ML16172A186

Sequoyah Nuclear Plant, Units 1 and 2 - Updated Final Safety Analysis Report Amendment 26 Redacted Version

ML16172A186
Person / Time
Site:	Sequoyah
Issue date:	06/10/2016
From:	Tennessee Valley Authority
To:	Office of Nuclear Reactor Regulation
References
Download: ML16172A186 (3995)

Text

Sequoyah Nuclear Plant Final Safety Analysis Report Amendment 26 TENNESSEE VALLEY AUTHORITY

1 SQN-26 TABLE OF CONTENTS SECTION TITLE PAGE

1.0 INTRODUCTION

AND GENERAL DESCRIPTION OF PLANT

1.1 INTRODUCTION

1.1-1 1.2 GENERAL PLANT DESCRIPTION 1.2-1 1.3 COMPARISON TABLES 1.3-1 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4-1 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1.5-1 1.6 MATERIAL INCORPORATED BY REFERENCE 1.6-1 1.7 LIST OF ABBREVIATIONS 1.7-1 1.8 TECHNICAL QUALIFICATIONS OF APPLICANT (HISTORICAL) 1.8-1 2.0 SITE CHARACTERISTICS 2.1 GEOGRAPHY AND DEMOGRAPHY 2.1-1 2.2 NEARBY INDUSTRIAL, TRANSPORTATION, AND 2.2-1 MILITARY FACILITIES 2.3 METEOROLOGY 2.3-1 2.4 HYDROLOGIC ENGINEERING 2.4-1 APPENDIX 2.4A FLOOD PROTECTION PLAN 2.4A-1 2.5 GEOLOGY AND SEISMOLOGY 2.5-1

2.6 CONCLUSION

S 2.6-1 3.0 DESIGN CRITERIA - STRUCTURES, COMPONENTS EQUIPMENT AND SYSTEMS 3.1 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA 3.1-1 3.2 CLASSIFICATION OF STRUCTURES, SYSTEMS, 3.2-1 AND COMPONENTS 3.3 WIND AND TORNADO LOADINGS 3.3-1 3.4 WATER LEVEL (FLOOD) DESIGN 3.4-1 3.5 MISSILE PROTECTION 3.5-1 3.6 PROTECTION AGAINST EFFECTS ASSOCIATED 3.6-1 WITH THE POSTULATED RUPTURE OF PIPING 3.7 SEISMIC DESIGN 3.7-1 3.8 DESIGN OF CATEGORY I STRUCTURES 3.8-1 APPENDIX 3.8A SHELL TEMPERATURE TRANSIENTS 3.8A-1 APPENDIX 3.8B CONTAINMENT VESSEL PENETRATIONS 3.8B-1 APPENDIX 3.8C CONTAINMENT ANCHORAGE 3.8C-1 APPENDIX 3.8D COMPUTER PROGRAMS USED IN 3.8D-1 STRUCTURAL ANALYSIS APPENDIX 3.8E DESIGN PROCEDURE FOR REINFORCED 3.8E-1 CONCRETE BLOCK WALLS 3.9 MECHANICAL SYSTEMS AND COMPONENTS 3.9-1 3.10 SEISMIC DESIGN OF CATEGORY I INSTRUMENTATION 3.10-1 AND ELECTRICAL EQUIPMENT 3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND 3.11-1 ELECTRICAL EQUIPMENT 3.12 CONTROL OF HEAVY LOADS 3.12-1 3.13 FLEX RESPONSE SYSTEM 3.13-1

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• SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT 10 0 10 2 0 3 C LC- , SCALE OF MILES FEATURES WITHIN 50 MILES FIGURE 2.1.1-2 SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT , I FEATURES WITHIN 10 MILES I FIGURE 2.1.1-3 1 0 1 2 3 4 h-- I I I SCALE OF MILES

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 2.1 .2-2 SITE BOUNDARY (RrVTSrD RY AMrNDMrNT EXCLUSION AREA -" - ROUTINE LIQUID RELEASE POINT D1lSCHARGE CANAL ROUTINE GASEOUS RELEASE POINT (elevation above plant grade) POINT 1 Auxiliary Building vent exhaust el. 107 ft Shield Building vent exhausl el. 121 -5 A RELEASE POINT 2 Service Building vent exhaust el. 26 ft RELEASE POINT 3 Condenser air ejector exhaust vent el. 93.7 ft Best Available Historical Image

Pine 6m Hardwood- Mixed LOCATION MdP f SEQUOYAH NUCLEAR PLANT FMAL SAFETY ANALYSIS REPORT I FOREST TYPES AND COVER BRADLEY AND HAMILTON COUNTIES FIGURE 2.1.4-2 b

SEQUOYAH NUCW FINAL SAFETY ANALYSIS I Figure 2.3.2-23 (Sheet 1 of 9) Topography Within Five Radius 1

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S2-4.doc 2.4-8SQN-26 Potential storm amounts differing by seasons were analyzed in sufficient number to make certain that the March storms would be controlling. Enough centerings were investigated to assure that a most critical position was used.

Storms producing PMP above upstream tributary dams, whose failure has the potential to create maximum flood levels, were evaluated in the original FSAR analysis. Dam safety modifications at upstream tributary dams have eliminated these potential failures and subsequent plant site flood levels. A standard time distribution pattern was adopted for all storms based upon major observed storms transposable to the Tennessee Valley and in conformance with the usual practice of Federal agencies. The adopted distribution is shown on Figure 2.4.3-3.

The critical probable maximum storm was determined to be a total basin storm with downstream orographically fixed pattern (Figure 2.4.3-1) which would follow an antecedent storm commencing on March 15. Translation of the PMP from Report No. 41 to the basin results in an antecedent storm producing an average precipitation of 6.4 inches in three days, followed by a three-day dry period, and then by the main storm producing an average precipitation of 16.5 inches in three days. Figure 2.4.3-4 is an isohyetal map of the maximum three-day PMP. Basin rainfall depths are given in Table 2.4.3-1.

To evaluate the local plant drainage system for a PMP event, Hydrometeorological Report No. 56 was used to calculate a 1-hr storm totaling 16.21 inches. Three different temporal distributions were applied to the model, with peak intensity of 2.81 inches/5-min = 33.72 in/hr shifted between early, middle, and late occurrence. Depths for each 5-minute increment of the controlling late peak distribution were 0.68, 0.78, 0.87, 0.97, 0.97, 1.17, 1.26, 1.36, 1.55, 2.81, 2.04, and 1.75. Rainfall on plant building roofs was assumed to discharge to the ground surface.

2.4.3.2 Precipitation Losses Precipitation losses in the probable maximum storm are estimated with multivariable relationships used in the day-to-day operation of the TVA system. These relationships, developed from a study of storm and flood records, relate the amount of precipitation excess (and hence the precipitation loss) to the week of the year, an antecedent precipitation index (API), and geographic location. The relationships are such that the loss subtraction from rainfall to compute precipitation excess is greatest at the start of the storm and decreases to no subtraction when the storm rainfall totals from 7 to 16 inches. Precipitation losses become zero in the late part of extreme storms.

For this probable maximum flood analysis, median moisture conditions as determined from past records were used to determine the API at the start of the storm sequence. The antecedent storm is so large, however, that the precipitation excess computed for the later main storm is not sensitive to variations in adopted initial moisture conditions. The precipitation loss in the critical probable maximum storm totals 4.13 inches, 2.30 inches in the antecedent storm amounting to 36 percent of the 3-day 6.44-inch rainfall, and 1.83 inches in the main storm amounting to 11 percent of the 3-day, 16.46 inch rainfall. Table 2.4.3-1 displays the API, rain, and precipitation excess for each of the 45 subwatersheds of the hydrologic model for the SQN probable maximum flood. No precipitation loss was applied in the probable maximum storm on the local area used to test the adequacy of the site drainage system and roofs of safety-related structures. Runoff was made equal to rainfall.

S2-4.doc 2.4-12SQN-26 All gates were determined to be operable without failures during the flood. Gates on main river dams would be fully raised, thus requiring no additional operations by the last day of the storm, which is before the structures and access roads would be inundated.

Median initial reservoir elevations were used at the start of the storm sequence used to define the PMF to be consistent with statistical experience and to avoid unreasonable combinations of extreme events. As a result, 53 percent of the total reserved system flood detention capacity was occupied at the start of the main flood. This is considered to be amply conservative. The statement made in the PSAR and subsequent versions of the FSAR that 67 percent of the reserved system detention capacity was occupied at the start of the main storm was in error. The correct percentage was 33. The remaining reserved system detention capacity was 67 percent. This erroneous statement was first made in the PSAR and was copied in subsequent statements where the routings were the same. In the revised analysis submitted in Amendment 51, all reservoirs are higher or about the same elevation at the beginning of the main storm as a result of the revised starting levels explained in Section 2.4.3.4 of the FSAR. This conservative change results in 53 percent of the total reservoir system detention capacity being occupied at the start of the main flood rather than 33 percent in previous studies. Neither the initial reservoir levels nor the operating rules would have significant effect on maximum flood discharges and elevations at the plant site because spillway capacities, and hence, uncontrolled conditions, were reached early in the flood.

The procedures used to determine if and when an overtopped earth embankment would fail and the procedures for computing the effect of such failures are described in 2.4.4.2 and 2.4.4.3. In testing the adequacy of the yard drainage system, to safely pass the site PMP, all underground drains were assumed clogged and the surface drainage to be full. 2.4.3.5 Water Level Determinations The elevation hydrograph of the controlling PMF, cresting at elevation 719.6, is shown on Figure 2.4.3-12. Computation of both the probable maximum discharge hydrograph (Figure 2.4.3-11) and the corresponding elevation hydrograph was accomplished concurrently using the unsteady flow techniques described in Section 2.4.3.3. The less critical total area storm-producing PMP depths on the 7,980-square-mile watershed would produce crest elevation 718.9 at the plant site. Maximum water levels at buildings expected to result from the local plant PMP were determined using a transient flow (unsteady flow) model with hydraulically connected storage areas. Much of the plant site is flat, particularly at the switchyards, and a single flow path is not well defined. A transient model with interconnected storage areas very roughly approximates a two-dimensional model using one-dimensional methods by providing multiple simultaneous outlet paths for the exterior areas adjacent to plant buildings. The separate watershed subareas and flowpaths are shown on Figure 2.4.3-13a.

S2-4.doc 2.4-13SQN-26 The western plant site was evaluated as six interconnected storage areas with four primary weir-flow outlets and one connected transient flow stream-course model. Runoff from the western plant site will flow: Northwest to a channel along the main plant tracks and then across the main access highway (Area 7); to the West through a parking lot (Areas 6A, 6C, and 6E connected to transient flow model); Southwest through the vehicle barrier system directly to Chickamauga Lake (Area 6E); or South through the vehicle barrier system to the Yard Drainage and other Ponds (Area 6C). The maximum water surface elevations in Areas 6A and 6BS are below critical floor elevation 706.

The eastern plant site was evaluated as three interconnected storage areas with three weir-flow outlets and two connected transient flow stream-course models. Runoff from the eastern plant site will flow: North around the West and East ends of the multipurpose building to the intake channel (Area 5 connected to two transient flow models); South to the Condenser Circulating Water Discharge Channel (Areas 4 and 6D); or Southwest into the western plant site (Area 6D into 6C). The maximum water surface elevations in Areas 4, 5, and 6D are below critical floor elevation 706. Underground drains were assumed clogged throughout the storm. For fence sections, the Manning's n value was doubled to account for increased resistance to flow and the potential for debris blockage.

The only stream adjacent to SQN is the Tennessee River. There are no streams within the site. The 1 percent-chance floodplain of the Tennessee River at the site is delineated on Figure 2.4.3-14. Details of the analyses used in the computation of the 1-percent-chance flood flow and water elevation are described in a study made by TVA for the Federal Insurance Administration (FIA) and published in February 1979 [5].

The only structures located in the 1-percent-chance floodplain are transmission towers, the intake pumping station skimmer wall, and the ERCW pump station deck. The ERCW pumps are located on the pump station deck at elevation 720.5, well above the 1-percent-chance flood level. These structures are shown on Figure 2.4.3-14. The structures that are located in the floodplain will not alter flood flows or elevations. The 20,650-square-mile drainage area is not altered and the reduction in flow area at the site is infinitesimal and at the fringe of the flooded area. The site will be well maintained and any debris generated from it will be minimal and will present no problem to downstream facilities. 2.4.3.6 Coincident Wind-Wave Activity Some wind waves are likely when the probable maximum flood crests at SQN. The flood would be near its crest for a day beginning about 2-1/2 days after cessation of the probable maximum storm. The day of occurrence would most likely be in the month of March or possibly the first week in April. A conservatively high velocity of 45 miles per hour over water was adopted to associate with the probable maximum flood crest. A 45-mile- per-hour overwater velocity exceeds maximum March one-hour velocities observed in severe March windstorms of record in a homogeneous region as reported by the Corps of Engineers [6].

That a 45-mile-per-hour overwater wind is conservatively high, is supported also by an analysis of March day maximum winds of record collected at Knoxville and Chattanooga, Tennessee. The records analyzed varied from 30 years at Chattanooga to 26 years at Knoxville, providing samples ranging from 930 to 806 March days. The recorded fastest mile wind on each March day was used

T243-2.doc Table 2.4.3-2 UNIT HYDROGRAPH DATA Unit Drain Area, Duration, Q C T W W T AREA Name Sq. Miles Hours p p p 50 75 B 1 French Broad River at Asheville 945 6 15,000 .27 14 35 12 166 2 French Broad River, Newport to Asheville 913 6 35,000 .53 12 12 7 108 3 Pigeon River at Newporta 666 6 26,600 .56 12 11 6 78 4 Nolichucky River at Embreeville 805 6 27,300 .58 14 14 9 82 5 Nolichucky Local 378 6 10,600 .40 12 16 9 87 6 Douglas Locala 832 6 47,930 .27 6 8 6 60 7 Little Pigeon River at Sevierville 353 6 15,600 .62 12 10 6 102 8 French Broad River Localb 207 6 7,500 .51 12 11 8 60 9 South Holston 703 6 16,000 .53 18 24 17 100 10 Wataugab 468 6 17,700 .53 12 13 7 84 11 Boone Locala 669 6 22,890 .16 6 13 8 90 12 Fort Patrick Henry 63 6 3,200 .40 8 8 6 64 13 North Fork Holston River near Gate Citya 672 6 12,260 .60 24 33 25 108 14 Surgoinsville Localb 299 6 10,280 .48 12 13 9 66 15 Cherokee Local below Surgoinsvilleb 554 6 18,750 .48 12 14 7 66 16 Holston River Localb 289 6 6,800 .55 18 22 15 96 17 Little River at Mouthb 379 4 11,730 .68 16 14 8 96 18 Fort Loudoun Localb 323 6 20,000 .29 6 10 6 36 19 Little Tennessee River at Needmore 436 6 9,130 .49 18 23 12 126 20 Nantahala 91 6 3,770 .45 10 12 7 70 21 Tuckasegee River at Bryson City 655 6 26,000 .43 10 12 7 58 22 Fontana Local 389 6 16,350 .46 10 9 5 94 23 Little Tennessee River Local, Fontana-Chilhoweeb 406 6 16,900 .58 12 9 5 84 24 Little Tennessee River Local Chilhowee-Tellico Damb 650 6 17,000 .61 18 21 11 72 25 Watts Bar Local above Clinch Riverb 293 6 11,300 .30 8 9 7 84 26 Norris Dam 2912 6 43,300 .07 6 15 8 118 27 Coal Creekb 36.6 2 2,150 .64 8 9 5 40 28 Clinch Localb 22.25 2 1,350 .10 2 8 5 34 29 Hinds Creekb 66.4 2 3,620 .68 9 7 5 54 30 Bull Run Creekb 104 2 2,400 .47 14 21 14 84 31 Beaver Creekb 90.5 2 2,600 .58 14 14 10 88 32 Clinch Locals (5 areas)b 111.25 2 1,350 .10 2 8 5 34 33 Local above mi. 16b 37 2 4,490 .95 6 4 3 46 34 Poplar Creekb 136 2 2,800 .61 20 25 13 88 35 Emory River at Mouthb 865 6 34,000 .37 9 13 8 87 36 Local area at Mouthb 32 2 3,870 .95 6 3 2 46 37 Watts Bar Local below Clinch Riverb 427 6 16,300 .36 9 9 7 84 38 Chatuge Dama 189 6 13,570 .34 6 6 5 54 39 Nottely Dama 215 6 13,500 .29 6 5 4 80 40 Hiwassee Local 564 6 13,800 .36 12 18 12 124 41 Apalachia Local 50 6 2,900 .54 9 6 4 90 42 Blue Ridge Dama 232 6 11,920 .24 6 7 4 54 43 Ocoee No. 1 to Blue Ridgeb 363 6 17,000 .37 8 11 7 36 44 Lower Hiwassee 1087 6 32,500 .93 23 16 10 136 45 Chickamauga Locala 780 6 32,000 .38 9 14 7 36 Definition of Symbols Qp = Peak discharge in cfs Cp = Snyder coefficient Tp = Time in hours from beginning of precipitation excess to peak of unit hydrograph W50 = Width in hours at 50 percent of peak discharge W75 = Width in hours at 75 percent of peak discharge TB = Base length in hours of unit hydrograph a = Revised b = New

(0) I I i i I HISTORICAL I i 'Figure 2.4.2- 1 Flood Distribution , ! Diagram i Chattanooga, Tn. , : Revised by Amendment 17 I

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+/'//uydJ'J 931:'s c-c , 2 --. .. uN/TS - + liyd-05f3fi~ Hyd~~di/nd. .if-c,'cf Earfhqu~3 ks -- - Forc e Due 70 Ear f I) q /Me TYPICAL BLOCK-SP/LLWAY - NW- TW TL"/t f =3a~,,l~~~~~~,l 11 4 MG . AfC 7/ON -POWERHOUSE HN- TW UPLIFT DIAGRAM E/ 650 /. 0 a E e&r/hq,/&c /,*?er f/a K - 7Hf 7"=353 4, I ,I 11,11111~~~~ I I I III ' 11, 4 UPL/FT PRESSURE ASSUMLfD TO forces assumed d5 009 q UPLIFT DIAGRAM E/ 657 wc~lv,5.5~~/ h~r/zonfa//i/ snc' 0.06 q ACT ON /00% OFBASE AREA /]PL /F T PRESS URL A SSUMED TO verficor/ v af +he base ACT ON lOO% OF BASEAREA C and amp1;f ;ed up f he Nofe A: sfrucfure. 7%@ power house snd sp171r~ay GI 2. ~pil~way ga +es were fruc fures are we// keyed /n f o B Drain --, ~5sumec/ open $or- fh is fhe rock founddfioo. The rock fkrnaf/ur/s sre severe4 folded GW/y s h . w/ th fhe dp genera/& /h B dns Sr t direcf~on vd~yinq from /D " f~ 40 O- L 85.05' -)i Any fa//ure vuou/d requ1i.e cro.~~ bed shear of fhe rock. Rockof fhI-5 BASE PLAN A7 E/ 650.0 type hds cross bed shear sfrmqfh 32.79' much /n excess of the+ reqd for t----- 2 350' - 43.5' - - - - -1 dFscfor of Safefy of /. ' 5.0' BASE PLAN AT F/ 6570 11 t 66.25' % BASE PRLS S URES / k C 57+

• Shea~s, fhf 1s regd for Qs//s ~@/cu/ofedfrorn shear-fr/cf/orl Scale /"= 40' formula, Q= Y 0-65$$s4* A IS dss~med fo be enf/re base aped. 'D BASE PRLSS URES ,COWL,Y HOUSF & SPILL WA Y & (r IMR yep ica/ - IMP 1 cf fjCd/ ,QES UL 75 Of ANAL YS IS FOR I%I$~L~~~oJ. 1 I I__ Iy 2H z zM 9 Shear,SForQ=/ SRe9d/.mx/~=S.iSAe~ron / A"q SRr9d!~ffidA f3Jz-'[hcdl~n wv 75 SUCH* AM, LIFIE~ ---- =Xo P/aneA-A- - lShea/: SfiorQ=/ , ----A- p/sne A -A OP,RA T//G BAS IS LA R THOWkE CARTHQUhKE ANAL1',15 ::I mrt (YO(/M ~JI*IP/IYI/MU NO+, A* Nr, fe A ~rr un 12z 207 26,232 0.9L (entie 2z6psi W,~L 9 5ps;)*' PS; /56 /, 2/ 42 2ps; ' 244p.s; IZZPS;

• 52931 /.d4 29,~s; WATT3 3AR &AM om.,z, ~~TI~IP UM GYK 1 (44.7py')f % IUD 1)41#111 Lo"* - I 1 - th Snedi .5. re@/ for O- / c~nb/J~ii/)g partion of eJ3i :n ccrr7oresiid/. (no ref)dlof~)~ /05tt?aS of ei) ' c 6. LC a FIGURE 2.4.4- 1 Revised by Amendment 6.

Figure 2.4.4-4 . Powerhouse and Sp/ llway Fort Loud?v*q Dam ' Results of ~nalyshs for 112 SSE

X Shea~s, fhsf 1.i reqdfor &=/is C d/cu/df ed from shear- fr /c +/on f~fnl//d, d= d?.6521/+.yAp~ js LH dssumed fo be erlf/fe aped, *#Shear sfress,s, regd f~r a=/ ~ons/der/fig porf/bo of bd5 e io co/npress/on insSead of entire base area. 1 Spi//wdy ysf es op e fl 0.49q 0.66 q I, -' - -- - --- -- - f_r/E- -- f^/L/wE/8/7 -- - - \ - - - -. r I -7W-E/ 793 r _ i L - E/ ,777 f \a 'BI'LJ. I k -A 90.65' cl I ~drce. G c I OD* fd/ccbcJd Me- Accc/cr df/m k=/fh~~uKe. OfDdm TYP/CA L NONOYF@FL OH SfC 7/ON ~YP/CAL SP/L LWA Y SECT/UN HW- 7PP Twr ~--- K , Due A 4 = 513 L &2p/hqg * ~=4.56 %a ~IIIIII 11 ~~=475%# ~,,l,)~/~ilil , 1, A' @'dh!& ~6.25 OR /f T D/AGRAME/ 7-20 DR /FT D/A GPAM f/ 7/7 UflF7- PRESSURE ASSUMED 70 UPL /F i PAE55URE A SSLMED TO A C OM /UQ % O/C BASE ARk2 AC7 04 /DO% (3F6A.5: AREA NU72=5 /Ye// c a/ 4- e/ylL+f/ 9-1 cf non~vcr//or/ 1 2nd sp///wd J df &3e a3 sc/n~rd fd be 0 069. B/ bJhor L j/s/;li~~~fi/~~'/ d drte/c rd f ,-I <,Lv/e fie e WJ~ d- f e, *, e d to be 5, i29 2 f f9l~ &Or fhe //an~vzrf/ow :ecf/on d&D/yqd/ +he fop for +/ e . A / //t va~ . 2 /fuf/roo $d/' ,LC ~/crdf/s// of nu/ioverfhi~/ 1 ~nd~p//wdyrifhdrcrnsupcdfotr 1 c. .' 39, fiy dy//dn~ aodi/s/+ d@/ f/f/c_ dh-, 90.66 ' I of j~-ce/r/ ,, f rj 6.boye fise base &W- I P--- - --BASE PLANATE/ 720 d~+~f/fi/fiedf~/it~.~~ydffh~tc~~fc~ fhe r@/;l--vcr f/p~ recf/w and 0, 46g. BASEPLANA.rE/77/7 d'iOe~dpL3r/hespi//~ G p 3. Sp///wdy p~f c s olssu~ed open fc~ j+e/., und I "' a r 8 4 $1 v4@68' ,754 q$' .5c d/e /"= 40' NONOY~-/~FL ON& SPL L WA )/ ;~-5 i/L TS OF4 WAL YS/S FQ8 O/?f&ii T/#& EAS 1.5 ZilP ,~Hffbilkit TELL/CO DAM Revised FIGURE by Amendment 2.4.4-6 6. BASE PRESSURE h BASE PRESSb'RL- \ IIII 2 '-1 1 I. . 4 I I, \ I\ \ \ , j* a\,. -.tr*.tA\i i/crf/ca/ ZH 9 S'eqd fmar f5=- zMn Sesron dX fY Shed&SForL?=/ ZMo H.~GA-A I I /3,602K 97/5K. 0.72 24ps/ 7 @96p~/w err[ JWICU DYI pnalj64 cnw llCD (DW I 2N Avq 5 Aeqd "e/.t/cd/ ZY zH TY SheacSt-orQ=/ f ~S=~X'.She~?ron LMo ~/ane A-A //,'94Kl ZC?/fi ad9 bps,. 4jp5/'* dop5, , 6 40~~1 ...- 6 ~PS/)* YM.IIWW\M/UU Yn IUI? l*OlnrIn . .. Ill Revised by Amendment 1 Figure 2.4.4-7 Embankment - Tellieo Dam Results of Analysis for 1/2 SSE f ,. Earthquake. - '~L~vL % - /dD.B1 .- J a BASE PRESSURE *.Shear, r, /Ad+ is reg/ &v Q=/ is calcdd fed f/am dedr - fitc f/bn rormu/e, QQ-. 0.65XK* U , A ,s Z& UPL /fT PRES5UA' ~~~~ /. Yert~s/ OCC&P~//M a'm*cr/iwad p~ A cr ON 100 %O~BAJT~ AREA W/"W~Y " base essu-d be 4Q51. By &&mu tsw/ysis, &wpkf/cot/m, Of drce/efdf/on &ve the be w& i def&frn/ncd +o be d./lg s$ fhd #up m3.0 9' fdf I))e mfbrl J &#,'a yd q.r]* rt *e fo4 fh tAe s l//rufi , 2 accekpaL d-w~ and *//way e t bere a=#& to Pa 23J2% 0.0 9 g. by dynmhz e&p ls, 3-74 ewp/ifies4rm of dazekraS,m &eve tA# &se ~ss deSerm/n cd to d. 0.Hq of fhe fop * +ha -&,& %- BSE PRESSURE sec+/a? m/ 0.759 et SAe fw frr % fi. *;//uldy. k 'b 3. +///U~Y q&s BSYRH~ 40- fi~ *s ardysis. Figure 2.4.4-8 Spillway h Nonover~lov Norris Dam, Results of Analysis for 1/2 SSE PLAN @55~n*rg AFTER FMURE) Analysis for 1 2 SSE 6 One Maxi 4 a TYP/CAL SM L WAY SEC T/Cr/ Edrflqwde. - TW+ HLY- TH' UPLIFT DAGRAM E/ 5XW UPL /fT PffESSURf ASSUMED 70 ACT ON 100 % Of&45E AREA I 1 ---. . /3 8.9/ ' I BASE PLAN A 7 E/ 95'0

• Shear, s, f!/ is regd fir a=/ co/cu/dfed from sheer - f~/cf/hn p 50.7.1 formu/&, 8- 0-65s f H Yf A is dssc/med f- dr enhi-e drea. ' BBI/ *.S,f~edr rhcis,r, ~dfir o.1 3 ?L cons/der,ng porf~bn afhe ;fl BASE PRESSU' ~omprcss/bo /iu reed of cd/pe \4 bese wee. N07ES .- I /e/hca/ scc e/erdfk of he sp///my d & &sl ~mu/npd fa k 0.06 g. &y dywmi mr/rrljs r*rp/ific~f/bn of rj &M +/a ddP~ fie bare was def e k hsd fo br a//g d +h fdp. i 2. /forizm td wce/o~fn of JAo spl;uway rt He base 1essum.d 6. 0.09q. By ?!!-/c WI~, amp/M/csf/m of m e/e rsfbn above fA. &sr ~4'4s deferm'ned to be &5g &+ rHc few dW/wy gdes dsshmed open f#f #4/s &r/y+i'. 2f4.4-10

PLAN :3: - ? > SECTION B-9 2.4.4-1 2 Assumed of Failure, 112 SSt 112 ?iaximurn

DOWNSTREAM ELEVATION -a: SECTION A-A B-B NOT'. A : A// debris fiau fii/cd p@//un of &/n judged to be de/o~ /'a//ure ekprsf/bns. NU7f 8 : Powerhouse /frtdke blocks /2' 17 and A&noverf/ow Dd.31 b/~ck~ - 5 and 29 -35 judged fd re/;r~in in [ Figure 2.4.4-13 Douglas Dam, ha'sumed Condition of D m After Failure, 112 S E and Flood 112 Maximum Po 1 sible

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UPL /f 7 MESSW" ASSUMED TU ACT ON /UU%@ BASE AREA # Scd4s,fhaf /s rep& for Q;/ is ca/cu/&fed from shear - fr/c //on - fdrm u/e, 0.65 .ZY + SR , A is I# sssumed tobe enfire ercs. X # fir base s+ N 7# resu/fdnf 18505' fc~//s oufs/de b s~e un& r S $ E BASE PLANN 700 NO Tics: l Yer f/'ca/ a c e/erafi~ of sp~;Yhiy a+ bass sssumed f@ be d /Zq. By ~&I/.s/$ amp/,'ficof,;m sf ac c~~rl#kw &ova tAe bass ncjsf defermined fo be 484gd3 kp Z./S6r/-~nfd/ ec/&bu;, of ~7!.y sS6assseed to b. 0. /89 8y dymmb Wsis, amplific &fan of e ~~bkrd/b~ csboue /he base #sr defermined fe be L192jaf hj~ 3. Sp//way gafes ass 4 fee f fer +his en Tdspn , ysk. I I I Figure 2.4.4-24 Spillway, Fort for I Dam, Results of Analysis I sY &2#Y . Z8534X 2H Avq Sregd ZMR 2 P she& for Q= / Mx f%~o /. 62

/ PLAN Sp;//way judged fo fE/ 822.0 fm'/ above E/ 750.0 d= (see Note A) --, I / SECT A-A POWERHOUSE JUDGED ~'OT UPPER PART JUDGED MOT CF SP/LLWAY FA--) UGED TO FA/'-, EL E VA T/ON Note A : A// debrts fmm failed portion of dam judged fo 6e below fsi/ure elevation Nofe B: High portion of embankment ~udged fo fail during earfhgueke Mhef portion sssurned fo erode sfter fai/ure.

MMVERFL OW DOWNSTREAM ELEVATION B - ~/~D~..D~UFA/L. aN.muxw - with Year Figure 2.4.4-28 SSE + 25 Year Flood ~udged Condition of Dam After Failure, Norris Dam

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RELIEF SETTLING LOME WATER< - 695 BtOWDOWN VA LYE S ---- AND INTAKE CHANNEL =ATION flRUCTURE (1) (1 1 ERCW LOADS (4 LINES) REACTOR BLDG - ERCW PUMPS (0) DRAIN SUMP PUY PS WDS (r/unrr) _-- NEW ERC* I, HTAKE STATION 't' A SCREf NS NOTES: DRAWING DOE5 NOT INCLUDE: VENTILATION SYS.. EMERCENCY POWER SYS.. SAMPLING SYS., DETAILS OF ERCW LOADS, Sf'>OL PIECE CONNECTIONS.

