

• Table Table Number Number Title Title XV-II XV-1l SINGLE FAILURES FAILURES CONSIDERED IN LOCA ANALYSIS ANALYSIS XV-12 thru thru TABLES DELETED TABLES XV-21 XV-21 XV-21a XV-21a ANALYSIS ASSUMPTIONS FOR NINE MILE POINT 1 CALCULATIONS CALCULATIONS XV-22 XV-22 RELEASED TO THE REACTOR ACTIVITY RELEASED REACTOR BUILDING FOLLOWING FOLLOWING THE FHA (CURIES)

(CURIES) XV-23 XV-23 UNIFORM UNFILTERED UNIFORM UNFILTERED STACK DISCHARGE DISCHARGE RATES FROM 0 TO 22 HR AFTER THE FHA (CURIES/SEC) (CURIES/SEC) XV-24 XV-24 FUEL HANDLING ACCIDENT ACCIDENT DOSES DOSES XV-2S XV-25 FHA ANALYSIS ANALYSIS INPUTS INPUTS AND ASSUMPTIONS ASSUMPTIONS

• XV-26 XV-26 XV-27 XV-27 XV-28 XV-28 CRD ACCIDENT ACCIDENT ANALYSIS ANALYSIS INPUTS CRDA NOBLE GAS RELEASE CRDA HALOGEN RELEASE RELEASE RELEASE INPUTS AND ASSUMPTIONS ASSUMPTIONS XV-29 XV-29 CONTROL ROD DROP ACCIDENT DOSES CONTROL DOSES XV-29a XV-29a WETTING OF FUEL CLADDING BY CORE SPRAY XV-29b XV-29b POST-LOCA AIRBORNE POST-LOCA DRYWELL FISSION PRODUCT AIRBORNE DRYWELL PRODUCT INVENTORY (CURIES)

INVENTORY (CURIES) XV-29c XV-29c POST-LOCA REACTOR POST-LOCA REACTOR BUILDING BUILDING FISSION PRODUCT PRODUCT INVENTORY (CURIES) INVENTORY (CURIES) XV-29d XV-29d DISCHARGE RATES POST-LOCA DISCHARGE POST-LOCA RATES (CURIES/SEC) (CURIES/SEC) XV-30 XV-30 FISSION PRODUCT INVENTORY CORE FISSION INVENTORY XV-31 XV-31 INPUTS AND ASSUMPTIONS ANALYSIS INPUTS LOCA ANALYSIS ASSUMPTIONS

• XV-32 XV-32 LOSS-OF-COOLANT ACCIDENT LOSS-OF-COOLANT UFSAR Revision 21 21 xxxvii xxxvii DOSES ACCIDENT DOSES 2009 October 2009

Nine Mile Point Unit 1 UFSAR UFSAR LIST OF TABLES (Cont'd.) (Cont'd.) Table Table Number Title Title XV-32a XV-32a SIGNIFICANT INPUT PARAMETERS SIGNIFICANT INPUT PARAMETERS TO THE DBRDBR CONTAINMENT SUPPRESSION CONTAINMENT SUPPRESSION CHAMBER HEATUP HEATUP ANALYSIS ANALYSIS XV-33 TABLE DELETED XV-34 TABLE DELETED XV-34a XV-34a RELEASE/INTAKE ELEVATIONS RELEASE/INTAKE ELEVATIONS XV-34b XV-34b RELEASE/INTAKE RELEASE/INTAKE DISTANCE AND DIRECTIONS DIRECTIONS XV-3S XV-35 TABLE DELETED XV-3Sa XV-35a X/Q VALUES FOR THE CONTROL CONTROL ROOM XV-3Sb XV-35b X/Q VALUES FOR THE TECHNICAL SUPPORT CENTER XV-3Sc XV-35c XV-3Sd XV-35d XV-36 OFFSITE X/Q VALUES FOR GROUND-LEVEL GROUND-LEVEL RELEASES OFFSITE X/Q VALUES FOR ELEVATED RELEASES REACTOR BUILDING LEAKAGE PATHS PATHS RELEASES RELEASES

• XVI-1 XVI-l CODE CALCULATION

SUMMARY

CALCULATION

SUMMARY

XVI-2 STEADY-STATE - (100% STEADY-STATE (100% FULL POWER NORMAL NORMAL OPERATION) OPERATION) STRESSES PERTINENT STRESSES PERTINENT OR STRESS INTENSITIES INTENSITIES XVI-3 REACTIONS FOR REACTOR LIST OF REACTIONS REACTOR VESSEL NOZZLES NOZZLES XVI-4 EFFECT OF VALUE OF INITIAL FAILURE PROBABILITY XVI-S XVI-5 SINGLE TRANSIENT EVENT FOR REACTOR SINGLE TRANSIENT REACTOR PRESSURE PRESSURE VESSEL VESSEL XVI-6 POSTULATED EVENTS POSTULATED XVI-7 MAXIMUM STRAINS STRAINS FROM POSTULATED POSTULATED EVENTS EVENTS XVI-8 CORE STRUCTURE ANALYSIS RECIRCULATION STRUCTURE ANALYSIS LINE BREAK RECIRCULATION LINE UFSAR Revision UFSAR Revision 21 21 xxxviia xxxviia October 2009

Nine Mile Point Unit 1 UFSAR

• LIST OF TABLES (Cont'd.)

TABLES (Cont'd.) Table Table Number Number Title Title XVI-9 CORE STRUCTURE STRUCTURE ANALYSIS STEAM LINE BREAK XVI-9a CORE SHROUD REPAIR REPAIR DESIGN SUPPORTING DOCUMENTATION XVI-IO XVI-10 DRYWELL JET AND MISSILE HAZARD ANALYSIS ANALYSIS DATA XVI-11 XVI-1l DRYWELL JET AND MISSILE HAZARD HAZARD ANALYSIS ANALYSIS RESULTS RESULTS XVI-12 STRESS DUE TO DRYWELL FLOODING STRESS XVI-13 ALLOWABLE WELD SHEAR STRESS STRESS XVI-14 XVI-14 LEAK RATE TEST RESULTS RESULTS XVI-1S XVI-15 OVERPRESSURE TEST--PLATE OVERPRESSURE TEST--PLATE STRESSES STRESSES

• XVI-16 XVI-16 STRESS

SUMMARY

STRESS

SUMMARY

• UFSAR Revision 21 UFSAR 21 xxxviib October 2009

Unit 11 UFSAR Nine Mile Point unit LIST OF OF TABLES (Cont'd.) (Cont'd.) Table Number Numhber XVI-17 Tjtle Ti t I HEAT TRANSFER COEFFICIENTS AS A FUNCTIONFUNCTION OF DROP DROP DIAMETER .XVI-18 XVI-18 HEAT TRANSFER COEFFICIENT COEFFICIENT AS A FUNCTION OF OF PRESSURE XVI-19 RELATIONSHIP RELATIONSHIP BETWEEN PARTICLE SIZE AND TYPE OF OF SPRAY PATTERN XVI-20 ALLOWABLE STRESSES FOR FLOOR SLABS, SLABS, BEAMS, BEAMS, COLUMNS, WALLS, FOUNDATIONS, COLUMNS, WALLS, ETC. FOUNDATIONS, ETC. XVI-21 XVI-21 ALLOWABLE ALLOWABLE STRESSES FOR STRUCTURAL STRUCTURAL STEEL XVI-22 ALLOWABLE ALLOWABLE STRESSES - REACTOR VESSEL CONCRETECONCRETE PEDESTAL XVI-23 DRYWELL - ANALYZED ANALYZED DESIGN LOAD COMBINATIONS COMBINATIONS XVI-24 SUPPRESSION CHAMBER CHAMBER - ANALYZED ANALYZED DESIGN DESIGN LOAD COMBINATIONS COMBINATIONS XVI-25 ACI CODE 505 ALLOWABLE ALLOWABLE STRESSES AND ACTUAL ACTUAL STRESSES FOR CONCRETE VENTILATION STACK CONCRETE VENTILATION STACK XVI-26 XVI-26 ALLOWABLE STRESSES FOR CONCRETE ALLOWABLE CONCRETE SLABS, WALLS, SLABS, WALLS, STRUCTURAL STEEL, AND CONCRETE BLOCK BEAMS, STRUCTURAL STEEL, AND CONCRETE BLOCK WALLS BEAMS, WALLS XVI-27 XVI-27 SYSTEM LOAD COMBINATIONS COMBINATIONS XVI-28 XVI-28 HIGH-ENERGY SYSTEMS - HIGH-ENERGY INSIDE CONTAINMENT INSIDE CONTAINMENT XVI-29 HIGH-ENERGY SYSTEMS - OUTSIDE HIGH-ENERGY OUTSIDE CONTAINMENT CONTAINMENT XVI-3D XVI-30 SYSTEMS WHICH MAY BE AFFECTED SYSTEMS WHICH AFFECTED BY PIPE WHIPWHIP XVI-31 XVI-31 CAPABILITY RESIST WIND CAPABILITY TO RESIST PRESSURE AND WIND WIND PRESSURE VELOCITY XVII-l XVII-1 DISPERSION AND ASSOCIATED DISPERSION ASSOCIATED METEOROLOGICAL METEOROLOGICAL PARAMETERS PARAMETERS XVII-2 XVII-2 RELATION OF SATELLITE RELATION NINE MILE SATELLITE AND NINE MILE POINT POINT WINDS WINDS XVII-3 XVII-3 FREQUENCY OF OCCURRENCE FREQUENCY RATES -- LAPSE RATES OCCURRENCE OF LAPSE 1963 AND 1963 AND 1964 1964 UFSAR Revision 16 UFSAR Revision 16 xxxviii xxxviii November 1999 November 1999

Nine Mile Point Point Unit 1 UFSAR LIST OF FIGURES FIGURES (Cont'd.) (Cont'd.) Figure Number Title X-2 CLEANUP SYSTEM REACTOR CLEANUP SYSTEM X-3 CONTROL ROD DRIVE HYDRAULIC CONTROL HYDRAULIC SYSTEM SYSTEM X-4 REACTOR BUILDING CLOSED LOOP COOLING SYSTEM REACTOR SYSTEM X-5 TURBINE BUILDING CLOSED LOOP COOLING SYSTEM TURBINE SYSTEM X-6 SERVICE WATER WATER SYSTEM SYSTEM X-7 FIGURE DELETED X-8 SPENT FUEL STORAGE POOL FILTERING FILTERING AND COOLING SYSTEM SYSTEM X-9 BREATHING, BREATHING, INSTRUMENT, SERVICE AIR INSTRUMENT, AND SERVICE

• X-10 X-1O X-l1 X-11 XI-1 XI-l REACTOR REFUELING REFUELING SYSTEM PICTORIAL CASK DROP PROTECTION PROTECTION SYSTEM SYSTEM STEAM FLOW AND REHEATER VENTILATION STEAM VENTILATION SYSTEM SYSTEM XI-2 EXTRACTION STEAM FLOW EXTRACTION FLOW XI-3 CONDENSER AIR REMOVAL AND OFFGAS SYSTEM MAIN CONDENSER SYSTEM XI-4 CIRCULATING WATER SYSTEM CIRCULATING WATER SYSTEM XI-5 XI-5 CONDENSATE FLOW FLOW XI-6 XI-6 TRANSFER SYSTEM CONDENSATE TRANSFER SYSTEM XI-7 FEEDWATER FEEDWATER FLOW SYSTEM SYSTEM XII-1 XII-l RADIOACTIVE WASTE DISPOSAL DISPOSAL SYSTEM SYSTEM XIII-1 XIII-1 SENIOR LEVEL STATION MANAGEMENT ORGANIZATION STATION MANAGEMENT ORGANIZATION CHART CHART
• XIII-2 XIII-2 ENGINEERING ENGINEERING SERVICES Revision 21 UFSAR Revision SERVICES ORGANIZATION xlvii ORGANIZATION CHART CHART October October 2009

Nine Mile Point Point Unit 1 UFSAR LIST OF FIGURES (Cont'd.) (Cont'd.) Figure Number Number XIII-3 XIII-3 Title Title QUALITY ASSURANCE ORGANIZATION ORGANIZATION XIII-3a NUCLEAR SAFETY & SECURITY NUCLEAR SECURITY ORGANIZATION XIII-4 XIII-4 NINE MILE NINE MILE POINT NUCLEAR NUCLEAR STATION ORGANIZATION ORGANIZATION CHART CHART XIII-4a XIII-4a NINE MILE MILE POINT NUCLEAR NUCLEAR STATION ORGANIZATION ORGANIZATION CHART CHART XIII-4b XIII-4b NINE NINE MILE POINT NUCLEAR NUCLEAR STATION ORGANIZATION ORGANIZATION CHART CHART XIII-4c XIII-4c NINE NINE MILE POINT NUCLEAR STATION ORGANIZATION ORGANIZATION CHART CHART XIII-5 XIII-5 SAFETY ORGANIZATION XV-1 XV-I XV-2 XV-2 XV-3 XV-3 STATION TRANSIENT STATION FIGURE PLANT TRANSIENT DIAGRAM FIGURE DELETED PLANT RESPONSE DIAGRAM RESPONSE TO LOSS OF 100°F 100OF FEEDWATER FEEDWATER HEATING UFSAR Revision Revision 21 21 xlviia 2009 October 2009

• Mile Point unit Nine Mile Unit 1 UFSAR
• Figure Number Title Title LIST OF FIGURES FIGURES (Cont'd.)

(Cont'd.) XV-56E XV-56E LOSS-OF-COOLANT LOSS-OF-COOLANT ACCIDENT DRYWELL DRYWELL PRESSURE XV-56F XV-56F LOSS-OF-COOLANT LOSS-OF-COOLANT ACCIDENT SUPPRESSION CHAMBER SUPPRESSION CHAMBER PRESSURE PRESSURE XV-56G LOSS-OF-COOLANT LOSS-OF-COOLANT ACCIDENT CONTAINMENT CONTAINMENT TEMPERATURE TEMPERATURE

                 - WITH CORE SPRAY XV-57 XV-57           CONTAINMENT DESIGN BASIS CONTAINMENT            BASIS CLAD TEMPERATURE TEMPERATURE RESPONSE -- WITHOUT WITHOUT CORE SPRAY XV-58           CONTAINMENT CONTAINMENT DESIGN BASIS BASIS METAL-WATER METAL-WATER REACTION XV-59           CONTAINMENT DESIGN BASIS CONTAINMENT            BASIS CLAD PERFORATION PERFORATION WITHOUT CORE SPRAY XV-60           CONTAINMENT CONTAINMENT DESIGN BASIS CONTAINMENT       TEMPERATURE CONTAINMENT TEMPERATURE WITHOUT CORE SPRAY
                 - WITHOUT XV-60A          DBR ANALYSIS SUPPRESSION SUPPRESSION POOL AND WETWELL WETWELL AIRSPACE TEMPERATURE TEMPERATURE RESPONSE - CONTAINMENT     SPRAY CONTAINMENT SPRAY DESIGN BASIS DESIGN    BASIS ASSUMPTION XV-60B          DBR ANALYSIS ANALYSIS SUPPRESSION POOL AND WETWELL WETWELL AIRSPACE TEMPERATURE AIRSPACE    TEMPERATURE RESPONSE - EOP OPERATION ASSUMPTIONS ASSUMPTIONS XV-61           REACTOR BUILDING MODEL MODEL XV-62           EXFILTRATION VS.

EXFILTRATION VS. WIND WIND SPEED SPEED - NORTHERLY NORTHERLY WIND XV-63 REACTOR BUILDING DIFFERENTIAL REACTOR DIFFERENTIAL PRESSURE XV-64 EXFILTRATION VS. EXFILTRATION VS. WIND SPEED SPEED - SOUTHERLY SOUTHERLY WIND XV-65 XV-65 REACTOR BUILDING - REACTOR ISOMETRIC ISOMETRIC XV-66 XV-66 REACTOR BUILDING - CORNER SECTIONS CORNER SECTIONS XV-67 REACTOR BUILDING - ROOF SECTIONS REACTOR SECTIONS UFSAR Revision Revision 21 21 xlix xlix October 2009 October 2009

Nine Mile Point Point Unit 1 UFSAR LIST OF FIGURES (Cont'd.) (Cont'd.) Figure Number Number Title Title XV-68 XV-68 REACTOR BUILDING BUILDING - PANEL PANEL TO CONCRETE SECTIONS SECTIONS XV-69 XV-69 . REACTOR BUILDING BUILDING - EXPANSION EXPANSION JOINT SECTIONS SECTIONS XV-70 XV-70 REACTOR BUILDING REACTOR BUILDING EXFILTRATION EXFILTRATION - NORTHERLY NORTHERLY WIND XV-71 XV-71 REACTOR BUILDING BUILDING EXFILTRATION EXFILTRATION - SOUTHERLY SOUTHERLY WIND XV-72 XV-72 REACTOR BUILDING DIFFERENTIAL PRESSURE DIFFERENTIAL PRESSURE XV-73 XV-73 REACTOR BUILDING REACTOR PRESSURE VS. BUILDING PRESSURE VS. TIME BY REACTOR TIME REACTOR BUILDING ELEVATION BUILDING XV-74 XV-74 REACTOR BUILDING PRESSURE REACTOR PRESSURE VS. REACTOR VS. TIME BY REACTOR BUILDING ELEVATION BUILDING ELEVATION (FOCUSED (FOCUSED ON THE INITIAL 2.52.5 HR) HR) XVI-l XVI-1 XVI-2 XVI-2 SEISMIC ANALYSIS LUMPED ANALYSIS OF REACTOR LUMPED MASS REACTOR SUPPORT REACTOR MOMENT MOMENT REACTOR VESSEL MASS REPRESENTATION SUPPORT DYNAMIC VESSEL GEOMETRIC DYNAMIC ANALYSIS GEOMETRIC AND ANALYSIS - ELEVATION VS. VS. XVI-3 XVI-3 REACTOR SUPPORT REACTOR SUPPORT DYNAMIC DYNAMIC ANALYSIS ANALYSIS - ELEVATION ELEVATION VS.VS. SHEAR XVI-4 XVI-4 REACTOR SUPPORT DYNAMIC REACTOR DYNAMIC ANALYSIS ANALYSIS - ELEVATION ELEVATION VS.VS. DEFLECTION XVI-5 XVI-5 REACTOR SUPPORT REACTOR DYNAMIC ANALYSIS SUPPORT DYNAMIC ANALYSIS - ELEVATION ELEVATION VS.VS. ACCELERATION XVI-6 thru thru FIGURES DELETED XVI-8 XVI-8 XVI-9 XVI-9 REACTOR VESSEL SUPPORT SUPPORT STRUCTURE STRUCTURE STRESS

SUMMARY

XVI-I0 XVI-10 THERMAL ANALYSIS THERMAL ANALYSIS XVI-II XVI-11 FAILURE PROBABILITY FAILURE PROBABILITY DENSITY FUNCTION Revision 21 UFSAR Revision 21 1 October 2009 October 2009

Nine Mile Point Unit 11 UFSAR LIST OF FIGURES FIGURES (Cont'd.) (Cont'd.) Figure Number Number Title Title XVI-12 ADDITION STRAINS ADDITION STRAINS PAST 414% REQUIRED REQUIRED TO EXCEED DEFINED SAFETY MARGIN MARGIN XVI-12a XVI-12a SHROUD WELDS SHROUD WELDS XVI-12b XVI-12b CORE SHROUD STABILIZERS STABILIZERS XVI-12c XVI-12c CORE SHROUD WELDS WELDS XVI-12d XVI-12d V9/VIO VERTICAL WELD CLAMP ASSEMBLY V9/V10 VERTICAL XVI-13 LOSS-OF-COOLANT LOSS-OF-COOLANT ACCIDENT - CONTAINMENT CONTAINMENT PRESSURE NO CORE OR CONTAINMENT CONTAINMENT SPRAYS SPRAYS XVI-14 FIGURE DELETED FIGURE XVI-IS XVI-15 CONCRETE AIR DRYWELL TO CONCRETE AIR GAP GAP XVI-16 XVI-16 TYPICAL PENETRATIONS PENETRATIONS

• UFSAR Revision UFSAR Revision 21 21 la la October 2009

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Nine Mile Mile Point Unit 11 UFSAR

• A computer analysis A

building. building. analysis was made to determine seismic accelerations, accelerations, displacements, reactions acting on the RPV and its reactions support to the design earthquake included is determine the displacements, shears, its the maximum induced shears, moments and support and on the reactor The analysis includes the response of the RPV and earthquake and jet is the effect on the RPV of the displacement jet reaction reaction forces. forces. displacement of the and reactor and its Also its reactor building building and containment vessel due to the postulated earthquake. earthquake. Results of this analysis are contained on Figures VI-6 through VI-17. VI-17. Personnel access into the reactor reactor building is is controlled controlled from the track bay extension and from the turbine building. building. The track bay extension has a railroad entrance extension entrance and a personnel personnel access air air lock passageway from the outside. outside. The track bay extension consists of a 20-ft by 20-ft by 80-ft 80-ft long air lock, connected to the track bay compartment lock, connected compartment by a vertical liftlift inner door and an airtight airtight seal. seal. The track bay extension extension is is equipped with a motor-operated motor-operated double swing outer outer door 16 ft ft wide by 17 ft ft 6 inin high. high. The door can also be be operated operated manually and is is designed designed to resistresist an internal internal or or external load of 40 psf. external psf. The outer outer door closes against a closed cell sponge neoprene cell neoprene closure to provide an airtight airtight seal. seal. The The inner vertical lift vertical lift door bears against a one-piece one-piece inflatable seal of reinforced reinforced ethylene propylene diene monomer monomer around its its perimeter. The entire contact area of the inflatable seal will perimeter. inflatable will expand approximately 3/4 in expand approximately in under pressure. pressure. The seal material material will remain pliable and seal at temperatures temperatures of -20°F 210 0 F.

                                                                             -20OF to 210°F.

Containment integrity for the track bay compartment containment compartment and extension is provided is provided by an outside double outside double swing door, door, an inside inside vertical vertical lift lift door and personnel personnel doors connected connected by an airtight airtight accessaccess passageway. passageway. The track bay compartment compartment (with extension) and itsits access openings are shown shown on Figure 111-4. Typical door seals seals for the personnel personnel and equipment doors are shown on Figure Figure VI-18. VI-18. Interior Interior doors doors with air locks locks are provided provided in in the south south wall of of the reactor building reactor building leading into the turbine room at el 261, 261, as as shown shown on Figure Figure 111-4, and at el 340, 340, as shown on Figure III-8. Figure 111-8. The doors of the air lock have neoprene neoprene seals seals withwith sealing requirements requirements equivalent equivalent to those of the railroad railroad door. door. Details Details are shown on Figure VI-19. Figure VI-19. Procedures and alarms Procedures alarms are used used to control control access access and maintain building integrity. building integrity. Primary and secondary shielding secondary shielding is discussed is in Section in Section XII . XII.

• UFSAR Revision Revision 15 15 VI-19 VI-19 November 1997 November 1997

Nine Mile Mile Point Unit 11 UFSAR D. 1.0 CONTAINMENT CONTAINMENT ISOLATION Design Bases Design Isolation Bases ISOLATION SYSTEM Isolation valves are provided SYSTEM provided on lines penetrating penetrating the drywell drywell and pressure pressure suppression suppression chamber chamber to assure integrity integrity of the containment when required containment required during emergency emergency and post-accident post-accident periods. periods. Isolation valves which which must be closed to assure containment integrity containment immediately after a major integrity immediately accident are major accident automatically controlled automatically controlled by the reactor protection protection system (RPS) system (RPS) described in described in Section Section VIII.VIII. The drywell suppression chamber drywell and suppression chamber penetrations penetrations dedicated are dedicated to specific specific purposes purposes as shown in in Tables VI-l and VI-2, VI-2, respectively. respectively. The tables tables listlist the number, number, size, and type of of penetration associated with each purpose. penetration associated purpose. Containment isolation Containment isolation valves (also called isolation isolation valves) are defined as any valves which are relied upon to perform perform a containment isolation containment isolation function on lines penetrating penetrating the primary reactor containment containment and include include all all reactor coolant coolant isolation valves valves and all containment isolation valves. all primary containment valves. Test, vent Test, vent and drain (TVD) (TVD) valves located located on the containment containment pressure boundary containment isolation boundary are containment isolation valves but are not included

• in the tables of reactor coolant in isolation valves coolant isolation valves or primary containment isolation containment isolation valves.

valves. Reactor Reactor coolant isolation valves are containment coolant isolation containment isolation valves valves which are on lines penetrating the primary lines penetrating reactor primary reactor containment and are connected containment connected to the RCS (or a system containing reactor coolant) and function as reactor reactor reactor coolant coolant pressure boundary (RCPB) components. boundary (RCPB) components. Reactor coolant isolation valves Reactor coolant valves function function as primary containment primary containment isolation valves in in the event event ofof LOCA. a LOCA. Primary containment containment isolation valves valves are containment containment isolation valves penetrating the primary valves on lines penetrating reactor containment primary reactor containment connecting directly connecting directly to the free space enclosed space enclosed by the the containment. containment. Table Table VI-3a is is a listing listing of allall reactor reactor coolant coolant isolation valves, and Table valves, Table VI-3b lists lists primary containment isolation primary containment valves. valves. All lines which are part of the RCPB and penetrate penetrate the primary reactor containment reactor containment are provided with redundant isolation are provided isolation

• valves.

valves. As a general general rule,rule, one of each pair of isolation valves valves UFSAR Revision 21 21 VI-20 October 2009 2009

Nine Mile Point Unit 1 UFSAR

• in in series is located series is located inside the containment.

containment. outside the containment. and on the feedwater feedwater system where it both valves outside the containment, between the line and the containment between containment. On the emergency it was necessary The other other valve is emergency cooling system supply necessary to install containment, a guard pipe is containment vessel penetration install is installed installed penetration sleeve.sleeve. is This sleeve is is welded welded to the body of the first first isolation isolation valve outside the containment. This, in effect, containment. This, in effect, extends the extends containment containment to include include the body of the first first isolation isolation valve. valve. For the emergency emergency cooling cooling system supply, the two valve bodies bodies are welded welded end to end for greater integrity. integrity. For feedwater the feedwater system, the two valves system, separated by a 10-in valves are separated 10-in extension. extension. Lines which are partpart of the reactor reactor coolant coolant boundary boundary and may be be required required to have flow after an accident accident are provided with check valves. valves. The CRD and liquid poison systems systems have two check valves valves in series. in series. One valve valve isis inside the containment. containment. The feedwater feedwater described above, system, as described above, has two valves valves outside the the containment, containment, one of which is is a check valve. valve. shutdown cooling The cleanup and shutdown cooling systems each have redundant redundant isolation valves isolation valves with one valve inside the containment. containment. The The outer valve valve on the return return to the reactor line is is a check valve. check valve. Post-accident Post-accident thermal overpressurization protection overpressurization protection is provided is provided for the penetration penetration piping between between the isolation valves in in the shutdown shutdown cooling system. system. Instrument provided with Instrument lines are provided redundant redundant valving outside outside the containment. containment. Automatic flow check valves minimize minimize loss of of reactor coolant in in the event of an instrument line break. instrument break. All external external isolation valves valves are located as close close to the containment possible. containment as possible. Where guard pipes are used between between the containment penetration containment penetration and the line, line, the outer valve valve is is welded welded to the guard pipe. guard pipe. reactor coolant For reactor coolant isolation isolation valves on on low-temperature low-temperature lines where where no guard pipe pipe is is required, required, the outer outer valve is welded valve is welded directly to the penetrations penetrations sleeve.sleeve. Most lines which connect connect directly directly to the containment containment atmosphere and penetrate penetrate the primary reactor containment are reactor containment are provided with redundant redundant isolation valves. isolation valves. Two normally-closed normally-closed valves outside the containment containment are provided provided for systems which are not required to function under accident accident conditions. conditions. Lines which are not not equipped equipped with double isolation double isolation valves valves have been determined to be been determined be acceptable based acceptable ba~ed upon the fact that the system reliability reliability is is not compromised, compromised, the system is is closed outside containment, containment, and a

• single isolation isolation valve in UFSAR UFSAR Revision Revision 21 in the line.

21 line. accommodated with only one single active failure can be accommodated VI-21 VI-2l one October 2009

Nine Mile Point Point Unit Unit 1I UFSAR UFSAR Instrument Instrument lines connected to containment atmosphere which penetrate penetrate primary containment

barriers, barriers, such as Each containment containment are provided with two isolation as manual valves, containment spray line which is valves, caps, caps, oror diaphragm assemblies.

is required to be open under which isolation assemblies. under accident conditions conditions contains a check valve outside the the containment. containment. These check valves are installed to minimize check.valves minimize suppression during the initial pressure suppression bypassing of pressure initial pressure transient of the LOCA. LOCA. The oxygen sample return line and the nitrogen purge line for the traveling in-core probes use two check valves in in series outside the containment. outside containment. The traveling in-core in-core probe guide tubes use aa ball valve and manually-actuated manually-actuated explosive shear shear valve in containment. in series outside containment. Each line that penetrates containment and is penetrates primary reactor containment is neither part of the RCPB nor connected directly to the connected directly containment containment atmosphere, atmosphere, in in the case of the drywell cooling and recirculation pump cooling systems, recirculation systems, has-one isolation valve. has'one isolation valve. cooling water in circulate cooling These systems circulate These in a closed system into into containment. Each and out of the containment. carrying incoming cooling Each line carrying water water is is provided self-actuating check provided with aa self-actuating valve outside the check valve

• containment.

containment. Each line which carries water which carries water out of the containment containment has a MOV which is is actuated actuated by remote remote manual manual control. control. isolation system for each line is The isolation is designed accommodate designed to accommodate loss of power to an isolation valve. isolation valve. MOVs (ac or dc) are designed to fail fail in the mode in in in which they are when loss of of power occurs. power occurs. valves (AOV) Air-operated valves Air-operated (AOV) fail closed upon loss fail closed of power. Different power power. Different power sources for each each valve in in series series ensure that the isolation ensure function will not be defeated isolation function defeated by by single failure. single failure. Failure of a single power Failure power source does not not prevent isolation even where prevent isolation normally open MOV where a normally fails MOV fails open. open. Isolation Isolation is e~fected either is effected having a closed either by having closed piping system system which which does not communicate communicate with containment atmosphere containment atmosphere or by by having a redundant having powered valve in separately powered redundant separately in series series with the the failed valve. valve. In In the case case of systems which systems which are required required to be be open following an accident, open following normally open and valves are normally accident, valves and fail fail open, are open, are normally closed but fail normally closed open, or fail open, normally closed or are normally closed but fail but closed (as is) fail closed is) but have redundant valve path in have a redundant in parallel that is parallel fails open. and/or fails is open and/or open. Revision 20 UFSAR Revision UFSAR 20 VI-21a VI-21a October 2007 October 2007

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• UFSAR Revision UFSAR Revision 16 16 VI-21b VI-21b November 1999 1999

Nine Mile Point Unit 1 UFSAR Systems Systems which connect connect to the nuclear steam stearn supply supply system (NSSS)

• and may be required to have flow after after an accident are providedprovided either with two check valves or a check and a remote manually either manually controlled controlled valve in series.

in series. These are the feedwater, feedwater, the CRD hydraulic, and the liquid poison systems. hydraulic, systems. Instrument lines that run from the reactor Instrument reactor primary system primary system through the drywell are equipped equipped with shutoff shutoff valves and a flow flow check valve located outside valve located outside containment containment as indicated indicated on Figure VI-20. VI-20. The flow check check valves meet or exceed exceed the following design requirements: requirements: Design Conditions Conditions Operating Pressure Operating Pressure 1250 psig Operating Temperature Operating Temperature 575 0 F Specified Specified Flow to Close Valve gpm 25 gpm Horizontal Horizontal Acceleration Acceleration 0.20 g Vertical Acceleration Vertical Acceleration 0.10 g section of a typical A cross section typical 3/4-in check valve is is shown on on Figure VI-21. VI-21. The valve poppet poppet isis held open by the spring. spring. The The force generated pressure differential generated by the pressure differential over the seat seat area acts against against the spring. spring. Flow creates creates a pressure differential differential which overcomes overcomes the spring spring and closes the poppet. poppet. The The differential pressure differential pressure then acts on the poppet seating poppet seating area to to keep the poppet poppet closed. closed. A bypass arrangement is bypass arrangement is used on these instrument instrument lines as a means of equalizing equalizing line pressure pressure to open the flow check valve in the event it in it should close close and for blowdown purposes. blowdown purposes. Instrument line leaks leaks can be detected detected by one or a combination combination of of the following: following: 1.

1. Operator comparing readings Operator comparing readings with several several instruments instruments monitoring monitoring the same process variable variable such as reactorreactor level, recirculation level, recirculation pump flow, steam stearn flow, steam flow, and stearn pressure.

pressure. UFSAR UFSAR Revision Revision 21 21 VI-22 VI-22 October 2009

Nine Mile Point Unit 1 UFSAR Nine

• 2.

2. 3. 3. By annunciation annunciation of the control function, or low in in the control control room. general increase By a general readings increase in readings throughout the reactor room. either high in the area radiation monitor reactor building. building. monitor 4.

4. By audible noise either inside the turbine building building or or outside the reactor outside reactor building.

building. 5.

5. By alarms on the reactor building floor drain tank.

6.

6. By probable probable increase increase in in area temperature temperature monitor monitor readings in readings in the reactor building.building.