9 711) -- , Eiev. 6s; Note: flood .'fydrograph computed usi'ng watersbed mode/ aescrtbed in Paragraph --\ 2. 4.j3 as amended MARCH Piant grade elev 705 Critical flood elev. 703 C 0 - - 0 > Q - W Note - flood hydrograph computtd using watersbed modal described in Paragraph 2.4.5 . as amended. .--- i 1 ! 6 70 2 0 1 2 2 23 24 25 26 MARCH MOTE: T,mas shown allow 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for foracast commun~cat~ons computatioo. and SEQUOY,4H NUCL-EAR PLANT FLOOD PROTECTION PLAr.1 BASIS FOR SAFE SHUTDOWK FOR PLANT FLOODING Histroical These graphs would be impacted by the safety modifications to the dam. But their use, as is, results in equal or greater flood proection. Plant grade slsv. 705 - -- c:lil;ca/ flood elev. 703 /- 2.4A- 4 HISTORICAL Revised by Amendment

T252-1.doc SEQUOYAH PLANT HISTORICAL EARTHQUAKE LISTING 200 MILE RADIUS AROUND 85.1 W LON 35.2 N LAT YEAR MONTH DAY INTENSITY LOCATION NLAT WLON

1. 1776 Nov 5 IV Jackson Co.,NC 35.4 83.2 2. 1817 Dec 11 IV SC-GA 0.0 0.0 3. 1817 Dec 12 <IV KY 0.0 0.0 4. 1825 Mar 19 Columiba,TN 35.6 87.0 5. 1828 Mar 10 IV Southwestern VA 0.0 0.0 6. 1829 <IV Andrews,NC 35.2 83.8 7. 1843 Aug 9 IV Columbia,TN 35.6 87.0 8. 1844 Jun <IV Jackson Co.,NC 35.2 83.1 9. 1844 Nov 28 VI Knoxville,TN 36.0 83.9 10. 1848 <IV McDowell Co.,NC 35.7 82.0 11. 1851 Aug 11 V Asheville,NC 35.6 82.6 12. 1852 Oct 12 <IV Clinton,GA 33.0 83.5 13. 1852 Oct 23 <IV Clinton,GA 33.0 83.5 14. 1854 Feb 13 <IV Manchester,KY 37.2 83.8 15. 1860 Jan 20 NC-SC-GA 0.0 0.0 16. 1872 Jun 17 IV Milledgeville,GA 33.1 83.2 17. 1874 Feb 22 V McDowell Co.,NC 35.7 82.1 18. 1875 Jul 29 <IV Milledgeville,GA 33.1 83.2 19. 1875 Nov 2 IV Washington, GA 33.7 82.7 20. 1875 Nov 12 <IV Knoxville,TN 36.0 83.9 21. 1876 Jan 23 <IV McDowell Co.,NC 35.7 82.0 22. 1877 Apr 26 <IV Franklin,NC 35.2 83.4 23. 1877 May 25 <IV Knoxville,TN 36.0 83.9 24. 1877 Jun 3 <IV Stanford,KY 37.5 84.7 25. 1877 Oct 9 <IV Hendersonville,NC 35.3 82.5 26. 1877 Nov 16 IV Knoxville,TN 36.0 83.9 27. 1878 Nov 23 <IV Murphy,NC 35.1 84.0 28. 1880 Jan 28 <IV McDowell Co.,NC 35.7 82.0 29. 1882 Oct 15 <IV Murphy,NC 35.1 84.0 30. 1883 Jan 1 IV Ashwood,TN 35.6 87.1 31. 1884 Jan <IV McDowell Co.,NC 35.7 82.0 32. 1884 Mar 31 <IV Milledgeville,GA 33.1 83.2 33. 1884 Apr 30 <IV Ogreeta,NC 35.2 84.2 34. 1884 <IV Elk Mt.,NC 35.7 82.5 35. 1884 Aug 25 IV Knoxville,TN 36.0 83.9 36. 1886 Feb 5 IV Valley Head,AL 34.6 85.6 37. 1888 Mar 17 <IV Jonesboro,TN 36.3 82.5 38. 1889 Jun 7 IV Benton Co.,TN 35.9 88.1 39. 1889 Sep 28 <IV Parksville,TN 35.1 84.6 40. 1892 Dec 2 V Chattanooga,TN 35.0 85.3 41. 1895 Jul 27 Savannah,TN 35.2 88.3 42. 1898 Mar 30 <IV Mt. Hermon,KY 36.8 85.8 43. 1898 Jun 6 <IV Richmond,KY 37.8 84.3 44. 1902 May 29 IV Chattanooga,TN 35.0 85.3 45. 1902 Oct 18 V Chattanooga,TN 35.0 85.3 46. 1904 Mar 5 <IV Maryville,TN 35.8 84.0 47. 1909 Oct 8 <IV Dalton,GA 34.8 85.0 48. 1911 Apr 22 <IV Hendersonville,NC 35.3 82.5 49 1912 Oct 23 <IV Macon,GA 32.8 83.6 50. 1912 Dec 7 <IV West Springs,SC 34.8 81.8 51. 1913 Jan 1 VII West Springs,SC 34.8 81.8 52. 1913 Mar 13 <IV Calhoun,GA 34.5 85.0

T252-1.doc(Continued) SEQUOYAH PLANT HISTORICAL EARTHQUAKE LISTING 200 MILE RADIUS AROUND 85.1 W LON 35.2 N LAT YEAR MONTH DAY INTENSITY LOCATION NLAT WLON

53. 1913 Mar 28 VI Knoxville,TN 36.0 83.9 54. 1913 Apr 17 V Madisonville,TN 35.5 84.4 55. 1913 May 2 <IV Madisonville,TN 35.5 84.4 56. 1913 Aug 3 IV Knoxville,TN 36.0 83.9 57. 1914 Jan 24 IV Sweetwater,TN 35.6 84.5 58. 1914 Mar 5 IV Central GA 33.5 84.0 59. 1915 Jan 14 IV Briston,TN 36.6 82.2 60. 1915 Oct 29 IV Marshall,NC 35.8 82.7 61. 1916 Feb 21 VII Waynesville,NC 35.5 83.0 62. 1916 Mar 2 IV Anderson,SC 34.5 82.7 63. 1916 Oct 18 VII Irondale,AL 33.5 86.7 64. 1916 Nov 4 IV Birmingham,AL 33.5 86.8 65. 1917 Jan 2 IV McMillan,TN 36.6 83.9 66. 1917 Jan 25 Jefferson City,TN 36.1 83.5 67. 1917 Mar 5 Knoxville,TN 36.0 83.9 68. 1917 Mar 27 V Jefferson City,TN 36.1 83.5 69 1917 Apr 19 <IV southwestern VA 0.0 0.0 70. 1918 Jan 17 IV Knoxville,TN 36.0 83.9 71. 1918 Jun 22 IV Lenoir City,TN 35.8 84.3 72. 1920 Apr 7 II 36.3 88.2 73. 1920 Dec 24 IV Glen Alice,TN 35.8 84.7 74. 1921 Jul 15 V Mendota,VA 36.7 82.3 75. 1921 Sep 2 IV Statesville,TN 36.0 86.1 76. 1921 Dec 15 IV Glen Alice,TN 35.0 84.7 77. 1922 Mar 30 <IV Farmington,TN 35.5 86.7 78. 1922 Mar 30 <IV Arcadia,TN 36.6 82.5 79. 1923 Oct 18 IV Hendersonville,NC 35.3 82.5 80. 1924 Jan 1 IV Greenville,SC 34.8 82.4 81. 1924 Oct 20 IV Pickens,SC 34.9 82.7 82. 1924 Nov 13 V Bristol,VA 36.6 82.2 83. 1926 Jul 8 VII McDowell Co.,NC 35.7 82.0 84. 1927 Jun 16 IV Scottsboro,AL 34.7 86.0 85. 1927 Jul 20 V Knoxville,TN 36.0 83.9 86. 1927 Oct 8 IV Chattanooga,TN 35.0 85.3 87. 1928 Mar 7 IV Columbia,TN 35.6 87.0 88. 1928 Nov 3 VII Hot Springs,NC 35.9 82.8 89. 1928 Nov 20 IV Hot Springs,NC 35.9 82.8 90. 1929 Oct 28 IV Due West,SC 34.3 82.4 91. 1930 Aug 30 V Kingston,TN 35.9 84.5 92 1930 Oct 16 VI Knoxville,TN 36.0 83.9 93. 1930 Dec 10 Due West,SC 34.3 82.4 94. 1931 Apr 1 Hopkinsville,KY 36.9 87.5 95. 1931 May 5 VI Birmingham,AL 33.5 86.8 96. 1931 Nov 27 <IV Nashville,TN 36.2 86.8 97. 1935 Jan 1 V GA-NC 35.1 83.6 98. 1936 Jan 1 <IV Blue Ridge,GA 34.9 84.3 99. 1938 Mar 31 IV Tapoco,NC 35.5 84.0 100. 1939 May 5 V Anniston,AL 33.7 85.8 101. 1939 Jun 24 IV Huntsville,AL 34.7 86.6 102. 1940 Oct 19 IV Ryall Springs,TN 35.0 85.1 103. 1940 Dec 25 IV Hot Springs,NC 35.9 82.8 104. 1941 Mar 4 <IV Rockford,TN 35.9 83.9

T252-1.doc(Continued) SEQUOYAH PLANT HISTORICAL EARTHQUAKE LISTING 200 MILE RADIUS AROUND 85.1 W LON 35.2 N LAT

YEAR MONTH DAY INTENSITY LOCATION NLAT WLON 105. 1941 May 10 IV Asheville,NC 35.6 82.6 106. 1941 Sep 8 IV Lookout Mt.,TN 35.0 85.4 107. 1945 Jun 14 V Cleveland,TN 35.2 84.9 108. 1946 Apr 7 IV Cleveland,TN 35.2 84.9 109. 1947 Jun 6 IV Knoxville,TN 36.0 83.9 110. 1947 Dec 28 IV Ryall Springs,TN 35.0 85.1 111. 1948 Feb 10 VI Wells Springs,TN 36.4 84.0 112. 1949 Sep 17 V Pennington Gap,VA 36.8 83.0 113. 1950 Jun 19 IV Tapoco,NC 35.5 84.0 114. 1952 Feb 6 V Birmingham,AL 33.5 86.8 115. 1952 Jun 11 VI Johnson City,TN 36.3 82.4 116. 1953 Nov 10 IV Knoxville,TN 36.0 83.9 117. 1953 Dec 5 IV Knoxville,TN 36.0 83.9 118. 1954 Jan 1 IV Hazard,KY 37.2 83.2 119. 1954 Jan 2 VI Hazard,KY 37.2 83.2 120. 1954 Jan 14 IV Knoxville,TN 36.0 83.9 121. 1954 Jan 23 IV Etowah,TN 35.3 84.5 122. 1955 Jan 6 IV Bristol,TN 36.6 82.2 123. 1955 Jan 12 IV Maryville,TN 35.8 84.0 124. 1955 Jan 25 IV Knoxville,TN 36.0 83.9 125. 1956 Jan 5 IV Due West,SC 34.3 82.4 126. 1956 May 19 IV Due West,SC 34.3 82.4 127. 1956 May 27 IV Due West,SC 34.3 82.4 128. 1956 Sep 7 VI Maynardville,TN 36.2 83.8 129. 1956 Sep 9 IV College Grove,TN 35.8 86.7 130. 1957 Jan 25 IV Middlesboro,KY 36.6 83.7 131. 1957 Apr 23 VI Birmingham,AL 33.5 86.8 132. 1957 May 13 VI McDowell Co.,NC 35.7 82.0 133. 1957 Jun 23 IV Dixie Lee Junction,TN 35.9 84.2 134. 1957 Jul 2 VI Asheville,NC 35.6 82.6 135. 1957 Nov 7 <IV Powell,TN 36.0 84.0 136. 1957 Nov 24 VI Bryson City,NC 35.4 83.4 137. 1958 May 16 IV Asheville,NC 35.6 82.6 138. 1958 Oct 20 IV Anderson,SC 34.5 82.7 139. 1959 Jun 13 IV Tellico Plains,TN 35.4 84.3 140. 1959 Aug 12 VI Meridianville,AL 34.8 86.6 141. 1960 Jan 3 IV Spruce Pine,NC 35.9 82.1 142. 1960 Feb 9 VI Edneyville,NC 35.4 82.4 143. 1960 Apr 15 IV Maryville,TN 35.8 84.0 144. 1963 Apr 11 IV Greenville,SC 34.8 82.4 145. 1963 Nov 14 <IV Nashville,TN 36.2 86.8 146. 1963 Dec 5 <IV Beechmont,KY 37.2 87.0 147. 1963 Dec 15 <IV Beechmont,KY 37.2 87.0 148. 1964 Jan 20 IV Pensacola,NC 35.8 82.3 149. 1964 Feb 18 V Mentone,AL 34.6 85.6 150. 1964 Mar 13 IV Haddock,GA 33.0 83.4 151. 1964 Jul 28 <IV Inskip,TN 36.0 84.0 152. 1964 Oct 13 Knoxville,TN 36.0 83.9 153. 1965 Apr 7 McCormick,SC 33.9 82.3 154. 1965 Nov 8 <IV Canton,GA 34.2 84.5 155. 1966 Aug 24 IV Maryville,TN 35.8 84.0 156. 1969 May 5 GA-SC Border 33.9 82.50 T252-1.doc(Continued) SEQUOYAH PLANT HISTORICAL EARTHQUAKE LISTING 200 MILE RADIUS AROUND 85.1 W LON 35.2 N LAT YEAR MONTH DAY INTENSITY LOCATION NLAT WLON 157. 1969 Jul 13 V Knoxville,TN 36.0 83.9 158. 1969 Jul 24 Knoxville,TN 36.0 83.9 159. 1969 Dec 13 IV SC-NC Border 35.0 83.0 160. 1971 Jul 13 IV Kingston,TN 35.9 84.5 161. 1971 Jul 13 VI Newry,SC 34.7 82.9 162. 1971 Oct 9 V Gatlinburg,TN 35.7 83.5 163. 1973 Nov 30 VI Maryville,TN 35.8 84.0 164. 1974 Aug 2 V McCormick Co., SC 33.9 82.5 165. 1974 Oct 8 Clark Hill Reservoir,SC 34.0 82.3 166. 1974 Nov 5 Clark Hill,SC 33.7 82.2 167. 1974 Dec 3 Mt. Carmel,SC 34.0 82.5 168. 1975 Feb 10 Gatlinburg,TN 35.7 83.5 169. 1975 May 2 Oakdale,TN 36.0 84.6 170. 1975 May 14 Oak Ridge,TN 36.0 84.3 171. 1975 Jun 24 IV Fayette,AL 33.7 87.8 172. 1975 Aug 29 VI Palmerdale,AL 33.8 86.6 173. 1975 Oct 18 IV Jocassee Lake Dam,SC 34.9 83.0 174. 1975 Nov 7 Samantha,AL 33.4 87.6 175. 1975 Nov 25 IV Salem,SC 34.9 83.0 176. 1976 Jan 19 VI Knox Co.,KY 36.9 83.8 177. 1976 Feb 4 VI Conasauga,TN 35.0 84.7 178. 1976 Apr 15 V Sacramento,KY 37.4 87.3 179. 1977 Jul 27 V Athens,TN 35.4 84.6 180. 1978 Mar 1 III near Huntsville,AL 34.4 86.6 181. 1978 Oct 27 near Jasper,AL 33.8 87.5 182. 1979 Jan 19 IV Newry,SC 34.7 82.9 183. 1979 Aug 13 V near Cleveland,TN 35.2 84.4 184. 1979 Aug 26 VI Tamasee,SC 34.9 83.1 185. 1979 Sep 12 V Maryville,TN 35.8 84.0 186. 1980 Mar 23 IV Narrows,KY 37.6 86.7 187. 1980 Apr 21 Maryville,TN 35.8 84.0 188. 1980 Jun 25 IV Maryville,TN 35.8 84.0 189. 1980 Jul 12 III near Horse Branch,KY 37.3 87.0

T252-1.doc(Continued) SEQUOYAH PLANT HISTORICAL EARTHQUAKE LISTING 200 MILE RADIUS AROUND 85.1 W LON 35.2 N LAT YEAR MONTH DAY INTENSITY LOCATION NLAT WLON

87. 1928 Mar 7 IV Columbia,TN 35.6 87.0 7.

11. IV 92. 1930 Oct 16 VI Knoxville,TN 36.0 83.9 13. VI 26. <IV 29. <IV 32. IV 41. IV 59. VI 62. IV 73.

76. IV 92. V 116. IV 123. IV 124. IV 127. IV 131. IV 159.

164. V 165.

83. 1926 Jul 8 VII McDowell Co.,NC 35.7 82.0 15. <IV 27. <IV 34. <IV 37. <IV 139. VI 134. 1957 Jul 2 VI Asheville,NC 35.6 82.6 16. V 112. IV 144. IV 13. 1852 Oct 23 <IV Clinton,GA 33.0 83.5 17. <IV 16. 1872 Jun 17 IV Milledgeville,GA 33.1 83.2 24. <IV 38. <IV 79. 1923 Oct 18 IV Hendersonville,NC 35.3 82.5 31. <IV 54. <IV 29. 1882 Oct 15 <IV Murphy,NC 35.1 84.0 33. <IV 149. 1964 Feb 18 V Mentone,AL 34.6 85.6 42. IV 45. 1902 Oct 18 V Chattanooga,TN 35.0 85.3 46. V 50. IV 93. IV 163. 1973 Nov 30 VI Maryville,TN 35.8 84.0 52. <IV 130. IV 150. IV

T252-1.doc(Continued) SEQUOYAH PLANT HISTORICAL EARTHQUAKE LISTING 200 MILE RADIUS AROUND 85.1 W LON 35.2 N LAT YEAR MONTH DAY INTENSITY LOCATION NLAT WLON 162. IV 192. V 194.

195. IV 51. 1913 Jan 1 VII West Springs,SC 34.8 81.8 56. <IV 54. 1913 Apr 17 V Madisonville,TN 35.5 84.4 61. <IV 82. 1924 Nov 13 V Bristol,VA 36.6 82.2 65. IV 129. IV 138. 1958 Oct 20 IV Anderson,SC 34.5 82.7 68. IV 131. 1957 Apr 23 VI Birmingham,AL 33.5 86.8 70. IV 102. VI 121. V 68. 1917 Mar 27 V Jefferson City,TN 36.1 83.5 72.

177. 1976 Feb 4 VI Conasauga,TN 35.0 84.7 82. IV 144. 1963 Apr 11 IV Greenville,SC 34.8 82.4 87. IV 88. 1928 Nov 3 VII Hot Springs,NC 35.9 82.8 96. IV 110. IV 127. 1956 May 27 IV Due West,SC 34.3 82.4 97. IV 100.

132. IV 133. IV 91. 1930 Aug 30 V Kingston,TN 35.9 84.5 167. IV 145. 1963 Nov 14 <IV Nashville,TN 36.2 86.8 103. <IV 113. 1950 Jun 19 IV Tapoco,NC 35.5 84.0 106. IV 110. 1947 Dec 28 IV Ryall Springs,TN 35.0 85.1 109. IV 107. 1945 Jun 14 V Cleveland,TN 35.2 84.9 115. IV 119. 1954 Jan 2 VI Hazard,KY 37.2 83.2 125. IV 151. 1964 Jul 28 <IV Inskip,TN 36.0 84.0 142. <IV 147. 1963 Dec 15 <IV Beechmont,KY 37.2 87.0 153. <IV 164. 1974 Aug 2 V McCormick Co.,SC 33.9 82.5 163. 161. 1971 Jul 13 VI Newry,SC 34.7 82.9 189. IV 162. 1971 Oct 9 V Gatlinburg,TN 35.7 83.5 175. 1975 Nov 25 IV Salem,SC 34.9 83.0 180. IV RIYSIOGRAPHY BY N M. FENNEMAN \ SCALE: 5 10 Yi.. Figure 2.5.1-1 Area (464K33)

LEGEND: . Area SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Figure 2.5.1-3 GEW-CGIC INVESTIGATIONS GEOLOGIC tf4P OF PLANT SITE (822vW1946~3)

I SECTION W--30-00 SECTION W- .3,?+&7 SEC TIGN w'- .Z4..3," - - 6rq: . ,,,, ,,,,d ,,,,,,,, .., .,:.,*-; ..,. . .. ., ,,:..:,- - ... Pa,.:,< & ..,: ,.!, 7, ,.. ..,., kd'k ,.h.. > .,:& nN crr ,my -9 an'. h:.k-81. .I +.?;., ,. . :, I.. .... f I., Ll,C...7.7 i. . :L.. . l ,.I 725 700 675 650 SECTION W- 38f 00 SECTION A-A Lip par, fbm erp~hrll*. 1hw.hrU. /anrW.*d *I* - in was wlitic, .,,4 h/,mt- bond&, Nulh skak SECTION 8-8 ArOrEs: SCALES: 50 0 50 roc . ler- i.d-7-d FINAL SAFETY ANALYSIS REPORT. 2.5.1-6 GE~cGIC SECTIONS W-38+00, AA, (822~1180-2)

I.** .wax? sr).,o brw .A f& I.CI.J~ IIPIII:.~~ 1n8 rr,fbrr,a 'Jet ff- m mYrJldurd 1b1a ond Im~srpu ~41sof8~ tm e*GeCfdJ *r*.*.'ron 1. hc.4 r,i: b8 *nSJwI.r,d m01,rrEI .rd's-Au b &u*ilc;rrat .% 3-t rnwr,, Figure 2.5.1- 8 N-76+OO N-SO+OO (;, 2~1180-4)

Figure 2.5.1-10 S rage (41~1~702)

REACTOR UNIT I 8 E-W BASE~~NE N 65°-00'-~0" w ,, I I N71+00, 38-39 &A C.- W V) r REACTOR 3 I - 0.- .. . I BLDG UNIT I a REACTOR w INTAKE PUMP STA. BLDG UNIT 2 REACTOR BLDG 8 . . N-S BASELINE 37 - -- N 25°-00'-0~" E ESVR #2 ESVR #I w CONDENSATE DEMINERALIZER WASTE EVAP BLDG +- WASTE PACKAGING AREA -=-. A CT PUMP STA 8 6,~" . / 9 16 ' - A A COOLING - TOWER # I I0 ERCW PUMP STA -. . rb I8 SKIMMER WALL COOLING r- SCALE 400 400 800 Feet , , --. - ... - ... I LEGEND A Nx CORE HOLE I IN-SITU SOIL DYNAMICS I FSAR FIG 2.5.1-11 1 I /

-/b"er*UL- bt-bL _- - PmF/LE AZ- AZ 1 I 1 :-@ saw- w IN c.n urar GP-G2 IYm- :-. -, . - >- 0 xurw- / C /-- > FIGURE 2.5.1-12a Excavation & Backfill Category I Structures Revised by Amendment 13 L -- - -. S EQU OY AH N UCLE &R PLANT fmM SAfElY ANALY StS REPORT EXCAVATlON AND BACKFKL - CATEGORY I STRUCTURES. SHEE 3 FIGURE 2.5.1-I2b ~~lll,llll~~~i~~~~~~~~~~l""'~ Revised by Amendment 13 Urn. A SOiL BaRiNO FOR SPLIT-SPMN SIIMPLING 8 SOIL BORIIYG FOR SPLIT-SPMNANU UNOiSlURdCD .WMPLING SOIL BORiffi FOR UNDISTURBED SdMPLING fSOlL PERML48lLlTYl 0 BORING /IOVANCEO 8Y FISH7.4/l/NG n am,~yc ion spi,r-sm~ .W*IPLIM /CWIING 70WtHr /INO COOL WdTR WIVRM CHANNfll @ 8onING F28 SPl/7 SMN AN0 UNDlSTURO£O SAMPLING (COOLING TOWERS AND COOL WATCe WTURN CHANNEII BORlNG FOR SPLIT SPDDN/IND UNOIS7URBED 3.4MPLING - SIC NO7C A (row-rivii nmw~si~ STORAGE FdC,L/TVl n BORING m s~l,rsm~ SMPL~NG - SFF NOTK A /LOW-LEVEL RAOWASTE STOR&<* FdC/l/TYl M WRiW FOR UNOISTURBCD SAMPLING -SEE NOTE A IUW-LfVEL R40W/I57E STOMGE FAClL/7Yl b BDRlW mR SPLIT-SPOON AND UNO/SlURBED SAMPLING - 5°C NOTE A I~OOI~ONAL D/ESCL GtNfRaIOA BUflOINGI 6 FRCW MECH PJPING - - I MoRD,NAlfS II mc INTiRSECnDN Or ref #~S AND EW BASTIINES ART TENNSSEEUIMDE8T CMROIM7EJ 100 IRR- 0 100 - 200 SCALE - FEET A BUR/& FOR SPLIT-SPWN SAMPIIM Q BOR/ffi FOR SPLIT-SWON RND UNDf57URBED SAMPLING O BORlhC FOR UMURRBED S4MPLING n 6ORiNS FOR SPIN-SPOON UIMPLING /COOLING TOWm AND CooL WAZR RETURN CHANNZII 8 SOniK FOR rPL,7-SPOON AN0 UND,SiORB£D S4MPLING (CmLING TUWER3 AND CDaL WATER RfiURN CXANNCLI low m,rWIM; 6 SAMPLiNG Srai/ON 1' /- -- SEQUOYAH NUCLEAR PLANT Q 51 FINAL SAFETY ANALYSIS REPORT FIGURE 2.5.1-130 ESSENTIAL SOIL INVESTIGATIONS IN SITU (KtVIStU BY AMtNUMtNl 13)

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2.5.2-1

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TIME (DAYS)

S3-0toc.doc 3-7 SQN-26 TABLE OF CONTENTS Section Title Page 3.9.3.6 Methods and Results of Blowdown 3.9-38 Analysis (Mechanical) 3.9.3.7 Control Rod Drive Mechanisms 3.9-39 3.9.3.8 Evaluation of Reactor Internals for 3.9-39 Limited Displacement RPV Inlet and Outlet Nozzle Breaks 3.9.4 ADDITIONAL SUPPORT REQUIREMENTS 3.9-41 3.9.4.1 Support Welds 3.9-41 3.9.4.2 Allowable Loads for U-Bolts 3.9-41 and Unistruct Type Clamps 3.

9.5 REFERENCES

3.9-41 3.10 SEISMIC DESIGN OF CATEGORY I INSTRUMENTATION 3.10-1 AND ELECTRICAL EQUIPMENT 3.10.1 SEISMIC DESIGN CRITERIA 3.10-1 3.10.1.1 Instrumentation 3.10-1 3.10.1.2 Electrical Equipment 3.10-3 3.10.2 SEISMIC ANALYSES, TESTING PROCEDURES, 3.10-4 AND RESTRAINT MEASURES 3.10.2.1 Instrumentation 3.10-4 3.10.2.2 Support Structures 3.10-4 3.

10.3 REFERENCES

3.10-8 3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND 3.11-1 ELECTRICAL EQUIPMENT 3.11.1 EQUIPMENT IDENTIFICATION 3.11-1 3.11.2 ENVIRONMENTAL DESIGN AND ANALYSES 3.11-1 3.11.2.1 Environmental Design Criteria 3.11-1 3.11.2.2 Environmental Design Criteria for 3.11-2 ESF Equipment 3.11.2.3 Environmental Design of ESF Components 3.11-2 3.11.2.4 Environmental Design of 10 CFR 50.49 Scope 3.11-2 Equipment 3.11.3 LOSS OF VENTILATION 3.11-2 3.12 CONTROL OF HEAVY LOADS 3.12-1 3.

12.1 INTRODUCTION

/LICENSING BACKGROUND 3.12.1 3.12.2 SAFETY BASIS 3.12-1 3.12.3 SCOPE OF HEAVY LOAD HANDLING SYSTEMS 3.12-2 S3-0toc.doc 3-8 SQN-26 TABLE OF CONTENTS Section Title Page 3.12.4 CONTROL OF HEAVY LOADS PROGRAM 3.12-2 3.12.4.1 SQN Commitments in Response to 3.12-2 NUREG-0612, Section 5.1.1 3.12.4.2 Reactor Pressure Vessel Head (RPVH) 3.12-5 Lifting Procedures 3.12.5 SAFETY EVALUATION 3.12-7 3.13 FLEX RESPONSE SYSTEM 3.13-1 3.13.1 FLEX RESPONSE SYSTEM MITIGATION OF BEYOND 3.13-1 DESIGN BASIS EXTERNAL EVENTS 3.13.2 REFERENCE 3.13-1 S3-0toc.doc 3-9 SQN-18 LIST OF TABLES Number Title 3.2.1-1 Category I Structures 3.2.1-2 Summary of Criteria - Mechanical System Components (Excluding Piping) 3.2.1-3 Electrical Power System Equipment Designed to Operate During and After a "Safe Shutdown Earthquake" 3.2.2-1 Summary of Codes and Standards for Requirements for Sequoyah Mechanical System Components Excluding Piping 3.2.2-2 Summary of Codes and Standards for Requirements for Sequoyah Piping 3.2.2-3 Non-Nuclear Safety Classifications 3.5.1-1 Missiles 3.5.2-1 Missile Characteristics 3.5.5-1 Tornado Missile Spectrum for Category I Structures - Original Design 3.5.5-2 Sequoyah Nuclear Plant - Tornado Missile Spectrum A for Category I Structures 3.5.5-3 Tabulation of Walls and Roofs of Category I Structures Which are Less Than 2 Feet Thick 3.5.5-4 Sequoyah Nuclear Plant - Tornado Missile Spectrum B for ERCW Pumping Station 3.5.5-5 Sequoyah Nuclear Plant - Tornado Missile Spectrum D for Diesel Generator Equipment Doors 3.6.1-1 Piping Systems Inside Containment Where Energy Classifications Differ from Regulatory Guide 1.46 Definition 3.6.2-1 Postulated Design Basis Break Location for LOCA Analysis 3.6.7-1 Summary of Combined Stresses at Break Locations for Main Steam Lines 3.6.7-2 Checklist of Protection Provided Against Unacceptable Consequences of Main Steam and Feedwater Line Ruptures 3.6.7-3 Summary of Combined Stresses at Break Locations for Main Feedwater Lines S3-0toc.doc 3-10 SQN LIST OF TABLES Number Title 3.7.1-1 Category I Structures of the Original Plant Design 3.7.1-2 Periods for Spectral Values 3.7.1-3 Damping Ratios Used in Analysis of Category I Structures, Systems, Components, and Soil for Structures Listed in Table 3.7.1-1 3.7.1-3A Damping Ratios Used in Analysis of Category 1 Equipment, Components, and Their Supports for Structures Listed in Table 3.7.1-1 3.7.1-4 Soil Supported Category I Structures 3.7.1-5 Pile and Caisson Supported Category I Structures 3.7.2-1 Summary of Varied Parameter Ranges Used in the Seismic Analysis of the Ice Condenser Basket Support Frame 3.7.2-2 Category I Structures Affected by Concrete Modulus Change for Structures Listed in Table 3.7.1-1 3.7.2-3 Deleted per Amendment 6 3.7.2-4 Deleted per Amendment 6 3.7.2-5 Deleted per Amendment 6 3.7.2-6 Shield Building Element and Mass Point Properties 3.7.2-7 Shield Building Periods of Natural Modes of Vibration 3.7.2-8 Deleted per Amendment 6 3.7.2-9 Deleted per Amendment 6 3.7.2-10 Interior Concrete Structure Element Properties 3.7.2-11 Interior Concrete Structure Mass Point Properties 3.7.2-12 Interior Concrete Structure Periods for Natural Modes of Vibration 3.7.2-13 Steel Containment Vessel Element Properties 3.7.2-14 Steel Containment Vessel Mass Point Properties 3.7.2-15 Steel Containment Vessel Periods for Natural Modes of Vibrations S3-0toc.doc 3-11 SQN LIST OF TABLES Number Title 3.7.2-16 Auxiliary-Control Building Element Properties 3.7.2-17 Auxiliary-Control Building Mass Point Properties 3.7.2-18 Auxiliary-Control Building Periods for Natural Modes of Vibration 3.7.2-19 Element Properties Additional Equipment Building Unit 1 3.7.2-20 Element Properties Additional Equipment Building Unit 2 3.7.2-21 Mass Point Properties Additional Equipment Building Unit 1 3.7.2-22 Mass Point Properties Additional Equipment Building Unit 2 3.7.2-23 Normal Modes of Vibration Additional Equipment Building Unit 1 3.7.2-24 Additional Equipment Building - Unit 2 3.7.2-25 Pumping Station Element Properties 3.7.2-26 Pumping Station Mass Point Properties 3.7.2-27 Pumping Station Periods for Natural Modes of Vibration 3.7.2-28 ERCW Pumping Station Element Properties for Mathematical Model 3.7.2-29 ERCW Pumping Station Mass Point Properties for Mathematical Model 3.7.2-30 ERCW Pumping Station Normal Modes of Vibration 3.7.2-31 Diesel-Generator Building Element Properties 3.7.2-32 Diesel-Generator Building Mass Point Properties 3.7.2-33 Diesel-Generator Building Periods for Normal Modes of Vibration 3.7.2-34 Waste Packaging Area Element Properties 3.7.2-35 Waste Packaging Area Mass Point Properties 3.7.2-36 Waste Packaging Area Stiffness of Soil Springs 3.7.2-37 Condensate Demineralizer Waste Evaporator Building Stiffness for Soil Springs S3-0toc.doc 3-12 SQN-24 LIST OF TABLES Number Title 3.7.2-38 Condensate Demineralizer Waste Evaporator Building Element Properties 3.7.2-39 Condensate Demineralizer Waste Evaporator Building Mass Point Properties 3.7.2-40 Condensate Demineralizer Waste Evaporator Building Natural Periods of the Structural Model 3.7.2-41 East Steam Valve Room Element Properties 3.7.2-42 East Steam Valve Room Element Properties 3.7.2-43 East Steam Valve Room Mass Point Properties 3.7.2-44 East Steam Valve Room Spring Constants for Combined Caisson-Soil System 3.7.2-45 East Steam Valve Room Frequency Comparison for Three Soil Cases Horizontal Motion 3.8.1-1 Loading Combinations and Allowable Stresses for the Shield Building 3.8.1-2 Shield Building Equipment Hatch Doors and Sleeves Loads, Loading Combinations, and Allowable Stresses 3.8.2-1 Allowable Stress Criteria - Containment Vessel 3.8.2-2 Loading Combinations for Various Plant Conditions 3.8.3-1 Loading Combinations and Allowable Stresses for the Interior Concrete Structure 3.8.3-2 Loading Combinations and Load Factors 3.8.3-2A Loading Combinations, Load Factors and Allowable Stresses for SG Compartment Roof Modification (5)(6) 3.8.3-3 Seals Between Upper and Lower Compartments 3.8.3-4 Personnel Access Doors in Crane Wall 3.8.3-5 Ice Condenser Allowable Limits 3.8.3-6 Original Design Stress Margin Table 3.8.3-1 Criteria Versus Table 3.8.3-2 Criteria 3.8.3-7 Equipment Access Hatch S3-0toc.doc 3-13 SQN LIST OF TABLES Number Title 3.8.3-8 Escape Hatch - Divider Barrier Floor - Load Combinations - Allowable Stresses 3.8.3-9 Air Return Duct Penetration 3.8.3-10 Maximum Stress - Summary (DBA x 1.2) Per Table 3.8.3-1 Criteria 3.8.3-11 Selection of Steels in Relation to Prevention of Non-Ductile Fracture of Ice Condenser Components 3.8.4-1 Auxiliary Control Building Concrete Structure Loads, Loading Combinations, and Allowable Stresses 3.8.4-2 Auxiliary Control Building Structural Steel Loads, Loading Conditions, and Allowable Stresses 3.8.4-3 Condenser Cooling Water Intake Pumping Station - Loading Cases, Allowable Stresses, Factors of Safety and Material Properties

3.8.4-4 Retaining Walls - Loading Cases, Allowable Stresses, Factors of Safety, and Material Properties 3.8.4-5 Control Room Shield Doors - Loads, Loading Combinations, and Allowable Stresses 3.8.4-6 Auxiliary Building Railroad Access Hatch Covers - Loads, Loading Combinations, and Allowable Stresses 3.8.4-7 Railroad Access Door - Loads, Loading Combinations, and Allowable Stresses 3.8.4-8 Manways in RHR Sump Valve Room - Loads, Loading Combinations, and Allowable Stresses 3.8.4-9 Pressure Confining Personnel Doors - Loads, Loading Combinations, and Allowable Stresses 3.8.4-10 Diesel Generator Building Doors and Bulkheads - Loads, Loading Combinations, and Allowable Stresses 3.8.4-11 Diesel Generator Building - Loads, Loading Combinations, Allowable Stresses and Material Properties 3.8.4-12 Primary and Refueling Water Pipe Tunnels - Loads, Loading Combinations, Allowable Stresses, and Material Properties S3-0toc.doc 3-14 SQN-25 LIST OF TABLES Number Title 3.8.4-13 Class 1E Electrical Systems Structures - Loads, Loading Combinations, Allowable Stresses, and Material Properties 3.8.4-14 East Steam Valve Room - Loads, Load Combinations, Allowable Stresses, Factors of Safety and Material Properties 3.8.4-15 East Steam Valve Room - Structural Steel - Loading Combinations and Allowable Stresses for Structural Steel 3.8.4-16 ERCW Pumping Station Loads and Loading Combinations 3.8.4-17 Refueling Water Storage Tank Foundation 3.8.4-18 Spent Fuel Pool Gates Loads and Loading Combinations 3.8.6-1 Polar Cranes - Loads, Loading Combinations, and Allowable Stresses 3.8.6-2 Deleted 3.9.2-1 Codes and Other Criteria Governing the Analysis of TVA Class B, C, and D Components 3.9.2-2 Design Loading Combinations for Group Classes B, C, and D Components 3.9.2-3 Safety Class B, C, and D Component Loading Conditions and Stress Limits 3.9.2-4 Loading Combinations and Stress Limits for Safety Class B, C, and D Piping 3.9.2-5 Loading Combinations and Stress/Loading Limits for Safety Class B, C, and D Supports 3.9.3-1 Maximum Deflections for Reactor Internals Under Blowdown and Seismic Excitation 3.10.2-1 Load Combinations and Allowables for Cable Tray Supports 3.11.1-1 Electrical and Mechanical Equipment Required to Function During and/or After an Accident S3-0toc.doc 3-15 SQN-18 LIST OF FIGURES Number Title 3.3.2-1 Variations of Differential Pressure and Tangential Plus Translational Velocity as a Function of the Distance from the Center of the Tornado 3.5.2-1 Ice Condenser Lower Inlet Door Opening, Typical Missile Trajectory Orientation 3.5.2-2 Location of Main Steam and Main Feedwater Containment Isolation Valves at Azimuth 0 3.5.2-3 Location of Main Steam and Main Feedwater Containment Isolation Valves at Azimuth 180 3.5.4-1 Comparison of Missile Formulas 3.5.4-2 Comparison of Missile Formulas 3.5.4-3 Depth of Missile Penetration for Tornado 3.5.4-4 Depth of Missile Penetration for Tornado 3.6.1-1 High Energy Classification Per Sequoyah Nuclear Plant Pipe Rupture Analysis Definition and Per Regulatory Guide 1.46 Definition 3.6.2-1 Location of Postulated Breaks 3.6.4-2 Jet Expansion Models 3.6.7-1 Steam Generators 1 and 4 Postulated Break Locations and Fixes 3.6.7-2 Steam Generators 2 and 3 Postulated Break Locations and Fixes 3.7.1-1 Mathematical Model for Soil-Structure Interaction 3.7.1-2 ERCW Access Dike 3.7.2-1 Mathematical Model of Reactor Internals 3.7.2-2 First Mode of Vibration of Reactor Internals 3.7.2-3 Model of Horizontal Lattice Frame Structure 3.7.2-4 Group of Six Interconnected Lattice Frames S3-0toc.doc 3-16 SQN LIST OF FIGURES Number Title 3.7.2-5 Typical Model of Lattice Frame 3.7.2-6 Typical Multi-Level Horizontal Dynamic Model of Lattice Frame Basket Assembly 3.7.2-7 Lattice Frame Ice Basket Gap 3.7.2-8 Typical Displacement Time Histories for 12 ft. Basket With End Supports - Pluck Test 3.7.2-9 Typical Crane Wall Acceleration 3.7.2-10 Typical Crane Wall Velocity 3.7.2-11 Typical Crane Wall Displacement 3.7.2-12 Typical Ice Basket Displacement Response 3.7.2-13 Typical Ice Basket Impact Force Response 3.7.2-14 Typical Crane Wall Panel Load Response 3.7.2-15 Nonlinear Dynamic Model Results Wall Panel Seismic Load Versus Gap Size 3.7.2-16 Nonlinear Dynamic Model Results Wall Panel Seismic Load Versus Gap Size 3.7.2-17 Nonlinear Dynamic Model Results Wall Panel Seismic Load Versus Gap Size 3.7.2-18 Nonlinear Dynamic Model Results Wall Panel Seismic Load Versus Gap Size 3.7.2-19 Nonlinear Dynamic Model Results Wall Panel Seismic Load Versus Gap Size 3.7.2-20 Nonlinear Dynamic Model Results Wall Panel Seismic Load Versus Gap Size 3.7.2-21 Nonlinear Dynamic Model Results Wall Panel Seismic Load Versus Gap Size 3.7.2-22 Nonlinear Dynamic Model Results Wall Panel Seismic Load Versus Gap Size 3.7.2-23 Nonlinear Dynamic Model Results Seismic Impact Load Versus Gap Size S3-0toc.doc 3-17 SQN LIST OF FIGURES Number Title 3.7.2-24 Nonlinear Dynamic Model Results Seismic Impact Load Versus Gap Size 3.7.2-25 Nonlinear Dynamic Model Results Seismic Impact Load Versus Gap Size 3.7.2-26 Nonlinear Dynamic Model Results Seismic Impact Load Versus Gap Size 3.7.2-27 Nonlinear Dynamic Model Results Seismic Impact Load Versus Gap Size 3.7.2-28 Nonlinear Dynamic Model Results Seismic Impact Load Versus Gap Size 3.7.2-29 Nonlinear Dynamic Model Results Seismic Impact Load Versus Gap Size 3.7.2-30 Nonlinear Dynamic Model Results Seismic Impact Load Versus Gap Size 3.7.2-31 Original Accelerogram 3.7.2-32 Accelerogram After Integration and Differentiation 3.7.2-33 Nonlinear Dynamic Model 3.7.2-34 Section Through Reactor Shield Building Looking West, Lumped Mass Model for Dynamic Analysis 3.7.2-35 Flowchart of Operations for Response of the Dome 3.7.2-36 Shell Model for Dome Analysis - Shield Building 3.7.2-37 Lumped Mass Model for Dome Analysis - Shield Building 3.7.2-38 Reactor Building, Interior Concrete Structure Sectional Elevation Looking West, Lumped Mass Model for Dynamic Analysis 3.7.2-39 Steel Containment Vessel, Lumped Mass Model for Dynamic Analysis 3.7.2-40 Steel Containment Vessel, Finite Element Model 3.7.2-41 Sectional Elevation of Auxiliary Control Building Lumped Mass Model for Dynamic Analysis S3-0toc.doc 3-18 SQN LIST OF FIGURES Number Title 3.7.2-42 Additional Equipment Building - Unit 1 3.7.2-43 Additional Equipment Building - Unit 2 3.7.2-44 Sectional Elevation of Intake Pumping Station, Lumped Mass Model for Dynamic Analysis 3.7.2-45 ERCW Pumping Station - Model for Dynamic Analysis Sectional Elevation 3.7.2-46 ERCW Pumping Station - Key Plan 3.7.2-47 Sectional Elevation of Diesel Generator Building, Lumped Mass Model for Dynamic Analysis 3.7.2-48 Sectional Elevation of Waste Packaging Area, Lumped Mass Model for Dynamic Analysis 3.7.2-49 Response Acceleration Spectra 3.7.2-50 Condensate Demineralizer Waste Evaporator Building - Mathematical Model for Dynamic Analysis 3.7.2-51 Averaged Ground Response Spectrum 3.7.2-52 Concrete Caisson 3.7.2-53 East Valve Room - Lumped Mass and Spring Model 3.7.2-54 Interior Concrete Structure N-S Translational Motion, Translational Mode 1 3.7.2-55 Interior Concrete Structure N-S Translational Motion, Translational Mode 2 3.7.2-56 Deleted by Amendment 6 3.7.2-57 Interior Concrete Structure E-W Translation Plus Torsion, Translational Mode 1 3.7.2-58 Interior Concrete Structure E-W Translation Plus Torsion, Translational Mode 2 3.7.2-59 Interior Concrete Structure E-W Translation Plus Torsion, Translational Mode 3 S3-0toc.doc 3-19 SQN LIST OF FIGURES Number Title 3.7.2-60 Interior Concrete Structure E-W Translational Plus Torsion, Translational Mode 4 3.7.2-61 Interior Concrete Structure E-W Translational Plus Torsion, Translational Mode 5 3.7.2-62 Interior Concrete N-S Translation Response, Maximum Acceleration, Safe Shutdown Earthquake 3.7.2-63 Interior Concrete N-S Translation Response, Maximum Deflection, Safe Shutdown Earthquake 3.7.2-64 Interior Concrete N-S Translation Response, Maximum Shear, Safe Shutdown Earthquake 3.7.2-65 Interior Concrete N-S Translation Response, Maximum Bending Moment, Safe Shutdown Earthquake 3.7.2-66 Interior Concrete E-W Translation Plus Torsion Response, Maximum Acceleration, Safe Shutdown Earthquake 3.7.2-67 Interior Concrete E-W Translation Plus Torsion Response, Maximum Deflection, Safe Shutdown Earthquake 3.7.2-68 Interior Concrete E-W Translation Plus Torsion Response, Maximum Shear, Safe Shutdown Earthquake 3.7.2-69 Interior Concrete E-W Translation Plus Torsion Response, Maximum Bending Moment, Safe Shutdown Earthquake 3.7.2-70 Interior Concrete E-W Translation Plus Torsion Response, Maximum Angular Acceleration, Safe Shutdown Earthquake 3.7.2-71 Interior Concrete E-W Translation Plus Torsion Response, Maximum Angular Displacement, Safe Shutdown Earthquake 3.7.2-72 Interior Concrete E-W Translation Plus Torsion, Response Maximum Torque, Safe Shutdown Earthquake 3.7.2-73 Response Acceleration Spectrum, Reactor Building Interior Concrete, NS Mass Point No. 1, Safe Shutdown Earthquake 3.7.2-74 Response Acceleration Spectrum, Reactor Building Interior Concrete, EW Mass Point No. 1, Safe Shutdown Earthquake 3.7.2-75 Response Acceleration Spectrum Reactor Building Interior Concrete, NS Mass Point No. 6, Safe Shutdown Earthquake S3-0toc.doc 3-20 SQN-18 LIST OF FIGURES Number Title 3.7.2-76 Response Acceleration Spectrum Reactor Building Interior Concrete, EW Mass Point No. 6, Safe Shutdown Earthquake 3.7.2-77 Flow Chart for Development of Floor Response Spectra 3.7.2-78 Comparison of Time History and Response Spectrum Response 3.7.2-79 Replacement Steam Generator One-Stick Model 3.8.1-1 Temperature Gradient 3.8.2-1 Structural Steel Containment Vessel 3.8.2-2 Layout of Containment Shell 3.8.2-3 Containment Panel Analysis - Finite Element Model 3.8.2-4 Pressure - Time Function of Local Detonation 3.8.2-5 Graphical Representation of Pressure Function 3.8.2-6 Hydrogen Explosion - Containment Panel Analysis - Finite Element Model of Shell Segment 3.8.2-7 Structural Steel Containment Vessel Anchor Bolt Plan and Base Details 3.8.2-8 Steel Containment Vessel, Lumped Mass Model for Dynamic Analysis 3.8.2-9 Steel Containment Vessel, Finite Element Model 3.8.2-10 Steel Containment Vessel, Finite Element Model 3.8.2-11 Expansion Bellows For Personnel Locks 3.8.3-1 Seals Between Ice Condenser and Containment Vessel Arrangement 3.8.3-2 Temperature Gradient 3.8.4-1 Concrete Floor Design Data 3.8.4-2 D. G. Building Doors and Bulkhead Arrangement 3.8.4-3 Concrete-Pipe Tunnels and Tank FDNS Outline 3.8.4-4 Concrete-Pipe Tunnels and Tank FDNS Outline S3-0toc.doc 3-21 SQN LIST OF FIGURES Number Title 3.8.4-5 Category 1 Yard Electrical Manholes and Handholes 3.8.4-6 Concrete-Manholes and Handholes - Outline 3.8.4-7 Concrete - ERCW Pumping Station and ERCW Channel 3.8.4-8 Concrete - ERCW Pumping Station 3.8.4-9 Concrete ERCW Skimmer Wall and Underwater Dam 3.8.4-10 Concrete ERCW Discharge Box 3.8.4-11 Spent Fuel Pool Gate 3.8.6-1 RB 175 Ton Polar Cranes 3.8.6-2 RB 175 Ton Polar Cranes, Trolley 3.8.6-3 RB 175 Ton Polar Cranes, Trolley 3.8.6-4 RB 175 Ton Polar Cranes, Bridge 3.8.6-5 RB 175 Ton Polar Cranes 3.8.6-6 AB 125 Ton Crane 3.8.6-7 AB 125 Ton Crane Trolley 3.8.6-8 AB 125 Ton Crane Trolley 3.8.6-9 AB 125 Ton Crane Bridge 3.8.6-10 AB 125 Ton Crane Limit Switch 3.8.6-11 AB 125 Ton Crane Mechanical Stop 3.8.6-12 Sectional Elevation View Through the Spent Fuel Pit 3.9.1-1 Upper Internals Assembly 3.9.1-2 Upper Internal Support Model 3.9.1-3 Computer Geometry Plot of Lower Internals Support Model 3.9.1-4 Lower Internals Support Structure Comparison Between Experimental and Theoretical Vertical Deflections S3-0toc.doc 3-22 SQN LIST OF FIGURES Number Title 3.9.1-5 Reactor Vessel and Internals Vibration 3.9.1-6 Thermal Shield, Mode Shape n=4 Obtained from Shaker Test 3.9.1-7 Thermal Shield, Maximum Amplitude of Vibration During Preoperational Tests 3.9.1-8 Time-History Dynamic Solution for LOCA Loading 3.9.2-1 Reactor Coolant Piping Pressurizer Surge Line

SQN-26 Table 3.2.1-2 (Sheet 5) SUMMARY OF CRITERIA - MECHANICAL SYSTEM COMPONENTS (EXCLUDING PIPING) Scope Safety Class Code QA Reqd Location Rad Source Seismic. Component (1) (2) (3) (4) (5) (6) Category( 7)

Control Building Ventilation

• Fans (cleanup, pressurizing) T (10) AMCA X CB - I

• Filters T (10) (18) X CB P I

• Air Conditioning Unit (Elec. Board Room) T (10) - X CB - I Air Conditioning Unit (MCR) T (10) - X CB - I Diesel Building Ventilation Fans T (10) AMCA X DB P I Main Steam System

• Isolation Valves T B P&V-II X AB - I

• Isolation Bypass Valves T B III-2 X AB - I Feedwater System

• Stop Valves T B P&V-II X AB - I Auxiliary Feedwater System Auxiliary Feedwater Pumps

• Motor Driven T C P&V-III X AB - I

• Steam Turbine Driven T C P&V-III X AB - I Feedwater System

• Steam Generator W B III-B X C - I Feedwater Nozzle Thermal Liner (Unit 2, Loops 2, 3, & 4; Unit 1, Loops 1, 2, 3, & 4)

S/G Main Steam System Relief Valves T B P&V-II X AB - I Safety Valves T B P&V-II X AB - I Spent Fuel Pool Cooling and Cleanup System

• Spent Fuel Pool Heat Exch. (Tube) W C III-C X AB X I (Shell) W C VIII X AB - I

S3-03.doc 3.3-2 SQN-26 negative differential pressure defined as a trapezoidal step function, its magnitude varies from zero to -3 psi in 3.0 seconds, stays at -3 psi for 3 seconds then decreases to zero in 3.0 seconds, and the tornado-generated missiles described in Section 3.5. These loadings are considered to act concurrently. Coincident wind velocities and pressure drops for the design tornado are shown in Figure 3.3.2-1. The relationship between wind velocity and pressure in the design tornado shown in Figure 3.3.2-1 was developed based on Hoecker's studies of the Dallas tornado of 1957 (References 2 and 3).

3.3.2.2 Determination of Forces on Structures The methods used to convert the tornado loadings into forces on Category I structures, including the distribution across the structures, was determined by following the recommendations of ASCE Paper 3269, "Wind Forces on Structures" as outlined in Section 3.3.1.4. The provisions for gust factors and variation of wind velocity with height are not applied. The dynamic wind pressure, q, is defined as q = .00256V2, where q is in lb/ft2 and V is in mph. The wind pressure, p in lb/ft2, is defined as p = Cq, where C is the shape coefficient (CD).

A 1.3 shape coefficient is included for box-shaped structures with vertical walls normal to the wind direction. The dynamic pressure load, p = 1.3q, due to tornadoes is applied to the structure walls and roof in the same manner as the wind loads in Section 3.3.1.

Cylindrical structures and tanks have the same shape coefficients applied as for wind loads in Section 3.3.1. The pressures are applied over the structures as shown in Table 4(f) of ASCE Paper 3269. The effect at various combinations of tornado loadings were studied with respect to each Category I structure. The most adverse combination was selected individually for the design basis of each structure. The tornado loadings are not considered to be coincident with accident or earthquake loadings.

Venting is utilized to reduce the effective tornado-generated differential pressure in portions of the Auxiliary Building. Four hundred square feet of relief panel area are provided in the roof over the Spent Fuel Pool Room and Cask Loading Room at Elevation 791.75 for venting purposes during the tornado. The relief panels are held in place by gravity. An upward pressure of 0.25 lb/in2 is sufficient to offset the weight of the panels and cause them to be lifted from their normal positions. Two corners of each panel are chained to the roof to prevent the panel from becoming a missile after it relieves.

The Shutdown Board Room and, in general, the area between columns q and u at Elevation 734.0 are not part of that portion of the Auxiliary Building intentionally vented for tornado depressurization; however, this area and the remainder of the building will depressurize through the vent area provided by the air intake openings, through ventilation penetrations, and through the 734 foot elevation equipment hatch. In addition, the ERCW Pumping Station depressurizes due to the vent areas provided by the ventilation openings. The Diesel Generator Building depressurizes due to the vent areas provided by the ventilation openings and a Main Control Room manual action to open the Air Intake Dampers during a tornado event.

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S3-06.doc 3.6-23 SQN-26 3. Render inoperative any engineered safeguard system.

4. Cause failure of any other steam or feedwater line that could result in an uncontrolled blowdown of more than one steam generator.

5. Reduce flow capability of the Auxiliary Feedwater System below minimum requirements.

3.6.7.6.3 Postulated Break Locations Figures 3.6.7-1 and 3.6.7-2 show the approximate routing of the main steam lines from each steam generator to the flued head anchors in the outside wall of each main steam valve room. Also shown in these figures are locations of the break points postulated for each main steam piping run considered in this report. Table 3.6.7-1 shows the combined stress values for the postulated main steam line ruptures. (Note: The stress values listed are for example and historical purposes only) 3.6.7.6.4 Protective Measures Pipe rupture restraints, sleeves, and jet deflectors are used to mitigate the consequences of postulated main steam line ruptures.

Figures 3.6.7-1 and 3.6.7-2 indicate the approximate locations and the type of mitigation component for each postulated break location. Table 3.6.7-2 is a checklist showing that protection is provided for all unacceptable consequences of main steam line ruptures. 3.6.7.7 Main Feedwater 3.6.7.7.1 Description of Piping System The feedwater system is designed to supply a sufficient quantity of feedwater to the steam generator secondary side inlet during all normal operating conditions. The feedwater system delivers water to the steam generators at an elevated temperature and pressure. These lines are high energy from the feedwater pumps to the inlet of the steam generators.

The major portion of the feedwater piping evaluated herein is 16" OD lines designated as a TVA Class B system. The evaluated portion begins at the flued anchor at the exterior wall of the main steam room. The piping continues through the valve rooms (and isolation valves) and through guard pipes and bellows, penetrates the shield wall, the containment liner, crane wall, and terminates at the inlet of the steam generator. The piping termination at the Unit 2, Loops 2, 3, and 4 and Unit 1, Loops 1, 2, 3, and 4 steam generator feedwater nozzle is a specially designed elbow with an integral thermal liner for the purpose of mitigating thermal fatigue cracking. For further information regarding the elbow/liner, see Table 3.2.1-2.

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AUXILIARY CONTROL BUILDING STRUCTURAL STEEL LOADS, LOADING CONDITIONS, AND ALLOWABLE STRESSES

1

HYWSTAT/C LOAD - CASE lT Yydrostefic /d onvrr when pressure and hmp st nw-1 m confa,nment o~erdhno rrsM m-d,?i-- and annu/- - CASE /B 4 aot/om OF ice ~ed HYDROSTATIC LOAD - CASE 8 18 SEISM/C PRESSRE Hpierat,c temp *r boeh had oppcr occur, and ."/,en /ewer p.eS5-urr c"mnpa,fm"LT 8nd a- /t'pr, and 170.K A//owLi/e dess sk// mflorrn fo /ordordi,g cmdifion /or tdbie / in EE CoNDENSEp DUCT LMDS append?l A oP fbe rpczPiidiiini IYWM aoEPATiN6 CCYVO/NONS INS& rm cmr-nt-mr vurt~ Mar ~rarzurs 03orro -180' ANNULUS coi~apse pru 050pa~g oesrg,, /enip- , summer /?O' , , Wlntar 60' Sprel h&s 5~,"9 /me N 799 Ger,gn kmp 60F LOWER MWAPTEENT UPPER COM&R bsiniperi anrnufh ANNULUS "em 0.pr.q TBmserafw* 60' to /CO'F A-A CO~~AINMEN~ otjn~ PRE~UREES C% TEMPER~IUR~S L/N/TI AS SMWN UN/T 2 OPP HMO SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 3.8.2-1 STRUCTURAL STEEL CONTAINMIENT VESSEL (REVISED BY AMENDMENT 13) CON r*n "NIT AN" PART "i THF PnorAoAM nrr*n*ci

'I I TOP DECK OOORS-, I INTERMEDIATE DECK DOORS~I L= TIGHT SEAL 0.ETWEEN LINER AND COMPARTMENTS FIGURE 3.8.2-2 LAY OUT of CONTAINMENT SHELL Revised by Amendment 13 Elev. 778'06" CONIAI-T PANEL ANALYSIS FINIK EL= MODEL FIGURE 3.8.2-3 Revised by Amendment 13 0 w- r4 0, - I 4 f 1 8- r( 0- cD 0- u 0 - 2 . .- .- 2.. N 0' 1 I I 1 1 2 1 -5 *r ' TIHE sec ) 3.8.2-4 PBESSURE - TZME FUNCTION OF LOCAL DETONATION Revised by Amendment 13 A-A - FIGURE 3.8.2-5 GRAPHICAL REPRESENTATI ON OF FUNCTION H- Revised by Amendment 1 3 3.8.2-6 r-" SEOUOYRH NUCLEAR PLANT---HYDROGEN EXPLOSION---CONTRINHENT PRNEL RNALYSIS FINITE EXBENT MODEL OF SXELL SEGMENT e- Revised by Amendment 13

3.82-8 Steel Containment Vessel, Lumped Mass Model Dynamic Analysis Revisdby Amendment 13 3.8.2-9 Steel Containment Vessel ni te El emen t Model H~ Revised by Amendment 13 3 3.2- 10 Containment Vessel Finite Model r* ' Revised by Amendment 13 RAD 62'-6 1/2" -__----- 2'-9 1/4" 7 I1 I! W' G AIRLOCK-- ' I t' m A ---- -- - FOUNDATION ELEVATION OF BELLOWS FOR PERSONNEL LOCKS SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 3.8.2-11 EXPANSION BELLOWS FOR PERSONNEL LOCKS (REVISED BY AMENDMENT 13)

El 28'-4' KEYNr!LAN El DEVELOPED INTERIOR ELEVATION UNIT 1 SHOWN UNIT 2 SIMILAR d OPPOSITE HANO CONTAINMENT ONTAINMENT ESSLL SHELL h EL 797 -8 I/2' VENT BAR. CLAMP. AND ............... CURTAIN BY WESTINGHOUSE NORMAL OPERATING CONDITIONS DIFFERENTIAL PRESSURE ON SEAL NLGLIGIBLF PSI M+,X,,""M TLMPER*T"RL ........................., 20. r .............................. EL 728' 3 1/4' D1-D1 R*OI*T,ON RATE 2 R*OB,,,R RAOIATlON FOR 40-YR PROJECT LlrL............ 7 0 X 10' RADS EL 728 -4' EI-EI ACCIDENT CONDITION FOR 12-HR SEAL DESIGN LIFE MAY MOVEMENT OF VESSEL RELATIVE OIFIERLNTIAI PRESSURE ON SEAL ........... ......................... TO CONCRETE AT EL 737'-0'-798'-0' (LOWER COMPARTMENT PRESSURIZED1 15 PSI ALIGN SURFACES TO TOR COMBlNlD ACCIDENT AND MAXI,,,",, TEMPERATURE 327. F ......................... EARTHQUAKE CONDITION. ALIGN SU RADIATION FOR 12 HR 8 0 X 10' RAOS ?I/,' 5A POSSIBLE MAX MOVEMENT OF VLSSFL RLLATlVF TO CONCRETE AT VESSEL SHELL TlP SECTION AT VENT CURTAIN CATEGORY I SEAL FOR COMBINED ACCIDENT AND EARTHQUAKE CONDITION F WELOLO SEAL hS SECTIONS Al-A1 . B1-B1 & C1-C1 Dl-Dl & El-El A1 -Al AS SHOWN Cl-CI AS SHOWN ROTATED 90' CW 01-01 AS SHOWN AND NOTED 81-B1 OPP HAND. ROTATED 90' CCW El-EI AS NOTED SEAL NOT SHOWN DETAIL A-1 TYPICAL CONFIGURATION SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 3.8.3-1 SEALS BETWEEN ICE CONDENSER AND CONTAINMENT VESSEL ARRANGEMENT (REVISED BY AMENDMENT 13) THIS CONTICURATION CONTROL DRAWING IS MAINTAINED BY THE SON CAD UNIT &NO IS PART OF THC TYA PRCCADAM DATABASE 3.8.3-2 TEMPERA'IZTRE GRADIENT Revised by Amendment 13

/ d:SUYED *.;:.C, % hmrn P H/N6F PIN N- S BAS£ LINE B S REACrOnS ,-> 7---- I 90'- I S~CT/ON A -A Il 6. BACK 70 a*cx fM&EDaFD SrWL . - - -- - . -- CC CC , \I E -B ,&- MOYEYtNr rm TWS LEAF \ , RESTRKrED Br £NCIOSuR* D -D DfPU.4 A TIPICAI MDUNI,NG KEY PLAN \ <L FOR PERSONNEL DWR \'. REMOVABLE BUlUHtAD sru mr e 6xccpr SECONDARl LCAF r 8 ELEVATION C-C EL f VATION f - E LLEvAr/oN F-F SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 3.8.4-2 D.G. BUILDING DOORS AND BULKHEAD ARRANGEMENT (REVISED BY AMENDMENT 13) SON CAD UNIT &NO 15 PART DT THC TYA PROCADAM DATABASE

z 0 L t tL. V) CSI z 3 a z P"4 M M r

• P sg ~3 SU' 03 # a SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 3 8 4-5 CATEGORY 1 YARD ELECTRICAL MANHOLES AND HANDHOLES (REVISED BY AMENDMENT 13) bd 7'-0' El L PLAN - TOP SLAB SYM ABOUT C \ -FIN GRADE 3'-4 ,,2' IRAME d COVER 7 *'.9 ~oYl'IBC,AH~ ,\kAECTRICAL MANHOLES PLAN - BASE SLAB T@F- -MISSILE PROTECTION ,E-W BASE _LIME & c U"/7 / I SEQUOYAH NUCLEAR PLANT I FINAL SAFETY I ANALYSIS REPORT I I FIGURE 3.8.4-6 I CONCRETE-MANHOLES & HANDHOLES- OUTLINE (REVISED BY AMENDMENT 13) TIilE COIIFICURATIOH CONTROL ORA.INC 15 MAINTAINED 81 THE en., .." , ,.,.. ..m .e "."? "r .,.- .,,. ""-."... "...".*7 I I CAD MAINTAINED DRAWING I I ERCW PUMPiNG STATION PART FL AN EL 685.0 NT5 SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 3.8.4-8 CONCRETE-ERCW PUMPING STATION (REVISED BY AMENDMENT 13) SON CAD UNIT &NO 15 PART DT THC TYA PROCADAM DATABASE C 678 5 ~75----_-- -_----- A5 - A5 85- B5 NATURAL SLOPE, PLAN OF DIKE AT ERCW PUMG STATION Cp~*7 I032 STOM UaR SLAB VM Ca41PAC17OY REWIRED SEQUOYAH NUCLEAR PLANT DETAIL A5 B B5 FINAL SAFETY ANALYSIS REPORT &5-*5 FIGURE 3.8.4-9 CONCRETE ERCW SKIMMER WALL & UNDERWATER DAM (REVISED BY AMENDMENT 19)

JUTION AZAZ & 82 82 c2 J AP-A2 dS JNMV 82-82 OPPHINO SECIIDNAL PLAN 5. STPUCTL TnE tKCW3ISCXARGEBOX G* AND ?UAL /TI' IS A5SURPNCL 4 CArELOfiYZ /5 DdOUIRED, SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 3.8.4-10 CONCRETE ERCW DISCHARGE BOX (REVISED BY AMENDMENT 13)

CCATE AND SLOT NOTE DESIGN BASED ON WATER ASSUMED 4 LOAD PLUS A001TIONAL PRFSSURF $-*--- DUE TO EARTHOUAKE. /' B AIR SUPPLY INLETS TOP OF GATE EMBLDOFO PL SUPPORT EL 700'-7 PRESSURE DIAGRAM FLOOR EL 754'-0' KEY PLAN B P w," e> UPPER GATE GUTOF OUTSIDE SEAL RETAINERS SHOWN REMOVABLE Y-Y AIR SUPPLY &UPPER MIDOLL GUlDF EL 711'-3 13/16' DETAIL B H-H I SEAL LOWER GATE GUIDE SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 3.8.4-11 A-A DETAIL C SPENT FUEL POOL GATE GATE MK 1 (REVISED BY AMENDMENT 13) THIS CONTICURATION CONTROL DRAWING IS MAINTAINED BY THE SON CAD UNIT &NO IS PART OF THC TYA PRCCADAM DATABASE I z-z SEALS NOT SHOYN

ELEVATION HWKS AND RIEVINC NOT SHOWN CROSS SHAFT NOT SHOWN SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 3.8.6-2 RB 175 TON POLAR CRANES TROLLEY (REVISED BY AMENDMENT 13) THIS CONTICURATION CONTROL DRAWING IS MAINTAINED BY THE DOUBLE END DOUBLE ACTING 30 STROKE /W":!~"I"N':FRISZ~~:N"O'~~~~~R'N'A"? FcGRoLs. SEE SPECIFICATION r -I 2 -8 B B A FOR "E "om REEVING SW ABOUT C r D MAIN HOOK DETAIL E-E 35 TON HOOK OUTLINE ---SHEAVE 21' MIN DIA C t DOUBLE END, DOUBLE ACTING 20 STROKE HIDPLULlC CYLINDER. PIS-ON TYPE SECTION A-A ;::ks;:E;;~;~,:;; EXTERNAL CONTROLS SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 3.8.6-3 RE 175 TON POLAR CRANES TROLLEY (REVISED BY AMEhDMENT 13) THIS CUNIICUPATIVI CONTROL DRIYINC 15 WAINTIINFO BY THE i- DR,, PAN B-B HEAD CAPSCREWS CAPPCAEYS TO BE LOCK WlRFO (TYPICAL ALL SHEAVE PINS) C-C '175* CRINL RAIL F-F TROLLEY OR ,Rum FMME ,*LL"W >/a cLr***uci ALL *nouno DRILL AND TAP SAFTEI BLOCK AND ATTACH HOLD DOWN LUGS WITH BOLTS IACH HOL3 DOWN SMTE" ULOCR -- LUC IS TO RESTRAIN AN UPWARD FORCE OF 10 000* a B.L.5 -JJ LL I 1- rc>,i* CRANE RAIL BY SAFETY BLOCIKREUWDI:! ,!:,L,! DOWN LUGS LUGS FOR BRIDGE TRUCKS SIMILAR c.nd yslkclrd rover5 ar neceslsry for access to 011s. ebovt F crane E -J PLAN SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 3.8.6-4 RB 175 TON POLAR CRANES BRIDGE (REVISED BY AMENDMENT 13) SON CAD UNIT &NO 15 PART DT THC TYA PROCADAM DATABASE L3J ELEVATION GIRDER AND WALKWAY CROSS SECTION $ crane ra,, ,' ar r orrase - GIRDER AND END TIE CONNECTION WALKWAYS NOT SHOWN

/*rays /adder, and &,ails trpkal at four cdmerr oFbrld9e excepf as noted. SEC TlON A-A TYPIC4L FOR BRIDGE DRIVES 4 REQD PER CRANE PLAN P rail at E girde D-D 5-6 C - 6 CRANE DRIVC NOT SHOWN ICE BLANKET BUMPERS NOT SHOWN WHEEL ASSEMBLY 8 CRANE DRIVE NOT SHOWN SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 3.8.6-5 RB 175 TON POLAR CRANES (REVISED BY AMENDMENT 13) SON CAD UNIT &NO 15 PART DT THC TYA PROCADAM DATABASE KEY PLAN C-C CAPACITIES AND SPEEDS MAIN HOIST AUXILIARY HOIST 118 TESTING LOAD AT HWK 312.5001 20,000" ZJ.000" HOISTING SPEED AT FULL LOAD 0 $4 IMINI/5 357 (MAXI FPM O 85 (MlNl/3+ (MAXI FPM MOTOR (MIN SILL1 50 HP 22.5 HP BRlOGL 8 8 (MINl/42 6 (MAXI FPM 4 8 (MINl/25 (MAX1 FPM TRAVEL MOTORS (MIN SILL1 5 HP, ,200 RPM. 2 RLOO 3 HP. 1200 RPM YHEEL TREAD OlA (MINI 24' 24' OUALlTI ASSURANCE CRANE IS CLASS ?(LIB SLISMLC EQUIPMENT FACE OF BUMPER STOP ,&COLLECTOR BRACKET YlTH BUMPERS COMPRESSED PHWKS AT MAX CRANE TRAVEL. BUMPERS FREE 2. CYATEL REVEL EL 726 12' - - -- - -- - -- ELEVATION A-A SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 3.8.6-6 AB 125-TON CRANE (REVISED BY AMENDMENT 18)

AUX HOIST SECONDARY BRAKE AUX HOIST CFARFD ROTAR SECONDARY LlUlT SWITCH 21'-0' OVERALL PLAN TROLLEY CONDUCTOR SISTTU.CLTASON POIIRTRIK (OR FOUALI ELEVATION HWIS AND REEYING NOT SHOWN YAlN HOIST DRUM SAFETY SUPPORTSIPILLOW BLOCK FRAMES NOT SHOW TOP OF Rlll EL 773 -0' -LOWER BLOCKFOR 1318'014 WIRE ROPE MAIN HOOK- --- A-A I I CAD MAINTAINED DRAWING I I C-C Trolley or , Attow truck fmmr 70 surf' all around Weld to trolley \ or truck frame -* Hold down tug f /I~"cN"~ rar/ - Note: Dr,ll and tao ssletv bbrk and atrach hold &wn tugs with baltn and lock wes*e,s Eech holddown lug is to resfrain en yeward farce of ,O.OOO - SAFETY SL OCK W1TH HOLD DOWN LUGS 4 RfQ'D FOR TROLLEY LUGS FOR BRlDGF TRUCKS S/M/LAR (UPGRADED MAIN HOOK)

!~ b! Bolfcd rs11 cfip x,,n ~f"led shrar~bar GItiUER AND WALKWAY CROSS SECTION PLAN F'Povidc opwriiqs w,!h Oi//rd rower c in rr,d he webz end ends OF OPERATORS CAB girder, as nece,rary lor access lo Do/?+ FIE L'A TION GIRDER & END TIE CONNECTION WALKWAY NOT SHOWN SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 3.8.6-9 AB 125-TON CRANE BRIDGE (REVISED BY AMENDMENT 13) SON CAD UNIT &NO 15 PART DT THC TYA PROCADAM DATABASE /,-eg.ma# c-c pa, hoi uepfh dmphragms (tp)

P TRIPPER EAR FOR LIMIT E TRIPPER BAR FOR LIMIT SWITCH A AND LIMIT SWITCH A' (SWITCH O AND LIMIT SWITCH D' /M'-0" NOTE A : TRlPPER BARS LIMIT SWITCH, CAT* TO SUIT PLAN 7 P CRANE RAIL /E LIMIT SWITCH c E LIMIT SWITCM C' Is.- s,l.. C -- 4 1, ,,-TOP OF BRIDGE GIRDER L 1 +--+ ', D-D NOTES: GLNTRI(L OPEWLTlON OF LIMIT SWITCHES CONDITION I BRIDGE IS LOCATED EAST OF LIMIT SWITCH A TRfP EAR OR WEST Or LIMIT SWITCH B TRlP BAR. I. POWER FOR ALL MOTIONS IS AVAILABLE AN0 TROLLEY Un TWLMSvERsr ENilRE BRIDGE. TROLLEY IS LOClTED NORTH OF LIMIT SWITCH C ITHIS CONOlTlOY IS NECESSARY CONOIT IONS) FOR THE BRIDGE TO PASS OVER THE POOL UNOLR HOWL I POWER FOR ALL MOTIONS IS AVAIUBLE AND rnt BRIDGE UN TRANS- VERSE jrs ENTIRE RUN. SECT/ON A-A, 8-B d E-E -SPLICE Lb r4r) 70 EXIST/NB ANGLES L- AS SHOWN OR FABRICATE NEW BRACKETS TO ACCOMMODATE SWITCHES. PLAN LlMlT SWITCHES NOT SHOWN A.J ELEVATION LIMIT LIMIT LIMIT I I SWITCH A, / ,-LIMIT SWITCH A* A-A BRIDGE LOUTED BETWEEN LIMIT SWITCH A TRIP WR AND LIMIT SWITCH 0 TRlP mR AN0 TROLLEY IS LOCATE0 NORTH OF LIMIT SWITCH C A10 TPAVELING SOUTH. I. WHEU iinlr SWITCH c OR C' IS TRIPPED POWER FOR FURTHER SOUTH- WRD TRbVEL OF THE TROLLEY IS INTERR~PTEO I. TROLLE~ UN BE novcu IN THE REVERSE DIRECTION (NORTH) TO ESTABLISH CONPlTlOY 11. CONDITION Y SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 3.8.6-10 AB 125-TON CRANES LIMIT SWITCH (REVISED BY AMENDMENT 13)

KEY PLAN 1 140 PLAN , 20'-10' TO E NORTH RAIL NOTE 'A': FIELD ADJUST SO THAT SOLENOIO PLUNGER IS COMPLETELY SEATED WHEN BUMPER PLATE IS IN THE VERTICAL FUSITION NOTE '8'. FlEL THAT SWITCH BUMPER PLATC IS LIFTED AND OPENS WHEN PLATE IS LOWERED SECTION A - A LOCATE STOP m PROVIOE J/B. MINIMUM CLEARANCE WITH E TROLLEY WHEELS ON 6 RAIL RELOCATE TROLLEY BUMPER If ARI CLtilRANCE SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 3.8.6-11 AB 125 TON CRANE MECHANICAL STOP (REVISED BY AMENDMENT 13) SON CAD UNIT &NO 15 PART DT THC TYA PROCADAM DATABASE

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S3-13 3.13-1 SQN-26 3.13 Flex Response System 3.13.1 Flex Response System Mitigation of Beyond Design Basis External Events TVA has developed and installed diverse and flexible mitigation strategies (FLEX) that will increase defense-in-depth for beyond design basis scenarios to address an extended loss of alternating current power and loss of normal access to the ultimate heat sink occurring simultaneously at Units 1 and 2. This development is in response to orders from the Nuclear Regulatory Commission regarding the Fukushima Daiichi, Japan nuclear accident and is further described in SQN-DC-V-48.0; Reference 3.13.2.

The FLEX Response System components are designed to interface with existing safety related components. FLEX connections to plant systems are described in Reference 3.13.2.

3.13.2 Reference SQN-DC-V-48.0, FLEX Response Systems Design Criteria

S4-1.doc 4.1-1 SQN-25 4.0 REACTOR (WESTINGHOUSE FUEL) 4.1 SUMMARY DESCRIPTION This section addresses Westinghouse Standard (STD) and Vantage 5H (V5H) fuels. See 4.5.0 for information regarding AREVA fuel. Information concerning the power level uprate associated with the implementation of the Caldon Leading Edge Flow Meter is also included in Section 4.5.

This chapter describes 1) the mechanical components of the reactor and reactor core including the fuel rods and fuel assemblies, reactor internals, and the control rod drive mechanisms, 2) the nuclear design, and 3) the thermal-hydraulic design.

The reactor core is comprised of an array of fuel assemblies which are similar in mechanical design, but different in fuel enrichment. Three enrichments were employed in the initial core. Reload cores employ multiple enrichments.