Routine Routine surveillance surveillance as indicated indicated in in items 1 through 6 is is felt felt to be a sufficient sufficient program program for the periodic periodic testing and examination of the valves examination valves in in these small-diameter instrument these small-diameter instrument lines. lines. At each each major refueling refueling outage, instrument line outage, each instrument line flow check valve will be tested for operability. operability. The engineered safeguards systems which engineered safeguards which may be required required to to operate operate following accident originally following an accident originally had no specific specific isolation requirements. requirements. These systems, which These systems, which consist of core and containment containment sprayspray and the emergency system, were emergency cooling system, were designed designed as containment extensions containment extensions and diligent diligent efforts were were made to meet the intent of Section 111-1965 of the ASME Code. Code. Valves were provided provided in in the lines from the suppression suppression chamber chamber in ~hose and in those into the drywell to provide system isolation isolation for for maintenance or testing. maintenance testing. Isolation valvesvalves for these systems are shown in in Tables VI-3a VI-3a and VI-3b.VI-3b. The opening opening times, times, failure failure modes, and normal position of the valves in modes, spray, in the core spray, containment spray and emergency containment emergency cooling cooling systems are based based on the individual system individual system operational requirements as discussed operational requirements discussed in in Sections V V and VII. VII. In general, In general, the closure closure time of all all isolation isolation valves is is such release of fission products that the release products to the environment environment is is minimized. minimized. As described described in in Section XV, XV, no large-scale fission large-scale fission product release occursoccurs before 1 min has elapsed. elapsed. The valve closure times are are thus set for a 1-min maximum unless unless operational operational restrictions restrictions are more severe. severe. The closure closure times of all all valves on lines in in systems connecting to the NSSS are based on preventing preventing fuel damage damage from overheating with no feedwater feedwater makeup following a line break break in the in the

• particular system.

particular (MSL) is line (MSL) system. The valve closure is based on the MSL break UFSAR Revision 21 21 closure time for the main steam break accident discussed VI-23 discussed in steam in October 20092009

Nine Mile Point Point Unit 1 UFSAR Section xv.XV. By keeping the valve closure time less than about about

• 10 sec, sufficient sec, sufficient coolant will remain in in the reactor vessel to to provide adequate core provide adequate core cooling.

cooling. The valves are designeddesigned to close and to be leak-tight during the worst worst conditions pressure, conditions of pressure, temperature temperature and steam steam flow following a break in in the MSL outside pressure suppression system. the pressure system. The codes used in in the design of Class I system containment containment isolation isolation valves valves at the time of construction construction included included ASMEAS ME Section 1-1962 Section 1-1962 or ANSI B31.1-1955 and ANSI B16.5-1955, B16.5-1955, with requirements of ASME Section 111-1965 requirements nondestructive testing 111-1965 for nondestructive testing (NDT). (NDT). For subsequent modifications, modifications, Regulatory Regulatory Guide (RG) (RG) 1.29 recommendations are followed. recommendations segments Piping system segments penetrating penetrating containment considered susceptible containment and considered susceptible to thermal thermal overpressurization overpressurization are analyzed in accordance analyzed in accordance with the criteria criteria of the ASME Boiler & Pressure Vessel Vessel Code, Section III, Code, Section III, Appendix Appendix F (1986 Edition). Edition). The design criteria criteria for containment containment isolation isolation valves consist of of normal normal and special loadings, loadings, load combinations, combinations, and load combination combination limits. limits. Seismic design Seismic criteria design criteria are listed listed in Table in Table VI-4. VI-4. 1.1 Containment Containment Spray Appendix Table VI-3b lists containment containment. lists containment spray system containment. These These Appendix J Water containment isolation primary containment system which enter the free space lines have an Appendix Appendix J Requirements Water Seal Requirements isolation valves of the space of the water seal seal by by virtue of system operationoperation following following the design basis LOCA. LOCA. The system system design basis is is continuous operation following the DBA as continuous operation as UFSAR Revision Revision 21 21 VI-23a VI-23a October 2009

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• 17 UFSAR Revision 17 VI-23b October 2001

Nine Mile Mile Point Unit 11 UFSAR documented in documented Group (BWROG) operation. operation. in Section XV-5.3. (BWROG) Emergency (EPG/SAG) restrict BWROG EPG/SAG spray the BWROG XV-5.3. The Boiling Water Reactor Procedure and Severe Emergency Procedure restrict drywelldrywell and suppression The emergency emergency operating Severe Accident operating procedures spray limitations, limitations, are intended Reactor Owners' chamber spray suppression chamber procedures (EOP) Owners' Guidelines Accident Guidelines (EOP), , based intended to provide based onon Operator guidance to prevent Operator guidance prevent beyond beyond design basis evaporative evaporative cooling cooling conditions from developing. developing. The evaluation, evaluation, which which determined determined the impact of the EOP assumed assumed actions upon the licensing basis, basis, concluded concluded that the radiological radiological impact impact of the potential leakage from the primary potential leakage containment for the primary containment conditions conditions where the water seal is is secured secured would result in in less than 20 percent percent of the 10CFR100 10CFR100 regulatory regulatory limits, limits, and less than 65 percent percent of the control control room regulatory regulatory limits per per 10CFR50 10CFR50 Appendix A, Appendix A, General Design Criterion (GDC) (GDC) 19. 19. The drywell spray spray limitations limitations werewere developed developed to address address evaporative evaporative cooling conditions conditions which which are beyond beyond the Unit 1 design basis. basis. Therefore, the conditions Therefore, conditions which interruptinterrupt the the 10CFR50 Appendix 10CFR50 Appendix J water seal are evaluated evaluated as beyond design basis conditions. basis conditions. In this respect, In respect, the maximum maximum potential potential leakage assumed leakage assumed in in this evaluation evaluation is is not included as part of included of design basis primary the design primary containment leakage. The leakage containment leakage. leakage is is only used to compare compare the maximum potential leakage relative to potential leakage to 10CFR100 and 10CFR50 10CFR100 10CFR50 Appendix Appendix A,A, GDC 19.19. In order to ensure In that assumptions used in in this evaluation evaluation remain valid, surveillance tests surveillance tests required to monitor are required packing degradation monitor packing degradation and ensure ensure minimal system cross-tie cross-tie leakage (see Section VI-F.l.2 VI-F.1.2 and VII-B.4.0). VII-B.4.0). Post-accident Post-accident secondary containment conditions secondary containment conditions are defineddefined based based on the integrity integrity of the containment containment spray system pressure boundary and the containment boundary containment isolation check check valves. valves. The The secondary conditions are defined based containment conditions secondary containment based on total total leakage of 1.5 percent leakage percent per day as defined in in Section Section VI-F. VI-F. This This is based on the integrated is integrated leak leak rate testing (ILRT) (ILRT) discussed in in Section VI-F.1.2. VI-F.1.2. Therefore, post-accident Therefore, equipment post-accident equipment conditions or post-accident qualification conditions qualification post-accident vitalvital area access is is not affected affected by the potential potential leakage used to evaluate evaluate the beyond beyond design basis EOP conditions conditions which which terminate terminate the containment sprays. containment sprays. 2.0 System

System Design

list A list of all all isolation valves isolation penetrating the valves on lines penetrating containment vessels and their pertinent containment pertinent modes and characteristics is characteristics is given given inin Table except those on instrumentation valves, except

valves, UFSAR Revision 21 21 VI-24 VI-3.

Table VI-3. Figure VI-22 shows all lines. instrumentation lines. all October 2009 2009

Nine Mile Point Unit 11 UFSARUFSAR

• service service water-cooled maintain temperatures at 85°F maintain temperatures accessible areas and 100°F accessible inaccessible areas.

inaccessible areas. exchanger units are designed water-cooled heat exchanger 0 85 F maximum and 70OF 100OF maximum and 50°F 50OF minimum ~nin to designed to 70°F minimum inin Both the main supply and exhaust Both exhaust ducts are equipped equipped with two leak-tight isolation leak-tight isolation valves in in series, series, which close automatically automatically upon detection detection of high radiation radiation levels within the building. levels within building. exhaust fans trip The supply and exhaust immediately. trip immediately. The closure sequence normal supply and exhaust isolation sequence of the normal isolation valves valves ensures ensures that reactor reactor building building negative pressure is negative pressure is maintained maintained during the transition transition from normal normal to emergency emergency ventilation ventilation for for events which are not accompanied accompanied by a loss of offsiteoffsite power power (LOOP) (see Section XV for LOCA/LOOP discussion). (LOOP) discussion). They also maymay be controlled controlled manually from the main control control room. room. inlet The inlet and outlet duct penetrations through the building penetrations through building walls are sealed against leakage. sealed leakage. A steel pipe sleeve is is integrally integrally cast cast concrete, and the outer in the concrete, in sleeve has a gasketed outer end of the sleeve connects to the first flange which connects first isolation valve. isolation valve.

• UFSAR Revision 21 21 VI-29 VI-29 October 2009 October 2009

Nine Mile Point Unit unit 1 UFSAR F. F. TEST AND INSPECTIONS A program A developed Leakage includes overall INSPECTIONS developed based on Appendix containment system program of testing the primary containment Appendix J of 10CFR50, Leakage Testing for Water overall ILRTs, Water Cooled Power system has been "Reactor Containment 10CFR50, "Reactor Reactors." Power Reactors." leakage rate ILRTs, local leakage Containment The program program rate tests and isolation leakage tests. valve leakage 1.0 Suppression Chamber Drywell and Suppression Chamber 1.1 Preoperational Preoperational Testing construction of the drywell and suppression chamber, Following construction chamber, tested at 1.15 times its pressure tested each was pressure pressure. its design pressure. Penetrations were sealed with welded end caps as were the Penetrations downcomers from the drywell to the suppression downcomers chamber. suppression chamber. The The suppression chamber to the drywell were relief lines from the suppression were blanked. Following also blanked. strength test, the drywell Following the strength drywell and chamber were tested for leakage suppression chamber leakage rate at design pressure; each met the criterion pressure; leakage at this stage criterion for leakage of stage of construction of less than 0.1 percent construction pressure. percent per day at design pressure. The suppression chamber was also tested while half filled with suppression chamber simulate operating water to simulate conditions. operating conditions. After complete complete installation penetrations, ILRTs of the installation of all penetrations, suppression chamber, drywell, suppression drywell, associated penetrations chamber, and associated were penetrations were conducted. conducted. The tests were conducted several test pressures conducted at several pressures up

• including 35 psig to establish to and including curve. The leakage rate curve.

establish a leakage The necessary instrumentation necessary installed in instrumentation was installed in the containment containment systems to provide the data to calculate calculate the leakage rate. leakage rate. Table Table VI-5 summarizes VI-S summarizes the initial preoperational initial preoperational tests conducted. conducted. 1.2 1.2 Postoperational Testing postoperational An integrated rate Type A test is leakage rate integrated leakage performed to is performed to leakage through the primary demonstrate that leakage demonstrate containment and primary containment systems and components systems penetrating primary components penetrating containment does not primary containment not exceed allowable leakage exceed the allowable specified in leakage rate specified Technical in the Technical Specifications. Specifications. integrated leakage Type A test shall include The integrated containment spray system piping in the containment operating mode in its operating mode configuration. Specification configuration. Technical Specification Technical is based on the This is analysis analysis discussed in in Section VI-D.1.1 which takes credit Section VI-D.I.l credit for the ILRT ILRT as confirmation containment spray confirmation of containment system integrity (i.e., spray system (i.e., minor components leak-tight, cross-tie components are leak-tight, leakage is cross-tie leakage is minimal). minimal). The integrated leakage test is integrated leakage is conducted analyzed maximum conducted at the analyzed maximum accident pressure, Pa. accident pressure, Pa. The test pressure, pressure, as required 10CFR50 required by 10CFRSO Appendix J, is based on the design basis LOCA conditions. J, is conditions. The The ' pressure following aa LOCA would be 35 containment pressure peak primary containment 35 psig. 10CFR50 acceptance Appendix JJ to 10CFR50 The Appendix states that the criteria states acceptance criteria maximum allowable leakage rate allowable leakage exceed 1.S rate shall not exceed weight 1.5 weight ., percent percent of the contained contained air in in 24 hr at 35 psig. allowable psig. The allowable UFSAR Revision lS 15 VI-3 VI-300 1997 November 1997

• Nine Nine Mile Point Unit 1 UFSAR UFSAR TABLE VI-3a V1-3a (Cont'd.)

(Cont'd.) NOTES: NOTES: (4) These These valves are provided valves are provided with with aa water water seal. seal. Valves shall Valves shall be tested consistent be tested consistent with with Appendix Appendix JJ water water seal testing requirements. requirements. Under 10CFR50, Appendix under 10CFRSO, Appendix J, J, Option B, through RG 1.163, 1.163, water-sealed water-sealed C1V CIV test frequency may be set using a performance performance basis inin a manner similar to that described described inin NEI NE1 94-01, 94-01, Revision 0,0, dated 7/26/9S, 7/26/95, for Type B and Type C C test intervals. intervals. Leakage rates shall be conservatively conservatively limited to 0.5 O.S gpm per nominal inch inch of valve diameter up to a maximum of S 5 gpm. gpm. (5) (5) These valves are These valves are tested in accordance tested in accordance with with Technical Technical Specification Specification Section 4.2.7.1a. Section 4.2.7.1a. (6) (6) The The self-actuating self-actuating flow fuse is flow fuse tested in is tested in accordance accordance with Specification Section with Technical Specification Section 4.3.4c. 4.3.4c. (7) Two 1i globe Two 1" globe valves (38-206 and valves (38-206 and 208) are provided 208) are outside in provided outside in the the seal water (core seal water (core spray) flow test line spray) flow line and one 3/4" globe valve (38-209) is provided (38-209) is provided outside in in the seal seal water water supply line drain, which also serve serve as RCS isolation valves. valves. (8) (8) One 3/4" check One 3/4" check valve valve (38-216) (38-216) isis provided provided inside inside primary primary containment containment around around isolation valve 38-01. isolation valve 38-01. This valve isis provided provided with a water seal and tested under under the Appendix JJ program program for limited limited flow inin the open direction, and under the 1ST IST Program, Program, exercised exercised closed for isolation capability. isolation capability. (9) (9) Reactor coolant Reactor isolation valves coolant isolation valves function as primary function as primary containment containment isolation isolation valves valves inin the event of aa LOCA. LOCA. UFSAR Revision Revision 2121 4 of of 4 October 2009 October 2009

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• UFSAR 20 UFSAR Revision 20 VII-Tb VII-7b 2007 October 2007

Nine Mile Point Unit 11 UFSAR UFSAR B. B. 1.0 CONTAINMENT CONTAINMENT SPRAY SYSTEM Licensing SYSTEM Licensing Basis Requirements The following regulatory Requirements documents are applicable regulatory documents applicable to the containment containment spray system (CSS) (CSS) and,in general terms, and, in general terms, form the basis on which the system system is is designed and operated. operated. 1.1 10CFR50.49 - Environmental 10CFR50.49 Environmental Qualification Qualification of Electric Equipment Important to Safety Safety for Nuclear Power Plants Plants An EQ program program for electrical electrical equipment has been conducted in been conducted in accordance with 10CFR50.49. accordance 10CFR50.49. Consequently, Consequently, electrical electrical equipment equipment important to safety in in the CSS system has been qualified qualified toto operate operate in environment. in the LOCA environment. 1.2 10CFR50 Appendix 10CFR50 Appendix A - General General Design Criteria Criteria for Nuclear Nuclear Power Plants The Technical Technical Supplement Supplement to Petition Petition for Conversion Conversion from Power Power Operating License to Full Term Operating Operating Operating License covered the Unit 1 positions relative relative to the GeneralGeneral Design Criteria (GDC). Criteria (GDC). Those portions of the documentation documentation that cover both the description of the requirements description requirements and NMPC's positions relative relative to to

• these requirements, these requirements, as they pertain directly pertain directly to the CSS system, system, extracted and are shown below:

have been extracted below: Criterion Criterion 16 16 Containment Containment Design Reactor containment and associated Reactor containment associated systems systems shall be provided provided to establish establish an essentially essentially leak-tight leak-tight barrier barrier against the uncontrolled against uncontrolled release of radioactivity radioactivity to the environment, and to assure environment, containment design assure that the containment conditions important conditions important to safety safety are not exceeded exceeded for as long as as postulated accident conditions require. postulated accident require. A suppression containment A pressure suppression containment system consisting consisting of a drywell, suppression chamber (torus), drywell, suppression chamber (torus), and interconnectinginterconnecting vent vent piping piping is containment for the main coolant system. is the primary containment system. During normal operation, operation, the reactor reactor building, building, containing containing the pressure suppression system, pressure provides a secondary system, provides secondary containment containment barrier. barrier. To ensure ensure the integrity integrity of the primaryprimary containment, containment, integrated leak tests tests were performed performed prior to Station Station operation operation and periodically thereafter, periodically thereafter, as provided in in the Technical Technical Specifications. Specifications. The results demonstrated demonstrated that the containment containment UFSAR UFSAR Revision Revision 21 21 VII-8 VII-8 October 2009

Nine Mile Point Unit 11 UFSAR

• met the design leak rate 35 psig and, therefore, barrier.

containment rate of 0.5 percent percent per day at a pressure therefore, provides an essentially The design essentially leak-tight design basis LOCA was evaluated containment maximum allowable per day at 35 psig. psig. allowable accident The analysis pressure of leak-tight evaluated at the primary accident leak rate of 1.5 percent analysis demonstrates percent demonstrates that the offsite offsite of doses from this accident accident would be well within the limits of limits of 10CFR50.67. 10CFR50.67. Criterion Criterion 3838 Containment Containment Heat Removal A system to remove A system remove heat from the reactor containment shall be provided. reactor containment provided. The system safety function shall be to reduce rapidly, consistentconsistent with the functioning of other associated associated systems, systems, the containment containment pressure pressure and temperature temperature following any LOCA and maintain maintain them at at acceptably low levels. acceptably levels. Suitable redundancy Suitable redundancy in in components components and features, and suitable interconnections, leak detection, isolation, and containment interconnections, containment capabilities capabilities shall be provided provided to assure that, for onsite electric power electric operation (assuming offsite power is power system operation is not not available), and for offsite available), electric power offsite electric power system system operation (assuming onsite power is is not available), available), the system safety function can be accomplished. accomplished. Two CSS system loops are provided to remove heat, heat, reduce reduce pressure, and restore pressure, restore the pressure pressure suppression suppression system system temperature temperature following a LOCA. a LOCA. Each loop is capable is capable of removing all all the decay heat and, and, in in addition, addition, the energy energy from any credible credible metal-water metal-water reaction at a rate that will prevent prevent containment pressures and temperatures containment pressures temperatures from exceeding their exceeding their design values. values. The power power for the pumps is is provided from redundant redundant Station reserve reserve power power supply systems or from one of two emergency emergency diesel diesel generators. generators. One of the two spray loops is is automatically automatically actuated on the combined actuated combined condition condition of high drywell pressurepressure and low-low reactor water level. water level. The other other loop can can be manually manually controlled from the main control controlled control room. room. Criterion 39 Criterion 39 Inspection Inspection of Containment Containment Heat Removal Removal System The containment containment heat removal removal system shall be designed to permit appropriate appropriate periodic periodic inspection inspection of important important components, components, suchsuch as the torus, torus,

• sumps, spray nozzles, sumps, capability UFSAR nozzles, and piping to assure capability of the system.

UFSAR Revision 21 21 system. VII-9 VII-9 assure the integrity integrity and October 2009

Nine Mile Point Unit 11 UFSAR Essential CSS system Essential system components are inspected periodically components are periodically to to

• ensure the integrity integrity and capability capability of the system.

system. system The system tests tests and inspections inspections are described described in in Section VII-B-6.0 VII-B-6.0 and in in the Technical Specifications. Technical Specifications. Criterion 40 40 Testing of Containment Heat Heat Removal Removal System The containment containment heatheat removal removal system shall be designed to permit permit appropriate appropriate periodic pressure pressure and functional testingtesting to assure: 1) the structural

1) structural leak-tight integrity and leak-tight its integrity of its components, components, 2) 2) the operability and performance performance of the active components components of the system, system, and 3) 3) the operability operability of the system as a whole,whole, and under conditions conditions as close close to the design as practical, practical, the performance performance of the full full operational sequence operational sequence that brings the system into operation,operation, including including operation operation of applicable applicable portions of the protection system, the transfer system, between normal transfer between normal and emergency emergency power sources, power sources, operation of the associated and the operation associated cooling water system. system.

The CSS system is CSS_system is designed designed to permit appropriate appropriate periodic pressure pressure and functional testing. Pumps Pumps are periodically periodically tested developed pressure for flow, developed pressure and automatic automatic initiation. initiation. Containment spray injection Containment injection valves valves are normally open and are not not

• required operate.

required to operate. The testing program demonstrates, demonstrates, underunder simulated conditions, simulated conditions, that pump sets can be relied upon to to function as they are designed to operate operate under accident accident conditions. conditions. spraying of water Periodic spraying water into the containment containment is is not not practical. practical. Therefore, water is Therefore, recycled back to the suppression is recycled pool during tests. tests. Air tests tests are used used to ensure ensure flow through header and nozzles. the header nozzles. Testing emergency power Testing of emergency sources for containment power sources containment cooling cooling isis periodically performed. periodically performed. The power power systems systems are tested for for automatic pickup automatic pickup of load required required for the LOCA. LOCA. Criterion Criterion 44 44 Cooling Water A system to transfer heat from structures, Cooling Water structures,

systems, systems, and components components important important to safety safety to an ultimate heat ultimate heat sink (UHS)

(UHS) shall be provided. provided. safety function shall The system safety shall be to transfer the combined combined heat load of these structures, structures, systems, and components systems, components under under normal operating and accident normal operating accident conditions. conditions. UFSAR Revision Revision 2121 VII-10 October 2009 2009

Nine Mile Point unit Unit 1 UFSAR

• Suitable redundancy Suitable shall redundancy in interconnections, interconnections, leak in components leak components and features, detection, and provided to assure that for onsite electric shall be provided operation (assuming offsite system operation for offsite offsite power is electric power system operation offsite electric features, and suitable isolation capabilities capabilities power electric power is not available),

available), and operation (assuming onsite power power is is not available), available), the system safety safety function can be be accomplished accomplished assuming a single single failure. failure. Heat removal removal from containment containment following a LOCA, LOCA, and transferring that energy energy to the UHS, UHS, isis achieved and assured assured through the use use of redundant pump trains trains drawing drawing suction on the suppression suppression pool pool and removing heat through a heat exchanger supplied by the raw exchanger supplied raw water pumps. pumps. The system is is designed and suitably suitably sized to to maintain maintain the torus below the NPSH temperature temperature limits of core spray containment spray. spray and containment spray. 2.0 Design BasesBases 2.1 Design Basis Functional Functional Requirements Requirements The CSS system shall perform the following functions important important to safety in in order to prevent containment pressure prevent containment pressure and temperature temperature from exceeding exceeding its its design values values for reactor coolant coolant (RCS) leaks system (RCS) leaks up to and including including the DBA,DBA, double-ended double-ended break break of a reactor coolant coolant recirculation recirculation line: 1.

1. Functional Requirement Requirement - Remove energy energy from the the drywell drywell and torus following vessel vessel leaks, leaks, up to and including a LOCA, including LOCA, to reduce containment containment temperature temperature and pressure pressure and maintain them below containment containment design pressure temperature limits.

pressure and temperature limits. Basis Basis - A A means of removing energy from containmentcontainment following a LOCA and of transferring transferring energy energy to the UHS UHS is required is required by GDC 38 and GDC 44.

44. The CSS system system provides provides the primary means of energy energy removal removal from from containment after a LOCA.

containment LOCA. 2.

2. Functional Requirement Requirement - Ensure Ensure the torus water water temperature temperature does not exceed that required to satisfy satisfy containment spray and core spray NPSH requirements.

containment requirements. Basis - Inadequate Inadequate NPSH can limit the containmentcontainment spray and containment containment raw water pump performance performance and reliability. reliability. Without adequate NPSH, Without adequate NPSH, the ability ability of of

• UFSAR UFSAR Revision the system to remove energy diminished.

diminished. Revision 21 21 VII-II VlI-II containment may be energy from containment October 2009 be 2009

Nine Mile Point Nine Point Unit Unit 1 UFSAR UFSAR 3.

3. Functional Requirement Functional isolate isolate CSS containment Basis Requirement -- Provide system containment boundary.

boundary. piping Provide the that the capability piping that penetrates Unit 1 did not commit Basis - Unit capability to penetrates the providing isolation commit to providing to valves inin the CSS system as as would be be required required toto satisfy satisfy GDC 56.

56. Containment Containment spray spray was was originally originally designed as an extension designed extension ofof primary primary containment.

containment. However, Unit However, Unit 1 has committed committed to maintaining maintaining a water water seal in seal in lieu lieu ofof leak leak rate rate testing testing of of the isolation valves. valves. 4.

4. Requirement - The Functional Requirement Functional The CSS system piping must must provide an essentially provide essentially leak-tight barrier against the leak-tight barrier uncontrolled release uncontrolled release of radioactivity radioactivity to the environment.

environment. Basis - The CSS system originally designed as an system was originally an extension of extension of primary containment. containment. As such, such, the containment spray containment spray piping must satisfy satisfy the intent of of 16 and provide GDC 16 provide an essentially leak-tight barrier essentially leak-tight barrier against against the the uncontrolled release of radioactivity uncontrolled release radioactivity to to the environment. environment.

55. . Functional Requirement Functional airborne fission Requirement - Remove airborne fission atmosphere following a LOCA products from the drywell atmosphere products which results in damage, to limit significant fuel damage, in significant limit fission product releases releases from containment leakage containment leakage paths.

paths. radiological analysis Basis - The LOCA radiological implementing analysis implementing alternative source term* the alternative term. (AST) methodology methodology described in Regulatory Guide in Regulatory Guide (RG) 1.183 credits (RG) 1.183 airborne fission product removal by the CSS system. The AST analysis is is described described in in Section XV.xv. 2.2 Controlling Parameters Controlling Parameters requirements of Section VII-B-2.1, To meet the design requirements VII-B-2.1, the CSS CSS system must be capable of meeting operational meeting the following operational requirements: requirements: CSS pump flow through the drywell sparger nozzles must must be ~3300

               Ž3300 gpm.

gpm. CSS pump CSS through the flow through pump flow the torus torus sparger nozzles must sparger nozzles must be ~300

               Ž300 gpm.

gpm. UFSAR Revision 21 21 VII-12 October October 2009

Nine Mile Point Unit 1 UFSAR UFSAR ess drywell and torus sparger spray droplet size must CSS must be ~1000

1000 microns.

microns. ess CSS pump flow in in the torus cooling cooling mode must be Ž2800 ~2800 gpm. gpm. ess shell CSS shell side heat exchanger exchanger flow must be Ž3600 ~3600 gpm gpm containment spray) (during containment spray). . ess pump available CSS available NPSH must be Ž34.2 ~34.2 ft ft for the most most restrictive restrictive case (least NPSH margin) in in which two pumps are operating operating through separate separate strainer strainer assemblies assemblies at a flow rate of 3759 gpm. gpm. ess raw water CSS water pump flow, through the heat exchanger exchanger tube side, ~3000 gpm. side, must be Ž3000 gpm. ess CSS raw water available NPSH must be ~31 water pump available ft.

                                                                     Ž31 ft.

ess drywell and torus sparger CSS sparger nozzle pressure pressure must be

            Ž30
            ~30    psi   above  containment   pressure containment pressure       for    a sufficient sufficient number of nozzles to achieve minimum required                  flows.

required flows.

• ess CSS ess least spray header pressure must be 110 percent of containment containment pressure CSS heat heat exchangers exchangers must

                                           ~38.5 psig.

pressure or Ž38.5 psig. be capable must be capable of least 120 million Btu/hr, with two containment of removing at of removing containment spray at pumps operating pumps operating and a spray water temperature temperature reduction from 140°F reduction 100 0 F. 140OF to 100°F. 3.0 System Design 3.1 3.1 System Function The ess CSS system is is an engineered engineered safeguards safeguards system designeddesigned to to prevent overheating overheating and overpressurization overpressurization of the containment, containment, reduce reduce drywell airborne airborne fission product concentrations, concentrations, and control pressure suppression control the pressure suppression chamber waterwater temperature following a design basis LOCA. LOCA. The system system is is designed designed to to provide heat removal capabilities removal capabilities for vessel leaks up to and including the DBA, double-ended break of a reactor DBA, the double-ended reactor recirculation line, without recirculation line, without core spray system operation. operation.

• UFSAR UFSAR Revision Revision 21 21 Vll-12a VII-12a October 2009

Mile Point Unit 11 UFSAR Nine Mile 3.2 As shown redundant System Design Description shown on Figure redundant loops. the primary Figure VII-3, VII-3, the CSS system loops. The primary system isis designed primary loop (Loop 11) provides primary or inner drywell sparger and to the torus sparger. designed with two provides water to sparger. to The secondary secondary loop (Loop 12) 12) provides water to the secondary secondary or or outer drywell sparger and to the torus sparger. outer sparger. The torus sparger sparger is is common to both loops. loops. Each of the two loops are cross-connected through the test cross-connected test return lines such that each of of the loops can provide flow to both the primary primary and secondary secondary spargers. Each loop includes spargers. includes two redundant trains and consists redundant trains consists of two suction headers, headers, two containment spray containment spray pumps, pumps, two heat heat exchangers exchangers and the associated associated containment containment spray spray raw water pumps, pumps, a common test test return return line, and associated associated piping piping and control control valves. valves. All pumps in in a loop are powered powered from the same same emergency emergency power bus. power bus. Each loop is is electrically electrically independent independent from the other other loop. loop. The CSS system is is normally in in standby. standby. Containment spray pump Containment pump operation is automatically operation is automatically initiated initiated by two RPS signals--high drywell pressure pressure and low-low reactor reactor water level.level. Automatic Automatic initiation initiation of the containment containment spray spray pumps pumps occurs following following the the core spray spray pumps pumps and core spray topping pumps initiation. initiation. Upon Upon

• receipt of an actuating signal, the four containment actuating signal, containment spray pumps pumps are sequentially sequentially started started when poweredpowered from either either the reserve service or the diesel generators.

Station service generators. containment Upon containment spray pump initiation, initiation, self-actuating check self-actuating check valves open to allow allow containment containment spray water to flow through the system. system. The The containment spray raw water pumps must be manually containment initiated manually initiated automatic initiation following automatic containment spray pumps. initiation of the containment pumps. A 15-min delay can be tolerated in in starting starting a raw water water pumppump since it provides lake water it provides water to a containment containment spray heat heat exchanger for the purpose exchanger purpose of long-term long-term cooling cooling of the torus water. water. Each pump takes suction from the torus through individual through individual suction lines. suction lines. The water water inin each each suction suction line flows from the torus through through a suction strainer strainer assembly. assembly. Two strainers comprise each of two suction strainer assemblies. suction strainer assemblies. When When two

pumps, pumps, either 112 112 and 122, 122, or 111 and 121, 121, are operated, they operated, will take suction from the same suction suction strainer assembly.

strainer assembly. The The discharge discharge from each pump passes passes through through the shell side of a heat heat exchanger exchanger where it is cooled it is cooled prior to being being distributed to the drywell drywell and torus spray headers. headers. The spraying of the water in in the containment containment increasesincreases the heat removal removal rate, thereby decreasing decreasing containment headers UFSAR Revision 21 containment temperature headers inside the drywell 21 temperature and pressure. drywell and torus are arranged VII-12b VII-12b pressure. The spray arranged to distribute distribute October 2009

Nine Mile Point Unit 11 UFSAR

• water water as uniformly as possible direction of spray impact possible throughout spray from the nozzles throughout the free volume.

nozzles is impact on equipment and allow as much free-fall maximize maximize steam condensation. containment condensation. containment spray pump discharge In free-fall In addition, flow from the volume. arranged to minimize is arranged minimize as possible possible to directed to the torus discharge can be directed The The to via a 6-in testtest return line that provides suppression suppression poolpool cooling. cooling. Each of the containment containment sprayspray heat exchangers is heat exchangers is supplied cooling water from a dedicated cooling containment spray raw water pump. dedicated containment pump. Each containment containment spray raw water pump takes suction suction from the condenser condenser circulating circulating water water intake tunnel. tunnel. The pump discharge discharge passes passes through through a duplex strainerstrainer prior to entering the tube side of the containment containment spray spray heat exchanger. exchanger. After passing through the heat exchanger exchanger and coolingcooling the suppression suppression pool water, water, the raw water is released water is released to the discharge manifold. discharge manifold. In the event of a total In total loss of the containment containment spray primary water source source (suppression chamber chamber water below the containment containment spray spray pump suction suction level), level), raw water water pumps pumps 112 and 121 can be be aligned to supply the containment containment spray spargers to provide an an alternate source alternate source of containment cooling. containment cooling. Likewise, Likewise, raw waterwater ill and 122 can be aligned to supply the core pumps 111 core spray system. system. 3.3 System Design originally designed The CSS system was originally designed to operate operate with Loop Loop 11 and Loop Loop 12 flow paths in in the drywell totally drywell as totally independent independent redundant systems. redundant systems. However, in However, in order to satisfy 10CFR50 10CFR50 Appendix Appendix J,J, paragraph paragraph III.C.3(b) requirements, requirements, the current current standby standby configuration configuration of the system provides provides flow to both primary primary and secondary secondary spargers, spargers, with two pumps including either either train train 111 or 122 122 inin operation, operation, to form a water seal. seal. This isis accomplished accomplished by cross-connecting cross-connecting the two trains trains via the test test

• UFSAR UFSAR Revision Revision 21 21 VII-12c VII-12c October 2009

Mile Point Unit Nine Mile Unit 1 UFSAR THIS PAGE INTENTIONALLY INTENTIONALLY BLANK Revision 21 UFSAR Revision 21 VII-12d VII-12d October 2009 October 2009

Nine Mile Nine Mile Point Unit 1 UFSAR

• decay heat decay reaction.

reaction. heat and chemical analysis shows that more analysis exists in in the system. chemical energy from a 70-percent With a maximum possible more than sufficient 70-percent metal-water reaction of 27 percent, possible reaction sufficient heat system. This analysis requires satisfy the analysis input assumptions discussed in metal-water percent, the heat removal capacity requires the CSS system system to in Section to XV-C-5.3.2. XV-C-5.3.2. containment spray distribution of containment determine proper distribution To determine spray through the nozzles, nozzles, testing was performed performed on a sample sample spray spray nozzle nozzle of of the size and type used in containment spray. in containment spray. Water was run Water through the nozzle at various pressures pressures from 10 psig to 100 psig, and spray pattern pattern and spray particle particle fineness was was observed. observed. Pressure Pressure drops of 80 psig and 30 psig represent represent the original system configuration pressure conditions for two-pump configuration pressure two-pump operation operation and one-pumpone-pump operation, respectively. operation, respectively. The particle particle sizes for the two-pump operation are in two-pump operation in the range of 10 to 400 400 microns. microns. For one-pump operation, one-pump operation, particle particle sizes range range from 500 500 microns. to 1000 microns. The CSS system design design flow, spray distribution/droplet distribution/droplet size, size, and fall heights were used to determine fall determine the airborne fission product product removal rate for implementation implementation of the AST methodology methodology describeddescribed in in Section XV. xv. 4.2 Response System Response After initiation After an initiation signal is received, signal is received, there is is a time delay delay of of 20 sec to allowallow the core core spray and core core spray spray topping pumps to to start. start. At the 25-sec mark, mark, containment containment spray spray pumps pumps 111 and 121 121 receive a start will receive start signal, signal, and at 30 sec, containment sec, containment spray pumps 112 pumps 112 and 122 will receive their start start signal. signal. If If the core spray and core spray topping pumps do not start, spray start, a set of backup contacts will start timer contacts containment spray start start the containment start sequence sequence in 50 sec to allow the core spray starting in starting logic to be initiated initiated time. a second time. This will cause cause pumps 111 and 121 121 to start start at 55 55 sec, and pumps 112 and 122 to start sec, start in 60 sec. in sec. This interlock, interlock, delaying the starting of the containment delaying containment spray pumps, is spray pumps, is provided provided to avoid overloading overloading of the diesel generators. diesel generators. 4.3 Interdependency With Other Engineered Interdependency Safeguards Systems Engineered Safeguards Systems CSS system is The CSSsystem is used in conjunction with the core in conjunction core spray spray systemsystem described in described Section VII-A. in Section The core spray system removes heat heat from the corecore in event of a LOCA. in the event LOCA. In In the heat removal removal process, the core process, spray water core spray converted to steam, is converted water is steam, which is is

• released to the containment.

then released UFSAR Revision Revision 21 containment. VII-14e containment sprays The containment sprays October 2009 October

Nine Mile Point unit Unit 1 UFSAR condense the steam in condense containment in the drywell drywell and remove containment vessels through heat exchangers. The raw water pumps are interconnected system and the containment remove heat from the exchangers. interconnected with the core spray containment spray loops to provide provide an emergency emergency source of water. water. Raw water water pump 112 can supply water water to to containment spray containment spray train 122, and raw water pump train 122, pump 121 can supply water to containment containment spray train 111. The motor-operated motor-operated valves valves between raw water and containment containment spray water are interlocked with the heat exchanger exchanger raw water discharge discharge valves. valves. If If oneone valve valve is is open, open, the other must be closed. closed. In addition, raw water In addition, water pump 111 isis connected connected to core spray pump train 11 and raw water water pump 122 isis connected connected to core spray pump train 12. 12. The The air-operated valves located on the connection air-operated connection betweenbetween the two systems are also interlocked systems interlocked with the raw water discharge valves. valves. The following systemssystems must be in in operation operation to support the CSS CSS system: system: Instrument air Instrument air must must be be operational operational to to permit permit operation operation of the containment containment spray inlet inlet isolation valves and isolation valves bypass blocking valves. blocking valves. 4.16-kV 4.16-kV and and 600-V 600-V ac ac power power distribution distribution systemssystems are are required required to provide power to the containment containment spray pumps, raw water pumps, pumps, pumps, and isolation valves. isolation valves. The The RPS RPS system system isis required required toto provide provide automatic automatic initiation initiation signals to the containment containment spray pumps pumps and waste disposal isolation valves. disposal isolation valves. The The process process radiation radiation monitoring monitoring systemsystem must must be be operational operational to alert alert Operators Operators of leakageleakage of of contamination contamination into the raw water water system due to heat heat exchanger leaks. exchanger leaks. 5.0 System Operation Operation 5.1 Limiting Limiting Conditions for Operation Operation The limiting conditions for operation (LCO) (LCO) pertaining pertaining to the CSS system are listed listed inin Section 3.3.7 of the Unit 1 Technical Technical Specifications. Specifications. associated with generic Other LCOs associated generic equipment equipment and programs are are also applicable applicable and are listed listed in other in other sections. sections. The intent of the LCOs is of the system are operable UFSAR Revision 21 operable when is to ensure that both loops fuel VII-14f VII-14f is is inin the vessel and loops the October 200~ 200 9-

Nine Mile Point Unit 11 UFSAR Nine

• reactor containment containment coolant temperature reactor coolant cooling, airborne cooling, reduction redundancy spray temperature is loop will is greater provide airborne fission product removal, reduction for the DBA. DBA.

the 215 0 F. greater than 215°F. required However, to provide sufficient However, satisfy the single failure criterion, redundancy to satisfy One One containment required containment removal, and pressure sufficient criterion, both loops loops of the CSS system are required required to be operable. operable. If a redundant If redundant component component in in one loop of containment containment spray or its its associated raw water loop becomes inoperable, associated operation may inoperable, operation may continue provided continue provided the component component is is returned returned to an operable operable condition condition within 15 days. days. If a redundant If redundant component component in in both containment containment spray loops or their associated associated raw water loops loops becomes inoperable, becomes operation may continue inoperable, operation continue provided the component component is is returned returned to service service within 77 days.days. In both cases, In cases, additional surveillance requirements additional surveillance requirements are imposed. imposed. If a If containment containment spray loop or its its associated raw water associated water loop becomes becomes inoperable and all inoperable all components of the other loop are the components operable, operable, the reactor reactor may remain remain in in operation operation for a period not not to exceed exceed 7 days. days. If the LCOs are not met, If met, then a normal shutdown shall be normal orderly shutdown be initiated initiated within 1 hr and the reactor shall be placed placed in in cold shutdown within 10 hr. hr. 6.0 Tests and Inspection To ensure ensure that the performance performance of the CSS system continues to system continues to meet the design requirements, requirements, the following surveillance surveillance tests tests and inservice inservice inspections inspections requirements satisfied. requirements must be satisfied. ASME Section XI ASME Section inservice examination XI inservice examination of of components components ASME ASME OMOM Code Code inservice inservice testing testing of of pumps pumps and and valves valves ASME Section XI ASME Section XI system system pressure pressure tests tests Appendix Appendix JJ leakleak rate testing rate testing System operability System operability surveillance surveillance tests tests Several Several programs established to meet the requirements programs have been established requirements of the ASME Code and AppendixAppendix J. J. include: These include: 1) NMP1 lSI

1) ISI Program Program Plan, 2) 2) Inservice Pressure Pressure Testing Program Program Plan, 3) 3)

Pump and Valve Inservice Inservice Testing Testing Program Plan, and 4) 4) Appendix J Testing Program Plan. Plan.