Three Westinghouse fuel designs may be used: (1) STD fuel; (2) V5H fuel; (3) V5H fuel with certain Standard features (V5H with STD inconel grids and the corresponding guide tube diameters). AREVA fuel is also applicable for use and is discussed in Section 4.5.

In the discussion in the remainder of this chapter, the STD and V5H designs are discussed explicitly.

Where appropriate, the discussion should be understood to apply to V5H fuel with inconel grids.

The core is cooled and moderated by light water at a pressure of 2250 psia in the Reactor Coolant System. The moderator coolant contains boron as a neutron poison. The concentration of boron in the coolant is varied as required to control relatively slow reactivity changes including the effects of fuel burnup. Additional boron, in the form of Integral Fuel Burnable Absorbers (IFBA) or burnable absorber rods, is employed as needed to decrease the moderator temperature coefficient and to control the power distribution.

Two hundred and sixty-four fuel rods, twenty four guide thimble tubes and one instrumentation thimble tube are arranged within a supporting structure to form a fuel assembly. The instrumentation thimble is located in the center position and provides a channel for insertion of an incore neutron detector and thimble tube if the fuel assembly is located in an instrumented core position. The guide thimbles provide channels for insertion of either a rod cluster control assembly, a neutron source assembly, a burnable absorber assembly or a plugging device, depending on the position of the particular fuel assembly in the core. The fuel rods are supported in intervals along their length by grid assemblies which maintain the lateral spacing between the rods throughout the design life of the assembly. The grid assembly consists of an "egg-crate" arrangement of interlocked straps. The straps contain spring fingers and dimples for fuel rod support as well as coolant mixing vanes.

The fuel rods consist of slightly enriched uranium dioxide, ceramic cylindrical pellets contained in slightly cold worked Zircaloy-4 tubing which is plugged and seal welded at the ends to encapsulate the fuel. All fuel rods are pressurized with helium during fabrication to reduce stresses and strains which serves to increase fatigue life.

The center position in the assembly is reserved for the incore instrumentation, while the remaining 24 positions in the array are equipped with guide thimbles joined to the grids and the

/- SLOTTED THlMBLE PLUG BASE PLATE CUT OUT IN TOP OF NOZZLE ADAPTOR PLATE /- THIMBLE PLUG ASSY SPA^@ FOOT (4) IIEMOVABLE FUEL ROD I IwERloR THiHBiis (81 FINAL SAFETY Removable Fuel Rod Assembly Outline (Conceptual) 1 FIGURE 4.2.1-1 1 ; Revised by Amendment 13

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,-- SPRING SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 4.2.3-1 TYPICAL ROD CLUSTER CONTROL ASSEMBLY OUTLINE (REVISED BY AMENDMENT 17) \ LENGTH ABSORBER 80X SILVER 15% INOIUM 5% CAOMIUM

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SQN-26Note:LocationB3ofUnit2fortheMovableDetectorisabandonedinplaceperDCN23646A."PNMLKJHGFEOCB[*]wE3£3WE3E3E3mt3r3£3£*3£3i3£*3£*3£*3m£3C3£*3t3C3£3E3£3m£3£*3*11E3THERMOCOUPLE(65)*MOVABLEDETECTOR(58)(NOFLOWMIXINGOEVICES)£3£*3£3£*3£*1£*3E3E3£*3E3E3£3E3£3£3E3£*3E3E3£3£*3E3E3E3E3E3E3Figure4.4.5-1DistributionofIn-CoreInstrumentation£*3E3E3E3£*3£3£3-5-6-7-8-9-10*-111213141S

T452-2 SQN-25 Table 4.5.2-2 Reactor Coolant System Design Transient Group Core Power History for Fuel Rod Fatigue Analysis (Full Core Mark-BW) Transient Number of Fatigue Cycles Core Power History (% FP) Heatup and cooldown at 100 F/hr. (pressurizer cooldown 200 F/hr.) 200 + 200 100-HZP-0-HZP-100 Unit loading and unloading at 5% of full power/minute

Step load increase and decrease of 10% of full power (FP) Steady-state fluctuations. (18300x2)/2000x2/infinite 100-95-100-90-100 Large step load decrease (95% of FP with steam dump)

Loss of power (blackout with natural circulation in the reactor coolant system)

Loss of flow (partial loss of flow, one pump only)

Reactor trip from FP Inadvertent auxiliary spray 200/200/80/400/10 100-HZP Loss of load, without immediate turbine or reactor trip Loss of power (blackout with natural circulation in the reactor coolant system) 80/40 100-0 Minor loss of coolant accident or secondary steam line break 1 100-0 Terms: HZP - Hot Zero Power FP - Full Power Note: The cyclic/transient limits used for the fuel rod fatigue analysis, as summarized in this table, are bounding for the replacement steam generator cyclic/transient limits given in Table 5.2.1-1.

SQN-25 Table 4.5.2-3 Reactor Coolant System Design Transient Group Core Power History for Fuel Rod Fatigue Analysis (Full and Transition Core Advanced W17 HTP) Transient Number of Fatigue Cycles Core Power History (% FP) Heatup and cooldown at 100 °F/hr 200 + 200 0-HZP-100 100-HZP-0 Unit loading and unloading at 5% of full power/minute 18300x2 15-100 100-15 Step load increase and decrease of 10% of full power

Steady-state fluctuations 2000x2 Infinite 90-100 100-90 Large step load decrease (95% of FP)

Reactor trip from FP Pressurizer auxiliary spray actuation 200 400 12 100-HZP Loss of flow in one reactor coolant loop 80 100-HZP Loss of load, without immediate turbine or reactor trip

Loss of offsite A.C. electrical power (blackout with natural circulation in the reactor coolant system) 80 40 100-HZP Main reactor coolant pipe break or steam pipe break 1 100-0 Terms: HZP - Hot Zero Power FP - Full Power Note: The cyclic/transient limits used for the fuel rod fatigue analysis, as summarized in this table, are bounding for the replacement steam generator cyclic/transient limits given in Table 5.2.1-1.

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S5-0toc.doc 5-3 SQN-25 TABLE OF CONTENTS Section Title Page 5.4.3.3 Thermal Stresses Due to Gamma Heating 5.4-4 5.4.3.4 Thermal Stresses Due to Loss of Coolant 5.4-4 Accident (LOCA) 5.4.3.5 Stresses in UHI Adaptors 5.4-5 5.4.3.6 Heatup and Cooldown 5.4-5 5.4.3.7 Reactor Vessel Material Surveillance 5.4-5 Program Requirements 5.4.3.8 Capability for Annealing the Reactor Vessel 5.4-13 5.4.4 TESTS AND INSPECTIONS 5.4-13 5.4.4.1 Ultrasonic Examinations 5.4-13 5.4.4.2 Penetrant Examinations 5.4-14 5.4.4.3 Magnetic Particle Examination 5.4-14 5.4.4.4 Reactor Vessel Inservice Inspection 5.4-14 5.

4.5 REFERENCES

5.4-15 5.5 COMPONENT AND SUBSYSTEM DESIGN 5.5-1 5.5.1 REACTOR COOLANT PUMPS 5.5-1 5.5.1.1 Design Bases 5.5-1 5.5.1.2 Design Description 5.5-1 5.5.1.3 Design Evaluation 5.5-2 5.5.1.4 Tests and Inspections 5.5-6 5.5.2 STEAM GENERATOR 5.5-9 5.5.2.1 Design Bases 5.5-9 5.5.2.2 Design Description5.5-9 5.5.2.3 Design Evaluation 5.5-10 5.5.2.4 Tests and Inspections 5.5-15 5.5.3 REACTOR COOLANT PIPING 5.5-16 5.5.3.1 Design Bases 5.5-16 5.5.3.2 Design Description 5.5-16 5.5.3.3 Design Evaluation 5.5-18 5.5.3.4 Tests and Inspections 5.5-19 5.5.4 MAIN STEAM LINE FLOW RESTRICTORS 5.5-19 5.5.5 MAIN STEAM LINE ISOLATION SYSTEM 5.5-19 5.5.6 REACTOR CORE ISOLATION COOLING SYSTEM 5.5-19 5.5.7 RESIDUAL HEAT REMOVAL SYSTEM 5.5-19 5.5.7.1 Design Bases 5.5-19 5.5.7.2 System Description 5.5-20 5.5.7.3 Design Evaluation 5.5-25 5.5.7.4 Tests and Inspections 5.5-26 5.5.8 REACTOR COOLANT CLEANUP SYSTEM 5.5-26 5.5.9 MAIN STEAM LINE AND FEEDWATER PIPING 5.5-26

S5-1.doc 5.1-7 SQN-25 When the reactor power level is less than 15 percent, the reactor power is controlled manually. At power above 15 percent, the RCS controls automatically maintain an average coolant temperature, consistent with the power relationships, by control rod movement.

During the hot shutdown operations, when the reactor is subcritical, the RCS temperature is maintained by steam dump to the main condenser. This is accomplished by a controller in the steam line, operating in the pressure control mode, which is set to maintain the steam generator steam pressure. Residual heat from the core or operation of a reactor coolant pump provides heat to overcome RCS heat losses.

Plant Shutdown Plant shutdown is the operation which brings the reactor plant from no load power operating temperature and pressure to cold shutdown. During the plant cooldown, charging is provided to makeup for coolant contraction. During the initial phase of the cooldown, the makeup is provided from the boric acid tanks. The boric acid tanks should be used until at least the technical requirement manual minimum volume has been charged. At that point, operators can continue using the boric acid tanks if additional volume is available, or shift suction of the charging pumps to the refueling water storage tank. If the boric acid tanks are used, pure boric acid should be charged until the reactor coolant system reaches the desired cold shutdown concentration. The cooldown is completed by using blended makeup at the cold shutdown concentration. If the RCS is to be opened during the shutdown, the hydrogen and fission gas in the reactor coolant is reduced by degassing the coolant in the Volume Control Tank.

Plant shutdown is accomplished in two phases, the first is by the combined use of the RCS and steam systems, and the second by the Residual Heat Removal System. During the first phase of shutdown, residual core and reactor coolant heat is transferred to the steam system via the steam generator. Steam from the steam generator is dumped to the main condenser. At least one reactor coolant pump is kept running to assure uniform RCS cooldown. When the reactor coolant temperature and pressure are below approximately 350°F and 380 psig, respectively, the second phase of shutdown commences with the operation of the Residual Heat Removal System. With RHR in service, RCS pressure is maintained less than or equal to 380 psig. This provides sufficient margin to prevent the RHR suction relief valve from lifting at its setpoint of 450 psig. The pressurizer heaters are maintained energized while filling the pressurizer water solid to prevent stratification of the pressurizer water volume and to maintain an outflow of hot water from the surge line. With the heaters energized, the pressurizer is deliberately filled by creating a charging and letdown mismatch. After the pressurizer is filled, the heaters are deenergized to allow the pressurizer to cool. At approximately 140°F the last operating reactor coolant pump(s) may be turned off or cooldown may continue using a reactor coolant pump(s) and its associated steam generator to expedite cooldown to a reactor coolant system temperature of approximately 100°F. After the last reactor coolant pump is turned off, pressurizer cooldown is continued by initiating auxiliary spray flow from the Chemical and Volume Control System. Refueling Before removing the reactor vessel head for refueling, the system temperature has been reduced to 140°F or less and hydrogen and fission product levels are reduced. Installed plant instrumentation is provided to monitor RCS level during drain down activities. Draining continues until the water level is below the reactor vessel flange. The vessel head is then raised and the refueling cavity is flooded. Upon completion of refueling, the system is refilled for plant startup.

PRZ SAFETY COLD LEG (TYP) DETAIL C COLO LEG RTO THLRMOYLLL LOCATIONS (UNIT 2 ONLY1 STEAM GENERATOR LOOP NO 4 1. NSR DEFINES NON-SAFETY RELATED BOUNDARY 2. FAIL OPEN DENOTED BY 0 3. FAIL CLOSE DENOTED BY X SEE DETAIL B (UNIT li SEE DETAIL B (UNIT 11 OFT411 C (UNIT 21 OETAll C IUNlT 21 SEQUOYAH NUCLEAR PLANT ROM ACCUMULUOR NO 1 FINAL SAFETY ANALYSIS REPORT FIGURE 5.1-1 REACTOR COOLANT SYSTEM (REVISED BY AMENDMENT DETAIL A HOT LEG RTD THERMOWELl DETAIL B COLD LEG RTO THERMOUELL LOCATIONS (UNIT I ONLY1 START STOP 68-8A PULL TO LOCK NORMAL AUXILIARY - UNDERFREOUENCY TRIP - - MOTOR PROTECTION - OIL RESERVOIR OIL COOLER OIL LIFT PUMP REACTOR COOLANT PUMP 1 REACTOR COOLANT PUMPS 2.3 & 4 LOGIC TYPICAL SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 5.1-2 REACTOR COOLANT PUMP LOGIC (REVISED BY AMENDMENT 13) THIS CONTICURATION CONTROL DRAWING IS MAINTAINED BY THE SON CAD UNIT &NO IS PART OF THC TYA PRCCADAM DATABASE NOTES. 1 LOCATION OF CONTROLS *AUXILIARY CONTROL STATION 2. LOGIC SHOWN IS TYPICAL FOR UNITS 1 AND 2

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T521-1.doc SQN-26 TABLE 5.2.1-1 (Sheet 1) SUMMARY OF REACTOR COOLANT SYSTEM CYCLIC OR TRANSIENT LIMITS CONDITION CYCLIC OR TRANSIENT LIMIT DESIGN CYCLE OR TRANSIENT Normal 200 heatup cycles at < 100oF/hr and 200 cooldown cycles at < 100oF/hr per unit Heatup cycle - Tavg from < 200°F to > 550°F. Cooldown cycle -Tavg from 550F to 200F 200 pressurizer cooldown cycles at < 200°F/hr per unit Pressurizer cooldown cycle temperatures from > 650°F to < 200°F. 13,900 loading and unloading power changes per unit at 5% per minute, for RSGs (See Note 1) > 15% of Rated Thermal Power to 100% of Rated Thermal Power 2000 step load increases and decreases of 10% per unit (See Note 2) > 15% of Rated Thermal Power to 100% of Rated Thermal Power 200 large step load decreases of 95% From 100% of Rated Thermal Power 3.0E6 of steady state fluctuations for RSGs (See Note 3) RCS temperature changes of +/-6°F per minute at Tavg, (Tavg + 3°F) Upset 80 loss of load cycles, without immediate turbine or reactor trip > 15% of RATED THERMAL POWER to 0% of RATED THERMAL POWER 40 cycles of loss of offsite A.C. electrical power (blackout with natural circulation in the RCS) Loss of offsite A.C. electrical power source supplying the onsite ESF Electrical System 80 cycles of loss of flow in one reactor coolant loop Loss of only one reactor coolant pump at 100% of Rated Thermal Power 400 reactor trip cycles 100% to 0% of Rated Thermal Power. 10 pressurizer auxiliary spray actuation cycles Spray water temperature differential > 320°F and < 560°F 200 cycles of 1/2 Safe Shutdown Earthquake Reactor Vessel 50 cycles of 1/2 Safe Shutdown Earthquake Steam Generator and Pressurizer 10 low temperature water-solid overpressure events**Water-solid system actuation Faulted Conditions* Occurrences Main reactor coolant pipe break 1 Steam pipe break 1 Steam generator tube rupture Included in 400 reactor trip cycles from full power Safe Shutdown Earthquake 1 Test Conditions Occurrences Turbine roll test 10 Hydrostatic test conditions a. Primary side pressurized to 3110 psig 10 b. Secondary side pressurized to 1360 psig 10 c. Primary side leak test pressurized to 2485 psig 200 T521-1.doc SQN-26 TABLE 5.2.1-1 (Sheet 2) SUMMARY OF REACTOR COOLANT SYSTEM CYCLIC OR TRANSIENT LIMITS NOTES

• In accordance with ASME Boiler and Pressure Vessel Code,Section III, faulted conditions are not included in fatigue evaluations. ** Low Temperature Over Pressure events are not specified in the Upset Conditions for the Unit 1 RSGs. Refer to design specification S1RSG-CD-C201, Revision 2 (Reference 38) for details. 1. Since a limit of 13,900 cycles for loading and for unloading is permitted by fatigue calculations, the allowed normal loading and unloading transient cycles are NOT expected to be challenged during the extended life of the plant. Therefore, this transient is NOT recorded by monitoring instructions. 2. Over a 60-year plant life, 2,000 step load increases/decreases equates to more than one (1) allowed per week. Due to the stability of the TVA grid, step load changes of +/-10 percent seldom occur. Therefore, this transient will NOT be recorded by monitoring instructions. 3. This cyclic limit equates to approximately 6 cycles per hour for 60 years. Since 6 cycles per hour bounds SQN normal operation for 60 years of plant life, this transient will NOT be recorded by monitoring instructions.

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NOTE: THIS DRAWING IS NOT TO SCALE: ITS SOLE PURPOSE IS TO ILLUSTRATE THE PROVISIONS THAT HAVE BEEN MADE WITHIN THE FLOOR AND EQUIPMENT DRAIN SYSTEM TO ASSURE SUFFICIENT POST-LOCA LOWER COMPARTMENT INVENTORY INSIDE THE CRANE WALL. CRANE INSIDE OUTSIDE WALL S I PHON UPPER COMPARTMENT DRAIN EL 733 TYPICAL D a / -\ a a D a D a BREAKER ACCUM RM 3 & 4 FLOOR DRAINS 7 VENT ACCUM RM 3 & 4 FLOOD DRAINS ROUTED INSIDE CRANE WALL LOWER COMPARTMENT DRAIN EL 679.8 TYPICAL a EL 693 LJ~~~~~~ NMENT SUMP POCKET a a SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT A a a D a D a a a a a a D D D FIGURE527 -1 REACTOR BUILDING FLOOR AND EQUIPMENT DRAINS (REVISED BY AMMENDMENT 13) t 0 \v/ 0

S5-4.doc 5.4-9 SQN-25 5.4.3.7.1 Measurement of Integrated Fast Neutron (E > 1.0 Mev) Flux at the Irradiation Samples The use of passive neutron sensors such as those included in the internal surveillance capsule dosimetry sets does not yield a direct measure of the energy dependent neutron flux level at the measurement location. Rather, the activation or fission process is a measure of the integrated effect that the time- and energy-dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average flux level, and hence, time integrated exposure (fluence) experienced by the sensors may be developed from the measurements only if the sensor characteristics and the parameters of the irradiation are well known. In particular, the following variables are of interest:

1) the measured specific activity of each sensor;

2) the physical characteristics of each sensor; 3) the operating history of the reactor;

4) the energy response of each sensor; and 5) the neutron energy spectrum at the sensor location.

In this section the procedures used to determine sensor specific activities, to develop reaction rates for individual sensors from the measured specific activities and the operating history of the reactor, and to derive key fast neutron exposure parameters from the measured reaction rates are described. 5.4.3.7.1.1 Determination of Sensor Reaction Rates The specific activity of each of the radiometric sensors is determined using established ASTM procedures. Following sample preparation and weighing, the specific activity of each sensor is determined by means of a high purity germanium gamma spectrometer. In the case of the surveillance capsule multiple foil sensor sets, these analyses are performed by direct counting of each of the individual wires; or, as in the case of U-238 and Np-237 fission monitors, by direct counting preceded by dissolution and chemical separation of cesium from the sensor. The irradiation history of the reactor over its operating lifetime is determined from plant power generation records. In particular, operating data are extracted on a monthly basis from reactor startup to the end of the capsule irradiation period. For the sensor sets utilized in the surveillance capsule irradiations, the half-lives of the product isotopes are long enough that a monthly histogram describing reactor operation has proven to be an adequate representation for use in radioactive decay corrections for the reactions of interest in the exposure evaluations.

Having the measured specific activities, the operating history of the reactor, and the physical characteristics of the sensors, reaction rates referenced to full power operation are determined from the following equation: eeCPPFYNARdjttjrefjj10 S5-4.doc 5.4-10 SQN-25 where:

A = measured specific activity provided in terms of disintegrations per second per gram of target material (dps/gm).

R = reaction rate averaged over the irradiation period and referenced to operation at a core power level of Pref expressed in terms of reactions per second per nucleus of target isotope (rps/nucleus).

N0 = number of target element atoms per gram of sensor. F = weight fraction of the target isotope in the sensor material. Y = number of product atoms produced per reaction. Pj = average core power level during irradiation period j (MW). Pref = maximum or reference core power level of the reactor (MW). Cj = calculated ratio of (E> 1.0 MeV) during irradiation period j to the time weighted average (E> 1.0 MeV) over the entire irradiation period. = decay constant of the product isotope (sec-1). tj = length of irradiation period j (sec). td = decay time following irradiation period j (sec).

and the summation is carried out over the total number of monthly intervals comprising the total irradiation period.

In the above equation, the ratio Pj/Pref accounts for month-by-month variation of power level within a given fuel cycle. The ratio Cj is calculated for each fuel cycle and accounts for the change in sensor reaction rates caused by variations in flux level due to changes in core power spatial distributions from fuel cycle to fuel cycle. Since the neutron flux at the measurement locations within the surveillance capsules is dominated by neutrons produced in the peripheral fuel assemblies, the change in the relative power in these assemblies from fuel cycle to fuel cycle can have a significant impact on the activation of neutron sensors. For a single-cycle irradiation, Cj = 1.0. However, for multiple-cycle irradiations, particularly those employing low leakage fuel management, the additional Cj correction must be utilized in order to provide accurate determinations of the decay corrected reaction rates for the dosimeter sets contained in the surveillance capsules. 5.4.3.7.1.2 Corrections to Reaction Rate Data Prior to using the measured reaction rates in the least squares adjustment procedure discussed in Section 5.4.3.7.1.3, additional corrections are made to the U-238 measurements to account for the presence of U-235 impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation. In addition to the corrections made for the presence of U-235 in the U-238 fission sensors, corrections are also made to both the U-238 and Np-237 sensor reaction rates to account for gamma ray induced fission reactions occurring over the course of the irradiation.

5.4.3.7.1.3 Least Squares Adiustment Procedure Least squares adjustment methods provide the capability of combining the measurement data with the neutron transport calculation resulting in a Best Estimate neutron energy spectrum with associated uncertainties. Best Estimates for key exposure parameters such as neutron fluence (E> 1.0 MeV) or iron atom displacements (dpa) along with their uncertainties are then easily obtained from the adjusted spectrum. The use of measurements in combination with the analytical results reduces the uncertainty in the calculated spectrum and acts to remove biases that may be present in the analytical technique.

S5-4.doc 5.4-11 SQN-25 In general, the least squares methods, as applied to pressure vessel fluence evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross-sections, and the calculated neutron energy spectrum within their respective uncertainties. For example, RjRgiigiggg relates a set of measured reaction rates, Ri, to a single neutron spectrum, g, through the multigroup dosimeter reaction cross-section, ig, each with an uncertainty . The use of least squares adjustment methods in LWR dosimetry evaluations is not new. The American Society for Testing and Materials (ASTM) has addressed the use of adjustment codes in ASTM Standard E944, "Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance" and many industry workshops have been held to discuss the various applications. For example, the ASTM-EURATOM Symposia on Reactor Dosimetry holds workshops on neutron spectrum unfolding and adjustment techniques at each of its bi-annual conferences.

The primary objective of the least squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement. The analytical method alone may be deficient because it inherently contains uncertainty due to the input assumptions to the calculation. Typically these assumptions include parameters such as the temperature of the water in the peripheral fuel assemblies, bypass region, and downcomer regions, component dimensions, and peripheral core source. Industry consensus indicates that the use of calculation alone results in overall uncertainties in the neutron exposure parameters in the range of 15-20% (1). The application of the least squares methodology requires the following input: 1. The calculated neutron energy spectrum and associated uncertainties at the measurement location.

2. The measured reaction rate and associated uncertainty for each sensor contained in the multiple foil set.

3. The energy dependent dosimetry reaction cross-sections and associated uncertainties for each sensor contained in the multiple foil sensor set. For a given application, the calculated neutron spectrum is obtained from the results of plant specific neutron transport calculations applicable to the irradiation period experienced by the dosimetry sensor set. This calculation is performed using the benchmarked transport calculational methodology described in Section 5.4.3.7.2. The sensor reaction rates are derived from the measured specific activities obtained from the counting laboratory using the specific irradiation history of the sensor set to perform the radioactive decay corrections. The dosimetry reaction cross-sections and uncertainties that are utilized in LWR evaluations comply with ASTM Standard E1018, "Application of ASTM Evaluated Cross-Section Data File, Matrix E 706 (IIB)."

The uncertainties associated with the measured reaction rates, dosimetry cross-sections, and calculated neutron spectrum are input to the least squares procedure in the form of variances and covariances. The assignment of the input uncertainties also follows the guidance provided in ASTM Standard E 944.

S5-4.doc 5.4-12 SQN-25 5.4.3.7.2 Calculation of Integrated Fast Neutron (E> 1.0 MeV) Flux at the Irradiation Samples A generalized set of guidelines for performing fast neutron exposure calculations within the reactor configuration, and procedures for analyzing measured irradiation sample data that can be correlated to these calculations, has been promulgated by the Nuclear Regulatory Commission (NRC) in Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" [Reference 1]. Since different calculational models exist and are continuously evolving along with the associated model inputs, e.g., cross-section data, it is worthwhile summarizing the key models, inputs, and procedures that the NRC staff finds acceptable for use in determining fast neutron exposures within the reactor geometry. This material is highlighted in the subsection of material that is provided below. 5.4.3.7.2.1 Calculation and Dosimetrv Measurement Procedures The selection of a particular geometric model, the corresponding input data, and the overall methodology used to determine fast neutron exposures within the reactor geometry are based on the needs for accurately determining a solution to the problem that must be solved and the data/resources that are currently available to accomplish this task. Based on these constraints, engineering judgment is applied to each problem based on an analyst's thorough understanding of the problem, detailed knowledge of the plant, and due consideration to the strengths and weaknesses associated with a given calculational model and/or methodology. Based on these conditions, Regulatory Guide 1.190 does not recommend using a singular calculational technique to determine fast neutron exposures. Instead, RG-1.190 suggests that one of the following neutron transport tools be used to perform this work.

Discrete Ordinates Transport Calculations a) Adjoint calculations benchmarked to a reference-forward calculation, or stand-alone forward calculations.

b) Various geometrical models utilized with suitable mesh spacing in order to accurately represent the spatial distribution of the material compositions and source.

c) In performing discrete ordinates calculations, RG-1.190 also suggests that a P3 angular decomposition of the scattering cross-sections be used, as a minimum.

d) RG-1.190 also recommends that discrete ordinates calculations utilize S8 angular quadrature, as a minimum.

e) RG-1.190 indicates that the latest version of the Evaluated Nuclear Data File, or ENDFIB, should be used for determining the nuclear cross-sections; however, cross-sections based on earlier or equivalent nuclear data sets that have been thoroughly benchmarked are also acceptable. Monte Carlo Transport Calculations A complete description of the Westinghouse pressure vessel neutron fluence methodology, which is based on discrete ordinates transport calculations, is provided in Reference 2. The Westinghouse methodology adheres to the guidelines set forth in Regulatory Guide RG-1.190.

S5-4.doc 5.4-13 SQN-25 5.4.3.7.2.2 Plant-Specific Calculations The most recent dosimetry analyses for both Sequoyah units, including the vessel fluence assessment that was made to support the Sequoyah 1.3% Uprate Program, were based on discrete ordinates transport calculations. All of these calculations were conducted in accordance with the guidelines that are specified in Regulatory Guide RG-1.190.

5.4.3.8 Capability for Annealing the Reactor Vessel There are no special design features which would prohibit the onsite annealing of the vessel. If the unlikely need for an annealing operation was required to restore the properties of the vessel material opposite the reactor core because of neutron irradiation damage, a metal temperature of approximately 750°F maximum for a period of 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> maximum would be applied. This annealing operation would be performed with the use of a special electrical space heater assembly designed to raise the affected vessel area to the required temperature for the necessary holding period. This heater assembly will consist of an insulated vessel cover assembly below which is suspended the required space heaters positioned opposite the affected area of the reactor vessel shell. The heater assembly will contain provisions for sealing to the vessel flange, and waterproof electric connections. Hydraulic connections for emptying the reactor vessel of water after the assembly is in place are also required. A thermocouple assembly to monitor vessel metal temperature during annealing would also be included.

The reactor vessel materials surveillance program is adequate to accommodate the annealing of the reactor vessel. The remaining surveillance capsules at the time of annealing would be removed and given a thermal cycle equivalent to the annealing cycle. They would then be reinserted in their normal position between the core internals assembly and the reactor vessel wall. Subsequent testing of the fracture toughness specimens from the capsules would then reflect both the radiation environment before any annealing operation and after any annealing operation. 5.4.4 Tests and Inspections The reactor vessel quality assurance program is located in Table 5.4.4-1. 5.4.4.1 Ultrasonic Examinations

1. During fabrication, in addition to the Design Code required straight beam ultrasonic examination, an angle beam examination of 100 percent of the plate material was performed to detect discontinuities that may go undetected by the straight beam examination. 2. The reactor vessel was examined after hydro-testing to provide a baseline map for use as a reference document in relation to later in-service examinations.

3. The UHI adaptor and attachment welds to the closure head received both straight beam and angle beam ultrasonic examination. 4. A special pre-service examination of the reactor vessel nozzles was conducted to evaluate the extent of underclad cracking. The examination utilized a 70° angle beam and 0° beam manual contact technique. All indications found were demonstrated to be acceptable in accordance with ASME Code Section XI criteria.

S5-4.doc 5.4-14 SQN-25 5.4.4.2 Penetrant Examinations The partial penetration welds for the control rod drive mechanism head adaptors and UHI adaptors were examined by dye penetrant after the first layer of weld metal, after each 1/4 inch of weld metal, and the final surface. Bottom instrumentation tubes were examined by dye penetrant after each layer of weld metal. Core support block attachment welds were examined by dye penetrant after first layer of weld metal, and after each 1/2 inch of weld metal. This is required to detect cracks or other defects, lower the weld surface temperatures, cleanliness and prevent microfissures. All austenitic stainless steel clad surfaces were 100 percent liquid penetrant examined after the hydrostatic test. 5.4.4.3 Magnetic Particle Examination

1. All surfaces of quenched and tempered materials had the inside diameters inspected prior to cladding and the outside diameter 100 percent examined after hydro-testing. This serves to detect possible defects resulting from the forming and heat treatment operations.

2. The attachment welds for the vessel supports, lifting lugs and refueling seal ledge were examined after the first layer of weld metal and after each 1/2 inch of weld thickness. Where welds are back chipped, the areas were examined prior to welding.

5.4.4.4 Reactor Vessel Inservice Inspection The Inservice Inspection Program is addressed in Section 5.2.8.

The welds in the following areas of the installed irradiated reactor vessel are available for ASME Section XI required inspections.

1. Vessel shell - The inside surface. 2. Primary coolant nozzles - The inside surface.

3. Vessel heads - The inside and outside surface.

The lower head weld on each reactor pressure vessel is partially inaccessible for examination from the vessel inside diameter due to instrumentation tubes which penetrate the lower head. A 100 percent pre-service examination of the weld was conducted from the vessel outside diameter. This was accomplished by performance of a manual ultrasonic examination. A remote ultrasonic examination was conducted from the vessel inside diameter on all accessible areas of the weld. Accessible areas of the weld will be reexamined during the inservice intervals in accordance with Section XI of the ASME Boiler and Pressure Vessel Code. 4. Closure studs, nuts and washers. 5. Field welds between the reactor vessel nozzles, and the main coolant piping.

6. Vessel flange seal surface.

7. CRDM and UHI adaptors.

S5-4.doc 5.4-15 SQN-25 The design considerations which have been incorporated into the system design to permit the above examinations are as follows: 1. All reactor internals are completely removable. The tools and storage space required to permit these examinations are provided.

2. The closure head is stored dry on the reactor operating deck with the insulation capable of being temporarily removed during refueling to facilitate examinations.

3. Typically, reactor vessel studs, nuts and washers are removed to dry storage during refueling.

4. Removable covers are provided in the reactor cavity floor. The insulation covering the nozzle welds may be removed.

5. Irradiation specimen access holes are provided in the lower internals barrel flange to allow remote access to the reactor vessel internal surfaces between the flange and the nozzles without removal of the internals.

6. A removable manway cover is provided in the lower core support plate to allow access for examination of the bottom head without removal of the lower internals.

The reactor vessel presents access problems because of the radiation levels and remote underwater accessibility to this component. Because of these limitations on access to the reactor vessel, several steps have been incorporated into the design and manufacturing procedures in preparation for the periodic non-destructive examination which are required by Section XI of the ASME boiler and pressure vessel code. These are:

1. Shop ultrasonic examinations were performed on all internally clad surfaces to an acceptance and repair standards to assure an adequate cladding bond to allow later ultrasonic examination of the base metal from inside surface. The size of cladding bonding defect allowed was 3/4 of an inch in diameter.

2. The design of the reactor vessel shell in the core area is a clean, uncluttered cylindrical surface to permit future positioning of the examination equipment without obstruction.

3. After the shop hydrostatic testing, selected areas of the reactor vessel were ultrasonically examined and mapped to facilitate the Inservice Inspection Program. Vessel design data is in Table 5.4.2-1. Transients and anticipated number of cycles are in Table 5.2.1-1. The reactor vessel quality assurance program is in Table 5.4.4-1. 5.4.5 References 1 Regulatory Guide RG-1.190 (RG-1.19O), "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," United States Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001. 2. WCAP-15557, "Qualification of the Westinghouse Pressure Vessel Neutron Fluence Evaluation Methodology," S. L. Anderson, August 2000.

2701THERHALSHIELDFigure5.A.3-3SurveillanceCapsulePlanVie\90c

S5-5.doc 5.5-12 SQN-25 impurities in the steam generator water. This reduces the potential for formation of highly concentrated solutions in low flow zones. By restriction of the total alkalinity in the steam generator and prohibition of extended operation with free alkalinity, the AVT program will minimize the recurrence of intergranular corrosion in localized areas due to excessive levels of free caustic.

Laboratory testing has shown that the Inconel-600 tubing is compatible with the AVT environment. Organic Amine/Hydrazine Treatment - Organic Amines, in conjunction with hydrazine, have been used successfully at numerous domestic and foreign utilities as a pH control additive. In these applications, organic amines reduce erosion-corrosion in two-phase regions, since they are less volatile than ammonia and provide a higher liquid phase pH in wet steam regions. A benefit expected with organic amines pH control is a reduction in steam generator sludge accumulations, lower than would be expected with just ammonia treatment. Organic amines provide better two-phase pH control, which reduces corrosion product formation, resulting in less corrosion product formation and transport to the steam generators, minimizing sludge accumulation. Experience and testing documented by Westinghouse and EPRI demonstrates morpholine's acceptability as a secondary side water treatment additive (References 5.0 and 6.0). Additionally, Westinghouse has provided a Sequoyah specific evaluation documenting compatibility of morpholine with secondary side materials and its usefulness as a secondary side chemical treatment additive (Reference 5).

Ethanolamine - This chemical favorably distributes itself into the water phase of wet steam cycles, which makes it effective as cycle pH additive. In addition, ethanolamine has a high basic strength at elevated temperatures, which reduces the quantity of the amine required to provide protection. Tests have indicated that ethanolamine can reduce corrosion product transport with little negative effect on polishers (Reference 6). 5.5.2.3.3.2 Secondary Side Chemical Treatment Dimethylamine - For Unit 1 and Unit 2 RSG's Dimethylamine (DMA) has a high basicity for "at temperature" pH control in the steam cycle, but will not significantly decrease condensate polisher run times. DMA has other characteristics such as anti-scaling and anti-fouling properties, and inhibition of silica formation that makes it desirable as a chemical additive in systems that have no copper based materials. DMA acts as a solvent to copper and copper oxides and will remove them from surfaces and surface deposits so they can be removed by steam generator blowdown.

S5-5.doc 5.5-13 SQN-25 A comprehensive program of steam generator inspections, including the requirements of Regulatory Guide 1.83, will insure detection and correction of any unanticipated degradation that might occur in the steam generator tubing.