• UFSAR Revision 21 VII-14g October 2009 October 2009

Nine Mile Point Unit 1 UFSAR The following CSS system tests, are conducted Containment Containment Spray verifies valve, tests, inspections, conducted to meet the requirements. Spray System inspections, and surveillances requirements. System Quarterly Quarterly Operability valve, pump and total system operability surveillances Operability Test - operability and operation of valve limit switches verifies operation switches and solenoid-operated valves solenoid-operated valves Containment Spray Containment Header and Nozzle Spray Header Nozzle Air Flow Test - verifies header, header, header check valve, and nozzle check valve, operability Containment Spray Containment Spray System System Suction Suction Valve Operability Test - verifies verifies valve operability valve operability Containment Spray Containment Spray Valve Valve Remote Positionposition Indicator Indicator Verification Verification - verifies operability of indicators verifies operability indicators Containment Spray Pressure Containment Pressure Test - verifies integrity integrity of the system by VT-2 visual examination Containment Spray Raw Water Pressure Test - verifies Containment integrity of the system by VT-2 visual examination integrity examination Containment Spray Raw Water System Containment System Intertie Valve Operability Operability Test - verifies verifies the operability operability of the containment spray/core containment spray/core spray intertieintertie valves check valves Testing of the initiating initiating instrumentation and controls instrumentation controls portion of the system is discussed system is discussed in in Section Section VIII. VIII. The emergency emergency powerpower

system, system, which supplies electricalelectrical power power to containment containment spray in in the event that offsite offsite power power isis unavailable, unavailable, is is tested as as described in described in Section Section IX. IX. Visual inspections inspections of all all system system components located outside components located drywell can be made at any time outside the drywell operation.

during power operation. Components Components inside the drywell drywell can be be visually inspected inspected only during periods of access to the drywell. drywell. UFSAR Revision 21 21 Vll-14h October 2009

Nine Mile Point Unit 1 UFSAR ~ C. C. LIQUID POISON INJECTION SYSTEM POISON INJECTION SYSTEM 1.0 Design Bases Bases The liquid poison injection injection system is is provided provided to bring the reactor reactor to a cold shutdown condition condition at any time in in core life life independent independent of the control rod system capabilities. capabilities. Cycle-specific analysis results Cycle-specific results are contained contained in in the SRLR(7). SRLR(7). The primary primary requirement requirement imposed on the liquid poison poison injection system is to shut down the reactor from a full-power operating system is condition, assuming complete failure condition, failure of the withdrawn withdrawn controlcontrol rods to respond to an insertion signal. insertion signal. Injection of liquid poison is is also required required following a large break LOCA to maintain the suppression suppression pool water pH ~> 7.0 in pool water support of the AST in support AST methodology. methodology. For the design rating of 1850 MWt, concentration of 109.8 ppm MWt, a concentration ppm of boron-10 boron-10 isotope (equivalent to 600 ppm of natural natural boron) is is required required inin the reactor to meet the reactor reactor shutdown requirement. requirement. However, additional 25-percent However, an additional 25-percent marginmargin is is included included inin the calculation calculation of required required liquid poison poison tank concentrations concentrations to allow for nonuniform nonuniform mixing of the liquid poison poison as it it isis injected injected into the reactor. reactor. The same tank ~ concentration concentration level has been determined to adequately the AST support function for controlling controlling pH above adequately satisfy above 7.0. 7.0. satisfy The rate rate of reactivity reactivity compensation compensation provided provided by the liquid poison poison injection injection system is designed to exceed the rate of is designed of reactivity gain associated reactivity associated with reactor cooldown from the reactor cooldown full-power condition. full-power condition. The liquid poison system is is not intended intended to be capable capable of producing producing as rapid a shutdown as is is produced produced by by scramming scramming the control rods, control rods, and should should not be construed construed as a scram backup. scram backup. Following a large break LOCA, initiation LOCA, initiation of the liquid poison poison system system within 1.5 hr after the potential potential for for significant fuel failure has been identified significant identified will ensure ensure that that suppression pool pH is the suppression is controlled controlled for at least 30 days. days. The The liquid poison injection poison injection system is actuated is actuated only by remote manual remote manual action from the control room, room, hence deliberate action. hence a deliberate injection system can be powered The liquid poison injection powered from the diesel generators and, therefore, diesel generators therefore, will be operableoperable in in the event event of a loss of normal and reserve ac power. power. The liquid poison is required system is required to function for a maximum maximum of 3 hr following pipe break events events (accidents) that produce produce harsh environmental environmental conditions. conditions. ~ UFSAR Revision 21 21 VII-15 October 2009 2009

Nine Mile Nine Mile Point Point Unit Unit 1 UFSARUFSAR Accordingly, Accordingly, EQ in in accordance accordance with I0CFR50.49 for the 3-hr with 10CFR50.49 3-hr

• post-LOCA post-LOCA mission mission time time has has been been demonstrated demonstrated for for the the electrical electrical components important to safety that comprise the liquid poison components important to safety that comprise the liquid poison system.

system. EQ in in accordance accordance with IOCFR50.49 is 10CFR50.49 is not not required required for for anticipated transients anticipated transients withoutwithout scram scram (ATWS) (ATWS) that may produceproduce harsh environmental conditions harsh environmental conditions inside containment, containment, but not in in reactor building the reactor building where where the electrical components electrical components are located. located. All All portions portions of the the system system are are designed designed for earthquake earthquake loadsloads ofof horizontal and O.lg 0.3g horizontal vertical. 0.1g vertical. 2.0 2.0 System Design The liquid injection system, liquid poison injection system, shown on Figure Figure VII-6, VII-6, consists consists of an ambient pressure ambient pressure tank with immersion heater immersion heater for for low-temperature sodium low-temperature pentaborate solution storage, sodium pentaborate storage, two high-pressure high-pressure positive positive displacement displacement pumps pumps for injecting injecting the solution into solution into the reactor reactor core,core, two explosive-actuated explosive-actuated shear shear plug valves for isolating the liquid liquid poison poison from the reactor reactor until required, required, an in-vessel in-vessel sparger sparger ring, a test test tank, tank, two isolation check valves, isolation check valves, a buffer buffer system system and additional additional valves, valves, piping piping and associated associated instrumentation. instrumentation. The liquid poison poison is is stored in 4080-gal tank which is in a 4080-gal is designed for atmospheric atmospheric pressure. pressure. This tank is is complete complete with top cover, hatch with lid lid for adding chemicals, chemicals, immersion-type immersion-type electric electric heater, instrument heater, instrument connections, connections, and nozzles nozzles for outlet, outlet, recirculation, overflow, recirculation, overflow, air air sparger and drain. The tank outlet outlet nozzle is is outfitted outfitted with a strainer, strainer, which extends extends above the tank bottom, to prevent prevent solid particlesparticles from being being discharged to to the pump suction. suction. The airair sparger, sparger, which is is used for mixing mixing the solution for each initial solution initial batch, has air air holes directed directed toward the bottom of the tank for sweeping sweeping the deposit there. there. The top cover and hatch lid lid are designed so that the solution, when when agitated by the air air sparger, will not spillover. sparger, spill over. The neutron absorber in in the sodium pentaborate pentaborate liquid poison solution isis the boron-l0 boron-10 isotope. isotope. The relationship relationship between between concentration and boron-l0 liquid poison solution concentration boron-10 enrichment enrichment is is contained in in the equivalency equation. equation is: The equation is: C 628300 628300 Q Q E

                   ]3% wt    x X      M     x 86 gpm X X          x 19.8% Atom >1 13%    wt         M        86 gpm      19.8% Atom -

Revision 21 UFSAR Revision 21 VII-16 October 2009

Nine Mile Point Unit 11 UFSAR Where: Where: c = C Sodium pentaborate Sodium solution concentration pentaborate solution concentration (wt %)  %) M = Mass of water in in reactor reactor vessel vessel and recirculation piping at hot rated conditions piping conditions (501500 (501500 lb)lb) Q = Liquid poison pump flow rate (30 gpm nominal) nominal) E = Boron-10 enrichment enrichment (Atom %)  %) saturation temperature The saturation temperature variesvaries with solution concentrations concentrations of sodium pentaborate pentaborate as shown on Figure VII-7. VII-7. This saturation curve curve has 5OF5°F margin margin above the actual actual saturation saturation temperature temperature to to prevent precipitation prevent precipitation of sodium pentaborate pentaborate while in storage. in storage. The liquid poison contains a minimum volume of 1325 gal of poison tank contains of sodium pentaborate sodium pentaborate solution solution whose whose (boron-10) enrichment enrichment and concentration concentration conform to the equivalencyequivalency equation. equation. To compensate evaporation which compensate for evaporation precipitation, which could lead to precipitation, the storage tank was oversized. oversized. The nominal nominal tank capacity capacity of of 4080 gal allows allows additional water to be added to the solution as as a safety margin against evaporation evaporation losses. losses.

• Temperature and liquid level Temperature annunciated annunciated in The 50-kW, 50-kW, 550-V three-phase level alarms for the storage in the control room. room.

three-phase immersion controlled by a temperature immersion heater temperature indicator storage tank are heater is automatically is automatically controlled indicator controller. controller. High- and High-low-temperature low-temperature annunciators annunciators are providedprovided to assure that the solution isis above saturation saturation temperature. temperature. test Pump test results indicate indicate adequate adequate NPSH is is available available at solution solution temperatures temperatures up to and including 105 0 F. including 105°F. Solution temperatures Solution temperatures up to 130°F 130OF have been analyzed analyzed and also provide adequate NPSH. provide adequate NPSH. To increase increase the rate of sodium pentaborate pentaborate solution in in water, water, a manual override on the temperature controller permits temperature controller permits heater operation for 150°F heater operation 150OF solution temperature. temperature. This manual override override may render the system inoperable. inoperable. An indicator indicator lamp is is provided provided to denote when denote when heater element is the heater is shorted to the solution. Should the immersion heater fail immersion heater fail during during Station operation, operation, no action need need be taken. taken. Normally, Normally, the building building heating systemsystem will maintain maintain the required temperature. required tank temperature. immersion heater is The immersion is used to to supply the endothermic endothermic heat required during solution required during solution mixing and only incidentally incidentally to maintain temperature. maintain solution temperature. If a If failure of the building heating heating system occurs simultaneously simultaneously with a failure of the immersion heater, heater, the ambient ambient temperature in the liquid poison in poison system area will will Revision 21 UFSAR Revision 21 VII-17 October 2009

Nine Mile Point Unit 1 UFSAR decrease decrease very slowly building location. building provide slowly due to its temporary heating. provide temporary The sodium pentaborate Therefore, heating. pentaborate solution is its Therefore, there large large mass and its there will be ample its interior interior ample time to is delivered to the reactor by to by one of two 30-gpm, 30-gpm, positive positive displacement displacement pumps,pumps, with a design discharge pressure pressure of 1670 psig. The pumps and pipingpiping are protected from overpressure protected overpressure by two relief relief valves which valves which discharge back back to the poison poison storage storage tank. tank. The relief relief valves are set to to open at a pressure between 1455 and 1545 pressure between 1545 psig. psig. The injection injection pumps produce a flow rate sufficient sufficient to meet the injection requirements for all injection requirements conditions of reactor all conditions reactor operation up to the primary primary system design pressurepressure of 1250 1250 psig. psig. Two Two pumps are provided to give complete pumps complete redundancy. redundancy. The pumps are specifically specifically designed designed for standby service to be operated operated infrequently, infrequently, only during emergencies emergencies and testing. Each Each operation is operation is for 3 hr maximum. maximum. Since the liquid poison injectioninjection system system is is to be operable operable in in the event event of loss of normal and reserve reserve ac power, power, one pump pump is is connected to PB 102 and the other to PB 103. connected 103. These boards are powered from the diesel powered diesel generators generators in in the event of failure of of their normal normal supply as described described in in Section IX, IX, Electrical Electrical Systems. Systems. A radiant heat shield is installed between is installed between the two liquid poison pumps to prevent fire fire damage to the redundant redundant pump in in the event event of a fire fire in either pump. in pump. The explosive explosive valves valves are double squib-actuated shear double squib-actuated shear plug valves. valves. A low-current electrical A low-current electrical monitoring monitoring system gives gives visible (pilot visible (pilot light) and analog (ammeter) indication of circuit (ammeter) indication circuit continuity through both firing squibs in continuity in each valve. valve. Operation of one valve valve provides provides sufficient sufficient flow passagepassage to meet the required flow rate. rate. Two valves provided to give complete valves are provided complete redundancy. redundancy. The firing reliability reliability of these explosive valves is is in in excess excess of of 99.99 percent. percent. The approximate approximate firing current is is 2 amps and the operating operating time at 22 amps is is a nominal 0.002 sec. sec. The products products of the explosion completely contained. explosion are completely contained. The buffer system is is composed composed of gas-charged diaphragm gas-charged diaphragm accumulators accumulators of the capacity required capacity required to absorb fluid pulsation initiated.by initiated by the UPSAR Revision 21 UFSAR the positive 21 positive displacement displacement pumps. VII-18 pumps. Each is is located October 2009

Nine Mile Point Unit 11 UFSAR

• as close as possible to its pump loop has an accumulator.

accumulator. Containment isolation is Containment liquid poison pipe,pipe, one check its respective pump discharge. respective provided by two check valves is provided check valve just discharge. valves in just outside the drywell Each Each in the drywell penetration penetration and the other check valve inside the drywell. drywell. additional check valve is An additional is installed downstream of each each relief relief valve connection. connection. The purpose of each check valve is is to prevent prevent flow through through an assumed defective defective relief relief valve of the idle pump pump loop while the second second loop is in operation. is in operation. This ensures ensures that that the capacity second pump remains unaffected. capacity of the second unaffected. The liquid liquid poison sparger in in the reactor pressure pressure vessel (RPV) is is a I-in stainless 1-in stainless steel pipe which is is fastened to the inside of the vessel shroud below below the core support plate. support plate. This 360-deg 360-deg sparger has ten 1/4-in drilled holes which are distributed sparger and which spray toward equally around the sparger toward the bottom of of the vessel. vessel. The liquid poison is is thereby mixed with the reactor reactor recirculating recirculating waterwater as it it enters the reactor reactor fuel assemblies, assemblies, assuring uniform aa mixture assuring as uniform mixture of poison as practical. practical. During injection following aa LOCA, injection LOCA, the solution is is mixed mixed in in the vessel vessel bottom head with core core spray spray water flowing through the reactor reactor core and out the break. break. test A test demineralized water supply tank and demineralized supply are an integral part integral part facilitate of the system to facilitate system testing and flushing. All All piping in in the system has been designed designed in in accordance accordance with ASA ASA B31.1-1955 Piping Code. B31.1-1955 Code. Tanks are constructed constructed in in accordance accordance with API 650.650. pressure-bearing parts The pressure-bearing parts of the pumps are built built in accordance in accordance with ASME Code Section Section III,III, Class C-1965.C-1965. Actuation of the liquid Actuation liquid poison system system is is manually manually initiated initiated from from the control control room, assuring that poison injection is room, assuring is caused caused by a deliberate deliberate act.act. 2.1 Operator Assessment Operator Assessment The Operator Operator can assess operation operation of the liquid poison system by means of pressure indication and pump motor ammeters on the main pressure indication main control panels. control room panels. Each explosive explosive valve valve has a low-current low-current monitoring system with an ammeter and lights on the electrical monitoring electrical the main control control room panel. panel. The ammeter and lights provide indication circuit continuity through both firing squibs in indication of circuit in each valve; valve; the ammeter and lights ensure ensure firing readiness. readiness.

• When fired, the circuit is UFSAR Revision 21 broken, the ammeter is broken, VII-19 ammeter reads 0, 0, the October 2009 October

Nine Mile Nine Mile Point Point Unit 11 UFSAR indicating indicating lights go out, alarms. alarms. The pressure pressure transmitter provided provided with with suitable out, and the control transmitter is is in suitable valving valving so that control room annunciator in an accessible annunciator accessible location location and is it can be that it be tested tested at is at any time. time. The liquid liquid poison tank is is provided provided with control room with control room level level indication indication in in addition to alarmsalarms for high-high- and and low-level low-level and high-high- and low-solution temperature. and low-solution temperature. 3.0 Design Evaluation Evaluation The liquid liquid poison system is is designed designed to provide the capability to bring bring the reactor reactor from full full design design rating rating (1850 (1850 MWt) MWt) to a xenon-free shutdown condition cold, xenon-free condition assuming none none ofof the control control rods can can be inserted, and to buffer the suppression be inserted, suppression pool pool water water following a large break break LOCA. LOCA. To meet meet the shutdown objective, objective, the system system is is designed to injectinject a quantity quantity of which of boron which produces produces a concentration concentration of at least 109.8 least 109.8 ppm of boron-10 isotope inin the reactor reactor core. core. This concentration concentration will bring the reactor reactor from full full design design rating rating (1850 (1850 MWt) MWt) to a subcritical subcritical condition considering condition coolant coolant voids, samarium. samarium. considering the combined effects voids, temperature temperature change, The same quantity suppression pool pH ~Ž 7.0 for 30 days following suppression results results inin significant results are results significant fuel damage. are contained contained in damage. SRLR(7 ). in the SRLR(7). effects of the control change, fuel doppler, control rods, doppler, xenon, quantity of boron will maintain the xenon, and the following a LOCA that Cycle-specific that Cycle-specific analysis analysis The minimum minimum required required tank storage storage volume volume and conformance conformance of the concentration and boron-10 isotope liquid poison solution concentration enrichment assures that the expected enrichment expected liquid poison solution will will provide the required required 109.8 109.8 ppm of boron-10 isotope to the reactor core or the necessary necessary buffering solution to the suppression pool. suppression pool. The liquid poison storage tank volume concentration concentration requirements requirements assure that the above requirements requirements for boron solution insertion are met with one 30-gpm 30-gpm liquid poison pump. pump. Normal level is is maintained between 1400 and 1700 gal. gal. The quantity of boron-10 required to be stored in isotope required in solution includes an additional additional 25 percent margin beyond the amount needed to shut down the reactor to allow for any unexpected non-uniform mixing. mixing. The The requirements include consideration minimum tank volume requirements consideration for 197 197 gal of solution which is is contained contained below the point where the pump takes pump takes suction suction from inserted into the reactor. UFSAR Revision 21 21 from the tank and, therefore, VII-20 VII-20 therefore, cannot be October 2009

Nine Mile Point Unit 1 UFSAR

• maintained saturation temperature The solution saturation concentration concentration of sodium maintained by Technical saturation saturation temperature Temperature and liquid Temperature temperature varies sodium pentaborate.

pentaborate. varies with the Solution Solution temperature Specification at least 5°F Technical Specification temperature to guard liquid level temperature is 5 0 F above guard against precipitation. precipitation. level alarms for the system system are is annunciated in annunciated in the control room. control room. Equipment and piping piping are designed designed to withstand withstand the most severe conditions conditions of loads including earthquake. Nozzles including the design earthquake. Nozzles leading into the reactor leading reactor vessel have been been designed designed taking into account possible possible vessel movement due to an earthquake.earthquake. Availability of emergency Availability emergency diesel generator generator power to both of the injection pumps injection pumps assures operability operability of the system if if required during a loss of normal and reserve ac power. power. 4.0 Tests and Inspections Inspections The system has been been designed designed to permit periodic testing, maintenance, maintenance, and operation of the injection operation injection pumps pumps and appropriate valves. appropriate valves. The pumps and valves will be tested periodically to ensure operability. periodically operability. Monthly pump tests Monthly tests are performed during Station operation performed either with demineralized operation either demineralized water recirculated recirculated to the test test tank, tank, or with the solution recirculated to the poison tank. recirculated tank. The isolation isolation valves may be be tested tested only during shutdown. shutdown. For explosive explosive valve tests, valve tests, the valves valves are dismantled dismantled and inspected. inspected. The charges charges are removed removed and replaced replaced with new chargescharges periodically periodically and the old charges charges test are test establish a rational charge fired to establish charge replacement replacement frequency. frequency. A demineralized water A demineralized water purge purge system is is provided provided so that the remaining portion portion of the systemsystem may be tested by pumping demineralized water through the distribution demineralized distribution system and into the reactor reactor vessel once once each operating cycle. each operating cycle. concentration and boron-10 enrichment Boron concentration enrichment of the solution will will be periodically determined by analysis. periodically determined analysis. temperature of the The temperature solution solution will be monitored annunciated in monitored and annunciated in the control room room to assure assure that the solution solution isis above its saturation temperature. its saturation temperature. A A continuity continuity check check of the firing circuit circuit on the explosive explosive valvesvalves is provided by pilot lights in is provided in the control room. control room. The functional test test and other surveillance surveillance of components, components, along with the monitoring instrumentation, instrumentation, gives gives a high reliability reliability

• for liquid poison UFSAR UFSAR Revision poison injection Revision 21 21 VII-21 VII-21 operability.

injection system operability. October 20092009

Mile Point Nine Mile Point Unit 1 UFSAR 5.0 In In the list Alternate Alternate Boron the event event the list alternate One alternate methods One method, method, referred Boron Injection the liquid liquid poison poison system methods for injecting referred to as system is injecting boron as the alternate is not available, available, the boron into the alternate boron the reactor. boron injection the EOPs reactor. injection system, EOPs system, provides provides for a portableportable pneumatic pneumatic hydrohydro pump connection connection between between overflow/drain line and poison tank overflow/drain the liquid poison and the liquid liquid poison injection injection line drain drain valves, valves, as shownshown on on Figure Figure VII-6. VII-6. BoratedBorated water is then suctioned water is then suctioned from the liquid poison poison tank through through the pump and discharged hydro pump discharged into the the existing liquid poison existing liquid poison injection line. injection line. air supply required The air required for the pneumatic pneumatic hydro pumppump can be be provided from a 1-in connection to the house I-in connection service air, house service air, or or from the instrument instrument air air system if if house house air is not available. air is available. The portable portable hydro pump pump hashas a design flow rate of 7.5 7.5 gpm at at 1460 1460 psig. psig. The design design pressure pressure and flow rate of of the hydro pump pump sufficient to provide are sufficient provide flow to the vessel under under a worst-case worst-case vessel vessel pressure of 1339 1339 psig psig using using an enriched enriched boron boron solution. The hoses (suction, discharge discharge and air air hose) and the portable hydro pump for the alternatealternate boron boron injection injection system are stored

• in the vicinity of the 55-gal in 55-gal drum in in the reactor reactor building.

building. The hoses are in in a locked compartment compartment to assure assure their availability. availability. The alternate alternate boron injection injection system is is nonsafety nonsafety related and nonseismic, as the additional hoses, nonseismic, hoses, pump, valves, and hose pump, valves, hose connections do not perform connections safety-related function and are perform a safety-related downstream of the safety-related downstream safety-related portions of the liquid poison system. system. The other other alternate boron injection method, method, using the reactor reactor cleanup system, provides for filling cleanup filling filter the cleanup filter with a boron solution and injectinginjecting the solution into the vessel vessel by placing the filter filter in service. in service. Neither alternate system is Neither alternate is expected to be available available for for injection following a LOCA, injection LOCA, since harsh environments environments in in the reactor reactor building will prevent the required operator required operator access.access. UFSAR Revision 21 21 VII-21a October 2009 2009

Nine Mile Point Unit 11 UFSAR INTENTIONALLY BLANK THIS PAGE INTENTIONALLY

• UFSAR Revision 21 21 VII-21b October 2009 2009

Nine Mile Point Unit 1 UFSAR D. CONTROL ROD VELOCITY LIMITER CONTROL 1.0 Design Bases Bases The control rod velocity velocity limiter, limiter, in conjunction with the rod in conjunction worth minimizer minimizer (RWM), (RWM) , is provided to limit any accidental is provided accidental reactivity addition to rates for which the resulting reactivity resulting excursion pressure vessel or impair rupture the pressure would not rupture operation of any impair operation safeguards equipment. safeguards equipment. reactivity addition The worst reactivity occurs addition occurs during the control accident (CRDA) control rod drop accident (Section XV), (CRDA) (Section XV), the consequences of which are reactivity consequences dependent. reactivity rate dependent. The control rod velocity velocity limiter is is an engineered safeguard that engineered safeguard that originally designed to limit the free-fall drop velocity was originally velocity of of control rod to 5 ft/sec or less and, the control thus, limit the rate of and, thus, of reactivity addition. reactivity addition. analysis Subsequent testing and analysis Subsequent maximum rod drop velocity of 3.11 ft/sec for use demonstrated a maximum demonstrated in in CRDA analyses(6). analyses(6). The CRDA can only happen in in the event event of of procedural violations simultaneous procedural simUltaneous violations and equipment malfunctions, a equipment malfunctions, separation mechanical failure in separation or mechanical in the drive line, sticking or line, sticking or control rod, binding of the control binding withdrawal of the detached rod, the withdrawal control rod drive (CRD) mechanism, and then the release of the (CRD) mechanism, means. The rod velocity unspecified means. control rod by some unspecified limiter velocity limiter is designed is consequences of the drop of the maximum designed to limit the consequences maximum worth control rod without significantly hindering the normal significantly hindering normal system. function of the system. The most probable probable threshold for potential mechanical damage potential mechanical to to the reactor core or other primary reactor core components cooling system components primary cooling is is a peak fuel enthalpy in cal/g. in excess of 425 cal/g. reducing the By reducing velocity of a free-falling free-falling rod, rod, and assuring that excessive excessive rod established, the CRDA will result in patterns are not established, worth patterns in peak fuel enthalpy values below the design limit of 280 cal/g.(') enthalpy values cal/g.(l) 2.0 System Design System The control control rod velocity limiter is is an integral part of the bottom of each control rod, as shown on Figure VII-9 (typical). (typical). It designed as a large is designed It is clearance piston which travels in large clearance in the control rod guide tube over the entire stroke. entire stroke. The original limiter assembly consists velocity limiter original velocity consists of two conical, machined from a single 304 stainless steel casting. elements machined elements casting. The The lower conical element is conical element relative to the upper 15-deg angle relative is at a 15-deg upper elements are separated element, and the two elements conical element, separated with fourfour apart. spacers 90 deg apart. There are no moving parts in in the velocity limiter. limiter. The rod velocity limiter provides velocity limiter streamlined profile provides a streamlined in the profile in direction and a nonstreamlined scram (upward) direction profile in nonstreamlined profile in the direction. (downward) direction. dropout (downward) It regarded as a It may be regarded nozzle-type since, during its nozzle-type limiter since, motion, aa high its downward motion,

• percentage of the total water percentage directly below the limiter water directly limiter flows up center of the limiter body and is through the center ejected radially is ejected UFSAR Revision 15 15 VII-22 November 1997 1997

Nine Mile Point Unit 1 UFSAR condenser hotwell, condenser hotwell, condensate transferred from the CSTs condensate will be transferred CSTs to the hotwell makeup. hotwell for makeup. The FWS system pumps operate on 4160 V. V. When the plant is in is in operation, the power is operation, generator through is supplied from the main generator Station service transformer the station generator is transformer when the generator is on-line on-line and grid. When the main generator connected to the grid. connected generator is off-line, the is off-line, feedwater pumps are supplied with normal offsite power feedwater power from the reserve transformers. 115 kV system through the reserve transformers. If a HPCI If HPCI initiation signal should occur, initiation occur, all HPCI/FWS system pumps would immediately with two feedwater pump trains available start immediately available for for HPCI injection injection using the single-element feedwater control single-element feedwater control system system control. If for reactor vessel level control. If a major power disturbance were to occur that resulted in 115-kV power supply to in loss of the 115-kV to bus, power would 115-kV bus, the Nine Mile Point 115-kV restored from a would be restored generator located generator Bennetts Bridge Hydro Station. located at the Bennetts station. This This generator would have the capacity generator supplying approximately capacity of supplying approximately 6,000 kVA which is operate one train of HPCI/FWS is sufficient to operate system pumps. If system pumps. If HPCI initiation were to occur, initiation were occur, the preferred feedwater train feedwater (feedwater pump 12, train pumps (feedwater feedwater booster pump 12, feedwater pump 13, 13, condensate condensate pump 13) nonpreferred train

13) would start. The nonpreferred train electrically locked out on a LOOP and not start pumps would be electrically start until the Operator Operator manually lockout by placing the manually reset the lockout backup pump control backup control switch in in the trip or close position.

position. If If a control switch had been preferred pump train pump control preferred been manually manually locked LOOP, it out prior to the LOOP, it would remain locked out and the nonpreferred train backup pump would automatically nonpreferred automatically startstart on HPCI HPCI initiation. If initiation. preferred and backup pumps are running, If both the preferred running, preferred pump would remain in the preferred backup pump service and the backup in service pump will trip. The use of a Bennetts Bridge hydro generator, Bennetts Bridge generator, while not equivalent equivalent to an onsite emergency power source, emergency power provides a source, provides highly reliable offsite power supply for the HPCI alternate offsite reliable alternate HPCI function of the FWS system. system. 4.0 Tests and Inspections Inspections Tests and inspections components are described inspections of the various components described in in section Steam-to-Power Conversion. Section XI - steam-to-Power Conversion.

• UFSAR Revision 15 15 VII-43 November 1997

Nine Mile Point Unit 1 UFSAR J. J. 1. 1. REFERENCES REFERENCES General General Electric Standard Application Electric Standard NEDE-24011-P-A-16, October NEDE-24011-P-A-16, Application for Reactor October 2007. 2007. Fuel, Reactor Fuel,

• 2.

2. R. D. Ackley, R.

R. D. R. E. E. Adams, Adams, and W. E. Browning, W. E. Browning, Jr., Jr.,

     "Removal  of  Radioactive   Methyl "Removal of Radioactive Methyl Iodide from Stearn Air Iodide   from  Steam  Air Systems,"   ORNL-4040, January 1967.

Systems," ORNL-4040, 1967. 3.

3. "Connecticut "Connecticut Yankee Charcoal Filter Yankee Charcoal Filter Tests," CYAP-101, Tests," CYAP-101, December 1966.

December 1966. 4.

4. R.

R. D. Ackley et aI, D. Ackley al, op cit. cit. 5.

5. Refer to (4),

Refer (4), pP VII-49. VII-49. 6.

6. Electric Report General Electric Report NEDO-10527, NEDO-10527, "Rod "Rod Drop Accident Accident Analysis for Large Boiling Boiling Water Reactors,"

Reactors," March March 1972. 1972. 7.

7. 0000-00B4-3226-SRLR, 0000-0084-3226-SRLR, Revision 0, 0, "Supplemental "Supplemental Reload Licensing Licensing Report for NMPl, NMP1, Reload 20, 19," December 20, Cycle 19," December 200B.

2008. 0 0 UFSAR Revision 2121 VII-44 October 2009

Nine Mile Point Unit 1 UFSAR

• 4.2 4.2.1 Offgas System Design Bases Bases Offgas system explosive Explosive Gas Monitoring System Explosive explosive gas monitoring concentration of potentially that the concentration monitoring isis provided provided to ensure potentially explosive explosive gas mixtures mixtures contained in contained in the waste gas treatmenttreatment system system is is maintained below maintained below the flammability flammability limits of hydrogen. Automatic control hydrogen. Automatic control features are included included in in the system to prevent the hydrogen concentration from reaching flammability limits.

reaching these flammability limits. Maintaining Maintaining the concentration concentration of hydrogen below flammability limits provides provides assurance that the releases of radioactive assurance radioactive materials materials will be be controlled controlled in in conformance conformance with the requirements requirements of General General Design Criterion (GDC) (GDC) 60 of Appendix Appendix A 10CFR50. A to 10CFR50. explosive gas monitoring program The explosive requirements are described program requirements in Technical in Technical Specifications. Specifications. The system is is designed designed toto withstand the effects effects of a hydrogen explosion. hydrogen explosion. The following surveillance requirements surveillance requirements and actions will be taken when when deficiencies deficiencies are identified. identified. Actions Actions

• 1.

1. A A minimum of one hydrogen channels channels operable operation system hydrogen monitor during offgas system operation.

monitor shall be operable operation. With the number operable less than the number operation of the main condenser system may continue provided gas samples operable number of number required, required, treatment condenser offgas treatment samples are of are collected and analyzed once per 8 hr. Restore the hydrogen hydrogen monitoring channel to operable monitoring channel operable status within 30 days days or outline outline in in the next Radiological Release Radiological Effluent Release Report the cause of the inoperability inoperability and how the monitoring channel was or will be restored monitoring channel restored to operable operable status. status. 2.

2. concentration of hydrogen in The concentration in the main condenser condenser offgas treatment treatment system shall be limited percent limited to 4 percent volume.

by volume. If the concentration If concentration of hydrogen hydrogen inin the main condenser condenser offgas treatment treatment system exceeds exceeds this limit, limit, restore the concentrations concentrations to within within the limit limit within within 48 hr. 4.2.2 Surveillance Requirements Surveillance Requirements 4.2.2.1 4.2.2.1 Hydrogen Monitor Operability Demonstration Operability Demonstration ~ Each hydrogen monitor shall be demonstrateddemonstrated operable operable by: by: UFSAR Revision 21 21 VIII-38a VIII-38a October 2009

Nine Mile Point Unit Unit 1 UFSAR 1.. 1 2. 2. Performance of a sensor Performance during main condenser operation. operation. sensor check check at least once per day condenser offgas offgas treatment Performance of a channel test Performance system treatment system test at least once once per month. month. 3.

3. Performance of a channel Performance calibration at least once per channel calibration per 3 months.

months. The channel calibration channel calibration shall include include the use of standard standard gas samples containing a nominal: samples containing nominal: a.

a. One volume volume percent of hydrogen, hydrogen, balance nitrogen, and b.

b. Four volume volume percent hydrogen, hydrogen, balance nitrogen.

balance nitrogen. 4.2.2.2 Hydrogen Concentration Hydrogen Concentration Requirement Requirement The concentration concentration of hydrogen in in the main condenser condenser offgas offgas treatment system shall determined to be within shall be determined percent within 4 percent hydrogen hydrogen by volume by continuously continuously monitoring the waste waste gases inin the main condenser condenser offgas treatment system in offgas treatment in accordance accordance with Section 4.2.1, Section 4.2.1, Item 1.1. UFSAR UFSAR Revision 21 Revision 21 VIII-38b October 2009 2009

Nine Mile Point Unit Unit 1 UFSAR 0 TABLE VIII-3 (Cont'd.) Type -~ EOP B C C D D E VARIABLE Category VARIABLE Category ~ 1 1 1 1 1 2 2 3 3 1 1 2 2 3 3 1 -F 1 27 2 3 3 1 1 2 2 T7* 3 19.

19. Reactor Coolant Reactor System Radioactivity Coolant System Concentration Concentration (Note 6) 6) 20,

20. Analysis of Primary Coolant (Gamma (Gamma Spectrum) (Note 7)

Spectrum) 7) 21. 2l. Primary Containment Area Primary Containment Area High Range X X X Radiation Level Radiation Level 22.

22. Containment Effluent Containment Effluent Radioactivity; Radioactivity; Noble Noble Gases (Note 8) 8) 23.

23. Radiation Exposure Radiation Exposure Rate (areas adjacent adjacent to X X X primary containment) primary containment) (Note 9) 9) 24.

24. Effluent Radioactivity; Noble Gases (from Effluent Radioactivity; (from areas areas adjacent adjacent to primary containment) containment)

(Note 8) 8) 25.

25. Feedwater Flow Feedwater Flow Rate X 26.

26. Condensate Storage Tank Water Level Condensate Storage Level X X

27.

27. Suppression Suppression Chamber Chamber (Torus) Spray Flow XX Rate, and Rate, and Valve Position Position (Note 10) 28.

28. Drywell Drywell Spray Flow Rate, Rate, and Position and Valve position (Note 10) 29.

29. Main Steam Steam Line Isolation Valve Leakage Leakage Control System Pressure Control Pressure (Note 11) 30.

30. Primary System (Note 27)

Safety/Relief Valve System Safety/Relief Valve Position Position X X I 31. 3l. Isolation Condenser Isolation Condenser Shell Shell Side Side Water Level Level X 32.

32. Isolation Condenser Condenser System Position System Valve position (Principal Flow Path) (Note 12) 12) 33.

33. Reactor Isolation Cooling Reactor Core Isolation Cooling System System Flow Flow (Injection (Injection to RPV) (Note 11)

UFSAR UFSAR Revision Revision 2121 22 of 10 of 10 October 2009 October 2009

Nine Mile Point Unit 1 UFSAR

• TABLE VIII-3 VIII-3 (Cont'd.)

(Cont'd.) NOTES (Cont'd.) (Cont'd.) Table), storage tank liquid level (Item 38 in in the Table), in in the Table), neutron flux level (non-LOCA Table), neutron (non-LOCA events) (Items 1, 12, 1, 12, and 13 in Table), and squib valve status in the Table), status (Item 67 in in the Table). Table). Therefore, monitoring system flow rate Therefore, monitoring rate is is not not considered to be necessary. considered necessary. In the SER addressing In conformance to RG 1.97 addressing Unit 11 conformance (Revision 2),2), dated November 19, dated November 1986, the NRC stated that 19, 1986, that identified instrumentation the identified instrumentation is is valid as an acceptableacceptable alternative alternative indication indication of liquid poison system flow rate. rate. 15.

15. At Unit 1 the shutdown shutdown cooling cooling system is is the functional functional equivalent equivalent of the residual residual heat heat removal removal system.

system. However, However, shutdown cooling system system flow rate is is not directly monitored. monitored. Shutdown cooling system Shutdown system flow rate rate isis adjusted as required required to control reactor coolant coolant cooldown cooldown rate (heat (heat removal) within removal) within applicable applicable limits. limits. The following parameters monitored to verify proper parameters are monitored proper shutdown cooling system operation: operation: Reactor vessel Reactor water level vessel water level (Item (Item 22 in in the the Table). Table) Shutdown system pump cooling system Shutdown cooling discharge pressure pump discharge (Item pressure (Item 68 inin the Table). Table) . cooling system Shutdown cooling Shutdown system heat exchanger tube heat exchanger side tube side (reactor coolant) inlet and outlet outlet temperatures temperatures (Item (Item 40 inin the Table). Table) . Shutdown cooling Shutdown system heat cooling system exchanger shell heat exchanger side shell side (cooling water) inlet and outlet temperatures temperatures (Item (Item 69 69 in in the Table) Table). .. - Shutdown Shutdown cooling system cooling system valve position -- flow valve position flow path path from and to the reactor reactor vessel vessel (Item 70 in in the Table). . Table) Additionally, the shutdown Additionally, shutdown cooling system is is not expected operated during to be operated during accident immediate post-accident accident or immediate post-accident conditions. conditions. It operated only in It would be operated in the long term term after the unit is is in in a normal stable shutdown normal stable shutdown condition. condition.

• UFSAR Revision 21 UFSAR 21 99 of of 10 10 October 2009

Nine Mile Point Unit unit 11 UFSAR TABLE VIII-3 (Cont'd.) (Cont'd.) NOTES (Cont'd.) (Cont' d. )

• In In the Safety Evaluation Evaluation Report (SER) (SER) addressing addressing Unit 1 conformance conformance to RG 1.97 1.97 (Revision 2), 2), dated November November 19, 19, 1986, 1986, the NRC stated stated that, based based on the identified alternate alternate instrumentation instrumentation and the design function of the shutdown shutdown cooling system, the deviation cooling system, deviation from the recommended recommended.

flow monitoring monitoring instrumentation instrumentation is is acceptable. acceptable. 16.

16. Cooling waterwater flow and cooling water temperature temperature for the core spray and containment containment spray pumps are not directly monitored. The cooling monitored. cooling water water isis recirculated recirculated pump discharge discharge flow. Pump suction is is normally from the suppression pool, suppression pool, thus torus water water temperature temperature (Item 4 in in the Table) provides indication indication of the temperature temperature of the cooling water cooling supplied to the pumps.

water supplied pumps. In addressing Unit 1 conformance In the SER addressing conformance to RG 1.97 (Revision 2), 2), dated November November 19, 1986, the NRC stated that, 19, 1986, that, based based on the identified identified plant-specific plant-specific system design features, features, the deviation from the recommended recommended cooling water water

• flow and temperature temperature monitoring instrumentation instrumentation is is acceptable.

acceptable. 17.

17. In addressing Unit 1 conformance In the SER addressing conformance to RG 1.97 (Revision 2), 2), dated November November 19, 1986, the NRC determined 19, 1986, determined that, because because Revision Revision 3 to RG 1.97 1.97 recommended recommended a Category 3 classification classification for this variable, variable, no deviation deviation in in Category exists.

Category exists. The NRC concluded concluded that the use of of Category 3 instrumentation Category instrumentation for this variable variable isis acceptable. acceptable. 18.

18. Included Included under Item Item 47 in in the Table.

Table. 19.

19. Included under Item 51 in Included in the Table.

Table. 20.

20. ability The ability determine/monitor bulk average temperature to determine/monitor temperature is necessary is necessary for this EOP Key Parameter.

Parameter.

21. specified in in NEDO-31558 NEDO-31558-A( 26) in lieu of those

21. Criteria specified -A (26) apply in specified specified in in RG 1.97.

1.97. See NMPC letters See letters NMPlL NMP1L 0765(13) and NMP1L 0813 0813(27), (27), and NRC letter letter dated February February 10, 1994 1994(28) (28) , additional information. for additional information. UFSAR Revision 21 10 of 10 10 10 October 2009 October 2009

Nine Mile Mile Point Unit unit 11 UFSAR

• TABLE VIII-3 NOTES (Cont' (Cont'd.)

(Cont'd.) (Cont'd.) d.) 22.

22. Neutron flux level below the APRM range is Neutron is not a keykey variable for accomplishing accomplishing mitigativemitigative actions for any DBA transient (including those or transient anticipated operational those anticipated operational occurrences required to be considered occurrences required considered in in the implementation implementation

[10CFR5O.62]); required Operator of the ATWS Rule [10CFR50.62]); Operator actions specified in specified in the plant EOPs for such events events can be be accomplished without reliance accomplished reliance on reactor power information below the APRM range. range. On this basis, designation of basis, the designation of Category Category 3 instrumentation instrumentation (in (in lieu of Category Category 1 instrumentation as recommended instrumentation recommended by RG 1.97) 1.97) is appropriate is appropriate for monitoring intermediate range and source monitoring intermediate source range neutron flux. 23.

23. Operator actions based on drywell Operator drywell water level would be a contingency action and, contingency and, therefore, therefore, do not meet the definition of a Type A variable.

definition variable. Since Since drywell drywell water level level is is not a RG 1.97 Revision 2 recommended recommended variable,variable, the drywell water water level recorder recorder does not need to meet meet the Category 1 criteria. Category criteria. Therefore, Therefore, a drywell drywell water level level recorder recorder is is not needed. needed. (29,30) (29,30) 24.

24. RG 1.97 recommends recommends that noble noble gas effluent effluent monitoring instrumentation instrumentation be designed designed with a range range of 1E-06 ~Ci/cc to 1E-06 paCi/cc to 1E+03 uCi/cc.

1E+03 ~Ci/cc. The range range of the offgas effluent effluent stack monitoring monitoring system (OGESMS) is system (OGESMS) is 1E-07 1E-07 ~Ci/cc uCi/cc to 1 ~Ci/ccpCi/cc (Xe-133) . The OGESMS (Xe-133). OGESMS lowerlower limit of detection of 1E-05 1E-05 pCi/cc

      ~Ci/cc     meets      the     NUREG-0737, NUREG-0737,         Item    II.F.l,       Attachment II.F.1, Attachment 1,       1, Position position (2) (2) criterion criterion of the instrumentation instrumentation range beginning beginning at normal conditions    conditions (as low as reasonably     reasonably achievable (ALARA)).

achievable (ALARA)). The OGESMS upper range limit of 1

      ~Ci/cc (Xe-133) provides
      ,Ci/cc                      provides aa safety safety margin          greater than a margin greater factor of two for the site-specific site-specific design    design basis effluent effluent release release which which occurs occurs at NMPl   NMP1 from a LOCA. LOCA.

RG 1.97 recommends recommends particulates particulates and halogens halogens instrumentation be designed instrumentation designed with a range of 1E-03 ~Ci/cc p Ci/cc to to 1E+02 pCi/cc, 1E+02 ~Ci/cc, with a 30-min sampling 30-min sampling time for detection of of significant releases, significant releases, release assessment, assessment, and long-term long-term surveillance. surveillance. With the use of OGESMS, OGESMS, the particulate samples would be collected collected by OGESMS OGESMS and taken to an onsite UFSAR Revision Revision 21 21 10a of 10 lOa 10 October 2009

Nine Nine Mile Mile Point Point Unit Unit 1 UFSARUFSAR TABLE VIII-3 TABLE VIII-3 (Cont'd.) (Cont'd.) NOTES (Cont'd.) NOTES (Cont'd.)

• facility.

facility. The onsite analysis analysis facility facility range of has a range of 1E-03 JICi/cc 1E-03 pCi/cc to 0.1 JICi/cc 0.1 uCi/cc with a 30-min sampling sampling time. time. The onsite The onsite analysis facility's upper analysis facility's upper range of of 0.1 uCi/cc 0.1 JICi/cc safety margin of provides a safety provides of two for a design basis basis effluent effluent release from a LOCA. release LOCA. NMP1's design Using NMPl's design basis effluent effluent release from a LOCA, release in lieu of 1E+02 LOCA, in uCi/cc as specified 1E+02 JICi/cc specified in NUREG-0737 in NUREG-0737 andand RG 1.97, 1.97, to determine determine doses to personnel personnel working with working the sampling media during with the during an accident, accident, the results in results in estimated estimated exposures exposures would be less less than the GDC GDC 19 limits. limits. summary, OGESMS In summary, In objective and OGESMS meets the objective and purpose of the NUREG-0737 NUREG-0737 and RG 1.97 guidance.guidance. deviations The deviations 34 from from NUREG-0737 and RG 1.97 are NUREG-0737 acceptable. (34)) are acceptable.( 25.

25. hydrogen monitoring A hydrogen system capable monitoring system capable of diagnosing beyond-design-basis accidents will be maintained beyond-design-basis accidents maintained in in 26.

26. accordance with License Amendment accordance letter letter dated October dated October 2, An oxygen monitoring Amendment No. 2006 (38)) . 2, 2006(38)). No. 191 (issued by NRC monitoring system capable of verifying inerted containment of the inerted function) will be maintained maintained in verifying the status containment (post-accident monitoring accordance with License in accordance License Amendment Amendment No. No. 191 (issued by NRC letter letter dated dated October October 2, 2, 2006(38)). 2006(38)). 27.

27. indication system for the primary The acoustic position indication relief safetyIvalves and relief system safety/valves valves has been downgraded downgraded I

Category 2 to Category from RG 1.97 Category Category 3 in in accordance accordance with NEDO-33160-A, Rev. BWROG LTR NEDO-33160-A, Rev. 1, 1, "Regulatory Relaxation for "Regulatory Relaxation for the Post Accident SRV Position Indication System." Indication System." UFSAR Revision 21 21 10b of lOb of 10 10 October 2009

Nine Mile Point Unit 1 UFSAR Table VIII-4 (Cont'd.)

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26. Main Steam Line Isolation Valves Partially (f) (f)

Closed

27. Loss Of Power To Auxiliary Bus Or Startup (e)

Transformer

28. Loss Of Power To Protection System Motor (g) X Generator Set

29. Rod worth Minimizer Prohibitive *

30. High Flux - Varied With Recirculation Flow

• 3l. Turbine Trips (u) X X

32. Neutron Monitors Off Normal

• 33 . Liquid Poison Initiation X
• 34.

35. 36. 37. 38. Scram (Automatic Only) High Steam Flow In Main Steam Line High Temperature In Main Steam Line Tunnel Anticipated Transients Without Scram High Radiation At Refueling Platform X (bb) X X X X X X

39. High Steam Flow on Condo Tube Side X

40. High Temp. Heat Exchanger Effluent - Cleanup X System 4l. High* Pressure At Cleanup Sys. Filters (a)

42. Low Flow Cleanup Pump Suction (a)

43. High Radiation At Stack Monitor X

44. High Radiation Control Room Ventilation (a)

KEY: (a) After Time Delay. (b) A Backup To The Procedures which Require These Valves To Be Closed At All Times During Plant Operation. (c) Bypassed On "Refuel" And On "Startup" When < 600 psi. (d) Permits Withdrawal Of One Rod. (e) Program Loading On Loss Of Both Auxiliary Buses. (f) May Be Bypassed On "Startup". (g) Eight- To Ten-Second Time Delay. (h) At Higher Drywell Pressure Than Scram Value In Combination With Low-Low Water Level 34-Second Time Delay. (k) Either IRM Or APRM In Startup And APRM In Run. (1) Bypass In Refuel Or Shutdown. (m) SRM, IRM, APRM. (p) With Reactor Pressure ~ 365 psig. (r) Manually Retractable After Short Time Delay. UFSAR Revision 21 2 of 3 October 2009

3500-DIESEL GENERATOR LOADING FOLLOWING LOSS-OF-COOLANT ACCIDENT 3400

• 3380 3200 3100 3000-2'100 2000 HR/YR RATING: 2838KW 2800 MANUAL START:

CONTAINMENT SPRAY AUTO RETURN TO AC POWER OF: }

                                                                                                                                                                 ~l 2700 BATTERY CHARGER 161A!B                                                                       RAW WATER PUMP 1\1 OR 112 INSTRUMENT & CONTROL POWER UPS 162 150KW                                                     450KW 2600                                                                                                 COMPUTER MG SET 167 AUTO START OF 25013-                                                                              CONTAINMENT SPRAY PUMP 112                                                                          \

300KW MANUAL TRIP OF: 2~00 CORE SPRAY TOPPING PUMP - 320KW CRD PUMP 11 - 230KW CONTAINMENT 2388 SPRAY PUMP AUTO START OF MANUAL START OF: .11\ OR 112 2200 CONTAINMENT 300KW SPRAY PUMP 111 RBCLC PUMP 13 - 1<l0KW 300KW EMERG. SERVo WATER PUMP 11 - 120KW 2100 WATER* CHILLER 11 - 5eJKW TIE BREAKER-PBISA & 8 CLOSED - 110KW 20013-AUTO START OF CORE SPRAY TOPPING PUMP 121 320KIi I 1800 o

 <I      1700 o
 ...J                                            AUTO START OF 1600                                    CORE SPRAY PUMP 121 400KW 15013-                                                                                                                                                                                                                                                        z I-1~00                                                                                                                                                                                                                                                       <I:
                                                                                                                                                                                                                                                                    ...J

l 1300 U

                                                                                                                                                                                                                                                                    ...J
                                                                                                                                                                                                                                                                    <I:

AUTO START OF U 1200 CORE SPRAY TOPPING PUMP III \oJ 320KW U 1100 Z

                                                                                                                                                                                                                                                                    \oJ a:
                                                                                                                                                                                                                                                                    \oJ 10013-                                                                                                                                                                                                                                                       I&.
                                                                                                                                                                                                                                                                    \oJ a:
          '100  AUTO START OF CORE SPRAY PUMP III                                                                                                                                                                       NOTES:.

400KW 800 AUTO RESTART OF 1. LOADING SHOWN FOR DG-102 CONTROL ROD DRIVE PUMP 11 250KW COG-103 HAS SIMILAR LOADING) 700 AUTO RESTART OF LI QUID POISON PUMP 11 680 30KW

2. ACTUAL KW VALUES ARE ~ THOSE SHOWN.

51313- REFER TO THE DIESEL GENERATOR LOADING CALCULATION FOR THE ACTUAL DIESEL GENERATOR

          ~00 LOAD VALUES.

300 POWERBOAROS ISIB. IS7. IS71A & 8. TRANSFORMER & CABLE LOSS. 250KW

               ~------------------------------------------------------.----------

200 100 134-~-,_~rT_r-.__._r._~-,_~-_r-,,_.-._~--.__,--_r--r__r--._rT--._~--~~r__r--.__.--~rT--.__,--.-_,r__r--.__.--._~--,__,r-_r--._~)~)--------------------~------_rl--------,I-------~I 1 o 2 4 6 B 10 12 .l14 16 IB 20 22 24 -'-26 2B 30 32 34 36 38 40 42 44 46 48 50 52 54 56 58 I 2 3 4 5 6 7 8 9 10 11 12 13 14 15 l ( I 2 3

               ~1.~----7-----13---------::CONDS--------~---------------~~I~.5------------------MINUTES------------------.+I.~--HOURS~

• TIME FIGURE IX-6 UFSAR ReVISIon 21 October- 2009

u.s. U.S. NUCLEAR REGULATORY NUCLEAR

• COMMISSION COMMISSION DOCKET 50-220 DOCKET 50-220 LICENSE DPR-63 NINE MILE POINT NINE NUCLEAR STATION NUCLEAR UNIT 1 UNIT FINAL SAFETY ANALYSIS REPORT ANALYSIS (UPDATED)

(UPDATED) VOLUME VOLUME 3 OCTOBER 2009 REVISION 21 REVISION

Nine Mile Point Unit 11 UFSAR TABLE OF CONTENTS CONTENTS Section Title Page CONTENTS TABLE OF CONTENTS i LIST OF TABLES TABLES xxxiv LIST OF FIGURES xxxix SECTION I INTRODUCTION INTRODUCTION AND

SUMMARY

1-1 I-I A. A. PRINCIPAL PRINCIPAL DESIGN CRITERIA CRITERIA 1-2 1.0 General 1-7 2.0 Buildings and Structures Structures I-8 1-8 3.0 Reactor Reactor 1-8 4.0 4.0 Reactor Vessel 1-10 5.0 Containment Containment 1-10 1-10 6.0 Control and Instrumentation Instrumentation 1-12 7.0 Electrical Electrical Power Power 1-14 8.0 Radioactive Waste Disposal Radioactive Disposal 1-14 9.0 Shielding Shielding and Access Control Control 1-14

• 10.0 B.

B. 1.0 2.0 Site Handling and Storage Fuel Handling CHARACTERISTICS CHARACTERISTICS Reactor Reactor Core Storage I-14a I-14a 1-15 1-15 1-15 1-15 3.0 4.0 Fuel Assembly 1-15 1-15 5.0 Control System Control 1-15 6.0 Core Design and Operating Core Conditions Conditions 1-16 7.0 Design Power Peaking Peaking Factor Factor 1-16 8.0 Nuclear Design Data 1-16 9.0 Reactor Vessel Reactor Vessel 1-17 10.0 Coolant Recirculation Recirculation Loops 1-17 11.00

11. Primary Containment Containment 1-17 12.0 Secondary Containment Secondary Containment 1-17 I-17 13.0 13.0 Structural Design Structural 1-18 14.0 14.0 Station Electrical Station Electrical System 1-18 15.0 15.0 Reactor System Instrumentation System Reactor Instrumentation 1-18 I-18 16.0 16.0 Reactor Protection Reactor Protection System I-18a I-18a C.

C. IDENTIFICATION IDENTIFICATION OF CONTRACTORS CONTRACTORS 1-19

• D.

UFSAR Revision GENERAL CONCLUSIONS GENERAL 21 Revision 21 CONCLUSIONS i l 1-20 1-20 October 2009

Nine Nine Mile Unit 1 UFSAR Point Unit Mile Point UFSAR CONTENTS (Cont'd.) TABLE OF CONTENTS (Cont'd.) Section Section Title Title Page E. E. REFERENCES REFERENCES 1-21 1-21 SECTION II SECTION II STATION SITE AND ENVIRONMENT STATION SITE ENVIRONMENT TI-i 11-1 A. A. SITE DESCRIPTION SITE DESCRIPTION IT-i 11-1 1.0 1.0 General General II-I 11-1 2.0 Physical Features Physical Features TI-i 11-1 3.0 Property Use and Development Property Use Development 11-2 11-2 B. B. DESCRIPTION DESCRIPTION OF AREA ADJACENT TO THE SITE 11-3 11-3 1.0 1.0 General General 11-3 11-3 1.1 i.i population Population 11-3 11-3 2.0 Agriculture, Agriculture, Industrial Industrial and Recreational Use Recreational 11-3 11-3 2.1 Agricultural Agricultural Use 11-3 11-3 2.2 Industrial Use Industrial Use 11-4 11-4 2.2.1 Toxic Chemicals Toxic Chemicals 11-4 11-4 2.3 C. C. D. D. Recreational Use Recreational METEOROLOGY METEOROLOGY LIMNOLOGY LIMNOLOGY II-4c II-4c 11-5 11-5 11-6 11-6 E. E. EARTH SCIENCES SCIENCES 11-7 11-7 F. F. ENVIRONMENTAL ENVIRONMENTAL RADIOLOGY RADIOLOGY 11-8 11-8 G. G. REFERENCES 11-9 11-9 SECTION III III BUILDINGS AND STRUCTURES III-i 111-1 A. A. TURBINE BUILDING BUILDING 111-3 1.0 Design Bases Bases 111-3 I.I 1.1 Loadings Wind and Snow Loadings 111-3 111-3 1.2 Pressure Pressure Relief Design 111-3 111-3 1.3 Seismic Seismic Design and Internal Internal Loadings Loadings 111-3 111-3 1.4 1.4 Heating and Ventilation Ventilation 111-4 111-4 1.5 1.5 Shielding and Access Control III-4a 2.0 Structure Design 111-5 111-5 2.1 UFSAR T1FAR

* * *

• Revision General General Structural Features 21 ii 11 111-5 111-5 October 2009 October 2009
• Nine Mile Point Unit 11 UFSAR TABLE OF CONTENTS (Cont'd.)

CONTENTS (Cont'd.) Section Section Title Page 2.2 Heating and Ventilation Heating Ventilation System 111-6 111-6 2.3 Smoke and Heat Removal Removal 111-7 2.4 Shielding Shielding and Access Control Control 111-7 2.5 2.5 Additional Additional Building Cooling 111-7 3.0 Safety Analysis Safety Analysis 111-8 B. B. CONTROL ROOM CONTROL ROOM 111-9 1.0 Design Bases Design Bases 111-9 1 1.1 Loadings Wind and Snow Loadings 111-9 111-9 1.2 Pressure Relief Design Pressure 111-9 1.3 Seismic Design and Internal Seismic Design Internal Loadings Loadings 111-9 1.4 Heating and Ventilation Ventilation 111-9 111-9 1.5 Shielding Shielding and Access Control 111-9 2.0 Structure Structure Design III-10 111-10 2.1 General Structural General Features Structural Features III-10 111-10 2.2 Heating, Ventilation and Air Heating, Ventilation Air Conditioning System Conditioning System III-11 III-II 2.3 Smoke and Heat Removal Removal 111-12 2.4 Shielding Shielding and Access Control Control III-12a III-12a 3.0 Safety Analysis Safety III-12a III-12a C. C. WASTE DISPOSAL BUILDING WASTE 111-13 1.0 Design Bases Bases 111-13 1.1 Wind and Snow Loadings 111-13 1.2 Pressure Relief Pressure Relief Design 111-13 1.3 Seismic Design and Internal Seismic Internal Loadings Loadings 111-13 1.4 Heating Heating and Ventilation Ventilation 111-14 1.5 Shielding and Access Shielding Access Control 111-14 2.0 Structure Design Structure 111-14 2.1 General Structural General Features Structural Features 111-14 2.2 Heating Ventilation System Heating and Ventilation 111-15 2.3 Shielding and Access Shielding Access Control 111-17 3.0 Safety Analysis Safety 111-17 D. D. BUILDING OFFGAS BUILDING 111-19 1.0 Design Bases Bases 111-19 1.1 Wind and Snow Loadings 111-19 1.2 Pressure Relief Design Pressure 111-19 1.3 Seismic Design and Internal Seismic Internal Loadings Loadings 111-19 1.4 Heating and Ventilation Heating Ventilation 111-19 1.5 l.5 Shielding and Access Control Shielding 111-19 III-19 UFSAR Revision 21 UFSAR Revision iii iii October 2 009 October 2009

Nine Mile Nine Unit i1 UFSAR Point Unit Mile Point UFSAR 0 TABLE OF TABLE CONTENTS (Cont'd.) OF CONTENTS (Cont'd.) Section section Title Title Pae 2.0 2.0 Structure Design Structure Design 111-19 111-19 2.1 2.1 General Structural Features General Structural Features 111-19 111-19 2.2 2.2 Heating Ventilation System and Ventilation Heating and System 111-20 111-20 2.3 2.3 Shielding Access Control and Access Shielding and Control 111-20 111-20 3.0 3.0 Safety Analysis Safety Analysis 111-20 111-20 E. E. NONCONTROLLED BUILDINGS NONCONTROLLED BUILDINGS 111-22 111-22 1.0 1.0 Administration Building Administration Building 111-22 111-22 1.1 1.1 Design Bases Design Bases 111-22 111-22 1.1.1 1.1.1 Wind Snow Loadings and Snow Wind and Loadings 111-22 111-22 1.1.2 1.1. 2 Relief Design Pressure Relief Pressure Design 111-22 111-22 1.1.3 1.1. 3 Seismic Internal Loadings Seismic Design and Internal Design and Loadings 111-22 111-22 1.1.4 1.1.4 Heating, Cooling and Heating, Cooling and Ventilation Ventilation 111-23 111-23 1.1.5 1.1. 5 Shielding Access Control and Access Shielding and Control 111-23 111-23 1.2 1.2 Structure Design Structure Design 111-23 111-23 1.2.1 1.2.1 Structural Features General Structural General Features 111-23 111-23 1.2.2

1. 2.2 Heating, Ventilation and Heating, Ventilation and Air Air Conditioning Conditioning 111-24 111-24 1.2.3

1. 2.3 1.3 1.3 2.0 2.0 2.1 2.1 2.1.1 2.1.1 2.1.2 Access Control Access Design Wind Control Safety Analysis Safety Analysis Treatment Building Sewage Treatment Sewage Bases Design Bases Building Snow Loadings and Snow Wind and Loadings Relief Design Pressure Relief Pressure Design 111-24 111-24 111-24 111-24 111-25 111-25 111-25 111-25 111-25 111-25 111-25 111-25 2.1. 2 2.1.3 2.1. 3 Seismic Internal Loadings Seismic Design and Internal Design and Loadings 111-25 111-25 2.1.4 2.1. 4 Electrical Design Electrical Design 111-25 111-25 2.1.5 2.1. 5 Fire Explosive Gas and Explosive Fire and Gas Detection Detection 111-25 111-25 2.1.6 2.1. 6 Heating and Ventilation Heating and Ventilation 111-26 111-26 2.1.7 2.1. 7 Shielding Access Control and Access Shielding and Control 111-26 111-26 2.2 2.2 Structure Design Structure Design 111-26 111-26 2.2.1 2.2.1 Structural Features General Structural General Features 111-26 111-26 2.2.2 2.2.2 Ventilation System Ventilation System 111-28 111-28 2.2.3 2.2.3 Access Control Access Control 111-28 111-28 3.0 3.0 Information Center Energy Information Energy Center 111-28 111-28 3.1 3.1 Design Bases Design Bases 111-28 111-28 3.1.1 3.1.1 Wind Snow Loadings and Snow Wind and Loadings 111-28 111-28 3.1.2 3.1. 2 Relief Design Pressure Relief Pressure Design 111-28 111-28 3.1.3 3.1. 3 Seismic Design and Internal Seismic Design and Internal Loadings Loadings 111-28 111-28 3.1.4 3.1.4 and Ventilation Heating and Heating Ventilation 111-29 111-29 3.1.5 Shielding Access Control.