5.5.2.3.4 Flow Induced Vibration In the design of the Westinghouse RSGs, consideration has been given to the possibility of degradation of tubes due to mechanical or flow induced vibration. This consideration includes detailed analyses of the tube support system as well as an extensive research program with tube vibration model testing. The primary cause of tube vibration in a steam generator is the propensity for the tubes to extract energy from the secondary fluid flow and convert it into tube motion. The extent of movement experienced by the tubes is dependent on the nature of the flow regime and the approach angle of the flow with regard to the tube span orientation. For vibration assessment purposes, in a typical inverted U-tube steam generator there are four types of secondary flow.

These are: 1. Entrance Cross-Flow - This is either subcooled or saturated liquid (recirculated water mixed with feedwater from the downcomer) cross-flow which occurs at the secondary face of the tubesheet on both the hot side and cold side. 2. Axial Flow - This flow can vary from subcooled to saturated liquid two-phase flow depending on elevation in the steam generator tube bundle. The direction of flow is parallel to the tube axis and this type of flow is the predominate flow in the unit. 3. Two-Phase Cross-Flow (Exit Region) - This type of flow exists for the tubes in the U-bend region of the tube bundle. The effects of this flow are similar to those produced by "entrance cross-flow" except that due to lower density, the energy available to the tubes is not as great. 4. Mixed Flow - This type of flow is a combination of axial flow and cross-flow which could exist in two distinctly different situations. a. Oblique Flow Across Straight Tubes - This could occur in regions of flow direction change such as directly above the recirculating entrance region.

b. Unidirectional Flow Across Curved Tubes - This occurs in the U-bend area of the tubes.

SQN-25Sincetheproblemofflowinducedvibrationintheidentifiedregionsisofsignificantconcerninthedesignofshellandtubeheatexchangers,Westinghousehasgivenconsiderationtotheexperimentalevaluationofthebehavioroftubearraysundercrossflow.ThevibrationanalysisoftheRSGtubesandtubesupportsisbasedonthefollowing:Fullscaleparametricflowtestdata.i.Multi-spandynamicresponsetestdata.ii.Bothgeneralandlocalshell-sideflowdistributionanalysis.v.State-of-the-artflowinducedvibrationanalysistechniques,includingAppendixNofASMECodeSectionIII.Thetestdatadescribedaboveprovidedthebasisfordeterminingacceptableshellsidefluidvelocitiesinthemostcriticalregionsofthetubebundlewhicharethefluidentranceandfluidexitregions.Basedonthetestdataandresultsfromtheshellsideflowdistributionanalyses,thedesigntubesupportspacinghasbeenshowntohaveconservativedesignmarginsagainsttheonsetoffluidelasticinstability.TheRSGdesignconfigurationindicatestubevibrationstabilityratios(i.e.,theeffectivecross-flowvelocity/fluid-elasticcriticalcross-flowvelocity)wellbelowthetheoreticalallowableof1.0andlessthanorequaltotheconservativedesignrequirementof0.75.DisplacementsduetorandomturbulentexcitationoftheRSGtubingbasedontheturbulentbuffetingmethodologyofAppendixNofSectionIIIoftheASMECodearewellbelowthedesigngoalof10mils.Reliablemethodsforcalculatingtubeweararenotavailableforstableflowvibrationsatthislevelofdisplacement.Experimentaltubevibrationdataindicatethattubewearisnotsignificantforvibrationdisplacementslessthan10mils.Additionally,theSequoyahplanttechnicalspecificationdefinestheSGtuberepairlimitas40percentthrough-walldegradationor0.0168inches.Historicallyandthroughouttheindustry,thisrepairlimithasbeendeterminedfromboundingvaluesofnon-destructiveexaminationsizinguncertaintyanddegradationgrowthratefromastructurallimitinconjunctionwithaconservativesafetymargin.Therepairlimitisdefinedsuchthattheperformancecriteriaforstructuralandleakageintegritywillbemetattheendofanintervalbetweeninspectionswiththisreductionintubewallthickness.ThestructuraleffectsofvibrationhavebeengivenconsiderationandthestresslimitationsforeachcategoryintheASMECodehavebeenmet.Thetubestressesduetoflowinducedvibrationsweredeterminedtobelessthan1.0kipspersquareinch,whichislessthanthelowerboundfatiguestresslimit.Therefore,therewillbenostructuralorfatiguedamageresultingfromflowinducedvibrationintheRSGs.Finally,itshouldbenotedthatsuccessfuloperationalexperiencewithseveralsteamgeneratordesignswithsimilartubesupportstructureshasgivenconfidenceintheoverallapproachtoaddressflowinducedvibrationintheRSGtubesupportdesign.S5-5.doc5.5-14 S5-5.doc 5.5-15 SQN-25 5.5.2.4 Tests and Inspections The steam generator quality assurance program during construction is given in Table 5.5.2-2. Radiographic examination and acceptance standards were in accordance with the requirements of Section III of the ASME code.

Liquid penetrant examination was performed on weld deposited tube sheet cladding, channel head cladding, tube-to-tube sheet weldments, and weld deposit cladding.

Liquid penetrant examination and acceptance standard were in accordance with the requirements of Section III of the ASME code.

Magnetic particle examination was performed on the tube sheet forging, channel head casting, nozzle forgings, and the following weldments:

l. Nozzle to shell 2. Support brackets 3. Instrument connections (primary and secondary)

4. Temporary attachments after removal

5. All accessible pressure containing welds after hydrostatic test. Magnetic particle examination and acceptance standard were in accordance with requirements of Section III of the ASME code. An ultrasonic examination was performed on the tube sheet forging, tube sheet cladding, secondary shell and heat plate and nozzle forgings. The heat transfer tubing was subjected to eddy current examination.

Hydrostatic tests were performed in accordance with Section III of the ASME code. In addition, the heat transfer tubes were subjected to a hydrostatic test pressure prior to installation into the vessel which is not less than 1.25 times the primary side design pressure multiplied by the ratio of the material allowable stress at the testing temperature.

Manways are to provide access to both the primary and secondary sides. Regulatory Guide 1.83, "Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes," provides recommendations concerning the inspection of tubes, which cover inspection equipment,

3.

+It g,: Z'; - 2 & 6 Ow LC, 0 0 ;; g 5 -0 1 74-505 "'"1 I I - MINIMUM FLOW LINE SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 5.5.7-1 RESIDUAL HEAT REMOVAL SYSTEM (REVISED BY AMMENDMENT 13)

S6-2.doc 6.2-95 SQN-26 Each containment vessel vacuum relief isolation valve is a pneumatically operated butterfly valve with an elastomer seat. The valve, including seat material, will withstand a post-LOCA temperature, pressure, and radiation conditions. Two separate trains of control air supplies are available to the two independent solenoid valves which power the isolation valve. The isolation valve, which is normally open, fails open, and will close when containment high pressure reaches the set pressure of 1.5 psid. The high pressure signal is developed from either of two independent sets of three pressure sensors and is completely independent of other containment isolation signals for other systems. Each isolation valve is equipped with a limit switch so that open and closed positions are indicated in the MCR. 6.2.6.4 Design Evaluation The containment vessel vacuum relief units are located in the annulus and thus are not affected by flood, wind, ice, snow, or tornado. The units are located such that they are also free from the danger of missile damage.

The relief capacities and set pressure of the VR system are based on conservative combinations of parameters. Extra margin is allowed for deviations in the actual vacuum relief flow rate. Based on the assumptions, analyses, and on the specified design limits, it was found that for all events requiring mitigation by the vacuum relief system, the integrity of the primary containment is assured. Figure 6.2.6-1 shows the upper compartment pressure transient during the initial vaporization stage as a function of time after inadvertent spray initiation. The CAR system is conservatively assumed to be operating during this period. The design case is for an initial upper compartment relative humidity of 4 percent. A principle aspect considered in the design of the vacuum relief system is that the initial containment vessel pressure is not necessarily the same as that in the annulus space. Thus, the pressure transient due to inadvertent spray operation must be considered as an addition to the initial pressure differential across the containment vessel. This consideration is, to a large extent, responsible for the low set pressure of 0.1 psid for the containment vessel vacuum relief valves. The annulus is assumed to have a negative pressure of 5 inches water gauge initial to the inadvertent spray saturation transient and the containment vessel is 0.1 psi below the annulus pressure (i.e., at the setpoint of the vacuum relief system). The analyses show that the design limit of 0.5 psid for the external pressure differential on the steel containment is not exceeded during the short term initial vaporization stage of an inadvertent spray operation. For the initial vaporization stage, the minimum upper compartment pressure occurs at about 12 seconds after the spray is initiated. After that time, the rise in water vapor pressure and the effects of the vacuum relief system are greater than the drop in pressure due to cooling. The upper compartment atmosphere becomes 100 percent saturated at about 24 seconds as indicated in Figure 6.2.6-3.

As indicated in Reference 92, the vacuum relief system is assumed to operate 2.2 seconds after the transient begins. One vacuum relief valve is assumed to fail to open in keeping with single failure criteria. The remaining two vacuum relief valves are assumed to be fully open at 2.2 seconds after the initiation of the inadvertent spray. Since the transient begins with the containment at the setpoint of the VR system, the vacuum relief units would begin to open at the initiation of the containment spray. However, no credit is taken for flow through the VR valves that would occur while they are opening. Figure 6.2.6-1 thru 6.2.6-3 show the short and long term transients for containment differential pressure, temperature, and relative humidity for inadvertent spray and air return fan operation. The compartment relative humidity is initially at 4 percent. A conservatively low containment VR system capacity of 28 pounds per second at 0.5 psid is used, assuming that the S6-2.doc 6.2-96 SQN-26 redundant unit is not operative. As indicated in Reference 92, an initial negative pressure of 5 inches water gauge is assumed for the annulus and the containment vessel initial pressure is again 0.1 psi below the annulus pressure. Figure 6.2.6-1 shows that for initial containment relative humidity at 4 percent, the worst pressure transient occurs during the initial saturation stage.

In the event of an inadvertent air return fan operation without operation of the containment spray, the containment atmosphere would eventually approach ice bed temperatures. However, even with both fans running, the pressure transient is much slower than that shown in Figure 6.2.6-1 because of the lower cooling rate. The containment external pressure (i.e., annulus pressure) is essentially equal to that of the vacuum relief unit set pressure. The annulus pressure, if unrelieved, would drive the shield building external pressure to approach the design value. However, the containment annulus vacuum control subsystem provides a flow path for air from the auxiliary building to go into the annulus, if necessary, so that the shield building external design pressure of 2 psid is never reached.

In setting the design flow rates and pressure setting for the VR system, considerable design effort and safety margins were made in providing proper annulus pressure reduction so that the integrities of both the containment vessel and the shield building are assured. Assurance of integrity is achieved by making certain that at a particular driving head both the maximum and minimum flow rates are within design limits, considering the operation and non-operation of the redundant units and considering all design basis accidents within the containment.

6.2.6.5 Testing and Inspection All components of the VR system are readily accessible for inspection, maintenance, and testing. The VR system is designed in accordance with the criteria set forth in section 3.1. A test connection between the vacuum relief valve and the isolation valve is provided in each unit for periodic pressure and leak testing, in conformance with Appendix J of 10 CFR 50, with approved exemptions. The system is designed and tested in accordance with applicable ASME Codes. Tests made during fabrication include hydrostatic pressure test, leak test across seals, and flow capacity test on VR system components. In-place tests include periodic tests on the actuator for its ability to move the disc and to operate the position indicating lights in the MCR. 6.2.6.6 Materials The VR system is not a safety feature system, although the valves will act as containment isolation valves on containment high pressure. The materials used meet the Class 2 requirements of the draft ASME Code for Pumps and Valves for Nuclear Power, 1968 Edition. The radiolytic or pyrolytic decomposition product, if any, of each material will not interfere with the safe operation of any Engineered Safety Feature System. 6.2.7 References 1. "Maximum Flow Rate of Single Component, Two-Phase Mixture," F. J. Moody ASME Publication, Paper No. 64-HT-35.

S6-2.doc 6.2-102 SQN-26 82. TVAN Calculation TI453, Revision 3, Iodine Loading on the EGTS Charcoal Absorbers and Temperature in the EGTS ACUs Following a LOCA-DBA. 83. Letter from NRC to S. A. White dated March 21, 1988, Hydrogen Analyzer Operability (RIMS A02 880323 007).

84. Westinghouse Letter TVA-03-19, dated February 12, 2003, "Steam Generator Compartment Pressurization Following a Steam Line Break for the Model 57AG Replacement Steam Generator." 85. R. Gridley to NRC, September 16, 1987, "Sequoyah Nuclear Plant - Containment Coatings."

86. NUREG-1232, Volume 2, "Safety Evaluation Report on Tennessee Valley Authority: Sequoyah Nuclear Performance Plan," May 1988.

87. Randy Douet to NRC, September 1, 2005, "Sequoyah Nuclear Plant - NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors - Second Response."

88. Nuclear Energy Institute Guidance Report No. NEI-04-07, "Pressurizer Water Reactor Sump Performance Evaluation."

89. SQN Memo, "Documentation of Historical Lower Containment Temperature and Ice Mass Inputs into the Sequoyah Containment Temperature Analysis of Record", (EDMS# B85111107001)

90. SQN Letter, "RWST Volume Requirement for Steamline Break", (RIMS# B25901226009)

91. Westinghouse Letter, TVA-09-69, dated September 8, 2009, "Reactor Coolant Loop Structural Analysis Evaluation of Unit 2 Replacement Steam Generators - Phase 2.

92. SQN Calculation, SQN APS2037 "Analysis of Inadvertent Containment Spray."

SQN-26 TABLE 6.2.4-1 (Sheet 2) CONTAINMENT PENETRATIONS CONTAINMENT ISOLATION VALVE STROKE TIME REQUIREMENTS VALVE NUMBER FUNCTION MAXIMUM ISOLATION TIME (SECONDS) 43 FCV-62-61 RCP Seals 10 44 FCV-62-63 RCP Seals 10 45 FCV-62-72 Letdown Line 10 46 FCV-62-73 Letdown Line 10 47 FCV-62-74 Letdown Line 10 48 FCV-62-77 Letdown Line 20 49 FCV-63-23 Accum to Hold Up Tank 10 50 FCV-63-64 WDS N to Accum 10 51 FCV-63-71 Accum to Hold Up Tank 10 52 FCV-63-84 Accum to Hold Up Tank 10 53 FCV-68-305 WDS N to PRT 10 54 FCV-68-307 PRT to Gas Analyzer 10 55 FCV-68-308 PRT to Gas Analyzer 10 56 FCV-70-85 CCS from Excess Lt Dn Hx 10 57 FCV-70-143 CCS to Excess Lt Dn Hx 60 58 FCV-77-9 RCDT Pump Disch 10 59 FCV-77-10 RCDT Pump Disch 10 60 FCV-77-18 RCDT and PRT to V H 10 61 FCV-77-19 RCDT and PRT to V H 10 62 FCV-77-20 N to RCDT 10 63 FCV-77-127 Floor Sump Pump Disch 10 64 FCV-77-128 Floor Sump Pump Disch 10 65 FCV-81-12 Primary Water Makeup 10 B. PHASE "B" ISOLATION 1 FCV-32-80 (U1), FCV-32-81 (U2) Control Air Supply 10 2 FCV-32-102 (U1), FCV-32-103 (U2) Control Air Supply 10 3 FCV-32-110 (U1), FCV-32-111 (U2) Control Air Supply 10 4 FCV-67-83 ERCW - LWR Cmpt Cirs 60 5 FCV-67-87 ERCW - LWR Cmpt Cirs 60 6 FCV-67-88 ERCW - LWR Cmpt Cirs 60 7 FCV-67-89 ERCW - LWR Cmpt Cirs 70 8 FCV-67-90 ERCW - LWR Cmpt Cirs 70 9 FCV-67-91 ERCW - LWR Cmpt Cirs 60 10 FCV-67-95 ERCW - LWR Cmpt Cirs 60 11 FCV-67-96 ERCW - LWR Cmpt Cirs 60 12 FCV-67-99 ERCW - LWR Cmpt Cirs 60 13 FCV-67-103 ERCW - LWR Cmpt Cirs 60 14 FCV-67-104 ERCW - LWR Cmpt Cirs 60 15 FCV-67-105 ERCW - LWR Cmpt Cirs 70 16 FCV-67-106 ERCW - LWR Cmpt Cirs 70 17 FCV-67-107 ERCW - LWR Cmpt Cirs 60 18 FCV-67-111 ERCW - LWR Cmpt Cirs 60 19 FCV-67-112 ERCW - LWR Cmpt Cirs 60 20 2-FCV-67-130 ERCW - Up Cmpt Clrs 60 21 2-FCV-67-131 ERCW - Up Cmpt Clrs 60 T624-1.doc

SQN-26 TABLE 6.2.4-1 (Sheet 7) CONTAINMENT PENETRATIONS CONTAINMENT PENETRATION LIST Penetration Description Inside Barrier Outside Barrier X-021 SI Pump Discharge to Hot Legs - Train B 63-547 63-549 63-167 63-157 Closed System X-022 Injection Tank Charging Pump Discharge 63-581 63-174 Closed System 63-25 63-26 X-023 PASF HL 3 Train B 43-310 43-309 X-024 SI Relief Valve Discharge 68-559 Closed System 72-512, 513 62-505 63-626, 627 63-534, 535, 536 63-511 X-025A Przr. Liquid Sample 43-02 43-03 X-025B Containment Sensor 30-311/44 N/A 30-311X 30-311Y 30-44X 30-44Y X-025C Rx Vessel Level N/A N/A X-025D Przr. Liquid Sample 43-11 43-12 X-026A Containment Sensor 30-310/43 30-310X 30-310Y 30-43X 30-43Y X-026B Control Air - Train B - Unit 1 32-297 32-102 32-295 X-026B Control Air - Train B - Unit 2 32-348 32-103 32-341 X-026C Rx Vessel Level N/A N/A X-027A Containment Sensor 30-30C N/A 30-30CY 30-30CX X-027B Containment Sensor 30-42 N/A 30-42Y 30-42X X-027C ILRT 52-504 52-505 X-027D Rx Vessel Level N/A N/A X-028 Spare N/A N/A X-029 CCS from RCP Coolers 70-89 70-92 70-698 T624-1.doc

T626-1.doc SQN-26 TABLE 6.2.6-1 DATA TABLE FOR THE VACUUM RELIEF SYSTEM Design Basis:

Maximum containment external pressure differential 0.5 psid Maximum shield building external pressure differential 2.0 psid Design Parameters:

Upper compartment free volume 651,000 ft3 Lower compartment free volume 253,114 ft3 Ice condenser free volume 110,521 ft3 Dead end compartment free volume 129,900 ft3 Upper ice plenum free volume 54,940 ft3 Annulus space free volume 375,000 ft3 Number of containment spray headers 2 Flow rate for each containment spray header 4,750 gpm Distance between spray headers and upper deck 152 feet Number of air return fans 2 Flow rate for each air return fan 40,000 cfm Maximum initial upper compartment dry bulb temperature 110 F Maximum initial lower compartment dry bulb temperature 120 F Minimum initial upper compartment relative humidity 4 percent Minimum containment spray water temperature 60 F Ice condenser temperature (dry air) 15 F Set pressure of ice condenser doors connected to the upper and lower compartment 1 psf Resultant Design:

Number of steel containment vacuum relief units 3 (1 redundant) Maximum initial external pressure differential on the containment (containment vacuum relief system set pressure) 0.1 psid Design flow rate of each containment vacuum relief unit at 0.5 psid 28 lbm/sec Maximum response time for any unit to be fully open for a design basis event 2.2 sec

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CONTAINMENT PRESSURE DIrFERNTIAL SWITCHES AUTO OR MANUAL SI I / / I\ NOR I TIME DELAY CONTAINMENT PURGE EXHAUNT RADIATION MONITORS r---------7 TRAIN A SHOWN WN B SIMILAR; ,TRAIN A SHOWN Z~IN B SIMILARJ NOTES 1. LOGIC SHOWN IS TYPICAL OF UNITS 1 AND 2 SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 6.2.4-1 MECHANICAL LOGIC DIAGRAM CONTAINMENT ISOLATION I (REVISED BY AMENDMENT 18) I I CAD MAINTAINED DRAWING I I

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NOTES: 1. NSR DEFINES NON-SAFETY RELATED BOUNDARY 2. FAIL OPEN DENOTED BY 0. 3. FAIL CLOSE DENOTED BY X. SYMBOLS. * - ECCS FLOW BALANCING ORIFICE

I h % D REFUELING SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 6.3.2-3 REACTOR BLDG ELEVATION ECCS ACTIVE SUMP AND INACTIVE SUMP LOCATIONS (REVISED BY AMENDMENT 13)

CRANE WALL WITH VENT HOLES FLE PLATE GRATING ON SIDES AND REAR WALL (NOT TO SCALE) OF SUMP C3 Z H 3 Q E n n W Z H Q t Z H Q 3 n < 0 SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 6.3.2-4 CONTAINMENT SUMP WITH VORTEX MODIFICATIONS SUPPRESSION (REVISED BY AMENDMENT 21)

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- PIPlNG TO UNIT Z SIMILAR TO UNIT 1 AIR DlSTRlBUTlON DUCT HEADER-- NOTES: 1. NSR DEFINES NON-SAFETY RELATED BOUNDARY SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 6.5.6-1 ICE CONDENSER GLYCOL SYSTEM (REVISED BY AMENDMENT 13)

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S7-1.doc 7.1-7SQN-26 18. The Institute of Electrical and Electronic Engineers, Inc., "IEEE Recommended Practices for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations," IEEE Std. 344-1975. (See Paragraph 7.1.2.11). 19. The Institute of Electrical and Electronic Engineers, Inc., "IEEE Standard for Electrical Penetration Assemblies in Containment Structures for Nuclear Fueled power Generating Stations," IEEE Std. 317-1976. (See Paragraph 7.1.2.4). 20. Regulatory Guide 1.105, "Instrument Setpoints for Safety-Related Systems", Revision 2 February 1986. 7.1.2.1 Design Bases The technical design bases for the protection systems are provided by Westinghouse equipment specifications which consider the functional requirements for these systems and applicable criteria such as IEEE 279-1971, IEEE 317-1971, IEEE 323-1971 and the NRC General Design Criteria. 7.1.2.1.1 Reactor Trip System The Reactor Trip System acts to limit the consequences of Condition II events (faults of moderate frequency such as loss of feedwater flow) by, at most, a shutdown of the reactor and turbine, with the plant capable of returning to operation after corrective action. The Reactor Trip System features impose a limiting boundary region to plant operation which ensures that the reactor safety limits analyzed in Chapter 15 are not exceeded during Condition II events and that these events can be accommodated without developing into more severe conditions.

The design requirements for the Reactor Trip System are derived by analyses of plant operating and fault conditions where automatic rapid control rod insertion is necessary in order to prevent or limit core or reactor coolant boundary damage. The design limits for this system are:

1. Minimum DNBR will not be less than limit value as a result of any anticipated transient or malfunction (Condition II faults).

2. Power density will not exceed the rated linear power density for Condition II faults. See Chapter 4 for fuel design limits.

3. The stress limit of the Reactor Coolant System for the various conditions will be as specified in Chapter 5.

4. Release of radioactive material will not be sufficient to interrupt or restrict public use of those areas beyond the exclusion distance or to exceed the guidelines of 10 CFR 20, "Standards For Protection Against Radiation," as a result of any Condition III fault.

5. For any Condition IV fault, release of radioactive material shall not result in an undue risk to public health and safety nor will it exceed the guidelines of 10 CFR 100, "Reactor Site Criteria." 7.1.2.1.2 Engineered Safety Features Actuation System The Engineered Safety Features Actuation System acts to limit the consequences of certain Condition II (upset conditions such as credible small steamline breaks) and Condition III events (infrequent faults such as primary coolant spillage from a small rupture which exceeds normal

7.1.3-1 1-4r.r rrr -rL-&am-w Yrra ~rr# r**m wyu. 4urrrrmrm um4rZ'. s-". AT. wr.~)?.-- 7.1.3-2 Sys ly r I em j 'a- y rum. ~@a* and- Figure

7.1.3-5 PLAN EL 732'-0" CONTROL PANEL NOMENCLATURE SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 7.1 .4-1 LAYOUT OF CONTROL BOARDS (REVISED BY AMENDMENT 16) UNlT I PANEL "NIT 2 NO. PANEL PANELS COWON TO BOTH UNITS O-hl-ZS ............... UONlTORING PANEL O-hl-26 ............... DIESEL GENERATOR CONTROL O-l-2iA .............. ESSENTIAL RAW CWLINC WATER 0-U-27B .............. CMPONFNT COOLING WATFR O-M-I2 ............... RAOIATION UONITORING AND RFCORDlNG O-,-ZBA SPARE 0.M.288 .............. SPARE O-hl-29 ............... FIRE DETECTION MONITOR PANEL

S7-2.doc 7.2-15 SQN-26 The accident analyses described in Section 15.2 demonstrate that the functional requirements as specified for the Reactor Trip System are adequate to meet the above considerations, even assuming, for conservatism, adverse combinations of instrument errors (Refer to Table 15.1.3-1). Safety limits associated with the reactor core and Reactor Coolant System, plus the Limiting Safety System Setpoints, are presented in the SNP Technical Specifications.

The Technical Specifications incorporate both nominal and limiting setpoints. Nominal settings of the setpoints are more conservative than the limiting settings. This allows for calibration uncertainty and instrument channel drift without violating the limiting setpoint. Automatic initiation of protective functions occurs at the nominal setpoints (plus or minus the allowed tolerances). The methodology used to derive the setpoints is documented in references 16 and 20. 7.2.1.2.5 Abnormal Events The malfunctions, accidents or other unusual events which could physically damage Reactor Trip System components or could cause environmental changes are as follows: 1. Earthquake (discussed in Chapter 2 and Chapter 3).

2. Fire (See Section 9.5). 3. Explosion (Hydrogen buildup inside containment). (See Section 6.2). 4. Missiles (See Sections 3.5 and 10.2.3). 5. Flood (See Chapter 2 and 3).

6. Wind and Tornadoes (See Section 3.3). All instrumentation, control and communication lines that will be required for operation in the flood mode are either above the design basis flood (DBF) or within a nonflooded structure or are designed for submerged operation. 7.2.1.2.6 Minimum Performance Requirements The performance requirements are as follows: 1. System response times: The reactor trip system response time shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The reactor trip system instrumentation response time values are provided in Table 7.2.1-5. 2. Reactor Trip accuracies are given in Table 7.2.1-4. 3. Protection system ranges: Range a. Power range nuclear power 1 to 120% full power b. Neutron flux rates +5% to -5% of full (positive and negative) power for rapid changes in power. c. Overtemperature T: Thot leg 530 to 650°F Tcold leg 510 to 630°F Tavg 530 to 630°F Pressurizer pressure 1700 to 2500 psig F(>) -60 to +60% T setpoint 0 to 150% power

S7-2.doc 7.2-32 SQN-26 17. L. E. Erin, "Topical Report Eagle-21 Microprocessor-Based Process Protection System," WCAP-12374, September 1989 (Westinghouse Proprietary Class 2).

18. J. F. Mesmigos, "Median Signal Selector for Foxboro Series Process Instrumentation Application to Deletion of Low Feedwater Flow Reactor Trip," WCAP-12417 (Westinghouse Proprietary Class 2).

19. The Institute of Electrical and Electronic Engineers, Inc., "IEEE Recommended Practices for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations," IEEE Std. 344-1975.

20. ISA-S67.04 - 1982, "Setpoints for Nuclear Safety-Related Instrumentation Used in Nuclear Power Plants".

(1)

LOGIC SYMBOLS ADDITIONAL SYhZBOLS DEVICE FUNCTION LETTERS AND NUMBERS SYMBOL LOC lC FUNCTION r/- INSTRUMENT CHANNEL BISTABLE FB FLOW Wbm INDICATES THAT ME DEVICE OR INSTRUMENT CHANNEL HAS A B ISTAB LE LOGIC 'I l.8 LEVEL W(UCL AND A DEVICE WHICH PRODUCES AN OUTPUT ONLY OUTPUT W EN1 N: WU) MIN~ A- PARAMETER MEASURED tS GREATER THAN A PRESET YALUE ?I--- PB PRESSWE CHUU*L WHEN EVERY INPUT EXISTS. RC RMIATION QUmR Z PARAMETER MEASURED IS LESS THAN A PRESET VALUE .59 SPECD CHVsR PARAKIETER MEASUfXD DEVIATES FRW A PRESET VALUE BY MORE THAN A TI TWERITUE WWa PRESET MtOUNl. Zll POSITION cxuwa $ -tf; OR 5 OR 5 SAME AS AEON EXCEPT WITH AN AUTOMATICALLY SET VARIABLE VALUE 20 ELECTR lC OPERATED VALVE NOT A DNlCE WH ICH PRODUCES AN OUTPUT ONLY 2 1 UNDERVOLTAGE RELAY WHEN THE INPUT DOES NOT EXIST. a OR J OR 1 SME AS ABOVE MCEPT WITH REQUIRED HYSTERES IS BETWEEN TURN ON 33 POSITION SWITCH AND TURN OFT. 51 AC CIRCUIT BREAULR v 63 PRLSrURE WITCH J -. NON- INSTRUMEM B lSTABE 1 I LEVEL SWITCH OR A DEVICE WH ICH PRODUCLS AN OUTPUT Bo FLOW BWITCM WHEN ONE INPUT (OR MORE1 MISTS. ?. OUTPUT INDICATOR SAME AS EXPLAINED ABOVE 8 1 ULIMRFREQUC*LCY RELAY 4 ALARM ANNUNCIATOR IAU\RMS ON THE SAhE SHEET W ITH THE SAME SUBSCRIPT SHARE A COMMON ANNUNCIATOR WINDOW) ADD\TLOMAL LQ~&$YMROL~ OFF RETURN A DEVICE WH ICH RCTAlNS THE CONDITION OF MEMORY OUTPUT CORRES WNDING TO THE LAST ENER- GIZED INPUT. EXCEPT UPON INTERRUPTION OF REACTOR TRIP 'FIRST OUT" ANNUNCIATOR SV.(MBOL LOCkIC FUuCT\O*J POWER IT RETURNS TO THE OFF CONDITION. NAB INE TRlP 'FIRST OUT" ANNUNCIATOR 4! k4N/c(Ol. + AFTER RKCIL\PT OF THU COhlTIZOL LOGIC \uPUT THf RETENTIVE tl DEVlCEYJHlCH RETAINS THE CONDITION OF @ - .INDICATOR LAMP hE\U/HOLD UUlt OUTPUT WILL BS ALLOWED TO MEMORY OUTPUT CORRES WNDINC TO THE LAST ENER- A ACTUATION STATUS LIGHTS DIGRUSE \P TW8 UNiT \UPUT DECZUTeS. \T GlZED INPUT IALSO UPON INTERRUPTION OF T TRIP STATUS LIGHTS t TWG \UPUI 10 THW UNIT \*JCIZILAC.-S, THE UU\T POWER). P PERMISSIVE STATUS LIGHTS OUTPUT W\LL HOLD THE MILJtMUW vALUC, $2 I BYPASS STATUS LIGHTS Slh)CC REG.\- Q= COrJTFZOL LO&LC,. ADJUSTABLE A DEVICE WHICH PROWCES AN OUTPUT El - COMPUTER INPUT TIME DELAY FOLLOW ING DEFINITI: INTENTIONAL TIME LOGIC INFORMATION TRANSMISS ION ENERC lZ ING DELAY AFTER RECEIV lNC AN INPUT, - - - - - - - ANALOG INFORMATIONTRANSMISSION % ADJUSTABLE A DEVICE WHICH CONTINUES TO PRODUCE AN ANALOG DISPLAY TITLE INDEX SHEET NO. 0- TIME DELAY OUTPUT FOR A DEFINITE INTENTIONAL PERIOD DE-ENERGIZING OF TIME AFTER THE INPUT HAS BEEN REMOVED. I ANALOG INDICATOR INDEX AND SYMBOLS ----------------. R RECORDER 1 R2 RECORDER 2 CHANNEL REACTORTRIP SIGNALS ------------- - 2 fq CO lNCl DENCE A DEVICE WHICH PRODUCES AN OUTPUT R3 RECORDER 3 CHANNEL NUCLEAR INSTR-AND MANUAL TRIP SIGNALS - --- 3 I2 OUT OF 3 SHOWN) WHEN THE PRESCRIBED NUMBER OF INPUTS R8 RECORDER 8 POINT NUCLEAR INSTR. PERMISSIVES AND BLOCKS ----- 4 EXIST IEXAMPLE 2 INPUTS MUST EXIST FOR 0- ANALOG SUMMER PRIMARY COOLANT SYSTEM TRIP S IGNALS --- --- 5 AN OUTPUT). PRESSURIZER TRIP 5 IGNALS ----------- b STEAM GENERATOR TRIP SIGNALS --------- I ANALOG SAFEGUARDS ACTUATION SIGNALS--------- 8 RETENTIVE A DEVICE HAV ING THE LOGICAL FUNCTION INPUT A DEVICE WHICH PERMITS AN ANALOC SIGNAL RODCONTROLS&RODBLOCKS ---------- 9 MEMORY AS INDICAED BY THE DIAGRAM BCLOW TO PASS IN AN ISOLATED CIRCUIT 1F TH CON- STEAM DUMP CONTROL -- - - - - - - -- - - 10 'ITH MANUAL ACTUATING S IGNAL MANUAL RESR CONTROL ANALOG LN'c GATE TROL LNIC INPUT EXISTS. PRESSURIZER PRESSURE (L LEVEL CONTROL----- 11 RESET PRESSURIZER HEATER CONTROL---- -- --- 12 INPUT NOTES: FEEDWATER CONTROl d( 45OlATION ------ -- 13 I. EXCEPT I)IERE lWlCATE0 OTKRIIISE. lX* HEFaLOlllNO Ism: FEEDWATER CONTROL & ISOLATION------- - 14 ANALOG UL L~IC CIRCUITS *RE RWT. AU BISTABES. CIRCUIT BREAKERS AUXILIARY FEEOWATER PUMPS STARTUP -- - --- WIATORS M INOICAIWW ARE NOTREOWSLHT. WK COHT~S m 15 OUTPUT NOT HLVE RE- UW~ORS. BUT m HAM REWT CWTACTS WHERE TURBINE TR IPS. RUNBACKS h OTHER S IGNALS --"- 16 LMilC IS Rmu.ouct. I@ REQUIREMENTS) 2. THIS sn w DRAWINGS IS IDENTICALFOR UNITS I & 2 EXCEPT EAM/TTD LOGIIC,PQOT~CIION SET L- - - \I FOR THE TAG NUMBERS. EA\~/ttca LOG~C,PSOT~C~IOU 607 - - 18 FOR UNIT 1. TAG NUMBERS ADD A "I". EXAMPLE: 1PC-455E. GAWTTD \-041C, PROTECT\OM --- 19 FOR UNIT 2, TAG NUMBERS ADD A "2': EXAMPLE: 2PC-455E. ~M~TD L~~IC,PROTECTIO~J SET E-- - 20 3. WHENEVER A PROCESS SIGNAL IS USE0 FOR CONTROL AND IS DERIVED FROM A PROTECTION CHANNn, ISOLATION MUST BE PROVIDED. 4. t~ IS SET OF DRAWINGS ILLUSTRATES THE FUNCTIONAL REQUIREMENTS OF THE REACTOR CONTROL AND PROTECTION SYSTEM. THESE DRAWINGS DO NOT REPRESENT ACTUAL HARDWARE IMPLEMENTATION. 5. SHEET NUMBERS REFER TO THE FSAR FIGURE 7.2.1-1 SHEET #. WPUT Sl MK SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 7.2.1-1 SHEET 1 FUNCTIONAL DIAGRAMS-INDEX AND SYMBOLS (REVISED BY AMENDMENT SON CAD UNIT hND 15 PART OF THC TYA PROCADAH DATABASE ROD DRIVE SUPPLY ONE LINE DIAGRAM TRAIN A REACTOR SHUNT TRlP SIGNALS H-G SET HlWUAL REACTOR TRlP SIGNAL (SHEET 3) MbWUAL SAFETY INJECTION SIGNAL (SHEET 8) REACTOR TRlP SWI TWGEAR LOGIC TRAIN A REACTOR TRlP SIGNALS RDD DRIVE POWER SUPPLY MANUAL TRlP SIGNAL (SHEET 3) SCURCE RANGE. HIGH FLUX (INTEUOCXED BY P-6 & P-10) ROD OR1 M PONER W1S (NOTE I) NEUTRON nux TRIP SIGNALS (SHEET 3) OVERTMPERANRE AT LOGIC TRAIN A PRIMARY CMLWT SYSTM TRlP SIGNALS (SHEET 5) BLOW LOGIC (SHEET 8) HIGH PRESSJRE PnESSURlZER TRIP SI WbLS LON PRESSURE ( INTERUKXEO BY P-7) HIM LEVEL ( INTERLOCKED BY P-7) ( WEET 6) STEM GFNERATOR TRlP SIGNALS ( WEE1 19) SAFETY INJECTION 51 WAL (WET TURBINE TRIP SIGNAL (SHEET 16) ( LOW AUTO STOP OIL PRESSURE OR LL STOP VALVES aoso (INTERLOCKED BY P-9) REACTOR TRlP S101L FOR TURBINE TRIP (SHEET 16) TO STEAM DUMP CONTROL LOGIC (SH~ZT 10) LOGIC TRAIN 5 REACTOR TRlP SIGNALS TO STEAM DUMP CONTROL LOGIC (SHEET 10) MWUAL TRIP SIGN& (SHEET 3) BLOW LOGIC (WEE1 8) NEUTRON FLUX TRlP SlMbLS (SHEET 3) TO FEEDWATER ISaATlON LOGIC (WEET 13) OVERTWERAN OMRPOnER CiT PRIUAW CMANT MSTM TRIP SIGNALS (SHEET 5) P-4 REACTOR TRIP SIGNAL FOR TURBINE TRlP (SHEET 16) UNOERFRECUEN HIGH PRESSUR PRESatRlZER TRlP SIWALS ( SHEET 6) LOGIC TMlN B STEAM GENERATOR TRlP SIGNUS (SHEET 19) SAFETY INJECTION SIGNAL (SHEET NOTES: TURBINE TRlP SIGNAL (SHEET 16) I. TRIPPING THE REACTDR TRlP BREAKERS 52/RTA AND $2/RTB RE(ZNOANlLY OE-ENERGIZES ME RW DRIVES. ALL FULL LENGTH CCNTAll RODS AND SHUiDOWN RODS ARE THEREBY RELEASED FOR GRAVITY INSERTION INTO THE REACTOR UX);. 2. NORHbL REACTOR OPERATION IS TO BE WITH REACTOR TRlP BREAKERS 52/RTA AND 52/RTB iN SERVICE PSID BY-PASS BREAKERS 52/BYA AN0 52/BYB WITHORAWN. CURING TEST, ONE BY-PASS BREAKER IS TO BE PUT IN SERVICE WD THW THE RESPECTIVE REACTOR TRlP BREAKER IS OPERATED USING A SlHlLATEO REACTilR TRlP SlPleL IN THE TRAIN UNDER TEST. ME REACTOR WILL NOT BE TRIPPED BY THE SIMLhATEO SIGNAL SINCE THE BY-PASS BREAKER IS CONTRaLEO FRCM THE OTHER TRAIN. ONLY ONE REACTOR TRlP BREAKER IS TO BE TESTED AT A TIME. MWAL REAClUR TRlP SIGNAL (SHEET 3 MWAL SAiETY INJECTION SIGNAL (SHE 3. AU CIRCUITS ON THIS SHEET ARE NOT REWNONjT BECAUSE BOTH TRAINS ARE SOW. 4. CPEN/CLOSEO INDlCATlM FOR EACH TRIP BREAKER NjO EAW BYPASS BREAKER IN CPITRa R(M. 5. CLOS~HG A BYPASS BREAKER WILL ACTUATE THE SOLID STATE PROTECTION SYSTEM GENERAL WARNING ALARM SYSTEM FOR THAT TRAIN.ACNATION ff THE GENERAL WARNING ALARM SYSTEM ON BOTH TRAIN A AND TRAIN 6 WlLL SEND TRlP SIGNALS TO ALL FOW BREAKERS (RTA. RTB, BYA. BYB.) THROUGH THE UMIERVOLTAGE COILS.' 6 SHEET NUMBERS REFER TO FIGURE 7.2 1-1 (SHEET NUMBER). SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 7.2.1-1 SHEET 2 FUNCTIONAL DIAGRAMS-REACTOR TRIP SIGNALS (REVISED BY AMENDMENT 13) THIS CONFICURArlON CONTROL DRAWING IS UAINTAINCO BY THE SQN CAD UNIT IN0 IS PART OF THC TYI PROCADAM DATA8hSE NOTES: I. N KWWT WWL BLW( mmns w rm, mmms M( mE W/WO FOR UW R&. WE FOR EIW TRAIN. 2. I/N SSP IS IN Laic TWIN A. Ifl 358 IS IN LOGIC TRAIN 8. 3. ~fi 381 \S {N COOIC Tmln A. Ifl 388 IS IN LOGIC TRAIN 8. 4. Ifl 471 IS IN LOGIC TRAIN A. Ifl 478 IS IN LOGIC TRAIN 8. TWO CDMRITER IWS ARE CCEEECTED TO THIS CIRCUIT INIlVlCWL FOR EACM TRAIN. 5: WAL RE R. CONTRCXPI WIST OF FOUI M*RI IN CCNTR[L Rm(. CNE CONT~L FOR EACH INSTWM CWm. CAD MAINTAINED DRAWING SOURCE RANGE REACTOR TRIP INTERMEDIATE RANGE REACTOR TRIP POWER RANGE REACTOR TRl P POWER RANGE HIGH NEUTRON I P I x IP FLUX RATE REACTOR TRlP I lx /, I A I I I I\ I --- 1 I I POWER RANGE I I BLOCK CONTROL I I HlGH NEUTRON f LUX HIGH NEUTRON FLUX MANUAL TRIP (MAIN CONTROL BOARD) HIGH NEUTRON (LOW SETPOINT) (HIGH SETPOINT) FLUX RATE REACTOR TRIP REACTOR TRIP REACTOR TRlP (SHEET 2) (SHEET 2) REACTOR TRIP REACTOR TRlP (SHEET 2) SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 7.2.1-1 SHEET 3 FUNCTIONAL DIAGRAMS-NUCLEAR INSTR & MANUAL TRIP SIGNALS (REVISED BY AMENDMENT 20)