Shielding and Access and Control, 111-29 111-29 3.1. 5 3.2 3.2 Structure Design Structure Design 111-29 111-29 Revision 21 UFSAR Revision UFSAR 21 iv iv October 2009 Octob~r 2009

Nine Mile Point Unit 1 UFSAR TABLE OF CONTENTS CONTENTS (Cont'd.) (Cont'd.) Section Section Title Page 3.2.1 General Structural General Features Structural Features 111-29 111-29 3.2.2 Heating and Ventilation Heating Ventilation System 111-30 111-30 3.2.3 Access Control 111-30 111-30 F. F. SCREENHOUSE, SCREENHOUSE, INTAKE AND DISCHARGE DISCHARGE TUNNELS 111-31 111-31 1.0 Screenhouse Screenhouse 111-31 111-31 I.i 1.1 Design Design Basis 111-31 111-31 1.1.1 11.1 Wind and Snow Loadings Loadings 111-31 111-31 1.1. 1.1.22 Pressure Pressure Relief Design 111-31 111-31 1.1.33 1.1. Seismic Design and Internal Seismic Internal Loadings Loadings 111-31 111-31 1.1.4 Heating and Ventilation Ventilation 111-31 111-31 1.1. 1.1.55 Shielding Shielding and Access Control 111-31 111-31 1.2 Structure Structure Design 111-31 111-31 2.0 Intake and Discharge Intake Discharge Tunnels Tunnels 111-33 2.1 Design Bases 111-33 111-33 2.2 Structure Design Structure 111-33 111-33 3.0 Safety Analysis Safety Analysis 111-34 111-34 G. G. STACK STACK 111-35 111-35 1.0 Design Bases 111-35 111-35 1.1 General General 111-35 111-35 1.2 Wind Loading 111-35 111-35 1.3 Seismic Design 111-35 111-35 1.4 1.4 Shielding and Access Control 111-35 111-35 2.0 Structure Structure Design 111-35 111-35 3.0 Safety Analysis Analysis 111-36 111-36 3.1 Radiology Radiology 111-36 111-36 3.2 Stack Failure Failure Analysis Analysis 111-37 111-37 3.2.1 Reactor Building Reactor Building 111-37 111-37 3.2.2 Diesel Generator Diesel Generator Building Building 111-38 111-38 3.2.3 Screen Pump House Screen and Pump 111-38 H. SECURITY WEST AND SECURITY BUILDING WEST SECURITY BUILDING BUILDING ANNEX 111-39 111-39 I. I. RADWASTE SOLIDIFICATION RADWASTE SOLIDIFICATION AND STORAGE BUILDING 111-40 111-40 1.0 Design Bases Design 111-40 111-40 1.1 Wind and Snow Loadings Loadings 111-40 111-40 UFSAR Revision 21 21 v V October 2009

Nine Mile Point Unit 1 UFSAR (Cont'd.) TABLE OF CONTENTS (Cont'd.) TABLE Section Section Title Page 1.2 Pressure Relief Design Design 111-40 1.3 Seismic Design and Internal Seismic Internal Loadings Loadings 111-40 111-40 1.4 Heating, Ventilation Heating, Ventilation and AirAir Conditioning Conditioning 111-40 1.5 Shielding and Access Control 111-40 2.0 Structure Structure and Design 111-41 2.1 General Structural General Structural Features 111-41 2.2 Heating, Ventilation Heating, Ventilation and AirAir Conditioning Conditioning 111-41 111-41 2.3 Shielding Shielding and Access Control 111-43 3.0 Use 111-43 J. J. REFERENCES REFERENCES 111-45 SECTION IV REACTOR IV-1 1V-l A. A. DESIGN BASES IV-1 1V-l 1.0 2.0 3.0 B. B. 1.0 General General Performance Objectives Performance REACTOR DESIGN REACTOR General General Objectives Design Limits and Targets Targets IV-i 1V-l IV-i 1V-l IV-2 1V-2 IV-3 1V-3 IV-3 1V-3 2.0 Nuclear Nuclear Design Technique IV-4 1V-4 2.1 Reference Loading Pattern Reference Pattern IV-5 IV-5 2.2 Final Loading Pattern IV-6 1V-6 2.2.1 Acceptable Reference Acceptable Deviation From Reference Loading Pattern Loading Pattern IV-6 1V-6 2.2.2 Reexamination of Licensing Reexamination Licensing Basis IV-6 IV-6 2.3 Refueling Cycle Refueling Cycle Reactivity Balance Reactivity Balance IV-7 1V-7 3.0 Thermal and Hydraulic Hydraulic Characteristics Characteristics IV-7 IV-7 3.1 Thermal and Hydraulic Hydraulic Design IV-7 1V-7 3.1.1 Recirculation Recirculation Flow Control Control IV-7 IV-7 3.1. 3.1.22 Core Thermal Thermal Limits IV-7 1V-7 3.1.2.1 3.1.2.1 Excessive Temperature Excessive Clad Temperature IV-8 1V-8 3.1.2.2 3.1.2.2 Cladding Strain IV-9 IV-9 3.1.2.3 3.1.2.3 Coolant Flow IV-9 1V-9 3.2 Thermal and Hydraulic Hydraulic Analyses IV-9 IV-9 3.2.1 Hydraulic Analysis Analysis IV-9 IV-9 3.2.2 Thermal Analysis Revision 21 UFSAR Revision UFSAR 21 Analysis vi vi IV-i1 IV-II October 2009

Nine Nine Mile Point Unit 11 UFSAR TABLE OF CONTENTS (Cont'd.) CONTENTS (Cont'd.) Section Section Title Page 3.2.2.1 3.2.2.1 Fuel Cladding Cladding Integrity Safety Limit Analysis IV-11 IV-11 3.2.2.2 MCPR Operating Analysis Operating Limit Analysis IV-12 3.3 Reactor Reactor Transients IV-13 4.0 Stability Analysis Stability Analysis IV-14 4.1 Design Bases Bases IV-14 4.2 Stability Analysis Stability Analysis Method IV-14 5.0 Mechanical Mechanical Design and Evaluation Evaluation IV-1S IV-15 5.1 Fuel Mechanical Mechanical Design IV-15 IV-1S 5.1.1 Design Bases Bases IV-1S IV-15 5.1. 5.1.22 Fuel Rods IV-15 5.1. 5.1.33 Water Rods Water IV-16 5.1. 5.1.4 4 Fuel Assemblies IV-16 S.1. 5.1.55 Mechanical Design Limits Mechanical Design Stress Limits and Stress Analysis IV-16 5.1. 5.1.66 Relationship Between Relationship Between Fuel Design Limits and Fuel Damage Damage Limits IV-16 5.1. Surveillance Testing Surveillance and Testing IV-16 5.1.77 6.0 Control Rod Mechanical Mechanical Design and Evaluation Evaluation IV-16 6.1 Design IV-16 6.1.1 Control Rods and Drives IV-16 6.1. 2 6.1.2 Standby Standby Liquid Poison System IV-19 6.2 Control System Evaluation Evaluation IV-19 6.2.1 Rod Withdrawal Errors Evaluation Withdrawal Errors Evaluation IV-19 6.2.2 Overall Overall Control System Evaluation Evaluation IV-21 6.3 Limiting Conditions Conditions for Operation and Surveillance Surveillance IV-23 6.4 Control Rod Lifetime Control Lifetime IV-23 7.0 Reactor Vessel Internal Reactor Vessel Internal Structure Structure IV-24 7.1 Design Bases IV-24 7.1.1 7.1.1 Core Shroud IV-25 IV-2S 7.1. 7.1.22 Core Support IV-27 7.1. 7.1.33 Top Grid IV-27 7.1.4 Control Rod Guide Guide Tubes IV-27 7.1. 7.1.5S Feedwater Sparger Feedwater Sparger IV-27 7.1. 7.1.66 Core Spray Spargers Spargers IV-27 7.1. 7.1.77 Liquid Poison Sparger Liquid IV-28 7.1. 7.1.88 Stearn Steam Separator Dryer Separator and Dryer IV-28 7.1. 7.1.99 Core Shroud Stabilizers Stabilizers IV-28 7.1.10 Core Shroud Vertical Vertical Weld Repair IV-30 IV-3~ 7.2 Design Evaluation Design Evaluation IV-30 IV-3~ Revision 21 UFSAR Revision 21 vii vii October 2009 October

Nine Mile Point Unit 11 UFSAR TABLE (Cont'd.) TABLE OF CONTENTS (Cont'd.) Section Section Title Title Page 7.3 Surveillance Surveillance and Testing IV-31 IV-3I C. C. REFERENCES REFERENCES IV-32 SECTION V V REACTOR COOLANT SYSTEM REACTOR SYSTEM V-I V-I A. A. DESIGN BASES DESIGN V-I V-I 1.0 General General V-I V-1 2.0 Performance Performance Objectives Objectives V-i V-I 3.0 Design Pressure V-2 V-2 4.0 Cyclic Loads (Mechanical (Mechanical and Thermal) Thermal) V-3 V-3 5.0 Codes V-3a V-3a B. B. SYSTEM DESIGN AND OPERATION OPERATION V-4 V-4 1.0 General General V-4 V-4 1.1 Drawings Drawings V-4 V-4 1.2 Materials of Construction Materials Construction V-4 V-4 1.3 1.4 1.5 2.0 3.0 4.0 Thermal Stresses Thermal Stresses Primary Coolant Leakage Coolant Chemistry Chemistry Reactor Vessel Reactor Reactor Reactor Vessel Leakage Reactor Recirculation Recirculation Loops Auxiliary Systems Reactor Steam and Auxiliary Systems V-4 V-4 V-5 V-S V-6 V-6 V-6 V-6 V-7 V-7 piping Piping V-8 V-8 5.0 Relief Devices Relief Devices V-8 V-8 c. C. SYSTEM DESIGN EVALUATION EVALUATION V-10 V-IO 1.0 General General V-10 V-IO 2.0 Pressure V-10 V-IO 3.0 Design Heatup Design Cooldown Rates Heatup and Cooldown Rates V-il V-II 4.0 Materials Radiation Materials Radiation Exposure V-12 4.1 Pressure-Temperature Pressure-Temperature LimitLimit Curves V-12 V-12 4.2 Temperature Limits for Boltup Temperature V-12 4.3 Temperature Limits for In-Service Temperature In-Service Pressure Tests System Pressure V-13 V-13 4.4 Operating Limits Operating Heatup, Limits During Heatup, Operation Cooldown, and Core Operation Cooldown, V-13 V-13 4.5 Predicted Predicted Shift inin RT~T RTNDT V-13 V-13 4.6 Fluence Calculations Neutron Fluence Calculations V-13 s.o 5.0 Mechanical Mechanical Considerations Considerations V-14 5.1 Revision 21 UFSAR Revision Reaction Forces Jet Reaction viii viii V-14 2009 October 2009

Nine Mile Point Unit 1 UFSAR TABLE OF CONTENTS (Cont'd.) (Cont'd.) section Section Title Page 5.2 Seismic Forces Seismic Forces V-14 6.0 Safety Safety Limits, Limits, Limiting Safety Settings Minimum Conditions for Settings and Minimum for Operation Operation V-14 D. D. TESTS AND INSPECTIONS INSPECTIONS V-16 V-16 1.0 Prestartup Testing Testing V-16 V-16 2.0 Inspection Inspection and Testing Following Startup V-16 V-16 2.1 Pressure Pressure Test V-16 V-16 2.2 Pressure Vessel Irradiation Irradiation V-16 V-16 E. E. EMERGENCY EMERGENCY COOLING SYSTEM SYSTEM V-17 1.0 Bases Design Bases V-17 2.0 System Design and Operation Operation V-17 3.0 Evaluation Design Evaluation V-19 3.1 Redundancy Redundancy V-19 Makeup Water 3.2 Makeup V-19 3.3 System Leaks V-19 V-19 3.4 Containment Containment Isolation Isolation V-19 4.0 Tests and Inspections Inspections V-20 V-20 4.1 Prestartup Prestartup Test V-20 V-20 4.2 Subsequent Subsequent Inspections Inspections and Tests V-20 V-20 F. F. REFERENCES REFERENCES V-21 V-21 SECTION VI SECTION CONTAINMENT SYSTEM CONTAINMENT VI-l VI-l A. A. PRIMARY CONTAINMENT PRIMARY CONTAINMENT - MARKMARK I CONTAINMENT PROGRAM CONTAINMENT VI-2 VI-2 1.0 General Structure General Structure VI-2 VI-2 2.0 Pressure Suppression Pressure Hydrodynamic Suppression Hydrodynamic Loads VI-2 VI-2 2.1 Safety/Relief Valve Safety/Relief Valve Discharge Discharge VI-2 VI-2 2.2 Loss-of-Coolant Accident Loss-of-Coolant Accident VI-3 VI-3 2.3 Summary of Loading Summary Loading Phenomena Phenomena VI-4 VI-4 3.0 Plant-Unique Plant-Unique Modifications Modifications VI-5 VI-5 B. B. PRIMARY PRIMARY CONTAINMENT CONTAINMENT - PRESSURE SUPPRESSION SYSTEM SUPPRESSION VI-6 VI-6 1.0 Design Bases Bases VI-6 VI-6 1.1 General General VI-6 VI-6 UFSAR Revision 21 21 ix ix October 2009

Mile Point Unit 1 UFSAR Nine Mile (Cont'd.) CONTENTS (Cont'd.) TABLE OF CONTENTS Section Section Title 1.2 1.2 Design Basis Basis Accident (DBA) (DBA) VI-6 1.3 Containment Heat Removal Containment VI-8 1.4 Isolation Isolation Criteria Criteria VI-8 1.5 Vacuum Relief Relief Criteria Criteria VI-9 1.6 Flooding Criteria Criteria VI-9 1.7 Shielding Shielding VI-9 2.0 Structure Design Structure Design VI-9 2.1 General General VI-9 2.2 Penetrations Access Openings Penetrations and Access Openings VI-Il 2.3 Jet and Missile Protection Protection VI-12 2.4 Materials Materials VI-13 2.5 Shielding Shielding VI-14 2.6 Vacuum Relief Relief VI-14 2.7 Containment Flooding Containment Flooding VI-14 C. C. SECONDARY SECONDARY CONTAINMENT CONTAINMENT -- REACTOR REACTOR BUILDING BUILDING VI-16 1.0 Design Bases VI-16 1.1 1.2 1.3 1.4 2.0 2.1 Wind and Snow Pressure Relief Shielding Structure Design Structure Design Loadings Snow Loadings Relief Design Seismic Design General Structural General Structural Features Features VI-16 VI-16 VI-16 VI-17 VI-17 VI-17 VI-17 D. CONTAINMENT ISOLATION SYSTEM CONTAINMENT ISOLATION VI-20 1.0 Design Bases VI-20 1.1 Containment Spray Appendix J Water Containment Water Seal Requirements Requirements VI-23a VI-23a 2.0 System Design VI-24 3.0 Tests and Inspections Tests Inspections VI-26 E. E. CONTAINMENT CONTAINMENT VENTILATION VENTILATION SYSTEM VI-27 1.0 Primary Containment Containment VI-27 VI-27 1.1 1.1 Design Bases VI-27 VI-27 1.2 System Design VI-27 2.0 Secondary Containment Secondary Containment VI-28 2.1 Bases Design Bases VI-28 2.2 System Design VI-28 F. F. TEST AND INSPECTIONS INSPECTIONS VI-30 1.0 1.0 21 Revision 21 UFSAR Revision Suppression Chamber Drywell and Suppression x X Chamber VI-30 October 2009

Nine Mile Nine Unit 11 UFSAR Point Unit Mile Point UFSAR TABLE OF TABLE CONTENTS (Cont'd.) OF CONTENTS (Cont'd.) Section Section Title Title Page 1.1 1.1 Preoperational Testing Preoperational Testing VI-30 VI-3D 1.2 1.2 Postoperational Testing Postoperational Testing VI-30 VI-30 2.0 2.0 Containment Penetrations and Containment Penetrations and Isolation Valves Isolation Valves VI-31 VI-31 2.1 2.1 Penetration and Penetration Valve Leakage and Valve Leakage VI-31 VI-31 2.2 2.2 Valve Operability Test Valve Operability Test VI-31 VI-31 3.0 3.0 Containment Ventilation System Containment Ventilation System VI-32 VI-32 4.0 4.0 Other Containment Other Tests Containment Tests VI-32 VI-32 5..0 5 .. 0 Reactor Building Reactor Building VI-32 VI-32 5.1 5.1 Reactor Normal Ventilation Reactor Building Normal Building Ventilation System System VI-32 VI-32 5.2 5.2 Reactor Building Reactor Isolation Valves Building Isolation Valves VI-32 VI-32 5.3 5.3 Emergency Ventilation System Emergency Ventilation System VI-33 VI-33 G. G. REFERENCES REFERENCES VI-33 VI-33 SECTION VII SECTION VII ENGINEERED SAFEGUARDS ENGINEERED SAFEGUARDS VII-2 VII-l

• A.

A. 1.0 2.0 2.0 2.1 2.1 2.2 2.2 Design SPRAY SYSTEM CORE SPRAY CORE General General Bases Design Bases

System Design

System SYSTEM Design Operator Assessment Operator Assessment VII-2 VII-2 VII-2 VII-2 VII-2 VII-2 VII-2 VII-2 VII-5 VII-5 3.0 3.0 Design Evaluation Design Evaluation VII-7 VII-6 4.0 4.0 Tests and Inspections Tests and Inspections VII-7 VII-7 B. B. CONTAINMENT SPRAY SYSTEM CONTAINMENT SPRAY SYSTEM VII-8 VII-8 1.0 1.0 Licensing Basis Requirements Licensing Basis Requirements VII-8 VII-8 1.1 1.1 10CFR50.49 -- Environmental IOCFR50.49 Environmental Qualification of Qualification Electric Equipment of Electric Equipment Important Important to Safety for Nuclear Safety for Nuclear Power Plants Power Plants VII-8 VII-8 1.2 1.2 10CFR5O Appendix A 10CFR50 Appendix General Design A -- General Design Criteria for Nuclear Criteria Power Plants Nuclear Power Plants VII-8 VII-8 2.0 2.0 Design Bases Design Bases VII-ll VII-11 2.1 2.1 Design Functional Requirements Basis Functional Design Basis Requirements VII-ll VII-11 2.2 2.2 Controlling Parameters Controlling Parameters VII-12 VII-12 3.0 3.0 System Design System Design VII-12a VII-12a 3.1 3.1 System Function System Function VII-12a VII-12a 3.2 3.2 System Design Description System Design Description VII-12b VII-12b VII-12c 3.3 3.3 System Design System Design VII-12c 3.4 3.4 Codes and Standards and Standards VII-14b VII-14b Revision 21 UFSAR Revision 21 xa xa 2009 October 2009 October

Nine Nine Mile Mile Point Unit 11 UFSAR UFSAR (Cont'd.) TABLE OF CONTENTS (Cont'd.) TABLE Section Title Page 3.5 3.5 System Instrumentation System Instrumentation VII-14b VII-14b 3.6 3.6 System Features System Design Features VII-14c VII-14c 4.0 4.0 Design Performance Evaluation Performance VII-14d VII-14d 4.1 4.1 System Performance Analyses System Performance Analyses VII-14d VII-14d 4.2 4.2 System Response System Response VII-14e VII-14e 4.3 4.3 Interdependency Other Interdependency With Other Engineered Engineered Safeguards Systems Safeguards Systems VII-14e VII-14e 5.0 5.0 System Operation System Operation VII-14f VII-14f 5.1 5.1 Limiting Conditions for Operation VII-14f VII-14f 6.0 6.0 Tests and Inspection Tests VII-14g VII-14g C. C. LIQUID POISON INJECTION LIQUID SYSTEM INJECTION SYSTEM VII-15 VII-IS 1.0 1.0 Bases Design Bases VII-15 VII-IS 2.0 2.0 System Design VII-16 VII-16 2.1 2.1 Operator Assessment Operator Assessment VII-19 VII-19 3.0 3.0 Design Evaluation Evaluation VII-20 VII-20 4.0 4.0 Tests and Inspections Inspections VII-21 VII-21 5.0 5.0 Alternate Boron Injection Alternate Boron VII-21a VII-21a D. D. 1.0 1.0 2.0 2.0 3.0 3.0 CONTROL ROD VELOCITY CONTROL Design Bases Design Bases

System Design

System VELOCITY LIMITER Design Evaluation General VII-22 VII-22 VII-22 VII-22 VII-22 VII-22 VII-24 VII-24 VII-24 0 3.1 3.1 General VII-24 3.2 3.2 Design Sensitivity VII-24 VII-24 3.3 3.3 Normal Operation VII-25 VII-25 4.0 4.0 Tests and Inspections Inspections VII-25 VII-25 E. E. HOUSING SUPPORT CONTROL ROD HOUSING SUPPORT VII-26 VII-26 1.0 1.0 Design Bases Bases VII-26 VII-26 2.0 2.0 System Design Design VII-26 VII-26 2.1 2.1 Deflections Loads and Deflections VII-28 VII-28 3.0 3.0 Design Evaluation VII-28 VII-28 4.0 4.0 Tests and Inspections Tests Inspections VII-29 VII-29 F. F. FLOW RESTRICTORS RESTRICTORS VII-30 VII-30 1.0 1.0 Bases Design Bases VII-30 VII-30 2.0 2.0 System Design Design VII-30 VII-30 3.0 3.0 Design Evaluation VII-30 VII-30 4.0 4.0 Tests and Inspections Inspections VII-31 VII-31 Revision 21 UFSAR Revision 21 xb xb October 2009 2009

Nine Mile Point Unit 11 UFSAR TABLE OF CONTENTS (Cont'd.) CONTENTS (Cont'd.) Section Section Title Title G. COMBUSTIBLE GAS CONTROL COMBUSTIBLE SYSTEM CONTROL SYSTEM VII-32 1.0 Design Bases Design Bases VII-32 2.0 Containment Inerting Containment Inerting System VII-32 2.1 System System Design VII-32 2.2 Design Evaluation Evaluation VII-33 3.0 Containment Atmospheric Containment Atmospheric Dilution System System VII-33 3.1 3.1 System Design System VII-33 3.2 3.2 Evaluation Design Evaluation VII-35 4.0 Tests and Inspections Inspections VII-35 H. H. EMERGENCY VENTILATION EMERGENCY VENTILATION SYSTEM VII-36 1.0 Design Bases VII-36 2.0 System Design System VII-36 2.1 Operator Assessment Operator VII-38 3.0 Design Evaluation Evaluation VII-39 4.0 Tests and Inspections Inspections VII-39

• I.

I. 1.0 2.0 3.0 4.0 Design Bases

System Design

System Design Evaluation COOLANT INJECTION HIGH-PRESSURE COOLANT HIGH-PRESSURE Evaluation Tests and Inspections Inspections INJECTION VII-41 VII-41 VII-41 VII-42 VII-43

• UFSAR Revision 21 UFSAR 21 xc xc October 2009 October

Mile Point Unit 1 UFSAR Nine Mile 0 PAGE INTENTIONALLY THIS PAGE BLANK INTENTIONALLY BLANK THIS 0 21 UFSAR Revision 21 xd xd October 2009

Nine Mile Point Unit 1 UFSAR

• Section Section Title TABLE OF CONTENTS (Cont'd.)

CONTENTS (Cont'd.) Page J. J. FUEL AND REACTOR COMPONENTS FUEL COMPONENTS HANDLING SYSTEM X-39 X-39 1.0 Design Bases Bases X-39 X-39 2.0 System Design System X-40 X-40 2.1 Description Description of Facility Facility X-40 X-40 2.1.1 2.1.1 Cask Drop Protection Cask Protection System X-42 2.2 Operation of the Facility Operation X-43 2.3 Control of Heavy Loads Program Control Program X-43a X-43a 2.3.1 Introduction/Licensing Introduction/Licensing Background Background X-43a X-43a 2.3.2 Safety Basis Safety X-43b X-43b 2.3.3 Scope of Heavy Scope Heavy Load Handling Handling Systems Systems X-43b X-43b 2.3.4 Control of Heavy Loads Program Program X-43c X-43c 2.3.4.1 NMPNS Commitments Commitments in Response to in Response to NUREG-0612, Phase I Elements NUREG-0612, X-43c X-43c 2.3.4.2 2.3.4.2 Reactor Pressure Vessel Head and Reactor Spent Fuel Spent Fuel Cask Lifts X-43d X-43d 2.3.5 2.3.5 Safety Evaluation Evaluation X-43d X-43d

• 3.0 4.0 K.

1.0 1.1 Evaluation Design Evaluation Tests and Inspections Inspections FIRE PROTECTION PROGRAM PROTECTION PROGRAM Program Bases Bases Nuclear Division Nuclear Division Directive Directive - Fire X-43d X-43d X-44 X-44 X-45 X-45 X-45 X-45 Protection Program Protection Program X-45 X-45 1.2 1.2 Nuclear Division Interface Nuclear Interface Procedure Procedure - Fire Protection Program Protection Program X-45 X-45 1.3 Fire Hazards Analysis Analysis X-45 X-45 1.4 Appendix RR Review Review Safe Safe Shutdown Analysis X-46 X-46 1.5 Fire Protection Protection and Appendix R Related Portions of Operations Related Operations Procedures (OPs, Procedures SOPs, and EOPs) (OPs, SOPs, EOPs) Repair Procedures and Damage Repair Procedures X-46 X-46 1.6 Fire Protection Protection Portions of the Emergency Emergency Plan X-46 X-46 2.0 Program Implementation Implementation and Design Aspects Aspects X-46 X-46 2.1 Fire Protection Implementing Protection Implementing Procedures Procedures X-46 X-46 2.2 Fire Protection Administrative Protection Administrative

• UFSAR Revision Controls 21 Revision 21 xvi xviii X-47 X-47 2009 October 2009

Nine Mile Mile Point Unit 11 UFSAR Section 2.3 Title TABLE OF CONTENTS (Cont'd.) Fire Protection (Cont'd.) System Drawings Protection System Drawings and Page Calculations Calculations X-47 X-47 2.4 Fire Protection Protection Engineering Evaluations (FPEEs) Evaluations (FPEEs) X-47 X-47 3.0 3.0 Monitoring Monitoring and Evaluating Program Evaluating Program Implementation Implementation X-47 X-47 3.1 Quality Assurance Program Quality Assurance X-47 X-47 3.2 Brigade Manning, Fire Brigade Training, Manning, Training, Responsibilities Drills and Responsibilities X-47 X-47 4.0 Surveillance Surveillance and Tests X-48 L. L. REMOTE SHUTDOWN SHUTDOWN SYSTEM X-49 X-49 1.0 1.0 Design Bases Bases X-49 2.0 System Design X-49 X-49 3.0 System Evaluation Evaluation X-49 X-49 4.0 Tests and Inspections Inspections X-50 X-SO M. HYDROGEN WATER CHEMISTRY NOBLE CHEMISTRY AND NOBLE CHEMICAL ADDITION METAL CHEMICAL METAL (NOBLECHEM) ADDITION (NOBLECHEM) SYSTEMS SYSTEMS X-51 X-S1 1.0 Design Basis X-51 X-S1 1.1 Noble Metal Metal Chemical Addition System Chemical Addition X-52 X-S2 1.2 Hydrogen Water Chemistry Chemistry System X-52a X-S2a 2.0 System System Design X-53 X-S3 2.1 Noble Metal Noble Metal Chemical Addition X-53 X-S3 2.2 Hydrogen Water Chemistry Chemistry System X-53 X-S3 2.2.1 HWC Feedwater Hydrogen Injection Feedwater Hydrogen Injection X-54 X-S4 2.2.2 HWC Offgas Oxygen Injection Injection X-54 X-S4 2.2.3 HWC Offgas Offgas Sample Sample X-54 X-S4 3.0 System Evaluation Evaluation X-54 X-S4 4.0 Tests and Inspections Inspections X-55 N. N. REFERENCES REFERENCES X-56 X-S6 APPENDIX lOA 10A HAZARDS ANALYSIS FIRE HAZARDS ANALYSIS APPENDIX lOB APPENDIX 10B SHUTDOWN ANALYSIS SAFE SHUTDOWN ANALYSIS SECTION XI SECTION STEAM-TO-POWER CONVERSION SYSTEM STEAM-TO-POWER CONVERSION SYSTEM XI-l XI-1 A. A. DESIGN BASES DESIGN XI-l XI-1 __ UFSAR Revision 21 Revision 21 xviia xviia October 2009

Nine Mile Point Unit 1 UFSAR

• Section Section TABLE OF CONTENTS Title (Cont'd.)

CONTENTS (Cont'd.) Page B. B. SYSTEM DESIGN AND OPERATION SYSTEM OPERATION XI-2 XI-2 1.0 1.0 Turbine Generator Turbine Generator XI-2 XI-2 2.0 Turbine Condenser XI-4 XI-4 3.0 Condenser Air Removal and Offgas Condenser Offgas System XI-5 XI-5 4.0 Circulating Water Water System XI-9a XI-9a 5.0 Condensate Pumps XI-9a XI-9a 6.0 Condensate Filtration System Condensate Filtration System XI-9b 6.0A 6.OA Condensate Demineralizer System Condensate Demineralizer System XI-9c XI-9c 7.0 Condensate Condensate Transfer System System XI-11 8.0 Feedwater Feedwater Booster Pumps XI-11 9.0 Feedwater Feedwater Pumps XI-lla XI-11a 10.0 Feedwater Feedwater Heaters XI-12 C. C. SYSTEM ANALYSIS ANALYSIS XI-13 XI-13 D. D. TESTS AND INSPECTIONS INSPECTIONS XI-16 XI-16

• SECTION XII SECTION A.

A. 1.0 1.0 1.1* 1.1i RADIOLOGICAL CONTROLS Design Bases Objectives Objectives CONTROLS RADIOACTIVE WASTES WASTES XII-l XII-1 XII-l XII-1 XII-l XII-1 XII-1 XII-1 1.2 Types of Radioactive Radioactive Wastes Wastes XII-l XII-1

• UFSAR UFSAR Revision 21 Revision 21 xviib October 2009 2009

Nine Mile Point Unit 1 UFSAR TABLE TABLE OF CONTENTS (Cont'd.) (Cont'd.) Section Section Title Page 1.2.1 1.2.1 Gaseous Waste Waste XlI-l XII-1 1.2.2 Liquid Wastes Wastes XlI-l XII-I 2.3 1.2.3

1. Solid Wastes Wastes XII-2 XII-2 2.0 system Design and Evaluation System Evaluation XII-2 XII-2 2.1 Gaseous Waste System Gaseous Waste XII-2 XII-2 2.1.1 Offgas System XII-3 XII-3 2.1.

2.1.2 2 Steam-Packing Steam-Packing Exhausting Exhausting System XII-3 XII-3 2.1. 2.1.33 Building Building Ventilation Ventilation Systems Systems XII-3 2.1.4 Stack Stack XII-3 XII-3 2.2 Liquid Waste System Liquid XII-4 XII-4 2.2.1 Liquid Waste Handling Processes Liquid Processes XII-4 XII-4 2.2.2 Sampling and Monitoring Liquid Wastes XII-6 XII-6 2.2.3 Liquid Waste Equipment Liquid Equipment Arrangement Arrangement XII-7 XII-7 2.2.4 Liquid Radioactive Liquid Radioactive Waste System System control Control XII-7 XII-7 2.3 Solid Waste System XII-8 XII-8 2.3.1 Solid Waste Handling Processes Processes XII-8 XII-8 2.3.2 Solid Waste System Equipment XII-9 XII-9 2.3.3 Process Process Control Program Program XII-9a XII-9a 3.0 Safety Safety Limits XII-9a XII-9a 4.0 Tests and Inspections Inspections XII-I0 XII-10 4.1 Waste Process Process Systems Systems XII-I0 XII-10 4.2 Filters Filters XII-I0 XII-10 4.3 Effluent Monitors Effluent XII-10 XII-I0 4.3.1 Offgas and Stack Stack Monitors Monitors XII-10 XII-I0 4.3.2 Liquid Waste Waste Effluent Effluent Monitor Monitor XII-I0 XII-10 B. B. RADIATION PROTECTION RADIATION PROTECTION XII-II XII-11 1.0 Primary Primary and Secondary Secondary Shielding Shielding XII-II XII-II 1.1 1.1 Design Bases Bases XII-!1 XII-II 1.2 Design XII-12

1. 2.1 1.2.1 Reactor Shield Wall Reactor XII-12

1. 2.2 1.2.2 Biological Shield Biological XII-12

1. 2.3 1.2.3 Miscellaneous Miscellaneous XII-12 1.3 Evaluation Evaluation XII-13 2.0 Area Radioactivity Radioactivity Monitoring Systems XII-13 XII-13 2.1 Area Radiation Radiation Monitoring Monitoring System XII-13 2.1.1 Bases Design Bases XII-13 2.1. 2 2.1.2 Design Design XII-14 2.1.

2.1.33 Evaluation' XII-15 XII-IS UFSAR Revision Revision 20 xviii October 2007 October 2007

Nine Mile Nine Unit 11 UFSAR Point Unit Mile Point UFSAR TABLE OF TABLE CONTENTS (Cont'd.) OF CONTENTS (Cont'd.) Section Section Title Title Page 2.2 2.2 Area Contamination Monitoring Air Contamination Area Air Monitoring System System XII-16 XII-16 2.2.1 2.2.1 Design Bases Design Bases XII-16 XII-16 2.2.2 2.2.2 Design Design XII-16 XII-16 2.2.3 2.2.3 Evaluation Evaluation XII-l-a XII-16a 3.0 3.0 Radiation Protection Radiation Protection XII-17 XII-17 3.1 3.1 Facilities Facilities XII-17 XII-17 3.1.1 3.1.1 Laboratory, Room and Counting Room Laboratory, Counting and Calibration Facilities Calibration Facilities XII-17 XII-17 3.1.2 3.1. 2 Change Room and Change Room Shower Facilities and Shower Facilities XII-18 XII-18 3.1.3 3.1.3 Decontamination Facility Personnel Decontamination Personnel Facility XII-19 XII-19 3.1.4 3.1.4 Tool and Equipment Decontamination Tool and Equipment Decontamination Facility Facility XII-19 XII-19 3.2 3.2 Radiation Control Radiation Control XII-20 XII-20 3.2.1 3.2.1 Shielding Shielding XII-20 XII-20 3.2.2 3.2.2 Access Control Access Control XII-20 XII-20 3.3 3.3 Contamination Control Contamination Control XII-21a XII-21a 3.3.1 3.3.1 Contamination Control Facility Contamination Facility Control XII-21a XII-21a

• 3.3.2 3.3.2 3.3.3 3.3.3 3.4 3.4 3.4.1 3.4.1 3.5 3.5 Personnel Contamination Control Personnel Airborne Personnel Radiation Contamination Dose Radiation Dose Control Contamination Control Airborne Contamination Control Dose Determinations Personnel Dose Determinations Radiation Protection Radiation Protection Instrumentation Instrumentation XII-22 XII-22 XII-23 XII-23 XII-24 XII-24 XII-24 XII-24 XII-24 XII-24 3.5.1 3.5.1 Room Instrumentation Counting Room Counting Instrumentation XII-24 XII-24 3.5.2 3.5.2 Radiation Instrumentation Portable Radiation Portable Instrumentation XII-25 XII-25 3.5.3 3.5.3 Air Sampling Instrumentation Air Sampling Instrumentation XII-25 XII-25 3.5.4 3.5.4 Monitoring Instruments Personnel Monitoring Personnel Instruments XII-25 XII-25 3.5.5 3.5.5 Emergency Instrumentation Emergency Instrumentation XII-25 XII-25 4.0 Tests and Inspections Tests and Inspections XII-26 XII-26 4.1 Shielding XII-26 XII-26 4.2 Area Radiation Monitors Area Radiation Monitors XII-27 XII-27 4.3 Area Contamination Monitors Area Air Contamination Air Monitors XII-27 XII-27 4.4 Radiation Protection Facilities Radiation Protection Facilities XII-27 XII-27 4.4.1 4.4.1 Ventilation Air Ventilation Air Flows Flows XII-27 XII-27 4.4.2 4.4.2 Calibration Well Instrument Calibration Instrument Well Shielding XII-28 XII-28 4.5 4.5 Radiation Protection Radiation Protection Instrumentation Instrumentation XII-28 XII-28
• UFSAR Revision 21 21 xix xix October October 2009 2009

Nine Mile Mile Point Unit 1 UFSAR TABLE OF CONTENTS (Cont'd.) TABLE (Cont'd.) Section Section SECTION XIII Title CONDUCT OF OPERATIONS CONDUCT OPERATIONS Page XllI-1 XIII-I A. A. ORGANIZATION AND RESPONSIBILITY ORGANIZATION RESPONSIBILITY XllI-1 XIII-l 1.0 Management Management and Technical Technical Support Support Organization Organization XIII-l XllI-1 1.1 Station Organization Station Organization XllI-1 XIII-l 1.1.1 1.1.1 Vice President Mile Point Nine Mile President Nine XIII-l XlII-1 1.1.22 1.1. Matrixed Reporting Reporting XIII-1 XIII-l 1.1.33 1.1. Qualifications Qualifications of Support Personnel Personnel XIII-2 XIII-2 2.0 Nine Mile Mile Point Nuclear Station, LLC, Station, LLC, Organization Organization XIII-2 XIII-2 2.1 Plant General Manager General Manager XIII-3 XIII-3 2.2 Other Functions Other Functions Reporting Reporting to the Nine Mile Point President Nine Vice President XIII-6 XIII-6 2.3 Supervisor Supervisor Engineering-Nuclear Engineering-Nuclear Fuels XIII-8 XIII-8 3.0 Quality Assurance Quality XIII-9 XIII-9 4.0 Operating Operating

                      \

Shift Crews Crews XIII-9 XIII-9 5.0 Qualifications Qualifications of Staff Personnel Personnel XIII-I0 XIII-10 B. B. 1.0 2.0 3.0 QUALIFICATIONS PERSONNEL PERSONNEL This section section deleted This section deleted This section deleted TRAINING OF QUALIFICATIONS AND TRAINING XIII-lI XIII-II XllI-l1 XIII-ll XllI-11 XIII-ll XllI-11 XIII-l1 4.0 Training Training of Personnel Personnel XIII-ll XllI-11 4.1 General Responsibility General Responsibility XlII-11 XIII-ll 4.2 Implementation Implementation XIII-II XIII-ll 4.3 Quality Quality XIII-II XIII-11 4.3.1 For Operator Operator Training Training XlII-11 XIII-ll 4.3.2 For Maintenance Maintenance XIII-12 4.3.3 For Technicians Technicians XIII-12 4.3.4 For General General Employee Employee Protection and Training/Radiation Protection Training/Radiation Emergency Emergency Plan XIII-12 4.3.5 For Industrial Safety Industrial Safety XIII-12 4.3.6 For Nuclear Nuclear Quality Assurance Assurance XIII-12 XIII-12 4.3.7 For Fire Brigade Brigade XIII-12 4.3.8 For Manager Manager Operations Operations and General General Supervisor Operations Operations XIII-12

                                                             ](111-12 UFSAR          21 UFSAR Revision 21                  xx xx                      October 2009

• Nine Mile Point Point Unit unit 11 UFSAR TABLE TABLE OF CONTENTS (Cont'd.)