POWER RANGE P-13 TURBINE IMPULSE CHAMBER PRESSURE POWER RANGE I -. INTERMEDIATE RANGE I A \ I: A- FROM IR BLOCK LOGIC (S H EET 3) FROM IR BYPA55 (SHEET 3) P- 6 (SHEET 3) NOT REDUNDANT I I HIGH NEUTRON FLUX ROD STOP (BLOCK ROD WITHDRAWAL) (SHEET 9) POWER RANGE I A \ nr %Pa55 BY PA5 5 (NOTE I) I POWER RANGE NOT REDUNDANT '. SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 7.2.1-1 SHEET 4 FUNCTIONAL DIAGRAMS-NUCLEAR INSTR PERMISSIVES & BLOCKS (REVISED BY AMENDMENT 13) SQN CAO UNIT AND 15 PART OF THE TYA PROCADAH OATABhSL I.THE BYPASS SIGNALS ARE MADE UP 8Y MEANS OF TWO THREE-POSITION SWITCHES ON A MIS RACK SWITCH I/N 49A BYPASSES EITHER NC-411 OR NC43L. SWITCH IIN498 BYPASSES EITHER NC-42L OR NC44L. C -2 OvERPow*R ROD STOP (BLOCK ROD w AUTOMATIC ITH DRAWAL) ( MANUAL. (SHEET 9) hoT REDUNDANT OVLRTRMPERATURE- A OVERPOWER AT <LE~ D/ CQNP~N~T~~J (~nolue ~WPEN~ATEW) LOOPL LOOP L LOOP 3 COOP 4 cWPZ COOP3 4 (wrcr 2) WT REDOIIOAIW (SWCCT 2) NOT REOUND&NT Lo-LO 'AV~ LOW ~AV& LOOP 1 LOQ* t? LOOP 3 LOOP 4 hlOtnC: I. sET-PO~NT FOR UNDBRJOLTPGE REUVS SIIOULO S*AP*ROY. 70%. 2. TnI: siT?O\hiT op?ut UIIUE*FR*VUEUCY <@LAY% WOULD 8F AD~u%T~~LE BEtWtCIJ 54 C?S 59 cPS. P-I2 3. THE a4rlMUM ALLOVdABLE RCP BRUKES TRIP TIME OELAY 15 0.1 5EC. (WEET 16) 10 FEEDW4TER ISDLATIOM 4 TUG UUDEQ\QLTA&E SENSGRS (WTENTIAL TIZP!~~FO'PMERS) (surrr 13) MUST SE+ LOCATED OU lUE MOTOR SIDE OF THE 2CP 6. REACTOR TRIP SYSTEN RESP3U56 TIME s\LQUlR+MCh)TS ClQCUIT BQEAKEPS TO THE TPlP CF PCP Fd? RCP UUDERVOLTAC~ IS 1.2 SEC.,< MS UI)065FREQUENCY LIQCUaT BREAKERS IU AC~DITIOU TO SUS uNEERKL-AGE 15 SO.& SEC..IHS RE5;t3NS'tTMfS \5 OEFINEO ASTHC: ELAPSED 5 THE UNDERFQEQUENLY SEUSORS MAY BE LOCATED CN TLME FROMTHE BlFaTABLE CHANGE OF -AT6 TO THE 9EblUl)l)JG THE MOTOR SIDE OF THE RCP ClPCUlT BPEAICERS OF COUTZ?OL ROO h40TIOhl. , REACTOD COQLAUT PUMP UNDER~LYAGE&JOTEL.) REALTOR.CO~LALLT PUMP UNDERFRE~UEMCY (NOTE&> \ / A \ ECP? RCP3 QCP 2 RCP 3 RCP 4 By I TVA -- fsur.r 1) TRIP 1 TRIP (TRIP RCPL RCP3 ,NOn,) RCPS BREAKER 8RCAKCR Fww WOO I FLOU LOLOOP z FLOW boor 3 FLOW LOOP 4 REACTOR TRIP REACTOR TRIP (SHEET 2) (SHEET 21 SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 7.2.1-1 SHEET 5 FUNCTIONAL DIAGRAMS-PRIMARY COOLANT (REVISED SYS TRIP BY AMENDMENT SIGNALS 13) ION CAD UNIT AND 15 PART OF THE TVA PROCADAM OATlBABL PRESSURIZER LOW PRE55URE (LEAD/ SHEET^) REACTOR 1RlP (SHEET 2) PRESSURIZER HlGH PGESSCIRE ?p+ - -- REACTOR TRIP (IHEETZ) PRE5SURlZER HlGH WATER LEVEL I ?I WES. I. THE REWNDANT WUAL BLOCK CONTRCC CaJSlSTS OF TWO MNTAOLS ON THE CDNTROL BOARD, ONE FOR EACH TRAIN P- SHEET^) 2. TWO COMPUTEfi INPUTS ARE CIWNECTED TO MIS CIRCUIT. INDIVIWAL FOR EACH TRAIN. LOW PRESSURltER PRESSURIZ'ER PRESSURI EER 5.1 PRESSURE PRESSbRE BLOCK CONTROL h 51 n Pa @I REACTOR TRIP (SHEET 2) &-@ SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 7.2.1-1 SHEET 6 FUNCTIONAL DIAGRAMS-PRZ TRIP SIGNALS (REVISED BY AMENDMENT 13) ** -m OTHER LOGIC TRAIN X ? TO SAFETY INJECTON (SHEET 0)

STM GEM, HI-HI LEVEL / ST& &EL) I SM GCN 2 SMW3 STW GEhl4 1 HIGH 5EAM PRESSURE RATE (RATE-LAG COMPLMLATEO) STEAMLINE 41 I LOOP I LOOP 7. LOW STEAMLINE. DRESSURE (LEAD-LAG COMPENSATED) r LOW FEEDWATER FLOW ALARMS I LOOP I LOOP 2 LOOP 3 LOOP 4 r \ I I , r LOW 5/G STMF/FWF MISMPllCH LOW I/G STMF/FWF MISMCICH LOW VG STMF/FWFMISYILXH LOW S/G SNF/FW MISMATW WATER LEVEL (LOW F/W FWW) ~ATZR LEVEL (WW F/W FLOW) WATER LEVEL (LOW YW FLOW) WATER LEVEL STEAMLINE ISOLAT IOU CSHEET 8) NOTES : I .THE REDUNDALLT MANUAL BLOCU CONTROL CONS15T'Z OF TWO CONTROLS ON THE CONTROL BOARD, ONE FOREACH 7RAIII. SUPPLIED 0Y OTHERS. 2. TWO COMPUTER INPUTS ARE WNUECTEO TO TH15 CIRCUIT, IUDWIDUAL FOR EaCHTRAIU 3. 4 RERRTO $MEET 19 FOR THE 6TCAM '5tNERATOR LOW-LOW WATER L*V'*L REACTOR TR\P- SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 7.2.1-1 SHEET 7 FUNCTIONAL DIAGRAMS-STEAM GENERATOR TRIP SIGNALS m*h4 GSdERATOR COl(TAIdMLHT PRLSSURE. MAML!AL ACTUATION FROM CDHTPOL BMRD I Low PRE~~URIZER Hl6H SEAM PRE55U RZ I TIO HOHMTIRI tDITRUS' WERATING EITHER CCNTRCL NlU ACTUATE. 2: THE MW SPRAY AC~A~ICW CCHSISTS OF FCUR CDITRCLS, MAIN rrrnnrm MS. UTUATIM WILL DCWR MY IF ASZDCIATED CCNTRUS ARE WERATED I. SERVICE IIATER %5m( IXLATID4 IS USED MY IF REWIREI). SIKLTMCUSLY. 9. w REDNDWT Y*NWL RESET WISTS w MTW CMTKUS M TIE. 1 WE WINTAINEO WTRa PER LW mrmCL gl*RO CNE FORM TRAIN 4: WAll+*wr PREWRE @ISTARES m~ PRAY ACN4TlCN ARE ENERGIZE m ID. YSO ~OSES TAE WPUS VUYE IN P~UU ll~ M US~IAT~ STEM UTUATE (OM 8ISlPBES ARE CE-D(Effi1ZE ACTUATE). LIM 5lW VUK. s EHQOSm CIWITRI IS HOT PAR7 OF WE SAFEWIRDS SUM IS HOT RELI*(H. 11. LlDm 9111LD BE PlWlaD IN WE CWrRX Rm* Fa EAM SWINE ?iTCC 8: motn ARE UL l~olVlwULY SEUD IN (LATCHED), ZD THAT LOSS OF WE VUVE m I~IUTE r~ac m~ vrn 1s FLUY aom M FULY WM. ACTUAT~M SIW UIU HOT CUE THESE CPPMENTS 70 REN(~( m TIE. mtmTlm 12. IWU REST IS EFFECTED car IF )LL IMIYIWU VUM C~~DLS m IN W PRIOR 70 WE #OMIT OF M WUATlD4 SIGNL. TEL USED P051T101. SEQUOYAH NUCLEAR PLAN1 FINAL SAFETY ANALYSIS REPORT FIGURE 7.2.1-1 SHEET 8 FUNCTIONAL DIAGRAMS-SAFEGUARDS ACTUATION SIGNALS IRFVTSFn RY AMFNDMFNT 131 C-,4 C-3 c-Z c- I T AVG 1 AVG BYPASS SELECTOR TAvG TAVG $ OVEP- OvEa- r,ux HIG,, COOP 1 SWITCW(IOMTROL LCWZ -COP 5 AT COOF ROO BCXTOM SIGNAL F0Ne;S WER TEMF: (VZ) 4111 FULL LENQTW ROO (WNE IMPULSE .3T 6T IN?'CRWED~~TE I FROM ROD P051TION INDIC~W SY5TEM CHAM= WES~UPQ (214) RANGE) 4AffiE I @MEET I&) FHEETS) (SHEET% eHEET4) ~~>HEET 4) I 1 J -- -ER SLNGL wEuT4D4 FLU* CU~YWELS I I u - . . '1 'I I ,NOT44) 1 c~OrE4) I <~m~l-), (NCTL~) I ----I J ' I I I L L ---- - ---J I - I . 0145 , 3 UPPEfi THC SUMML" LM-5 JUTPUTS HAYE -$XED MIYJAiLI iiiJU\TiHI- MANUAL ROD I I I I I 11 I BANK A P051nON - - I I FILED MANUN I ROD SPEED nuTo - MANU AL r-+ r- - - - -9 S RODS ANAL- IN OUT WD ~P+D %I C.N% " I FULL LENGTH CONTROL 84NU5 LOW LDLO LOW LO-LO LOW LO~O LOW LO-LO BANK A BANK B BANK C 8ANKD I - --

I SEQUOYAH NUCLEAR PLANT FINAL SAFETY I ANALYSIS REPORT I FIGURE 7.2.1-1 SHEET 10 FUNCTIONAL DIAGRAMS-STEAM DUMP CONTROL (REVISED BY 4MENDMENT 131 THIS CONIICURATION CONTROL DRAWING 15 *IbINTAINtD BY THC ?A" ,,MI, AM" ,$ PAR, "r ,"r ,"A PR""A"AM "A7ARA-r PRESSURIZER PRESSURE CHANNELS AUCTIONEERED TA.VG LEVEL C,'AiuhLLi -& CSHEFT 3) / \ r -A, I I TP III n I ;I: II I ADJUSTABLE f -0 LOAD - -- -- - - .~ -- - - - - - , ------- - - - - - - - - i---' SETWIIJT I WITHIN CUARGiHG I r-------- 1 ----- - I 5Te;TIDN I1 I I I - - - - - - - - -- -- ~. -- - I l ADJUSTABLE 1 I TPkE5s~~~ CHRNNEL' + / 'L PHLL5U H E I r4 I * ' I I I SELEC-OR - - -- REFERENCE LE~L I I I ~PO~ITOM z ~~RMALLI( PP~~~! I SFTPOlhlT -~ b I 1 SELECTED) WITHIN I I C I I I I 1 (P- P I REF.) I--- - f- - - - - - - I INTERLOCKI '1 I t 1 I ALL ORIFICE 1501 ATION I CONTR3LLtR TOE3TROLLER 'VALVES CLOSED I1 I I POWLR RELIEF VALVE POW~R RELIEF VALVE 1 I 1 1 Li L)<,Wk) LlMt I'50~AT3h)l CLTCOVJ~ .-ME 150LATIOAJ I CONTROL MODE CONTROL MODE I 1 VALVE CSNTROL VALVE COUTUOL SELECTOR SWITCH SCLECTOn SWITCH I1 1 5ELECTOh SbITCH 1 I 7 COLD OVERPRESSUKE OPEN AUTO CLOSE PEN AUTO CLOSE I I CONTROL 5YSTEM I I I INTERLOCK I I I I I COL3 OVERPRESSJRE CONTROL 5Y5-EM I INTERLOCK I 1 I I I I 1 I ORIFICE ~- CONTROL CONTROL COIJTFOL STATION 5Tn- b SCLATION ( NOTt 3)) I 1 I WWER RELIEF I VALVE PCV5s.A 1 (EJ jr E 2) I I STATION (NOTE 3) (NOTE 3) 1 I NUItS 1 - I I ALL CIHCJ7S UN :H 5 SHFtT CAI NC: ?t3UNnANT I c + t i i z LC~.ZL CONTPUL oir~li I nrs ALI UTI.L~: s :;NALS LCCAL O*,EPhl~lt ACTLA1 IS lil AHM IN :.JNIPJI HOrM TO TO MODUkLITE MOCU. A,TF CHARGINS T ,i TO 3 CP'_Y;SIUT .Ni!l..:ii ,CN IN CONTROL ROO* -URN-OM VAHIABLE SPRAY VALVE SPRAY VALV E FLOW TUiihl CU hELTER ALL CEFTER -1 '2 CO~~TROL 01.1 IIITERL OCII CLOSE OFEN '+ LI;HTS SHOULD EL PRUVIDEU IN THE LOhTROL ROOM FPR EACH SPHAY VALVL TO INDICATE WHtN IT IS NOT FULLY CLOSEC BACK-UP COUTROL pCV-455% PCV- 455C HAil(-LP BLOC< UI ! HEATER- 5lGNn~ 4) (NOTE 4) hELrER5 EX( r PT LOCYL :SHEET 12; :5nEET 12) (~~EET 12, ('IHFFT I,?) VAIJE VALVE ULL CPtN PULL SIGNAL SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 7.2.1-1 SHEET 11 FUNCTIONAL DIAGRAMS-PRESSIURIZER PRESSURE & LEVEL CONTROL (REVISED BY AMENDMENT 13) SON CAD UNIT AND 15 PART OF THE Ti* PROCADAM DATABASE NOTE 1 . ALL C I RCUI IS CN THIS SHEET ARE NO: REDUNCANT HEATER COMPCNSATED REMOTE CONTROL STATION AUTOMPTlC HEATER TURN- ON REMOTE ST4TlON HEOTER INTETLIICK DH-OFF PRESSURE FOR GROUP C HZATERS RFMOTE GXTROL STATION Fy% CROUPAA ~(EAEC~S (CO~~TKOL (SELECTOR SWITCH) 2. GROUP A A AND CROUP O 8 HEATCRS MUST BE CN SEPARATE VITAL POWER SUPPLIES WI 1H TbE LCCAL CONTnOL SEPAR4TELl SO THAT bNY SINGLE FAILURE DOtS NOT CtFFAT BOTH OFF 3 BACKLP IILATLP :TATUS INU LA-1 Ch IV 1 "8LlkUL HLI.IM. OFF ON AUTO ON 4. PRtCAUT1l:lri 'h'llll n Bk IAKtN 10 AVGl[J MANUAL HEAT:? OPEfiATlON. WHl CH NCLLD ICAUSC HtAltH CAMAGt. IF TH; AiiiEk LEbEL IJNCD'VFRS THE HEAXSS. (NOTE 4) -SUL CONTQZL 5TATICN TURN-OFF TURN-WF TURN-ON TU R P.l- OFF TURN- CONlROL CONTROL GR~UP ,&-A 2' GROUP A-A GROUPB-0 GWL)P GmLp SI$N&L FCQ GRoUPC HKATERS HEATERS HEATERS HEATERS HWTEe Hsx- IONlRDL G30UP "EATERS (NOTE 3) (NOTE 3) HEATLRS SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 7.2.1-1 SHEET 12 FUNCTIONAL DIAGRAMS-PRESSIURIZER HEATER CONTROL (REVISED BY AMENDMENT 13) ?ON C*" UNIT AN" 15 PART ni 7°F PnnraoAM "AranaTr

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 7.2.1-1 SHEET 15 FUNCTIONAL DIAGRAMS-AUXILIARY FEEDWATER PUMPS START-UP (REVISED BY AMENDMENT 13) -ON r*n "NIT AN" ri PART THF nrr*n*si b4FETY INJECTION ~T~AM G, N~R~IOR GENEPAICK 2 STECM GiNtRRI3R J 5TE4n C.EN~RAT'>R 4 SI\F-TY SIGNAL 213 LOW LOW -EYEL 213 LOW CkVEc 713 LOW COvl 'LEVFL 713 LOW LOW LCVF\ 5, ,NA'L (SHEET 81 (SHEETI~) (SHEET\?) (SMELT 19) (SUE~TI~) '-54 E *7 3) BY BY @ ~ -- -- - -- - - - - - - - ~- --- - -- - BY VK-OUT TRsP Or %a>+. FEEC FP-I FP-2 a0 70 7 .~IC,JL.L J - -- - - ~- -- - MAIJ'JAL START POL IlOOM MAUUAL START hlOT REDUL) WANT - -- LocnL !MOTES ze 3) . - - - -- - - MAhlUAL STOP CONTROL ROOM MANUAL STSP LOCAL f NOTES 2 3) NOTES I TRAIN TWIN B A COhTROLS CONTROLS MAFP MAFP 1A~A 18-B UPkAKltH BREAKER 2 LOCAL CWTHCL UVtRRIDLS ALL OlHt,R SIGNALS 3 LOCAL OVCRHICL ACNATzS ALARM IN CCWrGL HOW 4 OPCNiSHUT INK1 CAT lOlr Ill CONTROL ROW 5 MUIOR OPthAllN~ LIGHTS IN CON.ROlI 3L+dZIC-OUT SECLJENCE 51(.UAC 6. INDIVIIIUL FOR EACH VALVE. 7. INDIVIJUAL FOR EACH PJ'IP EGUAPD SEQUE~~CE S'CP~L:~H.M)~---- R. THE IUHRINE PEED C3hTROL IS TYPICAL ACllllU IMT ~MIN~A~ l3N HAY NOT INCLUOC YELO CONTROC AANUAL START, C0h)TiiO~ ROO~I(NOTE 71---- TURBINE CRIVEN 3 THER: IS CNE TVhHlNE DRIVFNAUXILIA9Y FbtDWATER I TURB I ME PLKP RECEIVING 5T4RT II(,NALS FHCU THNN A AND 4Ah)URL START, LOCAL(NOTE5 7,3$7) -- I TRAllU B. I AANUAL STOP, CONTROL ROOM (NOTE 7) -- SPEED I ~LNL~L -TOP i~oc n:. Z 5 $7) CONTROL - bu~s ,C~~Z,I-'V,~>G %> GP\>AL. -- -~ ~ -- -- START !,TART MOTOR 3RIVtN CLOSE B~oNi30'v~N AUX FEED I50cATION ANU 5AMnLE LibiE VALVES (NOTES 165) EOR ALL ST- SENERATORS

NARROW RANGE STEAM GENERATOR LOW LOW WATER LEVEL CONTAINMENT r A 1 ADVERSE ENVIRON S/G "2 S/G *3 EAY RESET S/G LEVEL SETPT I I I I .---- + ----. , - -- - + - - --. (NOTE I) ACTUATE (NOTE 7) I I I I I I I I I I @I @I @I @I @I @ I (NOTE 51 I I I I I I I I I I I I \]-I t (NOTE 6) I I [NOTE 21 - [NOTE 21 (NOTE 3) It I I 4 I 0 0 I I , I , I) , , 7 MIN/HOLO NOTES: I. THE EAM 9E5ET CONSISTS OF FOUR MCMENTA4Y SWiTC4ES LOCATED IN T I ONE PER P~OiECTION SET I 2. BISTABLE5 LB-529F AN0 539F PROVIDE THE ACYEfiSE STEAM SfVERATCfi t THE NORMAL LEVEL SETPOINT IS PROVIDED BY BISTABCES L0-5299 ANC 41 A2 (NOTE 4) m TM - ELAPSE0 TIME IF SETPOINT REACPED IN ChE OR MORE STEAM Gf PHL = POWER HIGH LIMIT. 0 = THERMAL PCWER. 7. THE ACVERSE EAM STEAM GEVERATGR LEVEL SETPOINT LATCH-1'4 CONTRC FOUR UOMENTARY SWITCHES LOCATED IN THE PROCESS CABINETS, ONE 5 PROTECTION SET. STEAM GENERATOR "2 PROTECTION SET I STEAM GENERATDR *J (SHEET 19) PROTECTION SET I [SHEET 191 SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 7 2 1-1 SHEET 17 FUNCTIONAL DIAGRAMS ENVIRONMENT ALLOWANCE MOD & TRIP TIME DELAYRkqIGs'& DY AMCNDMCNT 13) CON r*n "Nl, AN" ri PART oi THF ,"A nrr*n*si 3. ONE COMMON ANNUNCIATOR WINDOW 15 SPARED WITH ALARMS GEVERATED PROTiCTlON SETS. 4. ONE COMMCN ANNUNCIATOR WIVOOW FCR EACH STEaU GEhERATOS 15 SHAF ALARMS GEhERATEO IN THE OTHER PROTECTION SETS. I I 5. AT IS A SRECIAL TEMPERATLRE SIGkAL USED FOR PCWER IhDICATION C A3 CONVERSION TO POWER IS PERFORMED PRIOR TO THE TIMER FUVCTION C (NOTE 0 6. THE FOLLOWING DEFINlTIONS APPLY TO THE FUNCTION GENERATORI NOTES: I. THE EAM RESET CONSISTS OF FOUR MCMEYTARY SWITChES L9CATED IN TkE ?GCCE5S CASIhE'S. ONE PER PSOTECTION SET. 2. B!STAELES LB-519F AND 549F PROVIGE ?PE ADVE95E STEAM SENERATOR LEVEL SETPOINT. THE NORMAL LEVEL SETPOINT IS PROVlCED BY BISTABLES LB-519B AND 5458. 3. ONE COMMCN ANYUNCIATOR WINOOV IS SHARED WITH ALARMS GENERAYED IN THE OT4ER PROTECTIGN SETS. NARROW RANGE STEAM GENERATOR LOW LOW WATER LEVEL CONTAINMENT f A > ADVERSE ENVIRON S/G #I S/G *4 EAM RESET S/G LEVEL SETPT I I I I .---- * ----, (VOTE I) ACTUATE [NOTE 7) .----+----. I I I I I I I I I I @I @I @I @I @I 4. ONE COMMON ANNUNC:ATOR WIUOCY FOR EACH STEAM GENERATOR IS SPARED WITH ALAFiMS GENERATED :N THE OTHER PROTECTION SETS. [NOTE 2) 5, AT IS A SPECIAL TEMPERATURE SIGNAL USE0 FOR POWER IUDICATIOY ONLY. CONVERSION TO POWER IS PERFORMED PRIOR TO THE TIMER FUNCTION GENERATOR. 6. THE FDLLOWING DEFINITIONS APPLY TO THE FUNCTION GENERATORI TM - ELAPSED TIME IF SETPOINT REACHED IN ONE OR MORE STEAM GEVERATORS. PHL = POWER HIGH LIMIT. 0 = THERMAIL POWER. [NOTE 2) 7. THE AOVERSE EAM S'EAM GENERATOR LEVEL SEiPOINT LATCH-IN CChTROL CONSISTS OF FOUR MOMENTARY 5W:TCHES LOCATED IN THE PROCESS CAEINETS, ONE SWITCH PER PROTECTION SET. I I I I I I I I I I I I T Fi I [NOTE 6) 8 t I I I , Ql 41 [NOTE 4) STEAM GENERATOR *I PROTECTION SET STEAM GENERATOR *4 [SHEET 19) PROTECTION SET I1 [SHEET 191 SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 7.2.1-1 SHEET 18 FUNCTIONAL DIAGRAMS ENVIRONMENT ALLOWANCE MOD & TRIP TIME DELAY LOGIC ,KtVIStU BY AMtNUMtNl THIS CONFIGURATION CONTROL DRhWlAC 15 UhlNThINCD DI THC SON CAD UNIT AND IS PART OF THE Ti* PROCADAM DATABASE I 4) 1 # 4b ! I I , I 4 , I 1 I I 4 A -4 [NOTE 4)

VOTES: I. THE EAM 'ESET CONSISTS OF FOUR MOMENTARY SWITCHES LOCATED IN THE PROCESS CABINETS. ONE PER PROTECTION SET. 2. BISTABLES LB-518F. 528F, 538F AN0 54BF PROVIDE THE ADVERSE STEAU GENERATOR LCVEL SETPOINT. T'iE NOqMAL LCVEL SETPOINT IS PROVIDED BY LEVEL 01STABLES LG-5180. 5280, 5380 ANO 5480: 3. OVE COUUCN ANNUNCIATOR WINDOW IS SHARED WITH ALARMS GENERATED IN THE OTHER PRO-ECTICN 5ETS. 4. ONE C3UMCN ANNUNCIATOR WINDOW FOR EACH STEAM GENERATOR IS SHARFO WITH ALARMS GENERATED IN THE OTHER PROTECTION SETS. 5. THIS LOSIC 15 SPECIFIC FOR PROTECTION SET 111. SEE SHEETS 17. 8. AND 20 FOR PH3TECTION SET I. 11. AND IV LJGIC. 5. THIS LQGIC 15 REOUNDAhT AND IS PERFORMED IN THE SSPS. 7. 4T 15 4 SPECIAL TEMPERATURE SIGNAL USED FOR POWER INDICATION ONLY. CONYERSION TO POWER IS PEilFORMED PRIOR TO THE TIMER FUNCTION GENERATER. 8. THE FOLLOWING OEFINITIONS APPLY TO THE FUNCTION GENERATOR, TM = ELAPSEO TIME IF SETPOINT REACHED IN MULTIPLE STEAM GENERATORS 15 : ELAPSEO TIME IF SETPOLNT REACHEO IN ONE STEAM GENERATOR. PHL : POWER HIGH LIMIT. 0 = THERMAL POWER. 9. THE ADVERSE EAY STEAU GENERATOA LEVEL SETPOINT LATCH-IN CONTROL CONSISTS OF FOUR MOMENTARY SWITCHES LOCATED IN THE PROCE5S CABINETS, ONE SWITCH PER PROTECTION SET.

NOTES : 1. THE EAM RESET CONSISTS OF FOUR MOMENTARY SWITCHES LOCATED IN THE PROCESS CABINETS, ONE PER PROTECTION SET. NARROW RANGE STEAM GENERATOR LOW LOW WATER LEVEL CONTAINMENT ' > PRES?URE S/G *I S/G "2 S/O '3 S/G #4 ADVERSE ENVIRON I I I I I .----+----. .----+----. .----+----. .----. ----. : (NOTE EAM RESET S/G LEVE SETPT .-------------*------------, (NOTE 1) ACTUATE \NOTE 7) 1 , I I I I I I I I , I I I I I I 1 @J @I (NOTE 21 @I @I (NOTE 21 @I @I [NOTE 21 @I @I (NOTE 2) @I 2. BISTABLES LB-517F, 527F, 537F AND 547F PROVIDE THE ADVERSE STEAM GENERATOR LEVEL SETPOINT. THE NORMAL LEVEL SETPOINT IS PROVIDED BY LEVEL BISTABLES LB-5178. 5278, 5378 AND 5478. 3. ONE COMMON ANNUNCIATOR WINDOW IS SHARED WITH ALARMS GENERATED IN THE OTHER PROTECTION SETS. PHL PHL NOTE 61 I I I I - 4. ONE COMMON ANNUNCIATOR WINDOW FOR EACH STEAM GENERATOR IS SHARED WITH ALARMS GENERATED IN THE OTHER PROTECTION SETS. 5,AT IS A SPECIAL TEMPERATURE SIGNAL USED FOR POWER INDICATION ONLY. CONVERSION TO POWER IS PERFORMED PRIOR TO THE TiMER FUNCTION GENERATER. 6. THE FOLLOWING DEFINITIONS APPLY TO THE FUNCTION GENERATORI TM : ELAPSED TIME IF SETPOINT REACHED IN MULTIPLE STEAM GENERATORS. TS = ELAPSED TIME IF SETPOINT REACHED IN ONE STEAM GENERATOR. PHL = POWER HIGH LIMIT. 0 = THERMAL POWER. I; - 7. THE ADVERSE EAM STEAM GENERATOR LEVEL SETPOINT LATCH-IN CONTROL CONSISTS OF FOUR MOMENTARY SWITCHES LOCATED IN THE PROCESS CABINETS, ONE SWITCH PER PROTECTION SET. - T 1 T (NOTE 31 T .-------------*------------a I 1 I t l I C --------.--------.-----------A - - - - - - ttr I* 3 4 Y [NOTE 41 STEAM GENERATOR 'I STEAM GENERATOR *Z STEAM GENERATOR "3 STEAM GENERATOR "4 PROTECTION SET IV PROTECTION SET IV PROTECTION SET IV PROTECTION SET IV (SHEET 19) (SHEET 191 [SHEET 9) [SHEET 191 SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 7.2 1-1 SHEET 20 FUNCTIONAL DIAGRAMS ENVIRONMENT ALLOWANCE MOD & TRIP TIME DELAY LOGIC ,RCVISCD BY AMCNDMCNT i"N r*" UNlT AN" 11 PART OF ,Hi TY* PR"T"n* "O*A*R*Tr I I I

S7-6.doc 7.6-5 SQN-26 For the above interlock functions associated with the Fuel Transfer System, the appropriate provisions required by section 4.14 of IEEE 279 are specified.

7.6.4 DELETED BY AMENDMENT 8 7.6.5 Loose Part Detection System (LPDS) System Description The Loose Part Detection System consists of sensors capable of detecting acoustic disturbances within the reactor coolant pressure boundary, associated cabling, amplifiers, and a data acquisition system. Two sensors are located at each natural collection region on the exterior surface of the reactor coolant boundary (e.g., reactor vessel upper and lower plenum and each of the steam generator reactor coolant inlet plenums). Should a sensor or sensor channel fail at one of these natural collection regions, the failed sensor or channel can be removed from service with the remaining sensor or channel providing loose part monitoring coverage for that area. The online sensitivity of the system is capable of detecting a metallic loose part that weighs from 0.25 to 30 lbs. and impacts with a kinetic energy of 0.5 ft-lb on the inside surface of the reactor coolant pressure boundary within 3 feet of a sensor.

The data acquisition portion of the system has an automatic and a manual mode of operation. The manual mode allows users to view system operation, to set alarm discrimination parameters, and to perform diagnostic tests. The automatic mode provides continuous monitoring for loose part events, displays real time system status, and the ability to record raw data for later analysis. The automatic mode also provides filtering to prevent false loose part alarms. The system also allows for manual inhibiting of a channel's alarm functions.

The LPDS is capable of performing its function following all seismic events, up to and including the Operating Basis Earthquake (OBE), that do not require plant shutdown. While recording equipment may not function without maintenance following the seismic event, the audio or visual alarm capability will remain functional. Portions of the system located within containment are compatible with the operating environment and consistent with minimum maintenance requirements and low-failure rates. 7.6.6 Spurious Actuation Protection for Motor Operated Valves The design of Sequoyah Nuclear Plant is such that the failure of any single valve to operate on demand cannot result in the loss of capability to perform a system safety function. However, in the case of possible inadvertent valve misalignment, the following motor operated valves have been identified as valves whose spurious operation could result in the loss of a system safety function.

FCV 63-1 FCV 63-67 FCV 63-98 FCV 63-3 FCV 63-72 FCV 63-118 FCV 63-5 FCV 63-73 FCV 63-156 FCV 63-8 FCV 63-80 FCV 63-157 FCV 63-11 FCV 63-93 FCV 63-172 FCV 63-22 FCV 63-94

Means have been provided to preclude such spurious misalignment. The design consists of modified control circuits for these valves to ensure that no single failure will be able to energize

I Best Ava~lable H~storlcal Image m z - a LT !- > C = m CI = -I> -ax X-IO 3wa W W [r 3 v, v, w-0 acn . Z 0 Y mw oa w a WE PROT(%:!!!D BY AMENDMENT 13 1 QCLT J a Z 0 I- zv, xn OC CCL I- a I- LT 3 I- Q W E 3 0- --. Z> - r C3W ZQ a2 CKW I- w < no --I 30 O v, OD ZJI- wa=, 020 0 2-W 0m-l - m +aa OOI- I zLTm 3a- kwm > -IZ - - CJ a Cd 2x 00 ea 0C Jcn ow ew I-0 20 OLT oa a Z Y a a I- EZ a + 0. 60 >H I- - a4 3 zl- -0 z 46 (r m' I- I 0 Y + + O+ LTW am a* I- mo OLT LTa W > Z 0 - I- -1 + OI- CW W a 3 C3w ZQ <I LTW + W < OD -J 30 O v, OD f I- 6 + c 3 W a z-l+ WQ3 020 0 z-w 0m-l - m = I-C6 oo+ zcm 3cK- LLW~ I- a WE -Iz aw CKZ 6 I- 0 Z I z 0 00 Y z--m I- -I-W 0 +a0 w- cn J- I- OI- -o> xcnw cKw -C C >- 60 >H I- - 66 3 a = -0 a a 44 c. : -I>- E -ax I- - a 3x 00 X-IO 3wa awe LTQ c3a -lm OK' LTW I-0 20 oa SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 7.6.7-1 LOW TEMPERATURE OVERPRESSURE

Figure 7A- 1 Table I, Ins trumen t Identification, Combination of Letters c Refer to Section 40 tor aplrlut#j + nates for Tabla 11. Revised by Amendment 13 VAL YES ++ Ga +I va/rr @WRY& opm) $7 Mud vdl* + Four way vdve Check va/ve Sl'op check vdve Angle va/vr Excess f/o~ check va/e - & Wrench q~4rstad "dk8 & ~k wr8nch oprr~tcd H/ve p Vacuum re/i.f v&e Quick openlhg va/ve- -W- as// va/ve Vacuum c/eaning ~hkt v~/ve Vdwm ckdning ink/ vd/.-f.ka8 I- BufferF/r v~/ve(nbm//u open) VAL VZ OPERATORS n - SYMBOLS & D;=phrdp opvafed va/vr 1 '@ 1 SCN,?~ w/vl~l c~h;nr~ Mah process Jnes - (FIOM dire c tion) NO. /Z /tne 8 ~1.e cfric motor oper~jed Y~IC Sery/i:e comecf,'on Aux;/lbry process he . ndtirsirummt con&&, m qv Fire ho+~ ee~po~ed NO. 5 /in e fo p-ucess //be So/enoid opera fed vdve Ccvnmon pwt no fed) See Nose /) +1 9 fxlknofcs va/ve &;Is =/used -I5 Reducer No. -*+ 4 line -( l- 0r1'f;i:e or nozzle ______- No. 4 line 4 I b Rup f ure d/k~ ~7 (9) Wtk~ ~d/~e faik open frdiner --~+(LosJ of arf~tlhhg md/+ Bucket trap . b-1 B Cmbindhbn a/> fiyfer and ,$+ ~enok m/ve posii~ner pressure rspu/a f or L;J Brea4dow orifice cr &notes hand jack - % Sfrs+hfPn;ny vanes denotes Ow and chsd fi' Desuperhee+eP . kinif switches Self con fained pre&ure ---I t B/~hd fldnge :q -,A, reducing vd/ue 4 I4 I- Flexible connection '4 ' p Con fro/ air sylp/l &r~u/a+iw rp/ air ~CCLII~U/~+OT Primary can ?&men? -- penetration ~~~9s. see/ +ype) Capillary f ubiq (~i//ed system) Revised by Amendment 13 Figure 7A-3 Mechanical Flow and Control Diagram Symbols

FZOtV SYMBOLS TEMPERA TURE SYMBOLS 0 4.us.sum indica for LEVEL SYMBOLS Gage y/am . -. Level Controler *-* Tempe~afurk? bo/b ~ifh cap;//ary /h a fernperatwe we// 2 8 Level Switch @ Thermocoupk I/? a 4 fempergfu~e wdN w/iC/I femperd/~/~e re corder mounted on MCR pane/ Level Indicating Transmitter Revised by Amendment 13 Figure 7A-5 Mechanical Application of Basic Ins trurnentation Symbols

-M*WO -.I rqqss LICT~ NU I *-- I ODLL.ZIW n loK .s mu ar rr U nmluararm muu ~ISW ro mw -1% mms mA*o m rrls P -mala wn r ar~.cD Im .Imcrm - m/m ".a Dl- nu I*-m fi ma m WR 4mL mur-. w- I *m, no. rosnm r nur rmw-nKm~rurro~ (Dl - Figure 7A-6 Mechanical Digital Logic Symbols (andlor), General

S8-2.doc 8.2-1 SQN-26 8.2 OFFSITE POWER SYSTEM 8.2.1 Transmission Network Description The Sequoyah Nuclear Plant is connected into a strong existing transmission network supplying large load centers. One unit is connected into the 500-kV transmission network and the other unit is connected into the 161-kV transmission system. The two systems are interconnected at Sequoyah through a 1200-MVA, 500-161-kV intertie transformer bank. Preferred electric power to the emergency buses and to start up and shut down the generating units at the Sequoyah Nuclear Plant is supplied by two physically and electrically independent circuits from the Sequoyah 161-kV or 500-kV switchyard through separate transformers to the onsite electrical distribution system, (refer to Figure 8.2.1-1).

For Unit 1 and 2, the normal power supply to the emergency buses is typically supplied by unit power through the unit station service transformers (USSTs). For Unit 1, the normal power supply to start up and shut down the generator is typically supplied by the 500-kV system through the main bank transformers and USSTs. For Unit 2, the normal power supply to start up and shut down the generator is typically supplied by the 161-kV system through the main bank transformers and unit station service transformers. Power to the emergency buses and to start up or shut down the generating unit may be supplied by the common station service transformers (CSSTRs).

Five 500-kV transmission lines connect one generating unit into the 500-kV system. Except in the vicinity of Sequoyah Nuclear Plant, the lines are on rights of way which are sufficiently wide to preclude the likelihood of failure of one line causing failure of another.

The 161-kV switchyard is the terminus for the second nuclear unit, the 500-kV intertie transformer bank and eight 161-kV transmission lines. Four 161-kV transmission lines terminate on each bus section. Two fuseless 84 MVAR 161-kV capacitor banks are tied to the 161-kV switchyard through double bus-tie breakers. Each bank is independently switched. These capacitors provide reactive voltage support for the 161-kV offsite system. Of the eight 161-kV transmission lines emanating from the Sequoyah 161-kV switchyard, one connects to TVA's Chickamauga Hydro Plant; one connects to TVA's Watts Bar Hydro Plant; and six connect to 161-kV substations that are an integral part of the 161-kV transmission network. Nine hydro plants, one fossil-fueled plant, and one nuclear plant are located within a sixty-mile radius from the Sequoyah Nuclear Plant. These plants are strongly connected through the 161-kV and 500-kV transmission networks to Sequoyah and have an installed capacity of more than 4000 MVA.

The transmission line structures of the 161-kV and 500-kV systems are designed to withstand medium loading conditions as specified in The Bureau of Standards Handbook No. 10 (National Electrical Safety Code Part 2).

To reduce the total number of acres of easement right of way required for the line connections to the Sequoyah Nuclear Plant, a number of the 161-kV lines are constructed on double circuit towers and, also, on common wide right of way. The 161-kV switchyard is designed with two main bus sections and is arranged so that the supply to the onsite power system, as well as the connections to the generator and 500-161-kV intertie transformer bank is maintained to one bus section for a failure of the other section. Four of the 161-kV lines terminate on one bus section and connect to Hiwassee, East Cleveland, VW Chattanooga, and Moccasin 161-kV substations. The other four 161-kV transmission lines terminate on the other bus section and connect to Hiwassee and Concord 161-kV substations and Chickamauga and Watts Bar Hydro Plants.

S8-2.doc 8.2-2 SQN-26 To make the thirteen line connections into Sequoyah, a number of lines must cross each other. When lines of different voltages cross, the higher voltage line crosses over the lower voltage line. Crossings similar to these are common throughout the TVA service area. The Bradley 500-kV Line crosses over the Sequoyah-Moccasin 161-kV Transmission Line. Assuming the 500-kV line falls at the crossover point, this will result in the loss of both the 500-kV and 161-kV lines. The four remaining 500-kV connections at Sequoyah and the seven remaining 161-kV connections will stay in service.

The Sequoyah-Hiwassee 500-kV Transmission Line crosses under the Widows Creek 500-kV Line, the Franklin 500-kV Line, and the Watts Bar No. 1 Line. If one of the three physically higher lines fall, this will result in the loss of two 500-kV lines. Three 500-kV connections and eight 161-kV connections will stay in service.

The 161-kV transmission line crossover at Sequoyah in the 161-kV transmission grid system consists of the Concord No. 1 line crossing under the following five connections from Sequoyah. They are Hiwassee 2, Hiwassee 1, Chickamauga No. 1, Watts Bar Hydro, and East Cleveland. Only two of the 161-kV transmission lines would be involved if either of the five aforementioned lines were to fall. Those lines remaining in service will be five 500-kV connections and six 161-kV connections.

The Tennessee Valley Region is located in a high thunderstorm frequency area and interruptions due to lightning do occur. Most interruptions are momentary in duration and have no significant effect on the operation of TVA's network of lines. The lightning performance for the transmission lines connected to the 161- and 500-kV switchyards at Sequoyah indicates that for the period January 1, 1994, through December 31, 1998, there were twenty-one 500-kV line interruptions and twenty 161-kV line interruptions attributed to lightning. Of these interruptions, six 500 kV and no 161-kV interruptions resulted in outages in excess of one minute.

Localized heavy conductor icing has occurred on some of TVA's transmission lines in years past. TVA's lines are designed to withstand these heavy icing conditions and no mechanical failures have occurred due to icing of any of the lines being connected into Sequoyah.

Several of the existing transmission lines that will be connected into Sequoyah do traverse fairly rugged terrain. Construction across this type terrain is not unusual for TVA transmission lines. Conductor spans in excess of 2,000 feet are fairly common and construction of spans of this magnitude are handled routinely. The longest spans which normally require the tallest transmission towers are river crossing spans. The 3,400 foot river crossing span on the Watts Bar-Sequoyah 500-kV lines is the longest span in the lines being connected into Sequoyah. The overhead ground wire in this span is marked with aircraft hazard markers and the transmission line towers are lighted for aeronautical protection. TVA's transmission lines are designed and constructed to eliminate damaging conductor vibrations. Conductor galloping is a phenomenon which normally occurs on lines constructed of small conductors during conductor icing conditions in conjunction with a continuous low velocity wind. Since TVA's higher voltage lines utilize larger conductors, galloping on them is extremely rare and is no threat to the safe operation of the lines being connected into Sequoyah. 8.2.1.1 Preferred Power System The intent of GDC 17 has been implemented in the design of the Preferred Power System by providing two physically and functionally independent circuits for energizing safety related load groups. This section identifies these two circuits and describes the general provisions made to achieve functional independence between them. Paragraphs 8.2.1.2 through 8.2.1.4 describe measures taken to provide physical independence between them. The Preferred Power System

SQN-25

- z- PLAN- EL 7Z.0 I PLAN- ELfCTR/CAL CUIVTROL BOARD i COOLING TO*ER PANEL O~M-28d 2 BLAUK 3 GEN &UNIT ST* SEW TQANS WHM I COMMON 51A SIP" TRANS b PLANT BOdiN UiiM 5 250" D-C>UPPLI LIGHTS -- 6 48" U-C b ZOY A-C SUPPLY 7 FREDUENCT '~ECOROK&-GRPGHIC HETCR- 8 RCCORDING VOLTMCTCP BUS POT 9 ~a~s-5 TFVP R~~OP-IIR (. GRAPHIC YHM 0 RADIATION NONITOQIVG PhL M-30 A RADIATION UONITORING Ph. 2-M-30 DC OISTRI3UIIOU R3ARD SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 8.2.1-6 EQUIPMENT-ELECTRICAL CONTROL AREA PLAN-EL 732.0 (REVISED BY AMENDMENT 13) THIS CONFIG"RAT10H COHTROL ORAWING 1s Mh,MT>,MEO 8" THE SON CAD UNIT *NO IS PART OF THE TV* PROCADAM OIT"A5t

S8-3.doc 8.3-42 SQN-26 All circuit breakers have trip alarm contacts to alert the control room operator of a tripped breaker. The ground indicator has an alarm contact to warn the operator of a distribution system ground. Metering on the distribution board includes: battery current, bus voltage, main and spare charger voltage, board charging current, and ground current. Metering for battery current and bus voltage are also located on the main control board. Battery discharge alarm is provided in the MCR.

Vital Battery Board V This board consist of two metal enclosed panels. Mounted on these panels are: the main distribution bus, battery and charger input circuit breakers, battery board main breaker and fuse, one transfer switch and various instruments for monitoring the board.

The fifth vital battery board is used to connect the vital battery V to a normal vital battery board when the corresponding normal battery is out of service for any reason. For this purpose there is a transfer switch composed of two inter-connected non-automatic circuit breakers separated by a metallic barrier. One breaker is connected to a distribution panel which is associated with vital battery boards I and III and the other breaker is connected to a distribution panel which is associated with vital battery boards II and IV. These breakers are normally open and provide isolation between the two panels. Refer to Figure 8.1.2-2 for a functional key diagram of the system and the normal vital battery interfaces. During periods when the fifth vital battery is in service, the circuit breaker connecting the fifth vital battery charger is maintained in the open position.

Tests and Inspections Prior to placing the vital DC system in operation, the system components were tested to ensure their proper operation. The batteries are tested during preoperational testing by discharging them with a load which simulates their loading during an AC power outage. The test is performed in accordance with IEEE-450, Recommended Practice for Maintenance, Testing and Replacement of Large Stationary Type Power Plant and Substation Lead Storage Batteries. The actual discharge current for the test is determined using the worst case load data. The basis for each actual individual load current value is either the measured value from actual test or a value calculated from manufacturer's data.

The charger will be checked for normal and equalizing voltage adjustability, 100 percent output capability, specified regulation with and without the battery connected, and panel instruments calibration. For the distribution board, circuit breakers are tested for proper trip operation, fuses are checked to verify that the sizes and types specified have been installed, and the board instruments are calibrated. 8.3.2.1.2 Nonsafety Related DC Power Systems There are three nonsafety related DC power systems: (1) the 24-volt DC Power Distribution System, (2) the 48-volt DC Power Distribution System containing a 48-volt Telephone Battery and a 48-volt Plant Battery, and (3) the 250-volt DC Power Distribution System. These systems supply power primarily for balance-of-plant systems. 24-Volt DC Power Distribution System This system consists of: a 12-cell lead-acid battery, two 24-volt battery chargers, a 200 amp power board, and 24-volt DC distribution panels.

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORl FIGURE 8.3.1-1 WIRING DIAGRAM-6900V DIESEL GENERATORS SCHEMATIC DIAGRAMS SH-1 SYMBOLS P---D C RELAY PNL h----DIESEL BATTCRI DISTRIBUTION PANEL [ ---LOCAL ENGINE CONTROL PANEL SYMBOLS 8--EQUIPMENT LOCATED ON PROT RELAY PANEL 1- HYDRAULIC COVFRNOR SPLFD CONTROL HYDRAULIC COVERNOR CONTROL SPEED SWITCH RELAYS X CAD MAINTAINED DRAWING SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 8.3.1-3 WIRING DIAGRAMS 6900V DIESEL GENERATORS SCHEMATIC DIAGRI\RyvSL!E$ BY AMENDMENT 18)

DO NOT RESET IF ICIG IT 1% 1 1 I 1' 1 x DIESEL GENERATOR IA-A PROTECTIVE RELAYS (TYPICAL FOR OSLS 18-8, 2A-A d ZB-Bl SYMBOLS: + ---6900" SHTDN BO B ---DSL RFLAY PANEL * ---UNIT CONTROL ROW [ ---roc~r cawnor srarraN m---YFR PNL (EXCITER PNLl h---OX BTRI DISTR PNL SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 8.3.1-4 WIRING DIAGRAMS 6900V DIESEL GENERATORS SCHEMATIC D1AGRP~M~Sv~5EHD %Y AMENDMENT

UNIT CONTROL BOARD IN MAIN CONTROL ROOM COMP CWLINC REAC LOWER COUPT CONT ROO DRlVL MLCH SIS PUMP IA-A CWLING FAN ,A- CWLlNG FAN IA-A CONTAINMENT AIR AUX BLOC CEN CONTROL ROW ELEC BD RM && & RETURN FAN ,A-A & SUP FAN 1A & AHU *-A & AH" A-A & I *-AUTO PULL A-P AUTO PULL P-AUTO 1 START STOP PULL P-AUTO PULL TO LOCK PULL TO LOCK IN MANUAL PULL TO LOCK l' PULL TO LOCK PULL TO LOCI psTw PULL TO LOCK 6 0 0 5 18 20 21 22 FROM 690OV UV ANN UV ANN f 1 DC S1Al PNL 6 z- oo u m @ :;::DR:;;g OVERCURRENT TRIP OFVlCF WITH LONG DELAY AND INSTANTANEOUS @@ :P;;OR:;;:: OVERCURRENT TRIP DEVICE WITH LONG DELAY AN0 SHORT DELAY NOTES 1. 48OV SDB lA1-A SHOWN AND 2A1-A AS NOTED. 1B1-B AND 281-8 ARE SIMILAR. 2. NSR INDICATES NON-SAFETY RELATED EQUIPMENT. 3. SEQUENTIAL TIMER APPLIES LOAD TO DIESEL GENERATOR 4. THE NORMAL & ALTERNATE DC CONTROL SUPPLIES AS SHOWN ON THIS DRAWING FOR THE NORMAL & BACKUP BUSSES ARE THE EXACT REVERSE FOR BOARD 2A1-A. 5 RECEPTACLE FOR 48OV POWER DURING OUTAGES. 480V SD BOARD lA1-A. 2A1-A AS NOTED (1B1-B & 2B1-B SIMILAR) 6. ALL CIRCUITS BEING SUPPLIED BY THIS RECEPTACLE SHALL HAVE THEIR OWN PROPERLY SIZED ELECTRICAL CIRCUIT PROTECTION. SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 8.3.1-7 WIRING DIAGRAMS-480V SHUTDOWN 'OAR? R!~I's% 3: NAGMLEENDMLJNN: 17 )

CONTINUSO BELOW FUTURE (ED 1A2-A ONLY) CONTROL & SERVICE AIR COMPRESSOR A (ED 1A2-A ONLY) NSR SPARE (ED 2A2-A ONLY) SPARE (ED 2A2-A ONLY) FUTURE REACTOR LOWER COMPARTMENT COOLER FAN 1C-A, 2C-A FUTURE FIRE/FLOOD MODE PUMP A-A (ED 1A2-A SPARE (BD 2A2-A) COMPONENT COOLING SYSTEM o o PUMP C-S ALT FDR (BD 1A2-A 2 2 = = " " ONLY) c c m m m m m m z z + + + + = = CONTROL ROD DRIVE MECH - - 7 7 COOLER FAN 1C-A NSR 0 0 m m 5 5 0 0 5 5 + + 11 0 0 " " D D m m > > < < > > z z 0 0 SHUTDOWN BOARD ROOM WATER CHILLER PKG A-A (ED 2A2-A ONLY) SPARE (ED 1A2-A) FUEL HDLG EXH FAN A (ED 2A2-A ONLY) NSR SPARE (ED 1A2-A) FUTURE EO (ED 2A2-A) NOR FDR, REACTOR MOV ED 1A2-A ALT FDR REACTOR MOV BD lAl-A -c nmi mz NnOWm XVI I Fz ma= n a-in 52:: ~r mpm mm -man n~ L ZE <a= mn n -0 in c n-P ALT FDR CONT & AUX ELDG rum n I 3 ,m mr"? Rmg;4 $2 nOZ "Zo $ VENT ED 1A1-A +<-mi 7 Og: ana a z a zo m 9 00- z --on0 OX < "?r I I0 "1 xr 4 >-OP rnc 22- Z5;: T, COW 02 NOR FDR, CONT & AUX ++c . - ;Lo? BLDG VENT BD 1A2-A 0x9 a 4 -m o an an mmr - z 5; 7%;; P VI- a m ALT FDR REACTOR VENT BD 1A-A nn u) c- ca. nm -4 a"? 0 r FF 96: 2 -9 Vi m 858 mzqw 0 n> VI P P -010 a- "? - ++++ -m an 7 o man 3mm q, p~ z 0 0 n+o- SPARE x: ,2 2 nzn :=; 7 +r oz o am0 m FUTURE -r > Z mam9 z o ao VI &;;;& LI DL NOR FDR 48OV DIESEL AUX ED 1A2-A om ALT FDR 480V DIESEL AUX BD IA1-A in "7 TU l $2 $2 STANDBY LTG CAB NO 1 (ON BD 2A2-A) z z a STANDBY LTG CAE NO. 4 (ON ED 1A2-A) + NSR CVC SYSTEM HEAT TRACE XFMR A3 (ED 2A2-A ONLY) NSR SPARE (ED 1A2-A) NOR SPARE 125V VITAL BTRY CHGR I-S BD 1A2-A FDRiSPARE 125V VITAL BTRY CHGR 2-SiBD 2A2-AI FUTURE (BD 1A2-A ONLY) FUTURE L----- m=s 2 0- 973 0 V, XI- C 2 C +r - 0 0 =e Z > Z i

0 OVI -Vim < z --4 m CONTINUED CONTINUED ~7 ABOVE ABOVE 125" DC POWER 125V DC VIC POWER m, SAC BOY REL SAC BOX REL i? i? P P < x 0 Z m o m o T Z : T L $2 ;I 0 + =; C m m D 0 -80 AN REL -80 AN REL 7 7 *- -- I m=z-.l I?-- i-C <ZW m , 0 - aom - 0 > I Z Z i REFUEL FLOOR POWER OUTLET 0-PO-317-SK/llD 250V BTRY CHGR 02 NOR FOR 480V CONDENSATE DEMINERALIZER WASTE EVAPORATOR MC FUEL & WASTE HDL BD B, L----A NOR. FDR FUEL & WASTE HDL 80 A, ALT FDR CHEMICAL & VOL CONT BD NOR. FDR CHEMICAL & VOL CONT BD ALT FDR 250V SPARE BTRY CHGR NOR. FDR PCV-31C-601 GEN COOLING SYS SECONDARY LOOP PUMP B RAW COOLING WATER CONTROL & SERVICE AIR COMPRESSOR D LO OC ~m0 >>z+m LS-31C-626 PCV-31C-602 PCV-31C-601 CONTROL & SERVICE SECONDARY L TURBINE BLDG OI 7 0- - L;o 0 z- c 02 ;D m 0 2 zo m m m -m - P Z 5 OP A r' 2 mx W - m r D 0 > Z i

OSL GFN IA-A LNG CONT PNL LMLR DSL LNG HEAT FXCH SUPPLY YLY FROM HDR A FCY-67-65 TABULATION OF 480" SUPPLIES I BOAR0 NAME I NOR SUPPLY I ALT SUPPLY I DIESEL GFN RELAY PANEL I I PNL 0-M-27A IN MAlN CONTROL ROCU NOR SUP. FROM 480" 50 80 IAl-A (SEE TABLE) EUERC DSL ENC HEAT KXCHANGFR SUPPLY VALVE FRCN HDR A hLT SUP. FRW lBDV SO 80 111-A VIA 180'2 DIESEL AUX BD 1A1-A (SEE TABLE) kLT SUP TO 180" DIESEL AUX 80 111-A (SEE TABLEI !8,, L -- NOTES: 1. EQUIPMENT DESICNATIONS SHOWN FOR 1A2-A DESIGNATIONS FOR 182-8, 2A2-A d 282-8 SIMILAR EXCEPT #.S NOTED I 2 ALL EQUIPMENT ASSICNED TO BOARDS 1A2-4 d 2A2-A IS SUPPLIED FROM POWR TRAIN A ALL FOUIPMFNT ASSICNFD TO BOARDS IBZ-B d ZBZ-B IS SUPPLIED FRCM POWER TRAIN B ALL EQUIPMENT 15 SAFETY RELATED UNLESS NOTED BY THE SYMBOL A 3 POWER HAS BEEN REMOVED FROM 0-FCV-61-12. -11. -351 L 365 BY OPENING THEIR RESPECTIVE ACB'S I THESE HEATERS CAN BE LITHER 5 OR 5 KY MOTOR OPERATED DAMPER H i~k2EP,lL:E:':l:::LLEDl ANON SAFETY RELATED EQUIPMENT I SEQUOYAH NUCLEAR PLANT I FINAL SAFETY ANALYSIS REPORT WIRING DIAGRAMS 480V DIESEL AUX ED 1A2-A SINGLE LINE

FULL FULL OPEN CLOSED aoM 0 Ooas I bco 0 ucos. boo OOOOO~S 2 OC" onos. 2 GCO 010s. I ac- 010los. 2 bco 0 Mlos. 3 boo O1.3~. 4 boo -10s. 5 aoM 0 010s. 6 VALVE POSITION INDICATION ISOLIO LINE DENOTES CLOSED CONTACT1 2" :2 0 0- -- SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 8.3.1-15 WIRING DIAGRAMS-480V DIESEL AUX POWER SH 1 (REVISED BY AMENDMENT 19) COMPRESSOR 2 COMPRESSOR 1 DIESEL 480" DSL 180V DSL 1A2-A 181-8 182-8 2A-A 2A1-A 2A2-A 28-B ZB1-B 282-8 NOTE 1 DIESEL START SIGNAL AND TIME DELAY RELAY ARE hPPLICABLE TO TCV-67-66 AND FCV-61-61 ONLY SYMBOLS EQUIPMENT LOCATED ON ASSOCIATED EQUIPMENT 7 iOVlPUFNT LOCATFD ON OIFSFL CEN RFL PANFl 0 EOULPUFNT LccArro ON AUX BOILER FLAME SAFFCUARO PANEL EOUIPMENT LOCATED ON DIESEL CEN LOCAL CONTROL PANEL

• EOUlPUFNT LOCATrD ON UNIT CONTROL BO IN MAIN CDNT RLI t EOULPUFNT LOCATrD ON UOTOR CONTROL CLNTFR [ EQUIPMENT LOCATED ON LOCAL CONTROL STATION DIESEL GENERt430,RS,~~~ SUMP PUMP A DIESEL GENERATOR IA-A ENGINE 1Al WATER HEATER I!!P.,h!,B,I?fAC$!Pi!&ALT,t:9? OVERLOAD HTR OR JUMPER DIESEL GENERATOR 1A-A BATTERY HO?!CHE!!E%,'N &!EL OF THOSE LlSTLD) DIESEL GEN 1A-A ROOM EXHAUST FAN 1-A (SCHEMATIC SHOWN TYPICAL OF THOSE SHOWN) DIESEL GEN IA-A ENG 1Al AEc4,U,?i'i!'J,E,L ?iLTH2$F2nOqMP SYMBOLS *- EQUIPMENT LOCATED IN MAIN CONTROL ROW + - EQUIPMENT LOCATED ON MTOR CONTROL CENTER [ - EQUIPMENT LOCATED ON LOCAL CONTROL STATION c(- EQUIPMENT LOCATED ON LOCAL PANEL EQUIPMENT LOCATE0 AT OR NEAR LOCAL CONTROL STATION A- EQUIPMENT LOCATLO ON AIR DRIER CONTROL PANEL SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT WIRING DIAGRAMS-480V DIESEL AUX POWER SH 3 (REVISED BY AMENDMENT 16)

CURRENr IN AMYt:WFS q a I..I 1a an ro .a m en ss E 3 g 5 pEp% 1. I I I 11. ..I. I .I I I I iI i I I II.1I.I. I : 1 LI 1-1 .... .... 1 .... I...1...Ill.... I..I.I FIGURE 8.3.1-18 480 Volt Containment Penetration Protection Revised by Amendment 13 Best Available Historical Image 8.3.1-19 120 A. C Con tainrnen t Penetration Protection Revised by Amendment 13 I0 10 40 lOwloaow~ E J! Z sJZ!HSH CURRENT IN AMP& XI0 FIGURE 8.3.1-20 1'25 13 I ELECTRICAL PENETRATIONS I I 1 48 1 X-168E 1 12' 1 111'-6' 1 531' I NIS II PAM (INCORE T/C I 7 I I I I I I I 49 I X-ISIL I 12' 1 111'-6' 1 355' 1 PROCESS INSTRUMENTATION CH I 9 SO I X-ISIL 1 12' 1 113'-6' 1 209' 1 OlVlSlONAL INSTR I 2(TR A1 51 1 X-IZOF 1 12' 1 712'-5' 1 290' I IISC POWER CIRCUITS I I3 SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 8.3.1-21 ELECTRICAL PENETRATIONS (REVISED BY AMENDMENT 18)

I ELECTRICAL PENETRATIONS I I I &f$/e/[ We / 8 / REMARKS SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 8.3.1-22 ELECTRICAL PENETRATIONS (REVISED BY AMENDMENT 13)

FL 682. 48 NOTES : PLAN - EL 734.0 THE DOTTED LINE INDICATES THE 2. MOVEMENT OF MAIN HOOK SOUTH SEQUOYAH NUCLEAR PLANT APPROXIMATE REGION IN WHICH OF THIS LINE IS PROHIBITED FINAL SAFETY AUX BLDG CRANE IS PROHIBITED BY MECHANICAL STOP. WHEN ANALYSIS REPORT BY ELECTRICAL INTERLOCKS. HEAVY LOADS ARE BEING FIGURE HANDLED THE STOP IS APPLIED FUEL STORAGE & HANDLING ADMINISTRATIVELY. (REVISED BY AMENDMENT 13)

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FIGURE 9.i 2-1 ISOMETRIC VIEW OF A TYPICAL SPENT FUEL STORAGE RACK

-- (BASED ON NOMINAL POOL DIMENSIONS) + N SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 9.1 .2-2 MODULE LAYOUT IN THE SEQUOYAH SPENT FUEL POOL & CASK PIT (REVISED BY AMENDMENT FLOOD MODE CONNECTION TO SPOOL PIECES SPENT FUEL PIT CASK LOADING ARC* SPENT FUEL PIT NOTES: 1 THIS SIMPLIFIED FLOW DIAGRAM IS NOT INTENDED TO SHOW VALVE TYPES, BUT RATHER THAT A VALVE IS LOCATED AS SHOWN. 2 NSR DEFINES NON-SAFETY RELATED BOUNDARY. REFUELING WATER PURIFICATION PUMPS SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 9.1 .3-1 SPENT FUEL PIT COOLING SYSTEM (REVISED BY AMENDMENT 13)

CONTROLS SAME 0 N SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 9.1.3-2 SPENT FUEL PIT COOLING PUMP LOGIC (REVISED BY AMENDMENT 13) 0 F F SPENT FUEL PIT COOLING PUMP B-B LINKAGE - PWR TRANSFER BKR PWR TRANSFER BKR MANUALLY CLOSED MANUALLY CLOSED - FOR TRAIN A FOR TRAIN B - SPENT FUEL PIT COOLING PUMP A-A BLACKOUT/ UNDERVOLTAGE SPENT FUEL PIT COOLING BACK-UP PUMP C-S FUEL FUEL TRANSFER TRANSFER NEW FUEL ELEVATOR (UNIT 1 ONLY) WINCH DRUM E TRANSFER SYSTEM PLAN VIEW E GATE (AND THE PLANT) REACTOR BUILDING CELL CONTAINER TRANSFER CAR CONTAINER (UPENDED) (UPENDED) E EXISTING DRIVE SHEAVE STAND & SHEAVE ELEVATION VIEW (LOOKING EAST) UNIT 1 AS SHOWN UNIT 2 OPPOSITE HAND SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 9.1 .4-1 FUEL TRANSFER SYS LAYOUT (REVISED BY AMENDMENT 15)

Figure 9.1.4-3 Typical Manipulator Crane 9*1*4-4 Typical Spent Fuel Pit Bridge

-. .+/WEIGHTS OPERATING- FLOOR TOP OF TRACK CANAL FLOOR SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 9.1 .4-5 ROD CLUSTER CONTROL CHANGING FIXTURE (REVISED BY AMENDMENT 13)

! $ LIFTING L, LINK W w PART OF INTERNALS 0 LIFTING RIG LC SLING ASSEMBLY 10 I G W w 0 0 01 CO 7 PLATFORM- LL W w 0 0 0, CO 01 LL 0 0 LO MATING SURFACE ELEVATION SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE .4-6 REACTOR VESSEL HEAD LIFTING DEVICE (REVISED BY AMENDMENT 13)

LLLIFTING HOLES PLAN VIEW OF UPPER CORE SUPPORT STRUCTURE SEE SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 9.1 .4-7 REACTOR INTERNALS LIFTING DEVICE (REVISED BY AMENDMENT 18) BARREL DETAIL A

BLACKOUT PS-70-24W/A CCS PUMP 1A-A COMPONENT COOLING PUMPS 18-8. 2A-A, & 28-8 LOGIC TYPICAL TO COMPONENT COOLING PUMP 1A-A I LSWW%-lcPUT EXCEPT NO PRESSURE I 1----j I EXCEPT NO PRESSURE I LSJI? INPUT T-----~ BKR CLOSED MANUALLY FOR BKR CLOSED MANUALLY FOR CCS PUMP C-S SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 9.2.1-7 COMPONENT COOLING PUMP LOGIC (REVISED BY AMENDMENT 13) ,HIS CO"F,C"R~,,ON rnN,ROL DRAWING 1s MA,N,A,NED 8" ,HE

R.C.PUMP CONTROL ROD DRIVE MOTOR COOLERS VENT COOLERS UNIT 1 ALL VALVE NUMBERS PREFIXED WITH (UNIT-67-1

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 9.2.2-6 ERCW PUMP LOGIC (REVISED BY AMENDMENT 13) LOCATION Oi CONTRCLS