(Cont'd.)

• Section 4.4 Title Training of Licensed Licensed Operator Operator Candidates/Licensed NRC Operator Candidates/Licensed Operator Page Retraining Retraining XIII-12 5.0 Cooperative Local, Cooperative Training With Local, State and Federal State Federal Officials Officials XIII-14 C.

C. OPERATING OPERATING PROCEDURES PROCEDURES XIII-15 D. D. EMERGENCY PLAN AND PROCEDURES EMERGENCY XIII-17 E. E. SECURITY SECURITY XIII-19 F. F. RECORDS RECORDS XIII-20 1.0 Operations Operations XIII-20 1.1 Control Room Log XIII-20 1.2 Shift Manager's Manager's Log XIII-20 1.3 Radwaste Radwaste Log XIII-20 1.4 Waste Quantity Waste Quantity Level Level Shipped XIII-20 2.0 Maintenance Maintenance XIII-21 3.0 Radiation Protection Radiation Protection XIII-21 3.1 Personnel Exposure Personnel Exposure XIII-21 3.2 By-Product Material as Required By-Product Material Required by 10CFR30 10CFR30 XIII-21 3.3 Meter Calibrations Calibrations XIII-21 XIII~21 3.4 Station Radiological Condition~ Station Radiological Conditionsý inin Accessible Areas Accessible Areas XIII-21

• 21 Revision 21 UFSAR Revision xxa xxa October 2009

Nine Mile Nine Unit 11 UFSAR Point Unit Mile Point UFSAR THIS PAGE THIS INTENTIONALLY BLANK PAGE INTENTIONALLY BLANK Revision 17 UFSAR Revision UFSAR 17 xxb xxb October 2001 October 2001

• Nine Mile Point Unit 11 UFSAR Nine UFSAR TABLE CONTENTS (Cont'd.)

TABLE OF CONTENTS (Gont'd.)

• Section 3.S 3.5 Title Administration of the Radiation Administration Protection Procedures Protection Program and Procedures XIII-21 4.0 4.0 Radiochemistry Chemistry and Radiochemistry XIII-21 5.0 S.O Special Special Nuclear Materials Nuclear Materials XIII-22 6 .0 6.0 Calibration of Instruments Calibration Instruments XIII-22 7.0 7.0 Administrative Records Administrative Reports Records and Reports XIII-22 G.

G. REVIEW AND AUDIT OF OPERATIONS OPERATIONS XIII-23 1.0 Plant Plant Operations Committee Operations Review Committee XIII-23 1.1 Function XIII-23 2.0 2.0 Nuclear Safety Review Board Board XIII-23 2.1 2.1 Function XIII-24 3.0 3.0 Review Operating Experience Review of Operating Experience XIII-24 SECTION XIV INITIAL TESTING AND OPERATIONS INITIAL OPERATIONS XIV-1 XIV-1 A. A. TESTS PRIOR TO INITIAL REACTOR REACTOR FUELING XIV-XIV-11

• B.

B. 1.0 1.0 1.1 INITIAL CRITICALITY AND INITIAL CRITICALITY POSTCRITICALITY TESTS POSTCRITICALITY TESTS Initial Fuel Loading and Near-Zero Initial Atmospheric Pressure Power Tests at Atmospheric Power Requirements General Requirements General XIV-5 XIV-S XIV-5 XIV-S XIV-5 XIV-S 1.2 1.2 General Procedures General Procedures XIV-5 XIV-S 1.3 1.3 Core Loading Test Loading and Critical Test Program Program XIV-XIV-77 2.0 2.0 Heatup from Ambient to Rated Rated Temperature Temperature XIV-99 XIV-2.1 2.1 General General XIV-XIV-99 2.2 2.2 Tests Conducted Conducted XIV-9 XIV-9 3.0 3.0 From Zero to 100 Percent Initial Initial Reactor Reactor Rating XIV-10 XIV-10 4.0 4.0 Full-Power Full-Power Demonstration Run Demonstration Run XIV-XIV-1212 5.0 S.O Comparison of Base Conditions XIV-XIV-1212 6.0 6.0 Additional Tests at Design Rating Additional XIV-XIV-1313 7.0 7.0 Report Startup Report XIV-XIV-1313 XV SECTION XV SAFETY ANALYSIS ANALYSIS XV-1 XV-1 A. A. INTRODUCTION INTRODUCTION XV-XV-11

• Revision 21 UFSAR Revision 21 xxi xxi 2009 October 2009

Nine Mile Point Unit 1 UFSAR TABLE OF CONTENTS CONTENTS (Cont'd.) (Cont'd.) Section B. B. Title BOUNDARY PROTECTION Transients PROTECTION SYSTEMS Transients Considered Page XV-2 XV-2 XV-2 XV-2 1.0 2.0 Methods and Assumptions Methods Assumptions XV-3 XV-3 3.0 Transient Analysis Analysis XV-3 XV-3 3.1 3.1 Turbine Trip Without Bypass XV-3 XV-3 3.1.1 Objectives Objectives XV-3 XV-3 3.1. 3.1.22 Assumptions and Initial Assumptions Initial Conditions Conditions XV-3 XV-3 3.1. 3.1.33 Comments XV-3 XV-3 3.1.4 Results XV-3 XV-3 3.2 Loss of 100°F Feedwater Heating 100OF Feedwater XV-4 XV-4 3.2.1 Objective Objective XV-4 XV-4 3.2.2 Assumptions and Initial Assumptions Initial Conditions XV-4 XV-4 3.2.3 Results XV-4 XV-4 3.3 Feedwater Controller Feedwater Failure-Controller Failure-Maximum Demand XV-5 XV-5 3.3.1 Objective Objective XV-5 Xv-s 3.3.2 Assumptions and Initial Initial Conditions Xv-s XV-5 3.3.3 Comments Comments XV-5 Xv-s 3.3.4 Results Xv-s XV-5 3.4 Control Rod Withdrawal Control Withdrawal Error Error Xv-s XV-5 3.4.1 Objective Objective XV-5 XV-S 3.4.2 Assumptions and Initial Initial Conditions XV-5 XV-5 3.4.3 Comments XV-6 XV-6 3.4.4 Results Results XV-6 XV-6 3.5 Main Steam Line Isolation ValveValve Closure (With Scram) XV-6 XV-6 3.5.1 Objective Objective XV-6 XV-6 3.5.2 Assumptions and Initial Initial Conditions XV-6 XV-6 3.5.3 Comments XV-7 XV-7 3.5.4 Results XV-7 XV-7 3.6 Inadvertent Inadvertent Startup of Cold Cold Recirculation Recirculation Loop XV-7 XV-7 3.6.1 Objective Objective XV-7 XV-7 3.6.2 Assumptions and Initial Initial Conditions XV-7 XV-7 3.6.3 Comment XV-8 XV-8 3.6.4 Results Results XV-9 XV-9 3.7 Recirculation Pump Trips Recirculation XV-9 XV-9 3.7.1 Objectives Objectives XV-9 XV-9 3.7.2 Assumptions and Initial Assumptions Initial Conditions Conditions XV-9 XV-9 3.7.3 Comments XV-9 XV-9 3.7.4 Results XV-10 XV-10 3.8 Recirculation Recirculation Pump Stall XV-10 XV-10 A 16 UFSAR Revision 16 xxii xxii November 1999 1999

Nine Mile Point Unit 11 UFSAR CONTENTS (Cont'd.) TABLE OF CONTENTS (Cont'd.) Section Section Title Title Page 1.2.5 1.2.5 Subcooled Subcooled Liquid XV-30 XV-30 1.2.6 1.2.6 System Pressure Pressure and Steam-Water Steam-Water Mass XV-31 XV-31

1. 2.7 1.2.7 Mixture Impact Forces Mixture Forces XV-31 XV-31

1. 2.8 1.2.8 Core Internal Internal Forces Forces XV-31 XV-31 1.3 Radiological Effects Radiological XV-32 XV-32

1. 3.1 1.3.1 Radioactivity Radioactivity Releases Releases XV-32 XV-32 1.3.2 Meteorology and Dose Rates Meteorology Rates XV-33 XV-33 2.0 Loss-of-Coolant Loss-of-Coolant Accident XV-33 XV-33 2.1 Introduction Introduction XV-33 XV-33 2.2 Input to Analysis XV-35 XV-35 2.2.1 Operational Operational and ECCS Input Input Parameters Parameters XV-35 XV-35 2.2.2 Single Failure Single Failure Study on ECCS Manually-Controlled Manually-Controlled Electrically-Operated Valves Electrically-Operated XV-35 XV-35 2.2.3 Single Failure Failure Basis XV-35 XV-35 Pipe Whip Basis Basis XV-36 2.2.4 XV-36 2.3 section deleted This section XV-36 XV-36 2.4 Performance Appendix K LOCA Performance Analysis XV-36 XV-36 2.4.1 2.4.1 Computer Codes XV-36 XV-36 2.4.2 2.4.2 Description Description of Model Changes XV-37 XV-37 2.4.3 2.4.3 Analysis Procedure Procedure XV-37 XV-37 2.4.3.1 BWR/2 BWR/2 Generic Analysis Analysis XV-37 XV-37 2.4.3.2 unit 1-Specific Unit I-Specific Analysis Break Break Evaluation Spectrum Evaluation XV-38 XV-38 2.4.4 Analysis Results Analysis Results XV-38 XV-38 3.0 Refueling Refueling Accident Accident XV-40 XV-40 3.1 Identification of Causes Identification Causes XV-40 XV-40 3.2 Accident Analysis Accident XV-41 XV-41 3.3 Radiological Radiological Effects XV-44 XV-44 3.3.1 Fission Product Releases Releases XV-44 XV-44 3.3.2 Meteorology Meteorology and Dose Rates Rates XV-45 XV-45 4.0 Control Rod Drop Control Drop Accident XV-46 XV-46 4.1 Identification of Causes Identification XV-46 XV-46 4.2 Accident Analysis Analysis XV-46 XV-46 4.3 Designed Safeguards Safeguards XV-47 XV-47 4.4 Procedural Safeguards Procedural Safeguards XV-48 XV-48 4.5 Radiological Effects Radiological XV-48 XV-48 4.5.1 Releases Fission Product Releases XV-48a XV-48a 4.5.2 4.5.2 Meteorology and Dose Rates Meteorology Rates XV-49 XV-49 UFSAR Revision 21 UFSAR Revision 21 xxv XXV October 2009 2009

Nine Mile Point Unit 11 UFSAR CONTENTS (Cont'd.) TABLE OF CONTENTS (Cont'd.) Section Section Title Page 5.0 Containment Design Basis Accident Containment Accident XV-50 XV-50 5.1 Original Recirculation Line Rupture Original Recirculation Analysis -- With Core Spray Analysis Spray XV-50 XV-50 5.1.1 5.1.1 Purpose XV-50 XV-50 5.1. 2 5.1.2 Analysis Method Method and Assumptions Assumptions XV-51 XV-51 5.1. 5.1.33 Core Heat Buildup Buildup XV-51 XV-51 5.1.4 Spray System Core Spray XV-52 XV-52 5.1. 5.1.55 Containment Containment Pressure Immediately Immediately Following Blowdown Blowdown XV-53 XV-53 5.1. 5.1.66 Containment Containment Spray XV-54 XV-54 5.1. 5.1.77 Blowdown Effects Blowdown Effects on Core Components Components XV-55 XV-55 5.1.8 5.1. 8 Radiological Effects Radiological XV-56 XV-56 5.1.8.1 5.1.8.1 Fission Product Releases Releases XV-56 XV-56 5.1.8.2 Meteorology and Dose Rates Meteorology XV-58 XV-58 5.2 Original Containment Original Design Basis Containment Design Basis Accident Accident Analysis - Without without Core Spray XV-59 XV-59 5.2.1 Purpose XV-59 XV-59 5.2.2 5.2.3 5.3 5.3.1 5.3.1 5.3.2 5.3.2 Core Heatup Containment Response Containment Response Design Basis Reconstitution Reconstitution Suppression Chamber Heatup Suppression Introduction Introduction Input to Analysis Input Heatup Analysis XV-59 XV-59 XV-60 XV-60, XV-61 XV-61 XV-61 XV-61 XV-61a XV-61a 5.3.3 DBR Suppression Chamber Heatup Suppression Chamber Analysis Analysis XV-61a XV-61a 5.3.3.1 Computer Codes Codes XV-62 XV-62 5.3.3.2 Analysis Analysis Methods XV-62 XV-62 5.3.3.3 Analysis Results for Containment Containment Spray Design Basis Assumptions Basis Assumptions XV-64 XV-64 5.3.3.4 5.3.3.4 Analysis Results for EOPEOP Operation Assumptions Operation Assumptions XV-65 XV-65 5.3.4 Conclusions Conclusions XV-65a XV-65a 6.0 New Fuel Bundle Loading Error Error Analysis Analysis XV-66 6.1 Identification of Causes Identification Causes XV-66 XV-66 6.2 Accident Analysis Accident Analysis XV-67 XV-67 6.3 Safety Requirements Requirements XV-67 XV-67 7.0 Meteorological Models Used in Meteorological in Accident Analyses Accident Analyses XV-68 XV-68 7.1 Introduction Introduction XV-68 XV-68 UFSAR Revision 21 UFSAR 21 xxvi xxvi October 2009

Nine Mile Nine Unit 11 UFSAR Point Unit Mile Point UFSAR TABLE OF TABLE CONTENTS (Cont'd.) OF CONTENTS (Cont'd.) Section Section Title Title Page 7.2 7.2 Atmospheric Dispersion Factor Atmosphe~ic Dispersion Factor Calculations Calculations XV-68 XV-68 7.2.1 7.2.1 Offsite and LPZ EAB and Offsite -- EAB LPZ XV-69 XV-69 7.2.2 7.2.2 Control Room and Technical Support Control Room and Technical Support Center (Excluding MSLB) Center (Excluding MSLB) XV-69 XV-69 7.2.3 7.2.3 Control Puff Release MSLB Puff Room -- MSLB Control Room Release XV-70 XV-70 7.3 7.3 Summary of Summary of Results Results XV-70 XV-70 7.4 7.4 Exfiltration Exfiltration XV-70 XV-70 7.5 7.5 Containment Drawdown Secondary Containment Secondary Drawdown XV-76 XV-76 7.5.1 7.5.1 Introduction Introduction XV-76 XV-76 7.5.2 7.5.2 Analysis Analysis XV-76 XV-76 7.5.3 7.5.3 Results Results XV-77 XV-77

• UFSAR Revision 21 21 xxvia xxvia October October 2009 2009

Nine Mile Nine unit 11 UFSAR Point Unit Mile Point UFSAR THIS PAGE THIS INTENTIONALLY BLANK PAGE INTENTIONALLY BLANK 18 UFSAR Revision 18 UFSAR xxvib xxvib October October 2003 2003

• Point Unit Mile Point Nine Mile Unit 11 UFSAR UFSAR (Cont'd.)

CONTENTS (Cont'd.) TABLE OF CONTENTS Section section Title Title Page 5.2 5.2 NUREG-0737, Supplement 1, NUREG-0737, Supplement Section 1, Section 4.1.b 4.1.b XVIII-13 XVIII-13 5.2.1 5.2.1 Convenient Location Convenient XVIII-13 XVIII-13 5.2.2 5.2.2 Continuous Display Continuous XVIII-13 5.3 5.3 NUREG-0737, Supplement 1, NUREG-0737, Supplement Section 1, Section 4.1.cc 4.1. XVIII-13 5.3.1 5.3.1 Procedures and Training procedures XVIII-13 5.3.2 5.3.2 Isolation of SPDS from SPDS from Safety-Related Systems Safety-Related systems XVIII-13 5.4 5.4 NUREG-0737, Supplement 1, NUREG-0737, Supplement Section 1, Section 4.l.e e 4.1. XVIII-14 5.4.1 5.4.1 Incorporation of Accepted Human Incorporation Human Principles Factors Engineering Principles XVIII- 14 XVIII-14 5.4.2 5.4.2 Information Can Be Readily Information perceived Comprehended Perceived and Comprehended XVIII-14 5.5 5.5 NUREG-0737, Supplement 1, NUREG-0737, Supplement 1, Section 4.l.f, Sufficient Information 4.1.f, Sufficient Information XVI II- 15 XVIII-IS 6.0 6.0 Procedures procedures XVIII-IS XVIII-15 6.1 6.1 Operating Procedures Operating Procedures XVIII-15 XVIII-IS 6.2 6.2 Surveillance Procedures Surveillance Procedures XVIII-15 XVIII-IS 7.0 7.0 References References XVIII-16 XVIII- 16 APPENDIX A Unused Unused APPENDIX B POINT NUCLEAR NINE MILE POINT NUCLEAR STATION, LLC, QUALITY STATION, LLC, ASSURANCE PROGRAM TOPICAL REPORT, ASSURANCE PROGRAM NINE MILE POINT REPORT, NINE POINT NUCLEAR STATION NUCLEAR UNITS 11 AND STATION UNITS AND 2 OPERATIONS OPERATIONS PHASE PHASE APPENDIX C APPENDIX RENEWAL SUPPLEMENT LICENSE RENEWAL LICENSE AGING MANAGEMENT SUPPLEMENT -- AGING MANAGEMENT PROGRAMS AND TIME-LIMITED PROGRAMS AGING ANALYSES TIME-LIMITED AGING ANALYSES

• UFSAR Revision 20 UFSAR Revision 20 xxxiii xxxiii October 2007 October 2007

Nine Mile Point Unit 1 UFSAR Nine LIST OF TABLES TABLES Table Table Number Number Title Title I-i 1-1 COMPARISON COMPARISON TO STANDARDS - HISTORICAL HISTORICAL (PROVIDED WITH APPLICATION TO CONVERT CONVERT TO FULL-TERM FULL-TERM OPERATING LICENSE) OPERATING LICENSE) 1-2 ABBREVIATIONS AND ACRONYMS ABBREVIATIONS ACRONYMS USED IN UFSAR II-1 11-1 1980 POPULATION AND POPULATION 1980 POPULATION DENSITY FOR TOWNS POPULATION DENSITY TOWNS AND CITIES WITHIN 12 MILES OF NINE MILE CITIES WITHIN POINT - MILE POINT UNIT 1 UNIT 11-2 CITIES WITHIN WITHIN A 50-MILE 50-MILE RADIUS OF THE STATION 10,000 POPULATIONS OVER 10,000 WITH POPULATIONS 11-3 REGIONAL AGRICULTURAL USE REGIONAL AGRICULTURAL USE 11-4 REGIONAL AGRICULTURAL STATISTICS REGIONAL AGRICULTURAL STATISTICS - CATTLE AND MILK MILK PRODUCTION 11-5 11-6 11-7 INDUSTRIAL PUBLIC PUBLIC WATER PUBLIC FIRMS WITHIN 88 KM INDUSTRIAL FIRMS PUBLIC UTILITIES UTILITIES IN (5 OSWEGO COUNTY IN OSWEGO COUNTY MI) OF UNIT 1 (5 MI) WATER SUPPLY DATA FOR LOCATIONS LOCATIONS WITHIN WITHIN AN APPROXIMATE 3D-MILE APPROXIMATE 30-MILE RADIUS 11-8 RECREATIONAL RECREATIONAL AREAS IN IN THE REGION 11-9 SOURCES OF TOXIC CHEMICALS WITHIN 8 KM CHEMICALS WITHIN (5 MI) OF (5 MI) OF UNIT 1 SITE UNIT 11-10 II-10 PREDICTED VAPOR PREDICTED VAPOR CONCENTRATION CONCENTRATION IN THE UNIT 1 CONTROL ROOM V-1 REACTOR COOLANT SYSTEM DATA REACTOR V-2 OPERATING CYCLES AND TRANSIENT OPERATING ANALYSIS RESULTS TRANSIENT ANALYSIS V-3 FATIGUE FATIGUE RESISTANCE RESISTANCE ANALYSIS ANALYSIS V-4 CODES FOR SYSTEMS SYSTEMS FROM REACTOR REACTOR VESSEL CONNECTION 21 UFSAR Revision 21 ISOLATION VALVE TO SECOND ISOLATION xxxiv xxxiv VALVE October 2009

Nine Mile Point Unit 11 UFSAR LIST OF TABLES (Cont'd.) (Cont'd.) Table Table Number Number Title Title V-5 TIME TO AUTOMATIC AUTOMATIC BLOWDOWN BLOWDOWN VI-l DRYWELL PENETRATIONS PENETRATIONS VI-2 SUPPRESSION SUPPRESSION CHAMBER PENETRATIONS PENETRATIONS VI-3a VI-3a REACTOR COOLANT REACTOR COOLANT SYSTEM SYSTEM ISOLATION ISOLATION VALVES VALVES VI-3b PRIMARY CONTAINMENT PRIMARY CONTAINMENT ISOLATION VALVES - LINES ISOLATION VALVES LINES ENTERING FREE SPACE OF THE CONTAINMENT CONTAINMENT VI-4 SEISMIC DESIGN CRITERIA SEISMIC DESIGN CRITERIA FOR ISOLATION VALVES ISOLATION VALVES VI-5 VI-5 INITIAL TESTS PRIOR TO STATION OPERATIONOPERATION VII-1 VII-l PERFORMANCE TESTS PERFORMANCE

• VIII-1 VIII-l VIII-2 VIII-2 ASSOCIATION BETWEEN BETWEEN PRIMARY EMERGENCY OPERATING EMERGENCY PRIMARY SAFETY FUNCTIONS OPERATING PROCEDURES LIST OF EOP KEY PARAMETERS PROCEDURES PARAMETERS FUNCTIONS AND VIII-3 INSTRUMENT CATEGORY TYPE AND INSTRUMENT CATEGORY FOR UNIT 1 RG 1.97 VARIABLES VARIABLES VIII-4 VIII-4 PROTECTIVE SYSTEM FUNCTION VIII-5 VIII-5 NON-TECHNICAL SPECIFICATION NON-TECHNICAL SPECIFICATION INSTRUMENTATION INSTRUMENTATION THAT THAT INITIATES WITHDRAWAL BLOCK CONTROL ROD WITHDRAWAL INITIATES CONTROL IX-1 IX-l MAGNITUDE AND DUTY CYCLE OF MAJOR MAGNITUDE MAJOR STATION BATTERY BATTERY LOADS LOADS XII-1 XII-1 FLOWS AND ACTIVITIES OF MAJOR SOURCES OF GASEOUS MAJOR SOURCES GASEOUS ACTIVITY XII-2 XII-2 QUANTITIES QUANTITIES AND ACTIVITIES ACTIVITIES OF LIQUID RADIOACTIVE WASTES WASTES XII-3 ANNUAL SOLID WASTE WASTE ACCUMULATION ACCUMULATION AND ACTIVITY UFSAR Revision 21 21 xxxv XX~IV October 2009

Mile Point Unit 1 UFSAR Nine Mile LIST OF TABLES (Cont1d.) (Cont'd.) Table Table Number Number Title Title XII-4 XII-4 LIQUID WASTE LIQUID WASTE DISPOSAL DISPOSAL SYSTEM MAJOR COMPONENTS COMPONENTS XII-S XII-5 SOLID WASTE WASTE DISPOSAL SYSTEM MAJOR COMPONENTS COMPONENTS XII-6 XII-6 OCCUPANCY TIMES OCCUPANCY TIMES XII-7 XII-7 GAMMA ENERGY ENERGY GROUPS GROUPS XII-8 XII-8 RADIATION MONITOR AREA RADIATION MONITOR DETECTOR DETECTOR LOCATIONS XIII-l XIII-i ANSI STANDARD CROSS-REFERENCE CROSS-REFERENCE UNIT 1 XIII-2 XIII-2 MINIMUM SHIFT CREW MINIMUM CREW COMPOSITION XV-1 XV-l TABLE DELETED XV -2 XV-2 TRIP POINTS FOR PROTECTIVE FUNCTIONS FUNCTIONS XV-4 XV-S5 XV-thru XV-3 thru XV-4 TABLES DELETED BLOWDOWN RATES RATES XV- 6 XV-6 REACTOR COOLANT COOLANT CONCENTRATIONS CONCENTRATIONS (pCi/gm) XV- 7 XV-7 TABLE DELETED XV- 7a XV-7a MSLB ACCIDENT MSLB ACCIDENT ANALYSIS INPUTS AND ASSUMPTIONS ANALYSIS INPUTS ASSUMPTIONS XV-7b XV- 7b MSLB ACCIDENT MSLB ACCIDENT RELEASE RELEASE RATES RATES XV-8 XV-8 MAIN STEAM LINE BREAK ACCIDENT DOSESDOSES XV-9 XV-9 SIGNIFICANT PARAMETERS TO THE INPUT PARAMETERS SIGNIFICANT INPUT THE LOSS-OF-COOLANT LOSS-OF-COOLANT ACCIDENT ANALYSIS ANALYSIS XV-9a XV- 9a SYSTEM FLOW PERFORMANCE CORE SPRAY SYSTEM PERFORMANCE ASSUMED ININ LOCA ANALYSIS ANALYSIS XV-10 XV-10 ECCS SINGLE VALVE FAILURE ANALYSIS ANALYSIS UFSAR Revision 21 21 xxxvi xxxvi October 2009

Nine Mile Point UnitUnit 1 UFSAR LIST OF TABLES (Cont'd.) (Cont'd.)

• Table Table Number Number Title Title SINGLE FAILURES FAILURES CONSIDERED IN LOCA ANALYSIS ANALYSIS XV-11 thru XV-12 thru TABLES DELETED XV-21 XV-21 XV-21a XV-21a ANALYSIS ASSUMPTIONS ASSUMPTIONS FOR NINE MILE MILE POINT POINT.1 1 CALCULATIONS CALCULATIONS XV-22 XV-22 ACTIVITY RELEASED TO THE REACTOR BUILDING ACTIVITY RELEASED FOLLOWING THE FHA (CURIES)

FOLLOWING (CURIES) XV-23 XV-23 UNIFORM UNFILTERED UNFILTERED STACK DISCHARGE RATES FROM DISCHARGE RATES FROM 0 TO 2 HR AFTER THE FHA (CURIES/SEC) (CURIES/SEC) XV-24 XV-24 HANDLING ACCIDENT DOSES FUEL HANDLING DOSES XV-2S XV-25 FHA ANALYSIS FHA ANALYSIS INPUTS AND ASSUMPTIONS ASSUMPTIONS

• XV-26 XV-26 XV-27 XV-27 XV-28 XV-28 CRD ACCIDENT ANALYSIS ANALYSIS INPUTS AND ASSUMPTIONS RELEASE CRDA NOBLE GAS RELEASE CRDA HALOGEN RELEASE CRDA RELEASE ASSUMPTIONS XV-29 XV-29 CONTROL ROD DROP ACCIDENT DOSES CONTROL DOSES XV-29a XV-29a WETTING OF FUEL CLADDING BY CORE SPRAY XV-29b XV-29b POST-LOCA POST-LOCA AIRBORNE PRODUCT AIRBORNE DRYWELL FISSION PRODUCT INVENTORY (CURIES)

INVENTORY (CURIES) XV-29c XV-29c POST-LOCA POST-LOCA REACTOR BUILDING FISSION PRODUCT PRODUCT (CURIES) INVENTORY (CURIES) INVENTORY XV-29d XV-29d POST-LOCA POST-LOCA DISCHARGE (CURIES/SEC) DISCHARGE RATES (CURIES/SEC) XV-30 XV-30 PRODUCT INVENTORY CORE FISSION PRODUCT XV-31 XV-31 LOCA ANALYSIS LOCA ANALYSIS INPUTS AND ASSUMPTIONS ASSUMPTIONS XV-32 XV-32 LOSS-OF-COOLANT ACCIDENT DOSES LOSS-OF-COOLANT DOSES Revision 21 UFSAR Revision 21 xxxvii xxxv 1 1 October 2009 2009

Nine Mile Nine Unit 11 UFSAR Point Unit Mile Point UFSAR LIST OF LIST TABLES (Cont'd.) OF TABLES (Cont'd.) Table Table Number Number Title Title XV-32a XV-32a SIGNIFICANT INPUT SIGNIFICANT PARAMETERS TO INPUT PARAMETERS THE DBR TO THE DBR CONTAINMENT SUPPRESSION CONTAINMENT HEATUP ANALYSIS CHAMBER HEATUP SUPPRESSION CHAMBER ANALYSIS XV-33 XV-33 TABLE DELETED TABLE DELETED XV-34 XV-34 TABLE DELETED TABLE DELETED XV-34a XV-34a RELEASE/INTAKE ELEVATIONS RELEASE/INTAKE ELEVATIONS XV-34b XV-34b RELEASE/INTAKE DISTANCE RELEASE/INTAKE AND DIRECTIONS DISTANCE AND DIRECTIONS XV-35 XV-35 TABLE DELETED TABLE DELETED XV-35a XV-35a X/Q VALUES X/Q FOR THE VALUES FOR CONTROL ROOM THE CONTROL ROOM XV-35b XV-35b X/Q VALUES X/Q THE TECHNICAL FOR THE VALUES FOR SUPPORT CENTER TECHNICAL SUPPORT CENTER XV-35c XV-35c XV-35d XV-35d XV-36 XV-36 OFFSITE X/Q OFFSITE OFFSITE X/Q OFFSITE REACTOR VALUES FOR X/Q VALUES X/Q VALUES GROUND-LEVEL RELEASES FOR GROUND-LEVEL LEAKAGE PATHS BUILDING LEAKAGE REACTOR BUILDING PATHS RELEASES ELEVATED RELEASES FOR ELEVATED VALUES FOR RELEASES

• XVI-1 XVI-1 CODE CALCULATION

SUMMARY

CODE CALCULATION

SUMMARY

XVI-2 XVI-2 STEADY-STATE -- (100% STEADY-STATE FULL POWER (100% FULL NORMAL OPERATION) POWER NORMAL OPERATION) PERTINENT STRESSES PERTINENT OR STRESS INTENSITIES STRESSES OR STRESS INTENSITIES XVI-3 XVI-3 LIST OF LIST REACTIONS FOR OF REACTIONS FOR REACTOR VESSEL NOZZLES REACTOR VESSEL NOZZLES XVI-4 XVI-4 EFFECT OF EFFECT VALUE OF OF VALUE OF INITIAL FAILURE PROBABILITY INITIAL FAILURE PROBABILITY XVI-5 XVI-5 SINGLE TRANSIENT SINGLE EVENT FOR TRANSIENT EVENT REACTOR PRESSURE FOR REACTOR PRESSURE VESSEL VESSEL XVI-6 XVI-6 POSTULATED EVENTS POSTULATED EVENTS XVI-7 XVI-7 MAXIMUM STRAINS FROM MAXIMUM STRAINS POSTULATED EVENTS FROM POSTULATED EVENTS STRUCTURE ANALYSIS CORE STRUCTURE CORE LINE BREAK RECIRCULATION LINE ANALYSIS RECIRCULATION XVI-8 XVI-8 BREAK Revision 21 UFSAR Revision UFSAR 21 xxxviia xxxviia October October 2009 2009

Nine Mile Point Unit 1 UFSAR LIST OF TABLES (Cont'd.) (Cont'd.) Table Table Number Number Title Title XVI-9 CORE STRUCTURE STRUCTURE ANALYSIS STEAM LINE BREAK XVI-9a CORE SHROUD REPAIR DESIGN SUPPORTING CORE DOCUMENTATION DOCUMENTATION XVI-I0 XVI-10 DRYWELL JETJET AND MISSILE HAZARD ANALYSIS ANALYSIS DATA DATA XVI-II XVI-ll DRYWELL JET AND MISSILE HAZARD ANALYSIS ANALYSIS RESULTS RESULTS XVI-12 STRESS DUE TO DRYWELL FLOODING XVI-13 XVI-13 ALLOWABLE WELD SHEAR STRESS ALLOWABLE STRESS XVI-14 XVI-14 LEAK RATE TEST RESULTS RESULTS XVI-IS XVI-15 OVERPRESSURE TEST--PLATE STRESSES OVERPRESSURE TEST--PLATE

• XVI-16 STRESS

SUMMARY

• UFSAR 21 UFSAR Revision 21 xxxviib October 2009 October 2009

Point unit Nine Mile Point Unit 1 UFSAR LIST OF TABLES (Cont'd.) (Cont'd.) Table Number Number XVI-17

        'Title  TjtJe HEAT TRANSFER TRANSFER COEFFICIENTS AS A FUNCTION OF DROP DIAMETER DROP 0

XVI-18 XVI-18 HEAT TRANSFER TRANSFER COEFFICIENT COEFFICIENT AS A A FUNCTION FUNCTION OF OF PRESSURE XVI-19 RELATIONSHIP BETWEEN PARTICLE SIZE AND TYPE OF RELATIONSHIP BETWEEN OF SPRAY PATTERN SPRAY XVI-20 ALLOWABLE STRESSES FOR FLOOR SLABS, BEAMS, XVI-20 ALLOWABLE STRESSES FOR FLOOR SLABS, BEAMS, WALLS, FOUNDATIONS, COLUMNS, WALLS, COLUMNS, ETC. FOUNDATIONS, ETC. XVI-21 XVI-21 ALLOWABLE ALLOWABLE STRESSES FOR STRUCTURAL STRUCTURAL STEEL STEEL XVI-22 XVI-22 ALLOWABLE STRESSES - REACTOR VESSEL CONCRETE ALLOWABLE CONCRETE PEDESTAL XVI-23 DRYWELL - ANALYZED ANALYZED DESIGN LOAD COMBINATIONS COMBINATIONS XVI-24 XVI-24 SUPPRESSION SUPPRESSION CHAMBER -- ANALYZED ANALYZED DESIGN LOAD COMBINATIONS COMBINATIONS XVI-25 XVI-25 ACI CODE 505 ALLOWABLE ACI ALLOWABLE STRESSES AND ACTUAL ACTUAL STRESSES STRESSES FOR CONCRETE VENTILATION STACK CONCRETE VENTILATION XVI-26 XVI-26 ALLOWABLE STRESSES FOR CONCRETE CONCRETE SLABS, SLABS, WALLS, WALLS, BEAMS, STRUCTURAL BEAMS, STRUCTURAL STEEL, STEEL, AND CONCRETE CONCRETE BLOCK WALLS WALLS XVI-27 SYSTEM LOAD COMBINATIONS COMBINATIONS XVI-28 XVI-28 HIGH-ENERGY HIGH-ENERGY SYSTEMS SYSTEMS - INSIDE CONTAINMENT CONTAINMENT XVI-29 XVI-29 HIGH-ENERGY SYSTEMS - OUTSIDE HIGH-ENERGY OUTSIDE CONTAINMENT CONTAINMENT XVI-3D XVI-30 SYSTEMS WHICH MAY BE AFFECTED BY PIPE WHIP WHIP XVI-31 CAPABILITY CAPABILITY TO RESIST WIND PRESSURE PRESSURE AND WIND VELOCITY XVII-1 XVII-l DISPERSION AND ASSOCIATED METEOROLOGICAL METEOROLOGICAL PARAMETERS PARAMETERS XVII-2 RELATION OF SATELLITE SATELLITE AND NINE MILE POINT WINDS WINDS XVII-3 XVII-3 FREQUENCY OF OCCURRENCE FREQUENCY OCCURRENCE OF LAPSE RATES - 1963 AND 1964 UFSAR Revision 16 Revision 16 xxxviii xxxviii November 1999

• Nine Mile Point Unit 1 UFSAR FIGURES LIST OF FIGURES (Cont'd.)

(Cont'd.) Figure Number Title Title X-2 REACTOR CLEANUP REACTOR CLEANUP SYSTEM SYSTEM X-3 CONTROL ROD DRIVE HYDRAULIC HYDRAULIC SYSTEM SYSTEM X-4 REACTOR BUILDING CLOSED LOOP COOLING SYSTEMSYSTEM X-s X-5 TURBINE BUILDING BUILDING CLOSED LOOP COOLING SYSTEM SYSTEM X-6 X-6 SERVICE WATER SYSTEM SERVICE WATER SYSTEM X-7 FIGURE DELETED FIGURE X-8 X-8 SPENT FUEL STORAGE SPENT STORAGE POOL FILTERING FILTERING AND COOLING SYSTEM SYSTEM X-9 BREATHING, BREATHING, INSTRUMENT, AND SERVICE AIR INSTRUMENT,

• X-10 X-10 X-l1 X-11 XI-1 XI-l REACTOR REFUELING REACTOR REFUELING SYSTEM PICTORIAL CASK DROP PROTECTION PROTECTION SYSTEM STEAM FLOW AND REHEATER PICTORIAL SYSTEM VENTILATION SYSTEM REHEATER VENTILATION XI-2 EXTRACTION STEAM STEAM FLOW FLOW XI-3 MAIN CONDENSER MAIN CONDENSER AIR REMOVAL REMOVAL AND OFFGAS SYSTEM SYSTEM XI-4 CIRCULATING CIRCULATING WATER SYSTEM SYSTEM XI-5 CONDENSATE FLOW CONDENSATE FLOW XI-6 XI-6 CONDENSATE TRANSFER CONDENSATE TRANSFER SYSTEM SYSTEM XI-7 XI-7 FEEDWATER FLOW SYSTEM FEEDWATER SYSTEM XII-1 RADIOACTIVE RADIOACTIVE WASTE DISPOSAL DISPOSAL SYSTEM SYSTEM XIII-1 XIII-l STATION MANAGEMENT SENIOR LEVEL STATION ORGANIZATION MANAGEMENT ORGANIZATION CHART CHART
• XIII-2 XIII-2 UFSAR ENGINEERING 21 UFSAR Revision 21 SERVICES ORGANIZATION ENGINEERING SERVICES xlvii xlvii ORGANIZATION CHART CHART October 2009

Nine Mile Point Unit 1 UFSAR UFSAR LIST OF FIGURES (Cont'd.) (Cont'd.) Figure Number Number XIII-3 XIII-3 Title Title QUALITY QUALITY ASSURANCE ORGANIZATION ASSURANCE ORGANIZATION XIII-3a XIII-3a NUCLEAR SAFETY & & SECURITY SECURITY ORGANIZATION ORGANIZATION XIII-4 XIII-4 NUCLEAR STATION ORGANIZATION NINE MILE POINT NUCLEAR CHART CHART XIII-4a XIII-4a NINE MILE POINT POINT NUCLEAR NUCLEAR STATION STATION ORGANIZATION CHART CHART XIII-4b XIII-4b NINE MILE POINT NUCLEAR NUCLEAR STATION STATION ORGANIZATION CHART CHART XIII-4c XIII-4c NINE MILE POINT NUCLEAR NUCLEAR STATION ORGANIZATION ORGANIZATION CHART CHART XIII-5 XIII-5 SAFETY ORGANIZATION ORGANIZATION XV-1 XV-I XV-2 XV-2 XV-3 XV-3 STATION TRANSIENT TRANSIENT DIAGRAM FIGURE DELETED DIAGRAM PLANT RESPONSE TO LOSS OF 100°F 100OF FEEDWATER HEATING UFSAR Revision 21 UFSAR 21 xlviia October 2009

• Nine Mile Point Unit 1 UFSAR LIST OF FIGURES (Cont'd.)

(Cont'd.) Figure Number Title XV-56E XV-56E LOSS-OF-COOLANT ACCIDENT LOSS-OF-COOLANT ACCIDENT DRYWELL PRESSURE XV-56F LOSS-OF-COOLANT LOSS-OF-COOLANT ACCIDENT SUPPRESSION CHAMBER SUPPRESSION CHAMBER PRESSURE PRESSURE XV-56G XV-56G LOSS-OF-COOLANT LOSS-OF-COOLANT ACCIDENT CONTAINMENT CONTAINMENT TEMPERATURE TEMPERATURE

                 - WITH CORE SPRAY XV-57           CONTAINMENT DESIGN BASIS CLAD TEMPERATURE CONTAINMENT                         TEMPERATURE WITHOUT CORE SPRAY RESPONSE - WITHOUT

RESPONSE

XV-58 XV-58 CONTAINMENT DESIGN BASIS METAL-WATER CONTAINMENT DESIGN METAL-WATER REACTION XV-59 CONTAINMENT DESIGN CONTAINMENT DESIGN BASIS CLAD PERFORATION PERFORATION WITHOUT WITHOUT CORE SPRAY XV-60 CONTAINMENT DESIGN BASIS CONTAINMENT CONTAINMENT DESIGN TEMPERATURE CONTAINMENT TEMPERATURE

                 - WITHOUT CORE SPRAY XV-60A          DBR ANALYSIS  SUPPRESSION POOL AND WETWELL ANALYSIS SUPPRESSION               WETWELL AIRSPACE AIRSPACE TEMPERATURE TEMPERATURE RESPONSE - CONTAINMENT CONTAINMENT SPRAY DESIGN BASIS ASSUMPTION DESIGN XV-60B          DBR ANALYSIS  SUPPRESSION POOL AND WETWELL ANALYSIS SUPPRESSION               WETWELL AIRSPACE TEMPERATURE AIRSPACE                 RESPONSE - EOP OPERATION TEMPERATURE RESPONSE ASSUMPTIONS ASSUMPTIONS XV-61          REACTOR BUILDING MODEL MODEL XV-62          EXFILTRATION VS.

EXFILTRATION WIND SPEED - NORTHERLY VS. WIND NORTHERLY WIND XV-63 REACTOR BUILDING DIFFERENTIAL REACTOR DIFFERENTIAL PRESSURE XV-64 EXFILTRATION EXFILTRATION VS. VS. WIND SPEED - SOUTHERLY SOUTHERLY WIND XV-65 REACTOR BUILDING -- ISOMETRIC REACTOR ISOMETRIC XV-66 REACTOR BUILDING -- CORNER SECTIONS SECTIONS XV-67 XV-67 REACTOR BUILDING BUILDING -- ROOF SECTIONS SECTIONS UFSAR Revision Revision 21 21 X1tix xlix October October 2009

Nine Mile Point Unit 1 UFSAR LIST OF FIGURES FIGURES (Cont'd.) (Cont'd.) Figure Number Number Title Title XV-68 XV-68 REACTOR BUILDING -PANEL REACTOR CONCRETE SECTIONS

                                    - PANEL TO CONCRETE   SECTIONS XV-69 XV-69           REACTOR BUILDING REACTOR    BUILDING - EXPANSION JOINT SECTIONS SECTIONS XV-70 XV-70           REACTOR BUILDING EXFILTRATION REACTOR              EXFILTRATION - NORTHERLY NORTHERLY WIND XV-71 XV-71           REACTOR BUILDING EXFILTRATION REACTOR              EXFILTRATION - SOUTHERLY SOUTHERLY WIND XV-72 XV-72           REACTOR BUILDING DIFFERENTIAL       PRESSURE DIFFERENTIAL PRESSURE XV-73 XV-73                                PRESSURE VS.

REACTOR BUILDING PRESSURE REACTOR VS. TIME BY REACTOR BUILDING ELEVATION XV-74 XV-74 REACTOR BUILDING PRESSURE PRESSURE VS. VS. TIME BY REACTOR REACTOR BUILDING ELEVATION ELEVATION (FOCUSED (FOCUSED ON THE INITIAL 2.52.5 HR) XVI-l XVI-1 XVI-2 XVI-2 SEISMIC ANALYSIS SEISMIC REACTOR MOMENT MOMENT ANALYSIS OF REACTOR LUMPED MASS REACTOR VESSEL GEOMETRIC MASS REPRESENTATION REACTOR SUPPORT SUPPORT DYNAMIC ANALYSIS GEOMETRIC A] ANALYSIS - ELEVATION ELEVATION VS. AND ND V*S. XVI-3 XVI-3 REACTOR SUPPORT DYNAMIC ANALYSIS REACTOR ANALYSIS - ELEVATION ELEVATION VS.V*S. SHEAR XVI-4 XVI-4 REACTOR SUPPORT REACTOR SUPPORT DYNAMIC ANALYSIS ANALYSIS - ELEVATION ELEVATION VS.V*S. DEFLECTION XVI-5 XVI-5 REACTOR SUPPORT DYNAMIC REACTOR DYNAMIC ANALYSIS ANALYSIS - ELEVATION ELEVATION VS.V*3 . ACCELERATION XVI-6 thru thru FIGURES DELETED XVI-8 XVI-8 XVI-9 XVI-9 REACTOR VESSEL SUPPORT REACTOR SUPPORT STRUCTURE STRUCTURE STRESS

SUMMARY

XVI-I0 XVI-10 THERMAL ANALYSIS THERMAL ANALYSIS XVI-II XVI-11 FAILURE PROBABILITY DENSITY FUNCTION PROBABILITY DENSITY UFSAR Revision Revision 21 21 1 October 2009

Mile Point Nine Mile Point Unit I1 UFSAR (Cont'd.) LIST OF FIGURES (Cont'd.) Figure Number Number Title Title XVI-12 XVI- 12 ADDITION STRAINS PAST ADDITION PAST 4% REQUIRED REQUIRED TO EXCEED DEFINED SAFETY SAFETY MARGIN XVI-12a XVI-12a SHROUD WELDS SHROUD WELDS XVI-12b XVI- 12b CORE SHROUD STABILIZERS CORE STABILIZERS XVI-12c XVI-12c CORE SHROUD CORE SHROUD WELDS WELDS XVI-12d XVI-12d V9/V10 V9/Vl0 VERTICAL WELD CLAMP ASSEMBLY XVI-13 XVI-13 LOSS-OF-COOLANT ACCIDENT - CONTAINMENT LOSS-OF-COOLANT PRESSURE CONTAINMENT PRESSURE NO CORE OR CONTAINMENT CONTAINMENT SPRAYS SPRAYS XVI-14 FIGURE DELETED DRYWELL TO CONCRETE CONCRETE AIR GAP GAP XVI-IS XVI-15 DRYWELL XVI-16 XVI-16 TYPICAL PENETRATIONS PENETRATIONS

• UFSAR Revision Revision 21 21 la la October 2009

Nine Mile Point Unit 1 UFSAR INTENTIONALLY BLANK THIS PAGE INTENTIONALLY UFSAR Revision Revision 21 21 lb lb October 2009

Nine Mile Point Unit 1 UFSAR

• pressure pressure drop which maintains developed developed is by is manually adjusted The flow through valve is a

maintains the pressure motor-operated pressure motor-operated from the pressure main through this valve and the second-stage is substantially pressure for this stage is control control with isolation valves and a manual bypass valve for maintenance. room, room, valve. valve. and second-stage pressure constant and the valves, substantially constant is is This maintenance. pressure control valves, therefore, act to therefore, is provided valve control to maintain a constant constant differential differential reactor pressure. above reactor pressure. Changes Changes in in the setting of these these valves are required required only to adjust for for changes in requirements of the drive mechanism cooling requirements in the cooling mechanism as the seal characteristics characteristics change with time, changes in time, and for changes In pump pump flow characteristics. characteristics. ' The cooling water water is monitored by a flow indicator. is monitored indicator. A differential differential pressure indicator indicates pressure indicator difference between indicates the difference reactor pressure pressure and cooling water pressure. water pressure. 2.6 Exhaust Header Header The exhaust header takes water discharged by the drives during water discharged operation and by the third-stage pressure operation controller and conducts pressure controller conducts reactor. water to the reactor. this water The piping is is sized to maintain a low low differential differential (approximately 5 psi) above reactor (approximately reactor pressure pressure in in this header. A check valve header. permits isolating this line from the valve permits reactor reactor vessel automatically prevents reactor vessel and automatically reactor waterwater from from flowing into this line should the supply pressure line should pressure fail.fail. A A flow flow

• a point 2.7 indicator permit measuring element and an indicator during Station operation.

Station operation. upstream of this flow meter point upstream flows. flows. Accumulator Accumulator meter allows checking exhaust line flow measuring the exhaust A bypass line from the pump output to A checking of pump flow to accumulator on each The accumulator each drive is source of stored independent source is an independent energy to scram energy drive. scram that drive. accumulator contains The top of the accumulator contains water; the bottom water; bottom is initially is initially precharged approximately 600 precharged to approximately 600 psi with nitrogen. nitrogen. assure that it To assure is always capable it is producing aa scram, capable of producing scram, the accumulator accumulator is monitored for water leakage and for continuously monitored is continuously leakage for nitrogen pressure. pressure. float-type level switch A float-type switch will actuateactuate an an alarm if if water leaks past the nitrogen-water nitrogen-water barrier barrier and collects in the bottom of the accumulator. in accumulator. A pressure indicator and a pressure indicator pressure switch are connected pressure connected to the accumulator to monitor accumulator monitor nitrogen pressure. nitrogen pressure. During normal operation operation the accumulator accumulator barrier virtually zero pressure drop across barrier has virtually it. across it. If there If should be any loss of nitrogen, nitrogen, the barrier will move onto a stop and further loss will cause cause aa decrease decrease in in the nitrogen pressure. pressure. accumulator barrier will not move down beyond the stop and, The accumulator therefore, will not compress therefore, compress the reduced reduced amount of gas back back up to to pressure. A decrease pressure. decrease in nitrogen pressure will actuate in nitrogen actuate the pressure switchswitch and sound an alarm. alarm. An isolation valve allows allows pressure accumulator instruments to be, each of the accumulator be. isolated isolated and serviced serviced. . A connection A accumulator provides connection on the accumulator provides for precharging precharging and bleeding. bleeding. Revision 19 UFSAR Revision 19 X-9 X-9 October 2005

Nine Mile Point Unit 1 UFSAR charging its its 2.8 charging line allows The charging maintenance and prevents maintenance charging header. charge header. charge even if Scram Pilot Scram pilot Valves allows isolation prevents backflow It assures It if the supply Valves isolation of the accumulator accumulator for backflow from the accumulator assures that the accumulator accumulator to the accumulator will retain supply subsystem fails. fails. for retain

• During normal operation, each normal operation, each of the two parallel branches branches of of the RPS energizeenergize one of the two three-way three-way solenoid solenoid scram pilot pilot valves valves associated associated with each drive mechanism. mechanism. During During normal normal operation, operation, these pilot pilot valvesvalves are energized energized and supply supply instrument instrument air air to the operators operators of both the inlet scram valve and the outlet outlet scram valve, valve, holding holding both scram valves closed.

scram valves closed. During a full scram, scram, both of the RPS branches branches are de-energized de-energized and both pilot valves open, venting the scram valves' valves' operators operators and allowing allowing the scram valves valves to open. open. To protect againstagainst spurious scrams, scrams, the pilot pilot valves are interconnected interconnected so that that both pilotpilot valves valves must be de-energized de-energized to vent the scram valves' valves' operators. operators. On the other hand, hand, failure of either electric electric powerpower to both both solenoids solenoids or instrument air air will produce produce a scram.scram. The The pilot valves are selected selected based on simplicity of design, a minimum of moving parts, parts, fast opening time, and satisfactory opening time, statistical statistical operating history on similar operating units. similar units. protection, the instrument For added protection, instrument airair header to all all pilot the pilot valves has a pair of backup backup scramscram pilot valves. valves. Upon a scram scram signal these three-way three-way solenoidsolenoid valves close off the air air supply and vent the instrument air instrument air header. header. This will scram any drive should either either of its its scram scram pilot valves valves fail fail to vent. vent. A diverse reactor reactor trip trip system, alternate rod injection system, alternate injection (ARI), (ARI), has been added to provide an alternate alternate and diverse diverse method of of venting venting the instrument instrument air air header. header. An ARI initiation initiation signal, signal, high reactor reactor pressure, pressure, or low-low water level will actuate actuate the system. ARI system. 2.9 Scram Valves Valves The inlet inlet scram valve is is a globe valve which is is opened by the the force of an internal internal spring and closes when air air pressure pressure is is applied on top of the diaphragm operator. diaphragm operator. The opening force of of the spring is is approximately approximately 700 lb. lb. Each Each valve has a position indicator indicator switch which energizes energizes a light in in the control room as as soon as the valve starts valve starts to open. open. The scram valve valve is is selected based based on high operating operating force, force, fast opening opening time (approximately (approximately 0.1 sec) and satisfactory satisfactory operating operating history on similar units. units. UFSAR Revision 21 21 X-10 X-10 October 2009

Nine Mile Point Unit 1 UFSAR

• Both the inlet the inlet scram inlet and outlet scram valves are similar, except that inlet scram valve is scram valve is the outlet scram valve is is an angular angular pattern is a globe pattern.

pattern. pattern while the outlet The internal outlet internal spring preload is slightly greater valve to provide a faster opening that preload in in greater than the inlet scram opening characteristic. characteristic. scram

• UFSAR UFSAR Revision 21 21 X-10a X-lOa 2009 October 2009

Nine Mile Mile Point Unit 1 UFSAR THIS PAGE THIS PAGE INTENTIONALLY INTENTIONALLY BLANK BLANK 0 UFSAR Revision 21 21 X-10b X-lOb October 2009

Nine Mile Point Unit 1 UFSAR

• Demineralized Demineralized water is following:

normally provided directly is normally System Liquid Poison System directly to the Laboratories Laboratories and Sample Sinks Stator Winding Liquid Cooling Cooling System System System Air Compressor Filtration System Condensate Filtration Condensate Compressor 3.0 System Evaluation portable makeup system Operation of the portable Operation system is demand at routine is on demand infrequent infrequent intervals replenish demineralized intervals to replenish demineralized water in in storage tanks. tanks. With the system inoperable inoperable or when the portable portable demineralizer skid is demineralizer is not available, Station can continue available, the Station operation makeup water from the CSTs which have a combined operation with makeup combined capacity capacity of 400,000 gal. gal. Additional makeup water is Additional makeup is available demineralized makeup water from the demineralized water storage storage tank which has a 40,OOO-gal 40,000-gal capacity. capacity.

• conditions.

City water is lake water in Operators may take a supply of water As an option, Operators is an equivalent water from city processing, depending on the plant operating water for processing, conditions. better source for makeup equivalent or better terms of contaminants, city makeup than in terms contaminants, and delivery capacity capacity is is within or exceeds requirements for supply to the exceeds the requirements demineralized water system. demineralized system. 4.0 Tests and Inspections The demineralizer demineralizer effluent effluent isis controlled effluent controlled by effluent conductivity, periodic samples are taken of conductivity, conductivity, but periodic conductivity, TOC, silica, TOC, silica, chlorides, sulfates. . chlorides, and sulfates

• Revision 20 UFSAR Revision 20 X-29 X-29 October 2007 2007

Nine Mile Point unit Unit 1 UFSAR H. 1.0 SPENT FUEL STORAGE SPENT Design Bases This system is STORAGE POOL FILTERING AND COOLING SYSTEM Bases is designed to remove the spent fuel assemblies' SYSTEM assemblies' decay heat and the impurities impurities from the pool water water so as to to temperature and purity of the spent fuel pool water maintain the temperature water at acceptable acceptable levels, levels, assuring clarity clarity under all under all anticipated conditions. conditions. The pool water temperature temperature is is maintained maintained at or or below 140°F below 140OF during maximum anticipated anticipated storage storage conditions conditions and 110OF during reactor 110°F reactor power operation operation to maintain maintain the secondary containment containment licensing basis. basis. Normal refueling refueling conditions conditions are based based on refueling refueling the reactor every reactor every 24 months. months. During During certain instances, instances, it it may be necessary necessary to offload the entire core into into the spent fuel pool.pool. The maximum heatheat generation generation rate was was determined by assuming a full determined full core discharge discharge (532 bundles) after (532 bundles) after 24 months, months, with the maximum number previously discharged number of previously discharged fuel fuel bundles (3550) bundles being present (3550) being present in in the pool. pool. The greatest portion of the decay heat would would be produced produced by the bundles being discharged from the core, discharged core, rather rather than those bundlesbundles which have been stored been stored inin the spent fuel pool pool from previous previous discharges. discharges. The long-term long-term decay decay heat rate for GEl1 GEll fuel is is essentially essentially the same as for previous fuel designs.designs. Therefore, the decay heat Therefore, heat rate used as the basis for the spent spent fuel storage pool filteringfiltering

• and cooling cooling system design remains unchanged.

remains unchanged. Prior to Technical Specification Amendment No. Technical Specification 167, the spent No. 167, spent fuel pool was licensed for 2776 storage cells. cells. The north half half of the pool contained contained 1066 nonpoison nonpoison flux trap storage locations, and the south half provided locations, locations using provided 1710 locations Boraflex Boraflex as a neutron absorber. absorber. Currently, Currently, the spent fuel pool pool is licensed, is licensed, per Specification Amendment per Technical Specification Amendment No. 167, for No. 167, for 4086 spent fuel storage storage locations locations using the neutron absorber neutron absorber material Boral, material Boral, with 1840 storage storage locations locations in in the north half of of the pool and 2246 locations locations in in the south half. south half. The nonpoison nonpoison racks inin the north north half half of the pool were replaced replaced with new new poisoned poisoned racks after the 1999 1999 refuel outage. outage. The reracking reracking of of the south half of the pool has been partially completed. partially completed. Six six ofof existing Boraflex the eight existing Boraflex racks have have been replaced replaced with new Boral racks, increasing the capacity racks, increasing capacity from 1296 to 1656 1656 storage locations. locations. Two Boraflex Boraflex racks remain in in the south half,half, storage locations. providing 414 storage locations. The rerackrerack of the remaining. remaining two racks has been been deferred deferred until further capacity capacity increase increase is is warranted. warranted. UFSAR UFSAR Revision Revision 21 X-30 October 2009 October 2009

Nine MileMile Point Unit 1 UFSAR

• Unit 1 committed committed to the Nuclear Regulatory refueling and core offloading refueling it was determined it operable, Regulatory Commission (NRC) operations would not begin offloading operations determined that tpat the spent fuel pool cooling systems were operable, to ensure that the bulk pool temperature not be exceeded.

exceeded. (NRC) that begin until temperature limits would that until were would For a normal (full core offload or core core offload refueling, the core shuffle) refueling, offload offload time to the spent fuel pool and the RBCLC temperatures temperatures shall be verified verified to be consistent consistent with a bulk pool pool temperature not to exceed 140°F140OF with one cooling cooling train operating. train operating. For the case of an abnormal abnormal maximum heat load (such as a full full core offload offload shortly after after a normal refueling), refueling), this would would require verifying verifying that offload offload time and RBCLC temperatures temperatures were consistent with a pool temperature 0 F with both cooling consistent temperature <140<140°F both trains operating. operating. Based Based on past experience, experience, sufficient sufficient clarity clarity of the pool water water can be achieved achieved by a filter filter capable of removing capable removing particles particles asas small as 25 microns in in size. size. 2.0 System Design

• The system is is shown on Figure X-S.

pumps take suction pool water through two parallel and one heat exchanger. exchanger. X-8. Two full-capacity parallel loops consisting The water is the side opposite the surge tank skimmers. full-capacity (600 suction from the pool surge tanks and circulate water through (600 gpm) circulate the consisting of one filter is returned returned to the pool on skimmers. filter on The spent fuel pool cooling cooling (SFC)(SFC) system is is designed as seismic Category 1. 1. The SFC system bounding design conditions conditions are that, that, under full full core discharge discharge conditions conditions with with RBCLC coolant coolant water temperature temperature at its its maximum 95 0 F, and assuming maximum of 95°F, assuming the SFC heat exchangersexchangers are fouled to their design maximum and 5 percent of the tubes are plugged, plugged, a pool water temperature temperature of 140°F 140OF would be reached reached ifif a full core core offload offload began began 100S 1008 hr after reactor reactor shutdown, shutdown, and waswas completed 1129 hr after completed after reactor shutdown with one of the two reactor shutdown redundant cooling redundant cooling trains operating. operating. A more expedited expedited offload performed if offload may be performed plant if the plant conditions existexist to maintain maintain the pool water temperature temperature at oror below 140°F 140OF with one SFC train operating. operating.

• UFSAR Revision 21 X-31 October 2009 October 2009

Nine Mile Point Unit 1 UFSAR UFSAR mode. mode. Cooling control valves regulate Flow control use of a controller Cooling water is regulate the flow in controller that may be operated operated in supplied to the heat exchangers is supplied in each loop at 600 gpm by in the auto or manual exchangers from the RBCLCW 0 A manual RBCLCW by system at temperatures temperatures not exceedingexceeding 95°F. 95 F. A sample point is is incorporated incorporated to determine determine any tube leakage. leakage. Initial Initial filling filling and level maintenance maintenance in in the spent fuel pool and surge tanks was from the condensate condensate transfertransfer system. system. The total total volume of the surge surge tanks is is approximately approximately 2000 cu ft. ft. They They normally run at a level of approximately will normally approximately 1000 1000 cu ft. ft. The The difference in difference in surge surge tank volumevolume allows for the displacement displacement of of water from the spent fuel storage storage pool when a shipping cask (or (or any other object) is is placed in in the pool.pool. Makeup water is is provided by the condensate condensate transfer system. system. Normally, makeup is Normally, is directly to the spent fuel storage pool. pool. Makeup to the spent Makeup spent fuel storage storage pool is is automatically automatically initiated initiated when the surge tank volume decreases volume decreases to 800 cu ft ft and stops when when the volume reaches 1000 1000 cu ft.ft. If makeup to the spent fuel If the makeup fuel storage pool is is not sufficient sufficient to maintainmaintain surge tank volume, volume, makeup water can be provided directly to the surge tanks. provided directly tanks. The The

• condensate transfer system can provide a makeup condensate makeup rate of 75 gpm gpm or more to either the spent spent fuel storage storage pool or the surge tanks.

tanks. Makeup water can also be supplied directly directly to the spent spent fuel pool through fire fire hoses. water hoses. particles that enter the pool Any particles either sink to the bottom to pool either to be removed by a portable portable vacuum cleaner or float about in in the eventually enter the skimmers, pool and eventually skimmers, surge surge tanks and filtering filtering loop. loop. Provision Provision is transferring water to the is made for transferring liquid waste disposal disposal system system for processing processing if if the pool pool water water becomes contaminated. becomes highly contaminated. The precoat-type precoat-type filtersfilters use porous carbon elements. carbon elements. Precoat Pre coat material is material is powdered/crushed powdered/crushed resins. resins. One precoat pre coat mix tank and pump serves serves both filters. filters. The slurry is is circulated circulated through through the filter filter vessel and back to the tank until a uniform coating of a uniform coating of precoat pre coat material material covers all all elements. the elements. filter The filter is then is placed in in service service until differential differential pressure pressure signals need the need backwashing. The backwashing for backwashing. backwashing process process consists mainly of of first first valving off and draining draining the filter, filter, filling then filling the condensate from the condensate filter with condensate filter condensate transfer system. system. All All vents are closed during this filling filling and air air is is trapped in in the filter filter dome above the elements. elements. pressure in When the pressure in the filter filter

• approximately 80-100 psig, the drain valve is reaches approximately dome reaches is Revision 21 UFSAR Revision UFSAR 21 X-32 X-32 October 20092009

Nine Mile Point Unit 1 UFSAR UFSAR

• quickly openedopened and the filter the sludge tank the suspension the waste disposal system.

to the its filter impurities, washes into the fuel pool filter impurities, suspension of cake, together with trapped cake, filter sludge tank. of impurities and water is and purifying purifying the From From is pumped pumped Aside from its normal function of cooling .and spent fuel pool water, the system is is also used after reactor reactor refueling to drain the reactor internals storage pit pit and head cavity. cavity. Alternate lines allow transport of the water to either either the main condenser or to the waste disposal system for for processing. processing. In either case the water is In is filtered, filtered, demineralized demineralized and returned to the CSTs. CSTs. Each major piece of equipment is is designed to withstand seismic forces of 0.25g horizontally horizontally and 0.125g vertically. vertically. The ASME Boiler and Pressure Vessel Vessel Code, Code, Section VIII-1965, VIII-1965, is is specified for pump casings, casings, heat heat exchanger, filter exchanger, filter vessels, and the sludge tank, vessels, tank, as well as for for the fuel pool surge tanks. tanks. The fuel pool filters filters and the surge tanks are shielded shielded with concrete concrete to give a design radiation level of 5 mr/hr outside outside the shielded shielded area. area. 3.0 Design Design Evaluation Precoat-type filters Precoat-type filters capable of removing capable particles as small as 1 removing particles micron are provided, provided, although experience experience indicates indicates that that 25-micron particle 25-micron particle size filtration filtration should be sufficient sufficient to to maintain maintain pool clarity. clarity. Each pump filter filter heat exchanger exchanger loop is is adequately adequately sized to to handle handle the normal normal heat load of the spent spent fuel storage facility, storage facility, providing providing a complete complete standby loop. loop. The two loops are adequate adequate to handle the full full core discharge storage core discharge storage heat load. load. Various precautions are various precautions are taken to assure minimum minimum loss of water water from the system. system. All penetrations penetrations into into the pool are located located at at a minimum minimum height from the bottom such that there there will always be be at least least 1 ft ft of water water above above the fuel. fuel. Siphon breakers breakers are used where necessary where necessary and the pumps pumps areare sealed externally. sealed externally. For For flexibility, flexibility, either pump either pump may be used used with a given filter given filter heat heat exchanger exchanger loop. loop. Makeup Makeup water water to the the spent spent fuel storage storage pool is is provided provided by the condensate condensate transfertransfer system. system. The condensate transfer The condensate transfer system system can can be supplied emergency power from the diesel supplied emergency diesel generators, generators, ensuring the the supply supply of of makeup makeup water in in the the event of of loss loss of both normal normal and and reserve reserve ac ac power. power. UFSAR UFSAR Revision Revision 21 21 X-33 X-33 October 2009 October 2009

Nine Mile Point Unit 1 UFSAR Makeup Makeup water is through the fire The Operator water Operator is is also available level, water water level,* available to the spent fire protection The fuel pool cooling is controlled spent fuel storage controlled from a local is provided with indications storage pool protection system by the use of a water hose. cooling system is local panel. indications of system flow, pool temperature (on both sides of the heat water temperature heat pool hose. panel. pool exchangers), exchangers), sludge tank level, level, and valve positions. positions. Alarms provided on the annunciator Alarms are provided annunciator and the computercomputer for for low-pressure flow and temperature high- and low-pressure temperature where critical. where critical. The spent fuel pool pool system may be secured secured for maintenance maintenance for for limited periods periods as long as: 1) available for the

1) the time available maintenance maintenance activity activity has been predicted predicted by an approved approved calculation, which ensures calculation, temperature will remain ensures the pool temperature remain below 110 0 Fj 110°F; 2) 2) the pool temperature temperature is closely monitored during is closely monitored the maintenance maintenance activity to ensure the temperature temperature does not not exceed exceed 110°F 110OF (the maintenance maintenance time available available may be increased based on this empirical empirical data);

data) j and 3) condensate transfer

3) the condensate transfer system isis available available for makeup.

makeup. 4.0 Tests and Inspections Inspections All equipment equipment in in this system will be normally operated, operated, as spent spent

• fuel and other components are stored stored in in the pool.

pool. However, However, if if equipment equipment such as the spare spare pump filter filter heat exchanger exchanger loop should should stand idle for some time, time, itit will be exercised exercised to assure that it it operates operates properly. properly. UFSAR UFSAR Revision 21 21 X-34 X-34 October October 2009 2009

Nine Mile Point Unit 1 UFSAR with water to aa height of 48 ft ft 9 in in above the top of active fuel fuel

• (TAF).

{TAF}. storage This water is can be provided systems. is filtered filtered and cooled storage pool filtering and cooling system. provided by the reactor systems. These systems are described. X-B, X-B, respectively. respectively. cooled by the spent fuel system. shutdown cooling and cleanup reactor shutdown described in Sections X-H, in sections fuel Additional cooling X-H, X-A and A second second stainless steel-lined steel-lined transfer transfer canal connects the canal connects reactor reactor head cavity to the reactor reactor internals storage pit. During internals storage refueling, refueling, this canal is is filled with water to a depth of 19 ft 6 in. Steel-sheathed concrete in. steel-sheathed concrete plugs, plugs, 5 ft ft 5 in in thick, shield the refuel refuel floor from the reactor reactor head cavity during power operation. operation. Both Both canals are filled with similar plugs during power power operations. operations. The concrete concrete plugs provide approximately approximately 4 ft ft ofof shielding between shielding between the equipment equipment storage pit and the reactor reactor head cavity, and approximately cavity, approximately 4.5 ft ft of shielding between between the spent spent fuel pool and the reactor reactor head cavity. cavity. The reactor reactor internals internals storage storage pit is is a reinforced reinforced concrete concrete pit, completely completely lined with stainless stainless steel. The pit is flooded with is water to a depth of 24 ft ft 0 inin during refueling. refueling. This water water is is circulated circulated through the spent fuel storage storage pool filtering filtering and cooling system. cooling system. The pit is is large enough to accommodate accommodate the reactor reactor steam separator separator and the reactor steam dryer assemblies assemblies side by side. side. The refueling refueling platform is equipped with a 1200-lb is equipped 1200-lb capacity capacity main main hoist, 1,000-lb capacity hoist, and two 1,000-lb capacity auxiliary hoists. hoists. Each of these hoists can be positioned over over any pointpoint in in the reactor reactor head cavity or the spent fuel storage pool. pool. Protective interlocks, Protective interlocks, discussed in in the Refueling Accident, Refueling Accident, Section XV, Section XV, are installed in in the power supplies supplies to the refueling platform to prevent prevent inadvertent reactivity additions inadvertent reactivity additions to the core during refueling. refueling. operating floor is The operating is serviced serviced by the reactor crane, reactor building crane, which is is equipped equipped with a 125-ton 125-ton main hoist and a 25-ton auxiliary hoist. auxiliary hoist. TheseThese hoists can reach reach all areas areas of the operating operating floor. floor. The 125-ton125-ton main hoist is is also equipped equipped with a redundant redundant hoisting system, system, which will prevent prevent the dropping dropping of of heavy loads in in the event that a cable cable or other criticalcritical part of of the main hoist equipment equipment should fail. fail. 1/2-ton capacity Three 1j2-ton portable jib portable cranes are provided jib cranes operations in provided for operations in the fresh fuel fuel storage vault and the spent fuel storage pool. pool. Mountings (five Mountings {five in all) in all} for these cranescranes are provided around the periphery periphery of the pool. pool. A variety variety of tools for remote handling of fuel and reactor reactor internals and flow channel internals channel exchange are provided. provided. Fuel sipping may be required required to identify fuel assemblies that assemblies that

• contain failed fuel rods. rods. Additionally, Additionally, the fuel assemblies assemblies identified identified to contain contain failed rods may be inspected inspected and repaired UFSAR Revision 17 17 X-41 X-41 October 2001

Nine Mile Point Unit 1 UFSAR in the fuel prep machine. in machine. In order to perform In perform this work, it it is is ~ required required to store fuel sipping equipment equipment in empty control in empty control blade rack cells until such time that the sipping operation operation is is complete. complete. 2.1.1 2.1.1 Cask Drop Protection Protection System System The cask drop protection protection system has been been designed designed to 1) 1) prevent prevent loss of spent fuel pool integrity integrity as a result of certain certain types of cask drop accidents accidents which may occur occur over the spent fuel pool, pool, and 2)

2) minimize minimize damage damage to spent fuel and other other components components stored in the pool.

in Specifically, the system has been pool. Specifically, been designed designed to meet meet the following functional functional requirements: requirements: 1.

1. Prevent the cask from tipping into the spent fuel Prevent fuel pool.

pool. 2.

2. Guide the falling cask into the hydraulichydraulic dashpot dashpot section of the structure.

section structure. 3.

3. Control the attitude of the cask as it Control it falls falls through the guide structure structure and dashpot assembly.

assembly. 4.

4. Decelerate the cask Decelerate cask to a low impact impact velocity.

velocity. ~ 5.

5. Absorb the energy of the cask upon impact. impact.

6.

6. Limit loads transmitted transmitted to the floor of the spent spent fuel fuel pool to acceptable values.

acceptable values. This system consists consists of a circular circular basebase plate attached to the plate attached bottom of the shipping cask and a combinationcombination guide structure--dashpot structure--dashpot assembly assembly which is is permanently permanently installed installed in in the spent fuel pool (Figure X-ll). structural design X-II). The structural design of the cask drop protection protection system is is based worst-case based on the worst-case hydraulic, hydraulic, vertical vertical and lateral lateral loadings associated associated with a wide postulated cask range of postulated cask drop accidents. accidents. (4,5,6) (4','6) This design protection against a wide range of different provides protection different size and weight weight shipping casks. shipping casks. A summary summary description description of the basis for conducting conducting safe heavy heavy load movements is provided in Section movements is provided in Section X-J.2.3. X-J.2.3. Sufficient Sufficient protection from the risk associated protection potential heavy load associated with potential drops is is also provided provided by satisfying satisfying the guidelines guidelines of of NUREG-0612, NUREG-0612, Sections 5.1.1 5.1.1 and 5.3(7r8) 5.3(7,8).

                                                                                        ~

UFSAR Revision Revision 21 21 X-42 X-42 October October 20092009

Nine Mile Point Point Unit 1 UFSAR ~ 2.2 Facility Operation of the Facility Operation Fresh fuel is is brought into the reactor building through reactor building through the reactor building track bay extension shown on Figure reactor Figure 111-4, and hoisted to the operating operating floor through through the equipment equipment hatch utilizing utilizing the reactor building crane. reactor building crane. (See Figures 111-5 to to 111-9.) 111-9.) The fresh fuel is is removed from its its shipping containers, containers, inspected, flow channels inspected, attached, and stored in channels attached, fuel in the fresh fuel storage vault. storage vault. Normally prior to refueling, refueling, the fresh fuel is transferred to is transferred to the spent fuel storagestorage pool using the 25-ton auxiliary auxiliary overhead hoist. hoist. In preparation In refueling, the concrete preparation for refueling, concrete shield plugs in in the reactor reactor head cavity and the transfer canals removed by the canals are removed reactor building crane. reactor crane. The drywell head and reactor reactor vessel vessel head are removed removed using the same crane. crane. stearn dryer The steam dryer and the steam stearn separator separator assemblies assemblies are transferred transferred to the reactor internals storage pit. reactor internals storage pit. Water levels levels are controlled such that the steam are controlled separator is stearn separator is transferred ~ submerged. submerged. disassembly process, During the disassembly process, demineralized demineralized condensate condensate isis pumped into the reactorreactor until the head cavity and the reactor reactor internals storage pit internals pit are flooded to the normal normal level of the storage pool. spent fuel storage pool. storage pool gates are The spent fuel storage removed after the water water level level has reached the normal level of of spent fuel storage the spent storage pool. pool. Spent fuel is is removed removed from the reactor reactor using a grapple grapple attached to the refueling platform and placed placed inin racks in fuel in the spent fuel storage pool. storage pool. The same equipment equipment is is used to transfer transfer the fuelfuel from the spent storage pool to the reactor. spent fuel storage reactor. completion of reactor At the completion refueling, the moisture reactor refueling, separator, moisture separator, stearn steam dryer and reactorreactor head are put back into place following the proper proper maintenance procedures. maintenance procedures. The drywell head and shield blocks are then restored. concrete shield concrete restored. After refueling, refueling, the spent fuel bundles spent stored in bundles are stored fuel in spent fuel storage pool racks. storage racks. They will remain there until NRC resolution of disposal disposal problems is is finalized.

• UFSAR Revision 21 UF.SAR 21 X-43 October 2009

Nine Mile Nine Point Unit Mile Point Unit 11 UFSAR UFSAR 2.3 2.3 2.3.1 2.3.1 Control of Control NUREG-0612 provides NUREG-0612 Loads Program Heavy Loads of Heavy Introduction/Licensing Background Introduction/Licensing Program

Background

guidelines for regulatory guidelines provides regulatory for the control of the control of heavy loads to assure the safe handling heavy loads to assure the safe handlihg of heavy loads in areas of heavy loads in areas where aa load where load drop could impact drop could fuel, fuel spent fuel, stored spent impact stored fuel in in thethe reactor core, reactor core, or equipment that or equipment that maymay be required to be required achieve safe to achieve safe shutdown or shutdown permit continued decay or permit continued decay heat removal. heat removal. In a letter In a letter December 22, dated December dated 1980(11) (later 22, 1980(11) identified as (later identified as GL 80-113), as GL 80-113), as supplemented by GL 81-07(12) and GL 83-42(2o), supplemented by GL 81-07(12) and GL 83-42(20), the NRC requested the NRC requested that licensees that describe how licensees describe guidelines were these guidelines how these satisfied at were satisfied at their facility. their facility. This request was divided into This request was divided into two phases (Phase two phases (Phase Phase II). and Phase I and II). The Mohawk Power Niagara Mohawk The Niagara Corporation (NMPC) Power Corporation (NMPC) response to response to thethe Phase portion of Phase II portion of the request for the request for Unit Unit 1, I, addressing the guidelines of Section 5.1.1 of NUREG-0612, was addressing the guidelines of Section 5.1.1 of NUREG-0612, was initially provided initially provided in letters dated in letters dated May May 22, July 28, 22, July 28, and and September 22, September 22, 1981(13-5) information was Supplemental information 1981 (13-15). Supplemental was subsequently provided in NMPC letters subsequently provided in NMPC letters dated August 1, dated August I, 1982; 1982i September 30, September November 15, 30, November IS, and December 15, and December 1983i July IS, 1983; July 26, 26, 1984; 1984i January 18, and January and August 5., 18, August November 25, and November 5, and 25, 1985(16-19,21,22,24,2S) 1985(16-19,21,22,24,25). By letter dated By letter evaluation which evaluation Sections 5.1.1 Sections Phase I 85-11(S), the 85-11(), of 5.1.1 and the March 5, dated March and 5.3, NMPC 5, 1985( concluded response 1985(7), that

                                                  ), the for the Phase I of the NMPC response for Unit 1 was acceptable.

the NRC documented their NRC documented Unit 1 issued their NRC issued the NRC guidelines which concluded that the guidelines in NUREG-0612, 5.3, had had been satisfied for been satisfied was determination that their determination in for Unit their safety Unit 1, acceptable. a safety NUREG-0612, 1, andand that In GL In detailed that a detailed that GL Phase II Phase review of II review loads was heavy loads of heavy was not necessary and not necessary and thatthat Phase Phase II was considered II was considered completed. completed. letter dated By letter By dated MayMay 13, 13, 1996(27), provided the NMPC provided 1996 (27), NMPC the required required response to Bulletin response 96 - 02 (26) for Bulletin 96-02121) for UnitUnit 1.1. TheThe response response reiterated that reiterated movement of the movement that the of heavy loads over heavy loads critical areas over critical areas of the refuel floor and safety-related equipment is performed in of the refuel floor and safety-related equipment is performed in accordance with accordance controlled site with controlled site procedures developed in procedures developed in accordance with accordance with NUREG-0612. Additionally, the NUREG-0612. Additionally, the response response reaffirmed that the reactor building 125-ton crane is reaffirmed that the reactor building 125-ton crane is single-failure-proof (i.e., single-failure-proof (i.e., aa dualdual load load path, redundant hoisting path, redundant hoisting 8 system). The NRC's system). NRC I S April 23, 23, 1998, accepted the letter(2S)) accepted 1998, letter(2 the NMPCNMPC response and response and indicated completion of tasks indicated completion of tasks associated with associated with Bulletin 96-02. Bulletin 96-02. On July July 28, 2008(29), the 28, 2008)29), Nuclear Energy the Nuclear Institute (NEI) Energy Institute (NEI) transmitted NEI 08-05, Revision 0, Industry Initiative on transmitted NEI 08-05, Revision 0, Industry Initiative on Control of Control Loads to the Heavy Loads of Heavy the NRC. NRC. document was This document This issued to was issued to provide an provide an industry agreed-upon approach industry agreed-upon approach to providing additional to providing additional

• assurance of assurance compliance to of compliance to existing regulatory guidelines existing regulatory guidelines UFSAR UFSAR Revision Revision 21 21 X-43a X-43a October 2009 October 2009

Nine Mile Point Unit 1 UFSAR

• regarding control regarding safe heavy guidelines control of heavy loads at nuclear document includes NEI document contained in guidelines contained guidance associated includes guidance nuclear power plants.

associated with updating description of the basis for conducting to reflect a summary description heavy load movements. movements. The NRC safety evaluation in NEI 08-05, plants. updating UFSARs evaluation of the transmitted to NEI by letter 08-05, transmitted This This UFSARs letter dated September September 5,5, 2008(30), determined determined that the guidelines guidelines may be used used by licensees establish a revised licensing licensees to establish licensing basis for for handling of reactor handling reactor vessel vessel heads and other other heavy heavy loads,loads, subject subject to the clarifications clarifications and conditions conditions noted in in the NRC's safety evaluation. evaluation. 2.3.2 Safety Basis Basis Heavy load activities load handling activities pose a safety risk in in the areas of nuclear nuclear power plants where where load drops could could impact irradiated fuel or equipment necessary for safe shutdown. equipment necessary shutdown. Implementing Implementing the guidelines of NUREG-0612, guidelines NUREG-0612, Section 5.1.1, reduces Section 5.1.1, reduces the potentialpotential for heavy heavy load load drops and provides defense-in-depth provides a measure of defense-in-depth against such an occurrence. occurrence. associated with load handling The risk associated handling failures is is acceptably acceptably low based based on meeting the Phase I requirementsrequirements of NUREG-0612,NUREG-0612,

• Section Section 5.1.1, 5.1.1, and the use of the reactor building 125-ton single-failure-proof single-failure-proof crane and spent fuel casks.

casks. single-failure-proof single-failure-proof crane and has a redundant crane for lifting The 125-ton crane redundant hoisting as hoisting system lifting 125-ton reactor defined in the reactor vessel building crane reactor building NUREG-0612, defined in NUREG-0612, Appendix C, system which is is independently 125-ton vessel head crane is independently 2 3 ).. is a Appendix C, capable supporting the crane's rated load( capable of supporting load (23) 2.3.3 Scope Scope of Heavy Heavy Load Handling Systems Systems In NUREG-0612, In NUREG-0612, the scope of cranes cranes includes: includes:

       "Overhead handling "Overhead   handling systems that are used to handle heavy loads inin the area of the reactor reactor vessel or spent fuel in               in pool.

the spent fuel pool. Additionally, loads may be handled in Additionally, in other other areas where their accidentalaccidental drop may damage safe safe shutdown systems shutdown systems..." Based Based on the NMPC Phase Phase I responses in References 13 in References 13 through 19, 19, 21, 21, 22, 24, and 25, 22, 24, 25, the reactor reactor building 125-ton 125-ton crane crane isis within the scope of Section Section 5.1.1 of NUREG-0612. NUREG-0612. Heavy load movements movements in in areas of safe shutdown equipment safe shutdown equipment that are handled by other are handled other load-handling load-handling systems systems are also performed performed in in accordance accordance with the requirements of NUREG-0612 requirements NUREG-0612 as defined in controlled site in controlled site procedures. procedures. These other systems include, but are not limited systems include, to, the reactor reactor building building 25-ton auxiliary crane, 25-ton auxiliary crane, the turbine UFSAR Revision Revision 2121 X-43b X-43b October 20092009

Nine Mile Point Unit 1 UFSAR building 150-ton crane. crane. 2.3.4 ISO-ton crane, Control crane, and the screen and pump house Control of Heavy Loads Program Program house 25-ton Control of Heavy The Control Heavy Loads Loads Program Program consists of the following: following: 1.

1. NMPNS commitments commitments in in response to NUREG-0612, NUREG-0612, Phase Phase I elements, as described elements, described inin References References 13 through 19, 19, 21, 22, 21, 22, 24, 24, and 25.

25. 2.

2. For reactor reactor pressurepressure vessel head (RPVH) (RPVH) and spent fuel fuel lifts, cask lifts, single-failure-proof reactor building the single-failure-proof 125-ton crane 125-ton described in crane described in Section X-J.2.3.4.2 X-J.2.3~4.2 is is used.

used. 2.3.4.1 NMPNS Commitments Commitments in in Response to NUREG-0612, NUREG-0612, Phase I Elements Elements NMPNS has committed committed to controlling controlling the movement of heavy loads loads in accordance in accordance with the elements of Section 5.1.1 seven elements 5.1.1 of of NUREG-0612, NUREG-0612, as defined defined below: below: 1. 1. 2. 2. Safe load load paths for movement defined in in controlled Controlled plant Controlled movement of heavy loads are plant procedures implemented that control implemented procedures 1(l4) controlled plant procedures( procedures are developed control movement 4 ). developed and movement of heavy loads loads(14) (l4) . 3.

3. Crane operators are trained and qualified Crane operators qualified in in (14) accordance with controlled accordance controlled plant plant procedures procedures (l4) .

4.

4. Special lifting Special lifting devices devices follow the guidelinesguidelines of ANSI ANSI N14.6-197S(lS) .

N14.6-1978("). 5.

5. Lifting devices devices not specifically specifically designed follow the guidelines of ANSI B30. B30.9-197lt 1 5 .

guidelines 9-1971 (15) . 6.

6. The reactor reactor building 125-ton crane building 125-ton crane is is inspected, inspected, maintained consistent tested and maintained consistent with ANSI ANSI B30.2-1976(14)

B30.2-1976(' 4 ). 7.

7. The reactor reactor buildingbuilding 125-ton cranecrane is is designed to to CMAA-70 and meets the applicable criteria applicable criteria and guidelines of ANSI B30.2-1976('

guidelines 4 '1 5 ) B30.2-1976(l4,1S). Revision 21 UFSAR Revision 21 X-43c X-43c October 2009 2009

Nine Mile Point Unit 1 UFSAR ~ 2.3.4.2 Pressure Vessel Head and Spent Fuel Cask Lifts Reactor Pressure reactor building 125-ton The reactor crane is 125-ton crane is single-failure-proof, single-failure-proof, has has a redundant redundant load path, and is is designed designed to CMAA-70. CMAA-70. The The attributes were defined following attributes defined in design of the design in the of the crane(23) the crane (23) : 1.

1. Allowable stress limits are defined Allowable defined and conservative enough permanent deformation prevent permanent enough to prevent individual deformation of individual structural members structural members when exposed exposed to maximum load lifts.

lifts. 2.

2. The crane isis capable stopping and holding the load capable of stopping during a design basis earthquake.

earthquake.

33. . Automatic controls and limiting Automatic devices are designed limiting devices designed so that they fail-safe fail-safe and do not prevent prevent the crane stopping and holding the load safely.

from stopping safely. 4.

4. The design of the wire rope reeving system includes includes dual wire ropes.

ropes. 5.

5. switches are included to limit such Limit switches such items as as overspeed, overload overspeed, overtravel and cause the overload and overtravel

~ hoisting action action to stop when limits are exceeded. exceeded. 6.

6. The reeving system is designed against the destructive is designed effects effects of IItwo-blocking.1I "two-blocking."

7.

7. Safety devices such as limit switches are provided provided to to likelihood of a malfunction.

reduce the likelihood malfunction. 2.3.5 Safety Safety Evaluation Controls implemented implemented by NUREG-0612, NUREG-0612, Phase I elements,elements, together together single-failure-proof crane with the use of a single-failure-proof crane for RPVH and spent spent fuel cask lifts,lifts, make make the risk of aa load drop extremely extremely unlikely and acceptably acceptably low. low. The risk associated associated with the movement of of heavy loads is evaluated loads is evaluated and controlled controlled by station procedures. station procedures. 3.0 Design Evaluation Design Evaluation The spacing of fuel bundles in in the fresh fuel storage vault vault maintains maintains keff <0.95 even if even if flooded with water. water. The vault vault floor drain prevents prevents flooding. The spacing bundles in spacing of fuel bundles in the spent fuel storage pool maintains maintains keff <0.95. criticality

                                                              <0.95. A criticality monitor monitor in                          storage vault provides warning in in the fresh fuel storage                                            in the

~ unlikely event of a criticality incident. criticality incident. UFSAR Revision 21 21 X-43d X-43d October 2009

Nine Mile Point Unit 1 UFSAR Protective interlocks Protective interlocks prevent handling handling of fuel over over the reactor reactor ~ when a control rod isis withdrawn. withdrawn. Another Another set of interlocks interlocks prevents control prevents control rod withdrawal withdrawal when fuel isis being being handled overover the reactor. reactor. Limit switches switches on the refueling platform hoists refueling platform hoists interrupt power to the hoists when the TAF is is 8 ft ft below the surface surface of the water. water. Brakes all Brakes on all equipment lock upon loss of of power. power. Spent fuel will not be inadvertently handled with an inadvertently handled an inadequate depth of water shielding. inadequate depth shielding.

                                                                           ~
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UFSAR Revision Revision 21 21 X-43e X-43e October 2009 2009

Nine Mile Nine Mile Point Unit 1 UFSAR THIS PAGE INTENTIONALLY INTENTIONALLY BLANK

• UFSAR Revision 21 21 X-43f X-43f October 2009 2009

Nine Mile Point Unit 1 UFSAR The above above interlocks a significant interlocks can be bypassed significant portion bypassed to permit the unloading of portion of the reactor reactor core (full spiral offload) for such purposes as removal of temporary control curtains, requirements, requirements, examination maintenance, inservice curtains, CRD maintenance, (full inservice inspection examination of the core support Specification 3.5.3). support plate, core offload, core offload, plate, etc. etc. (lSI) inspection (ISI) of (Technical Specification 3.5.3) . Fuel stored in in the spent spent fuel storage storage pool is is covered covered by a minimum of 24 ft minimum ft of water. water. Irradiated Irradiated fuel being being moved is is at at all all times covered covered by a minimum depth depth of 8 ft 8 ft of water over TAF, TAF, except that the fuel preparation preparation machine is provided with is provided mechanical mechanical stops to ensure ensure that active active fuel remains remains under 7 ft ft water. of water. Spent fuel pool water level level is automatically is automatically controlled controlled to ensure ensure that during normalnormal operation, spent fuel fuel will be coveredcovered by a sufficient sufficient depth depth of water water to permit permit unrestricted unrestricted access access to the operating operating floor. The spent fuel storage pool cannot cannot be completely drained. completely drained. If If draining should be initiated initiated due to Operator Operator error, error, level alarms alarms will notify notify operating personnel and makeup water will be operating personnel be supplied supplied automatically. automatically. If no action were taken, If taken, the fuel would would still still be coveredcovered by approximately approximately 1 ft ft of water after the pool pool had drained drained down to the lowest lowest penetration. penetration. All reactor reactor servicing operations are carried servicing operations carried out within within the secondary containment, secondary containment, which is is described described in Section VI-C. in Section VI-C. A bypass aroundaround the refueling platform radiation radiation monitor will will allow the monitor monitor to be connected connected into the RPS during refueling operations operations or when recently recently irradiated irradiated fuel or aa fuel-loaded shipping shipping cask cask is is being handled. handled. This monitor provides a fast fast automatic isolation automatic isolation of the reactor reactor building ventilation system ventilation system initiation and initiation of the reactor building emergency reactor building emergency ventilation system. system. 4.0 Tests and Inspections During During testing prior to initial initial reactor fueling, fueling, the spent spent fuel fuel storage pool, storage pool, reactor reactor head cavity, and reactor reactor internals storage pit pit filled were filled with water water and checked for leakage. leakage. Dummy fuel fuel assemblies assemblies were run through a complete complete cycle from the fresh fuel fuel storage storage vault vault to the spent fuel storage pool. storage pool. During normal operation, operation, telltales telltales are examined examined for evidence of of potential leakage potential leakage from the spent fuel pool. pool. Prior to fuel fuel

handling, handling, all all hoists, hoists, cranes and tools are are inspected and tested to assure assure safe operation.

operation. UFSAR Revision 21 21 X-44 X-44 October 2009 October 2009

Nine Mile Point Unit 11 UFSAR 4.0 and Inspections Tests and preoperational test is The preoperational is performed to confirm the operability system components and piping configurations. of the installed system configurations. The startup test is is performed to verify the function of all all components and the capability system components capability of the system to control the process. process.

• UFSAR UFSAR Revision 17 Revision 17 X-55 X-55 2001 October 2001

Nine Mile Point Unit 11 UFSAR N. N. 1. 1. REFERENCES REFERENCES Application for Amendment Application 50-220, 50-220, from D. Amendment to Operating D. P. po. Dise (NMPC) License, Docket No. Operating License, (NMPC), , dated March March 21, 21, 1978. 1978. No.

• 2.

2. Letter from C. C. V. (NMPC) to Domenic B.

V. Mangan (NMPC) B. Vassallo (NRC), dated June 24, (NRC), 1983, Spent Fuel 24, 1983, Fuel Pool Modification. Modification. 3.

3. Letter from C. C. V. (NMPC) to Domenic B.

V. Mangan (NMPC) B. Vassallo Vassallo (NRC), (NRC), dated October October 5,5, 1983, 1983, Spent Fuel Storage Storage Capacity Capacity Expansion. Expansion. 4.

4. Letter from D. D. P.

P. Dise (NMPC) (NMPC) to V. V. Stello (NRC), July 17, Stello (NRC), 17, 1978. . . (Attached information: 1978 information: Movement of Heavy Loads Near Near the Spent Spent Fuel Storage Storage Pool or Reactor Core) Core). . 5.

5. Letter from T. T. J.

J. Brosnan (NMPC)(NMPC) to J.J. F. F. O'Leary O'Leary (AEC) (AEC), , September 29, September 1972. 29, 1972. (Attached report: NMP-l NMP-l Cask Drop Protection System) Protection System). . 6.

6. Letter Letter from P. P. D.

D. Raymond Raymond to D. D. L. L. Zieman (AEC), (AEC) , May 31, 31, 1973. 1973. (Attached report: NMP-l NMP-l Cask Drop Protection 7. 7. 8. System). . System) Letter Letter from D. March 5, 5, ... 1985, D. B. B. Vassallo (NRC) 1985, Control Generic Letter Control of Heavy (NRC) to B. Heavy Loads B. G. Hooten (NMPC), , Hooten (NMPC) (Phase 1). 1).

8. NRC Generic Letter 85-11, 85-11, June 28, 28, 1985, Completion of 1985, Completion of Phase Phase II II of "Control of Heavy Loads at Nuclear Nuclear Power Power Plants,"

Plants," NUREG-0612. NUREG-0612. 9.

9. DRFC 11-61, 11-61, "Calculation "Calculation of Free Volume Available for Scram Volume Available Scram with Leakage for SDV Long-Term," Long-Term," dateddated June June I, 1981.

1, 1981.

10. Technical Specification Technical Specification Amendment Amendment No. No. 167 for NMPINMP1 to to Reflect Reflect a Planned ModificationModification to Increase Increase the Storage Capacity of the Spent Fuel Pool, Capacity Pool, dated dated June 17, 1999.

17, 1999.

11. Letter Letter from D. D. G. Eisenhut (NRC) (NRC) to All Licensees, Licensees, December 22, December 22, 1980, 1980, Control of Heavy Heavy Loads (Generic Letter Letter 80-113). .

80-113) 12.

12. Letter Letter from D. G. Eisenhut (NRC)

D. G. (NRC) to All Licensees, Licensees, February February 3, 3, 1983, 1983, Control of Heavy Loads (Generic Letter Letter 81-07). 81-07) . UFSAR Revision 21 21 X-56 X-56 October 2009

Nine Mile Point Unit 11 UFSAR

• 13.

14. Letter from D. Letter May 22, 22, 1981. 1981. Letter from D. Letter July 28, July 1981. 28, 1981. D. P. P. Dise (NMPC) D. P. (NMPC) to P. Dise (NMPC) (NMPC) to to D. D. G. Eisenhut (NRC), to D. D. G.G Eisenhut (NRC), (NRC), (NRC),

15. Letter from D.

Letter D. P.P. Dise (NMPC) (NMPC) to to D. G. Eisenhut G Eisenhut (NRC), (NRC), September September 22, 22, 1981.

16. Letter from T.

Letter E. Lempges T. E. (NMPC) (NMPC) to to D D. G. G. Eisenhut (NRC), (NRC), August 1,1, 1982. 1982.

17. Letter from T.

Letter E. Lempges T. E. (NMPC) to D (NMPC) to D. G. G. Eisenhut (NRC), (NRC), September 30, September 30, 1983. 1983.

18. Letter Letter from T. T. E.

E. Lempges (NMPC) to D (NMPC) to (NRC), D. G. Eisenhut (NRC), November 15, November 15, 1983. 1983.

19. Letter Letter from T. T. E.

E. Lempges (NMPC) to D (NMPC) to D. G. G. Eisenhut (NRC), (NRC), December 15, December 15, 1983. 1983.

20. Letter from D.

Letter D. G.G. Eisenhut (NRC)(NRC) to All Licensees, Licensees, December 19, 1983, Clarification December 19, 1983, Clarification to Generic Generic Letter 81-07 Regarding Response to NUREG-0612, Regarding Response NUREG-0612, "Control "Control of Heavy Loads Loads at Nuclear Power Plants" (Generic Letter 83-42) 83-42). .

21. Letter from C.

Letter V. Mangan C. V. Mangan (NMPC) (NMPC) to D. B.B. Vassallo (NRC), (NRC), July 26, 1984. 26, 1984.

22. Letter from T. T. E.E. Lempges Lempges (NMPC)

(NMPC) to D. B. B. Vassallo (NRC), (NRC), January 18, January 18, 1985. 1985.

23. Letter from P. P. D.

D. Raymond (NMPC) to D. Raymond (NMPC) D. J. J. Skovholt Skovholt (NRC), (NRC), July 26, 1973. 26, 1973.

24. Letter from T. T. E.E. Lempges (NMPC) to D. B.

(NMPC) B. Vassallo Vassallo (NRC), (NRC), August August 5,5, 1985. 1985.

25. Letter from C. C. V.V. Mangan (NMPC)

(NMPC) J. A. to J. Zwolinski (NRC), A. Zwolinski (NRC), November November 25,25, 1985. 1985.

26. NRC Bulletin 96-02,96-02, Movement Movement of Heavy Heavy Loads Over Spent Spent Fuel, Fuel, Over Fuel in in the Reactor Reactor CoreCore or Over Safety Safety Related Related Equipment, April 11, 1996.

Equipment, 1996.

• UFSAR Revision 21 21 X-57 X-57 October 2009

Nine Mile Mile Point Unit Unit 1 UFSAR 27.

27. Letter Letter from C. C. D. (NMPC) to u.

D. Terry (NMPC) U. S.S. Nuclear Regulatory Nuclear ~ Commission, Commission, May 13, 13, 1996, 1996, NRC Bulletin Bulletin 96-02, 96-02, Movement of of . Heavy Heavy Loads Over Spent Fuel, Fuel, Over Fuel in the Reactor Core in or Over Safety Safety Related Related Equipment. Equipment. 28.

28. Letter from D.

Letter D. S. S. Hood (NRC) (NRC) to J. J. H. Mueller Mueller (NMPC), (NMPC) , April 23, 23, 1998, 1998, Completion Completion of Licensing Licensing ActionAction for NRC Bulletin 96-02 for Nine Mile Point Nuclear Bulletin Nuclear Station, Station, UnitUnit Numbers Numbers 1 and 2. 2. 29.

29. Letter from A. A. R. Pietrangelo (NEI)

R. Pietrangelo (NEI) to E. E. J. J. Leeds (NRC), Leeds (NRC), July 28, 2008, NEI 08-05, 28, 2008, 08-05, Revision 0, 0, Industry Initiative Initiative on Control Control of Heavy Loads. Loads. 30.

30. Letter from W. W. H. Ruland Ruland (NRC)

(NRC) to T. T. C. C. Houghton (NEI) (NEI), , September September 5, 2008, Industry 5, 2008, Industry Initiative Initiative on Control Control of Heavy Heavy Loads. Loads.

                                                                                     ~
                                                                                     ~

UFSAR Revision 2121 X-58 X-58 October 2009

CONTROL ROD DRIVE HYDRAULIC SYSTEM TO REFERENCE LEC BACKFILL

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• L REACTOR BUILDING FLOOR DRAIN *

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                                                                                                                                                                                                                                                                                                                                 '.-O<O.S.            FIGURE X-3 UFSAR ReViSion 21 October- 200')

Nine Mile Point Unit 1 UFSAR

• Section Title Title HAZARDS ANALYSIS FIRE HAZARDS ANALYSIS TABLE OF CONTENTS CONTENTS Page 1.0

1.0 INTRODUCTION

10A-1 10A-1 1.1 PURPOSE PURPOSE 10A-1 lOA-1 1.1.1 NRC Guidance Guidance 1OA-1 10A-1 1.1.22 1.1. Scope Scope 1OA-1 10A-l 1.2 1.2 DISCUSSION 1OA-1 10A-1 1.2.1

1. 2.1 Format Format 1OA-1 10A-l 1.2.1.1 1.2.1.1 Safe Shutdown Analysis Analysis 10A-1 lOA-1 1.2.1.2 1.2.1.2 BTP APCSB 9.5-1 Appendix A Comparison Comparison 1OA-1 10A-1 1.2.1.3 1.2.1.3 Loading Tables Hazard/Fire Loading Fire Hazard/Fire Tables 1OA-2 lOA-2 1.2.1.4 1.2.1.4 Analysis Building Analysis 1OA-2 10A-2 1.2.2 1.:2.2 Standards Codes and Standards 1OA-2 lOA-2

1. 2.3 1.2.3 Defense-in-Depth Defense-in-Depth 1OA-2 10A-2

1. 2.4 1.2.4 Use of Water on Electrical Cable Electrical Cable Fires Fires 1OA-3 10A-3

1. 2.5 1.2.5 Establishment and Use of Fire Areas Establishment Areas 1OA-3 lOA-3 1.3 1.3 DEFINITIONS 1OA-3 10A-3 2.0 2.0 9.5-1 APPENDIX A BTP APCSB 9.5-1 COMPARISON COMPARISON 1OA-8 10A-8 2.1 2.1 OVERALL PLANT FIRE OVERALL NUCLEAR PLANT PROGRAM REQUIREMENTS PROTECTION PROGRAM REQUIREMENTS 10A-8 10A-8 2.1.1 2.1.1 Personnel Personnel 1OA-8 10A-8 2.1.1.1 2.1.1.1 Organizational Responsibilities Organizational Responsibilities 1bA-8 10A-8 2.1.1.2 2.1.1.2 Personnel Qualifications Personnel Qualifications 10A-12 lOA-12 2.1.1.2.1 2.1.1.2.1 Action 10A-13 10A-13 2.1.

2.1.22 Basis Design Basis 1OA-13 10A-13 2.1. 2.1.33 Backup Backup 1OA-14 10A-14 2.1. 2.1.44 Single Failure Criterion Single Failure 1bA-14 10A-14 2.1. 2.1.55 Fire Fire Suppression Systems Suppression Systems 10A-15 10A-15 2.1. 2.1.66 Areas Fuel Storage Areas 1OA-15 lOA-IS 2.1. 2.1.77 Fuel Loading 10A-15 10A-15 2.1. 2.1.88 Multi-Reactor Sites Multi-Reactor 1OA-15 10A-1S 2.1.99 2.1. Simultaneous Fires Simultaneous 10A-16 10A-16

• 20 UFSAR Revision 20 10A-i 10A-i October 2007 October 2007

Nine Mile Point Unit 1 UFSAR Section Title FIRE HAZARDS ANALYSIS TABLE OF CONTENTS ANALYSIS (Cont'd.) CONTENTS (Cont'd.) Page 2.2 2.2 ADMINISTRATIVE PROCEDURES, ADMINISTRATIVE PROCEDURES, BRIGADE CONTROLS AND FIRE BRIGADE 10A-16 1OA-16 2.2.1 2.2.1 Administrative Procedures Administrative Procedures 1OA-16 10A-16 2.2.2 2.2.2 Bulk Combustible Combustible Material Material Storage 1OA-16 10A-16 2.2.3 2.2.3 Program/System Integrity Program/System 10A-16 1OA-16 2.2.3.1 2.2.3.1 Ignition Sources Ignition Sources 1OA-16 10A-16 2.2.3.2 2.2.3.2 Leak Leak Testing 1OA-17 10A-17 2.2.3.3 2.2.3.3 Combustible Material Combustible Material Storage 1OA-17 10A-17 2.2.4 2.2.4 Local Fire Department Support Department Support 1OA-17 10A-17 2.2.5 2.2.5 Fire Brigade 1OA-17 10A-17 2.2.5.1 2.2.5.1 Surveillance and Maintenance Surveillance Maintenance 1OA-17 10A-17 2.2.5.2 2.2.5.2 Fire Drills 1OA-18 10A-18 2.2.6 2.2.6 Fire Brigade Brigade Training 1OA-18 10A-18 2.2.7 2.2.7 Training Guidance}}