• p;;w;;;R" r"NTR0, A LOCAL NOTE ERCW PAIRED PUMPS J-A & 0-A MR&PB L-E & N-E K-A & K-A ERCW PUMP Q-A LOGIC TYPICAL rOR ERCW PUMPS N-B, P-B. K-A UtMlNtKALlitU WATFR STORAGF I w Et wm tt < "1 = 00 WU UNW > tQX Vl w -w< "WL I OQ SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 9.2.3-1 FLOW DIAGRAM DEMINERIZER WATER STORAGE AND DISTRIDUTION SYSTEV AUX BLDG "59-124 --- 1 59 567 NOTES 1 DW SUPPLY TO SERVICE CONNECTIONS, SAMPLE SINKS, CMCRGCNCY SHOWCRS. CYC WASHCR AND OTHCR NON-SAFCTY RELATED LOADS NOT SIOWN 7 NSR DFFlNFS NON SAFFTY RFI ATtD BOIINDARIFS 0N Y THE PRIMARY CONTAINMENT PENETRATIONS ARE SAFETY KLLAILU 0-59-551 1 PRIMARY WATER MAKEUP PUMPS kL SbY U AUX BLUG PkIMAKY WAltK LUAUS AUXILIARY ) SEOUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT 1 81 510 1 81 509 1 81 SO8 FIGURE 9 2.3-2 ~~-p ~-~~ ORIFICE 1 FLOW DIAGRAM DEMINERALIZED WATER & PRIMARY WATER (KtVlStU BY AMtNUMtNl 16)

RAW COOLING WATER PUMPS STRAINERS RACKFI USH I NOTES 1. ENTIRE SYSTEM SIIOWN IS NSR-NON SAIETY RELITED 2 COOLING WATER DRAINAGE AS FOLLOWS H - 10 IUKHINt HUILUING SIAIIUN UKAlNAGt AH - 10 AUX HUlLUlNG SIAllUN UKAINAGt DC - TO DISCHARGE CONDUIT OF RESPECTIVE UNIT 3 UNIT 1 AND COMMON iIOWN, UNIT 2 TYPICAL 73 UNIT 1 1 SHttl NUMHtKS KtttK 10 IHk tlGUKt 9.2 /-Sitkl IANK PUMP > MlNUK LUAUS NU1 SHCWN(e g SAMPLL COULLKS, SAMPLL 011 COO FRS CONNFCT~ONS, SAMPIF SINKS) 6 THIS SIMPLIIIED rLOW DIAGRAM IS NOT INTENDED TO SIOW VALVt IVPtS, BUI KAlHtK IHAI fi VALVt 15 LOCAltU AS SHOWN SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 9.2.7-1 CONDENSER VACUUM PUMP RECIRCULATION- FLOW DIAGRAM-RAW COOLING WATER (KtVlStU BY AMtNUMtNl 21)

I I CAD MAINTAINED DRAWING I I AUXILIARY BUILDING TURBINE BUILDING CONTROL BLOC i i i i i NOTES ADO EOPT BLOC. UNIT 2 i 1 ENTIRE SYSTEM SHOWN IS NOT SAFETY RCLATCD(NSR)CXCCPT CRCW SUPPLY FOR i rLooD MODE 2 SEE FIGURE 9.2 7-1 FOR GENERAL NOTES i 1-21-809 i i SEQUOYAH NUCLEAR PLANT i FINAL SAFETY i ANALYSIS REPORT ~ FIGURE 9.2.7-4 FLOW DIAGRAM-RAW COOLING WATER ADDITIONA FOPT Rl Dr. UNTT AUXI IARI RJllDTNC (KtVlStU BY AMtNUMtNl ZU)

PRCFILTCR C At ILKIILILK AIR RECEIVER 1 AIR RECEIVER 2 NSR (NOTE 3) TYPICAL CONTROL AIR CONTROL I SERVICE AIR LOMPHtbiOHS NOTE 1 EXACT FLOW PATHS OR VALVE TYPE NOT REPRESEUTED EY THIS SIMPLIFIED UIAGKAM NOTE 2 CONTROL AIR CONTAINMENT PENETRATIONS ARE SIMILAR TO THE TYPICAL SHOWN BUT WITH DIFFERENT VALVE NUMBERS NOTE 3 THIS NSR BOUNDARY APPLIES TO NON ESSENTIAL CONTAINMENT PENETRATIONS ONLY NOTE 4 NSR DEFINES NON-SAFETY RELATED BOUNDARY NDTF 5 X TNDTCATFS VA VF FAT 5 21 OSFD FINAL SAFETY ANALYSIS REPORT I I CAD MAINTAINED DRAWING 11 STATION AIR COMPRESSOR O FIGURE 9.3.1-1 TURBINE BLDG CONTROL AND SERVICE AIR SYSTEM SIMPLIFIED FLOW DIAGRAM (KtVl5tU BY AMtNUMtNl 11)

SERVICE AIR HEADER 1 NSR - SP,J"L P,_Ct REMOVED 4ND BLIND iL4NGE INSTALLED I I 1 I 1 , ,' TYPICAL REACT~$ITB~ILDING REACTOR BUILDING UNI I 1 NSR DEFINES NON-SAFTEY RELATED BOUNDARY I SEQUOYAH NUCLEAR PLANT FINAL SAFETY I ANALYSIS REPORT FIGURE 9.3.1-2 FLOW DIAGRAM SERVICE AIR SYSTEl I 7 C00~1NC WATER IN SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 9.3.1-3 AUX AIR COMPRESSOR LOGIC (REVISED BY AMENDMENT 17)

, TIT PASSIVE SUMP - JPPER COMPARTMENT /- rLANCED COVERS TO BE OWrR COMPARTMrNT REACTOR REFUELING CAVITY/ CANAL DRAINS ANNULUS DRAINS REACTOR BUILDING NSR ACCUMULATOR RM 3 ACCUMULATOR RM 4 ILOOR DRAIN FLOOR DRAIN /SCREEN FL ELEV 693 OUTSIDE-INSIDC CRANF WAI I CRANE WALL OUTSIDE-INSIDE CRANC WALL CRANF WA 1 GH LEVEL ALARM P EJECTOR PUMP CONTAINMENT PIT SUMP NSR I FVF A ARM ROTTOM SLIMP AUX R.B. FLOOR^ DRAIN SUMP di EQUIP DRAIN SUMP NSR (POCKET SUMP) NSR 1 NS7 DEIINES NON SAIETY RELATED CUMt'ONtNIS 7 TYPICA DRAW CONFlGLlRATlON SHOWN SPECIFIC INTERCONNECTIONS NOT CXPLICITLY DCTAILCD 3 X INDICATES VALVE FAILS CLOSED SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT

-- REACTOR BLDG UNIT 2 ANNULUS REACTOR BLDG UNIT 1 ANNULUS ELEV 653 ALL PIPING/EQUIPMENT NON SAFETY KtLAltU (NSK) TYPICAL DRAIN CONr IGURATION SIOWN SPECIFIC INTERCONNECTION NOT FXP ICTTI Y DFTATI FD AUXILIARY BUILDING FLOOR & EQUIPMENT DRAIN SUMP NSR SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FLOW DIAGRAM AUX BUILDING FLOOR & EQUIP DRAIN' I I CAD MAINTAINED DRAWING I I

HOLDUP TANK B HOLDUP TANK A PRIMARY IATLR -- -- PRIMARY IAT:R 1UN tXCH FTI TFR R NSR DErINES NON SArETY RELATED BOUNDARY SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 9.3.4-3 CVCS CHEMICAL CONTROL FLOW DIAGRAM (KtVlStU BY AMMtNUMtNl SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 9.3.4-4 CVCS CHEMICAL CONTROL FLOW DIAGRAM (KtVlStU BY AMtNUMtNl 15) BORIC ACID TANK B BORIC ACID TANK C BORIC ACID TANK A TRANS PUMPS TRANS PUMPS .3 w 1 NSR

SPOOL PltCt -~ F3 \ SPOOL 6:) PIECE 2-84-51 : 7 84 517 UNIT 2 CIARCINC H0USIkK PUMP DUTSTDi CONTATNMrNT INSIDE CONTAINMENT UNIT 1 NSR DEFINES NON-SAFTEY RELATED BOUNDARY SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT

1. 2.

3.

1. 2. 3.

4.

5.

S9-4.doc 9.4-19 SQN-26 and electrical panel ventilation fan is provided within each diesel room at the air supply opening to the room to deliver cooling air to the generator air intake and to the interior of the generator's electrical control panel. The original battery hood exhaust fans have been isolated from the system and abandoned in place. Each of the diesel generator room fans is connected to its respective diesel generator engineered safety power supply. One exhaust fan will automatically start upon diesel generator start. The generator and electrical panel ventilation fan will run when the diesel is running. Approximately 40,000 cfm of fresh air is routed through each diesel generator room when one exhaust fan is operating.

Each diesel generator unit is provided with a fan designed to exhaust approximately 3500 cfm of air from the elevation 740.5 electrical board room. A roof mounted air intake admits outdoor air to each electrical board room. Other building exhaust fans provide ventilation for the lubricating oil storage room, fuel oil transfer room, CO2 storage room, toilet room, radiation shelter room, and muffler rooms.

The oil and storage rooms are normally ventilated at all times while the electrical board rooms and muffler rooms are ventilated as required to remove heat during warm weather. However, the CO2 and lube oil storage room, diesel generator rooms and electrical board rooms exhaust fans are stopped during a CO2 initiation. Each exhaust fan and the corridor air intake vent is provided with motor-operated shutoff dampers designed to close tight when the fan is not running.

A backdraft damper is installed in the duct between the air intake room 1A-A and the CO2 storage room in order to prevent CO2 backflow into the diesel generator air intake room in the event of a CO2 system rupture.

Thermostatically controlled electric unit heaters are located within the diesel generator rooms, equipment access corridor, storage rooms, radiation shelter room, and electrical board rooms. These heaters are designed to maintain the rooms at not less than 50°F when 15°F outdoors.

Thermostats in each air exhaust room are designed to stop all operating diesel generator room fans upon a drop in room exhaust air temperature to below 68°F if the diesels are not running. The thermostats will automatically start the exhaust fans upon room temperature rise to 90°F. The thermostats will also start the standby exhaust fan, during diesel operation, when the room exhaust air temperature exceeds 90°F.

In the case of a tornado event the Diesel Generator Room Air Intake Dampers are opened by manual action as noted in section 3.3.2.2. For damper control see Figure 8.3.1-14.

9.4.5.3 Safety Evaluation The Diesel Generator Ventilating Systems are required to operate for maintenance of plant safety in the event of natural disasters or plant accidents. The diesel units are redundant to each other and the diesel generator room main exhaust fans for each diesel unit are provided in pairs for reliability. In the diesel generator building, the diesel generator room exhaust fans, electrical board room exhaust fans, and the generator and electrical panel ventilation

SQN

1. 2.

3.

4.

5.

<AUX BLDG EXH CONTROL ROD DRIVE LOPI KOOM ,, TYPTCAI FOR PDT-30-148 PDT-30-119 t C-30-1 PDT-30-1 58 PU I SO-213 PENETRATION -COO TNG COIl TURUINC DRIVCN AUX rEEDWATER PUMP KUUM ,707 - ~~ 1 10P9 NOTCS DAMPIR LCGCND F DENOTES FIRE DAMPER 1-31~-,098 ?-3,c-!?o? S DENOTES AUTOMATIC ISOLATING DAMPER A DENOTES AUTOMATIC OPERATING DAMPER B DENOTES BALANCING DAMPER 1 31,- TNJFCTTON UNIT-.. M DENOTES MANUAL OPEN/CLOSE DAMPER SYSTEM PUMP U UtNUltS BACK UKAt I UAMPtK KOUMS + / KM ULNOILS KLMUIL MANUAL OPLKAIING UAMPLK X DENOTES DAMPER IAILS CLOSED \ HOLD-UP -. ~-/ U UtNOltS UAMPtK tAlLS UPtN - TANK RMS 0 UtNUILS UAMPtK lHAl tAlLS AS 15 f B 1-31~-,08~ B 1-JIG-1081 2 AIR LOW PATIIS ARE REPRESENTATIVE AND NOT TNTFNDFD TO TIIUSTRATF FXACT ROOM (5) TRAVFRSFD f f f f WASTC HOLD-UP WASTC CVAP FCCD TANK ROOM PUMP ROOM PIPE 'f ,~ 11 613 1 UNIT ONE GENERAL VENTILATION SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 9.4.2-1 FLOW DIAGRAM AUXILIARY BUILDING HVAC (KtVlStU BY AMtNUMtNl 15) THIS CONFlCURITlON CONTROL ORAIINC 15 MAINTAINED BY -HE AIK COOLtU CONDENSING 480 VOLT TRANSrORMER KUUM 1 A 480~ BOARD RM 28 480~ EOARD RM I B rl 749 n CONDITIONING SUPPLY EMERGENCY AIR FROM CONTROL BLDG CONDITIONING SUPPI Y FROM CONTROL BLDG StALtU UOOKmm AUX CONT l-dlC-927 SEALED DOOR TSOl ATTON BOARD RATTFRY BOARD 9- ISULAI ION 480V SHUTDOWN ROOM 180 V SHU I DOWN 'OAK' KM 18UV SHUIUUWN , 480 V SHUTDOWN RDARD RM 1R1 BOARD RM 2A2 HUAKU KM lH2 R R + 1-31C-131 I-AIL-322 I' I 3IC-901 6 9 KV SHUTDOWN BCARD ROOM B COIL 1-11L-Bli 1-AIL-930- % AHU 16-6 ' AHU 1A-A FILTER # 1 ilC 911 -z AIK LONUI I IUNING / ts-s nt l-;llc-95+ 1 I-AIL-Y~I~ UNITS IJNTTS f' "" "j B +tz MECllAhICAL EOUIPMENT ROOM MtCHhNlCAL kUUIPMkNI KOUM tZ3Jb 480 V SHUTDOWN 480 V SHUTDOWN EL 734 0 BOARD RM 1A2 ROAR3 RM 1Al AIR FLOW DIAGRAM AUXILIARY BOARD ROOMS EL 749.0 SHUTDOWN BOARD ROOMS EL 734.0 NOTES 1 REiER TO iICURE 9 4 2-1 rOR GENERAL NOTES iL 719 0 5TH VITAL BATTERY ROOM SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 9.4.2-2 FLOW DIAGRAM AUXILIARY BLDG HVAC (REVISED BY AMENDMEhT THIS CONFlCURITlON CONTROL ORAIINC 15 MAINTAINED BY -HE OATTCRI ROOM IV [IATTCRI ROOM IATTCRI ROOM OATTCRI ROOM A1H HANDLING AIR HIIN", TNC UNIT 2*-A UNIT ?A-A 180 V OOiiRO ROCM 10 MECHANICAL IOUIPMLNT MECHANICAL EOUIPNLNT 51C-2153 2-31C-811 3 7-31C-808 il 119 L kL ilY 0 AIR FLOW DIAGRAM AUXILIARY BOARD ROOMS EL 749.0 ('A TRAIN) i 751 0 IIC-7PI +80 V BOARD ROOM 1A 1 AIR FLOW DIAGRAM AUXILIARY BOARD ROOMS & MECHANICAL EQUIPMENT ROOMS EL 749.0 d EL 763 0 B' !RAIN1 DAMPFR I FGFNO CMCRCCNCI CIS TRIATMCNT RM COOLCRS CLI d SPtNI kUtl "11 PUMP LOOLtHb 0-3IC-626 T PENETRATION RM COOLERS PENETRATIlON RM COOLERS hBLI5 HM ABCTS RN - AUTOMATIC OPCRATING '@A .OR ISOLATION 1 +, + RAl ANCTNG (MANUA ) i 754 0 NOTES 1 REFER TO FIGURE 9 1 2-1 FOR GENERAL NOTFS &/ + BACK DRAFT DAMPER + :RILLE 4- FIRE DAMPER EL 714 O AUX FIFOWATFR I RORlr CCW d AUX IEEDWATER PUMP COOLERS ACID TRANSFER PUMP PCNCTRATION RM COOLCRS PCNCTRATION RM COOLCRB I-P- 2-11 30-146 " + -P- -P- i i 6PO 0 BIZ PUMP SIS PUMP CENT CllARClNC EihFEANC PUMP COOLER COOLER COOLER ?P- PIPE CHPSE COOLLPS PIPE CHASE COOLERS -P- + PENETRATION RM COOLERS PFN~TRATION RM ins 2-31C-1830 -P- EL 659 0 SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT RiiR PUMP COOLER FLOW DIAGRAM AUXILIARY BLDG HVAC (REVISED BY AMENDMEhT I I CAD MAINTAINED DRAWING I I -P- '* + + AUXILIARY BLDG RIR PUMP COOLER C 5 PUMP COOLER f C 5 PUMP COOLER RiR PUMP COOLER RiR PUMP COOLER il 761- 1 ECTS ROOM UNIT 2 A1H LNIAKI - HOT REACTOR COOLANT SAMPLE SYiliM NODULL SAFETY INJECTION SYSTCI PUMP ROOMS FLOOR DRAIN C0 1 FrTnR TANK ROOM LUNlAlNMtNl biiHAY PUMP ROOMS ----- . . F, OOR DRAIN cot I 2-31C-17I1 NOTE -- 5 1 REFER TO FIGURE 9 4 2-1 FOR GENERAL NOTES + + RESIDUAL HEAT HtMOVAL PUMP HOOMb PlPi il 551 1 - SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 9.4.2-4 FLOW DIAGRAM AUXILIARY BLDG HVAC (REVISED BY AMENDMEhT UNIT TWO GENERAL VENTILATION WEST MAIN STEAM VAU T FXH FAN9 ROOF DFCK EAST MAIN STLAU VAULT EXH FANS AUX BUILDING HAOLAIION - AUX BLOC IXH VENT NU 1 LS (1) RCFCR TO FIGURE 9 4 2-1 FOR GCNCRAL NOTCS SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT I I CAD MAINTAINED DRAWING I I FUEL HANDLING AREA GENERAL VENTILATION FLOW DIAGRAM AUXILIARY BLDG HVAC (REVISED BY AMENDMEhT

, SHIELD BLOC NOTES 1 UNIT 1 SHOWN UNIT 2 SIMILAR 2 ALL DAMPERS FAIL CLOSE EXCEPT AS INCICATED BY 0 FOR FAILS OPEN. DAMPER LEGEND A AUTOMATIC OPERATING FOR NORMAL SERV:CE b IsoLATIoN MANUAL (FOR OPEN-CLOSE OPERATION) BALANCING - BACKDRliT SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT --\I- FIRE FLOW DIAGRAM-REACTOR BUILDING HEATING & VENTILATION AIR FLOW (REVISED BY AMENDMENT I I CAD MAINTAINED DRAWING I I POST ACCIDENT SAMPLE UNIT NO.l SAMPLE UNIT NO. 2 SIMILAR PASF FILTER SYSTEM r------------------------------------- FROM AD OUTSIDE IIR INTAKE I000 L------------------------------------- 2 BALANCING DAMPER GENERAL SUPPLY FAN C1 1DhlP-031-0107- RAt ANcrNC j DAMPER- TO POST ACCIDENT SAMPLINC ROOM NO ? X INDICATES DAMPER rAILS CLOSED SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 9.4.10-1 FLOW DIAGRAM HVAC (REVISED BY AMENDMEhT ? Rl D 031 0139 -~ - - *; 7' - SAMPLC HOT - - SlDF 5I"i t FILTER ISOLATION OlMPtHS - f -SUPPLY AIR FAN A-l

S9-5.doc 9.5-6 SQN-26 The sound-powered telephone systems are completely independent of power, each other, and all other systems provided. As long as a complete metallic path exists between instruments, communications can be maintained since the instruments supplied with these systems are very rugged and will successfully withstand high shocks, negligence and abuse. If permanently installed wires are rendered unusable for any reason, a temporary pair of wires can be used with the sound-powered instruments.

The design of the evacuation alarm system is such that it will not likely be inoperative for the following reasons:

1. Two independent widely separated operating centers are provided.

2. Duplicate timers located in widely separated bays are furnished with provision for manual override in case both fail. Each timer is powered by a separate alternating current source.

3. Duplicate actuating relays are provided in each remote control unit.

4. Independent contactors are provided, each controlling a group of sirens. Failure of one contactor will not affect the others.

5. Two sources of diesel-backed alternating current power are provided for most control units with provisions for annunciation upon failure of each source.

6. Power failure to the timers and to the remote control unit actuating relays is annunciated.

Refer to Table 9.5.2-1 for availability of intraplant communications during various postulated conditions.

9.5.2.4 Inspection and Tests Two communications systems were covered by preoperational test (TVA-11):

1. The sound-powered telephone systems provided for the backup control center, health physics office, and diesel building shielded room.

2. The evacuation alarm system.

All systems were also carefully installed and checked for proper operation initially by construction forces. Maintenance is performed on an as needed basis and includes such items as checking for proper switch operation, checking for proper operating levels, visual inspection, etc.

The most comprehensive testing, however, results from the heavy daily usage of the equipment and the subsequent reports by the users. Power failures in the systems are annunciated. 9.5.3 Lighting Systems 9.5.3.1 Design Bases There are three basic lighting systems in the plant (see section 9.5.3.3 for the Diesel Generator building lighting) designated as follows: normal, standby, and emergency. These systems are

ETSl ETS2 OIL STRAINER ESSENTIAL R COOLING WA MAIN LUBE OIL AND PISTON COOLING OIL PUMPS ro- c-- mmc -mmc n mx morr i H ir;co z- moru rn~ -v, 2 5 Y m 0 a I m yl = < +ffl 2 --I L4 m - < v, m > Q z c rH< <=> zF1 z "'c m%n m r amm oi> ;o<;o --I r > Z --I -WATER CONVECTION FLOW- IMMERSION HEATER TURBO SOAKBACK - WATER AND LUBE OIL - OPERATING FLOW OIL PUMPS 0 WbTER - LUBRICATING OIL -

, TO ENCINE COVCRNOR *4

• 3 1 1 7 7 L \ STARTING AIR TANKS AIR COMPRESSOR 1 (WEST END OF ROOM) AIR COMPRESSOR 2 (EAST END OF ROOM) STARTING AIR TANKS DIESEL GENERATOR SET IA-A, 18-8, 2A-A OR 28-8 CAD MAINTAINED DRAWING SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 9.5.6-1 DIESEL STARTING AIR SYSTEM FLOW DIAGRAM (KtVlStU BY AMMtNUMtNl 20)

P STEW TO MFPT TO .*A ec I 14.~11 L 101215501 12184H LP GTEAYI TO MFPT "A' IF TURRINC 'LV LI~~U~DIN~ .c 1 I' I WIIIIN~ (iY,"lCAl rOff 3 1.P. Tb'?RlNFS] liljw 30D3F ,om A,, 175 7F RH* w 1037811 OiU "'ACTOR .~- 405M51 1078 AH MAlN LONDrNSE" 3QRBP 444 di El LF .. EX( ii.nlti LP il iliATrI< em%" 53331, $ %: IISljBH XaSi 14116Wh 3B4 &4H 161 5P 'IOTWLIL 174 MH TO NO 7 i i C3lP:YSATE DTUYC"RLLi" UINLHNYC" il_l"ib" . .- VF iniw~, 111 311 ?>"SF 230 7ii 166011 1730' ,2021 BLOWMWNl 1i:K STFIW LITNE' ATCi 122 3F -- -. - *, SH I SEQUOYAH NUCLEAR PLANT FINAL SAFETY I ANALYSIS REPORT FIGURE 10.1-2 UNIT 1 HEAT BALANCE 100% MW THERMAL I I CAD MAINTAINED DRAWING I I

ANT I I CAD MAINTAINED DRAWING I I CONDENSATE TRANSFER PUMPS - CONDENSER VACUUM PUMPS FLOW DIAGRAM-CONDENSATE (KtVlStU BY AMtNUMtNl 11)

C I - B C EL 652 5 HOlltLL PUMP5 CONDENSER --, . , ,-, 1,-174 2-678 TO MFPT CONDENSERS t a 14-549 11 518 11-150 _ 1 ;'.I @ CONDENSATE DEMINERALIZER EL 706 0 OUTSIDE OF LNSLUt Oi TURBINE TURBINE BLOL HLUL ENTIRE SYSTEM SHOWN IS NO\ SAFETY RELATED il 667 i CONDENSATE DEMINERALIZER PUMPS SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 10.4.7-1 FLOW DIAGRAM-CONDENSATE (KtVlStU BY AMtNUMtNl 15) THIS CONFlCURITlON CONTROL ORAIINC 15 MAINTAINED BY -HE

TURBINE DRIVEN AUXILIARY FEEDWATER PUMP 1A-S NOTES SOLENOID TRIP OPERATES AT 11U SPEED AND RESETS AUTOMATICALLY M~CHANLLAL OVLHSF~~O OFLHAIL~ A! srwu nrro Mual ~tsk~ MANUAI 1 Y 2 MANUAL CONTRJL WHCN IN THC AUXILIARY CONTROL hlODC IS ACCOMPLISiED AT Tii CONTROLLER 1 ACCTDINT ZICNAl TO TRAIN A VAVIC IS TR-A WlTli TR-R RUriiRiD TR 3 VALVES WILL BE TR-B WITH TR-A BUFFERED 1 MANUAL PUSHBUTTON ON FRONT FACE OF CONTROLLER TAKES PRECEDENCE OVER REMOTE AUTOMATIC SIGNAL MCR INDICATING LICIITS STATUS ACTUAL CONTROLLER MOLE (AUTO OR MANUAL) SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 10.4.7-4 TURBINE DRIVEN AUX FEEDWATER PUMP LOGIC (KtVlStU BY AMtNUMtNl 19)

0- Y p 9 2 g 0 V) H CD -I 2 ow - I c= -1. 3 0 'J' <-a 7l m 30 -I 3- s -9, 0 c 7 Xru -i -2.0 m -rt V) -'a m --I q3. km c l-l- XI -rl a- < mn m row V) a- 5 --I rt C m I -z 3 m NPSH IN FEET

~evised by Amendment 13 "

- - - - - - TURBINE BUILDING VENTILATION .. . - - - - I I 2 0 3 0 4 0 BLOWDOWH RATE f 10.4.8-3 Effect d Bladom Rh 1-13! Release .-* Revised by Amendrneilt 13 NO. 3 HEATER DRAIN TANK PUMPS NOTES 1 FNTTRF SYSTFM SHOWN IS NON SAFFTY RELATED 2 UNIT 1 SHOWN, UNIT 2 SIMILAR 1 CONDENSER A 1 ('I\ SGR FI ASH TANK DETAIL A2 NOTFS 1 LNIIKL SYSItM SHOWN IS NUN SAttlY RELATED 2 UNIT 1 SIIOWN, UNIT 2 SIMILAR I SEQUOYAH NUCLEAR PLANT NO. 7 HEATER DRAIN PUMPS MFPT CONDENSER DRAIN PUMPS FINAL SAFETY I ANALYSIS REPORT I FIGURE 10.4.9-2 FLOW DIAGRAM LOW PRESSURE HEATER DRAINS a VENT (REVISED BY AMENDMENT I I CAD MAINTAINED DRAWING I I

CASK DECONTAMINATION COLLECTOR TANK 669 0 CASK DECON COLL TK FILTERS ENTIRE SYSTEM SHOWN IS NON-SAFETY RELATED SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 1 1 .2.2-2 FLOW DIAGRAM-WASTE DISPOSAL SYSTEM (KtVlStU BY AMtNUMtNl ZU)

TANK F 1 WASTE GAS COMPRESSOR PACKAGI BOUNDARY ' WASTE GAS COMPRESSOR PACKAGE B SIMILAR NUltS 1 NSR DErINES NON-SArETY RELATED BOUNDARY FINAL SAFETY ANALYSIS REPORT FIGURE 11 .3.2-1 FLOW DIAGRAM-WASTE DISPOSAL SYSTEM (KtVlStU BY AMtNUMtNl 16) CAD MAINTAINED DRAWING I I CAD MAINTAINED DRAWING I I

TEMPERATURE PRESSURE AND T5M~~~~~REPROFkE PRESSURE - - - - 700 -- - - -- 3500 I-------- 7 T 3OoO urn a I I \ I \ 2500 ,2 \ --- ---_ __------ -? - I -\ / 1 - 2000 a - E I I \ I \ / / \ lsoo \ \ I \ 1m I I 1 \ . I f 500 0 ------------ I/ i 0 - I 1. 1. 2. 2. 2. 2. 2. 2 ' 1 3 .. RHR 3. 3. J. 1 AN3 3. t AND 4. POST BUT i 5. AllD 5. REACTOR ST& @A# 6. INL~ DC IUCIOP ALARM 8. EEDU~ Trn 7. VMT UC PWTT SISrm 8. AND 8. AHD AND 8. 11. 13. AIARX FUEL 18. AIARX 12. RECEIPT AWD raEAMrXI AND W: 17. CARBON 12. AND ALIXILURY AND 15. ICS IMPI(AL KCAD 11.

I -- I 1,: LOU .. . . . - ---- .... . . . .-.. . . .-.--- - r.-.-. '-. . . .-.--...... . ...... ... . . . .. . PWER -- - I 6 b HOT I FUXi ( I 1 PL*WTKeUU- AND RDI Cm LWIW I I 14. WW AND raOee\NRE AND THEW rOrOI I(mUIID(M I AND 3. CLUSER 5. MD - WCLW AND -am : - 5. FU)U I I b. 1 / / 8 wwrn urn PaRl DEIn ' RCC* WOIW AND LC~A ABOVE WO( 1. SLX-8. ! I &. - I. Atm FLW MD - - I SO-8. &XU I SC-3.4 LIIW I I I FLAX - 1 POWFR I J. - 6 PrXU TEST XFASlIRLIWT b. *SD PMR Ahn .

POWER PROFILE - UNIT 2 I I I I I I 1 1 I I e..".. .---.. ..A 3-v nn-rn WINE SJ-ILVII IU LUU. ru--n - - . - -. . - - . . - - 1 TU 1 I I I I I I I I 1. HOT 4. AND m PUY (RTD) I 1 i / I PUKT Ill. SU-6.1 aJRE 12. su -6.2 PRERWUKS~ AWD QlEMF? SU-6.3 SYSTDI CORE 14. SU-8.5 WCLUR mERNVEE Ann IlIeUu ml'ex KCASU~ 2: - YORlli AND AND ANC b. WAD AM) AND IWD AND AND AND AtERWAl 2. ---- ~~-~ ~ .-.. I AND WAD WAD PWTT ( I FIGURE 14.1-4 STARNP TEST SEO~NCE - UNIT 2

1.61.822.20.70.91.11.3FHFractionofThermalPower 1.41.61.820.70.91.11.3FHFractionofThermalPower 1.61.822.20.70.91.11.3FH

00.20.4 0.60.81051015202530FractionofNominal(3455MWth)Time,seconds RCCA WithdFigurawal from SMarSQN-26 ure 15.2.1-3aubcritical - Furk - BW Fueluel/Clad Temmperature

050010001500 2000 25000510202530Temperature, F 15

00.20.4 0.6 0.811.2 1.4010203040506070Neutron Power Fraction Time (s) 00.20.4 0.6 0.811.2 1.4010203040506070Thermal Power Fraction Time (s)

22402260 2280 2300 2320 2340 2360 2380 2400010203040506070Pressurizer Pressure, psia Time (s) 6465 66 67 68 69 70 71 72010203040506070Pressurizer Level, % span Time (s)

580582 584 586 588 590 5920.0010.0020.0030.0040.0050.0060.0070.00Core Average Temperature, F Time (s) 1.002.00 3.00 4.00 5.00 6.00 7.00 8.00 9.0010.000.0010.0020.0030.0040.0050.0060.0070.00DNBR Time (s)

00.20.4 0.6 0.811.2 1.4020406080100120140160180200Neutron Power Fraction Time (s) 00.20.4 0.6 0.811.2 1.4020406080100120140160180200Thermal Power Fraction Time (s)

20002050 2100 2150 2200 2250 2300 2350 2400020406080100120140160180200Pressurizer Pressure, psia Time (s) 01020 30 40 50 60 70 80020406080100120140160180200Pressurizer Level, % span Time (s)

560565 570 575 580 585 590 595050100150200Core Average Temperature, F Time (s) 1.451.50 1.55 1.60 1.65 1.70 1.75 1.80 1.850.0020.0040.0060.0080.00100.00120.00DNBR Time (s)

Figure 15.2.2-7b Effect of Reactivity Insertion Rate on Minimum DNBR for a Rod Withdrawal Accident at 100% Power, Advanced W17 HTP Fuel (Full Core) 1.401.45 1.50 1.55 1.60 1.65 1.700.0010.010.1110100Minimum DNBR Rod Withdrawal Rate in pcm/sec Maximum feedbackMinimum feedbackOTDT Trip High Neutron Power Trip OTDT Trip High Neutron Power Trip

00.20.4 0.6 0.811.2 1.4012345678910HEATFLUX(FRACTOFNORM)TIME(SEC)

00.20.4 0.6 0.811.21.4012345678910HEATFLUX(FRACTOFNORM)TIME(SEC)

"

1.21.41.6 1.822.2 2.4 2.6012345DNBRTIME(SEC)

1.21.41.6 1.822.22.42.6012345DNBRTIME(SEC)

Figure 15.2.5-3c Partial Loss of Forced Reactor Coolant Flow, DNBR vs. Time - Advanced W17 HTP Fuel (Full Core)

"

01 2

3 45024681012DNBRTransienttimeinseconds

" "

"

1.601.70 1.801.90 2.00DNBR

Figure 15.2.11-3c Excessive Load Increase with Manual Reactor Control, at End of Life. DNBR, Nuclear Power and Pressurizer Pressure as a Function of Time - Advanced W17 HTP Fuel (Full Core) 1.601.70 1.80 1.90 2.00DNBR

1.801.90 2.00 2.100100200300400DNBR

1.601.70 1.80 1.90 2.00DNBR

1.801.90 2.00 2.100100200300400DNBR

1.601.70 1.80 1.90 2.00DNBR

" "

1.001.50 2.002.503.00 3.504.00051015202530354045505560DNBRTIME(SEC)

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"

00.20.4 0.60.811.2024681012COREFLOWFRACTION01 2

3 4024681012DNBR

1.01.5 2.0 2.5 3.0 3.50.03.06.09.012.015.018.021.024.0Calculated DNBR Time, s

TIME SEQUENCE OF EVENTS FOR STEAM LINE BREAK

Accident Event Time (Sec)

Major Secondary System Pipe Rupture 1. Case A Steam line rupture 0.0 Steam line isolation setpoint reached 2 Steam line isolation occurs 10 Feedwater isolation occurs 22 Safety Injection Flow initiated 30 Criticality attained 35 Pressurizer empties 57 Boron reaches the core 98

2. Case B Steam line rupture 0.0 Steam line isolation setpoint reached 2 Steam line isolation occurs 10 Feedwater isolation occurs 22 Pressurizer empties 30 Safety Injection Flow initiated 30 Criticality attained 32 Boron reaches the core 97

(continued) TIME SEQUENCE OF EVENTS FOR STEAM LINE BREAK

Accident Event Time (Sec)

Major Secondary System Pipe Rupture

3. Case C Steam line rupture 0.0 Steam line isolation setpoint reached 2 Steam line isolation occurs 10 Feedwater isolation occurs 22 Pressurizer empties 35 Criticality attained 53 Safety Injection Flow initiated 60 Boron reaches the core 125

4. Case D Steam line rupture 0.0 Steam line isolation set point reached 2 Steam line isolation occurs 10 Feedwater isolation occurs 22 Pressurizer empties 43 Criticality attained 48 Safety Injection Flow initiated 60 Boron reaches the core 127

(continued) TIME SEQUENCE OF EVENTS FOR STEAM LINE BREAK

% "

% " " " "%

%

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%

SS15-5.doc 15.5-21 SQN-25 11. During the postulated accident, iodine transferred to the secondary side in the three good steam generators is uniformly mixed with the water in the steam generators.

12. In the affected steam generator, all the water boils off and releases through the break immediately after the accident. The partition factor for the iodine released is assumed to be 1.0. After this initial release, further iodine is released due to the primary to secondary leakage in the affected steam generator. A partition factor of 1.0 is also assumed for this iodine release. 13. The primary pressure remains constant at 2235 psig for 0-2 hours and decreases linearly to atmospheric from 2235 psig during the period of 2-8 hours.

14. The 0-2 and 2-8 hour atmospheric diffusion factors given in Appendix 15A and the 0-8 hour breathing rate of 3.47 x 10-4 m3/sec are applicable.

The resulting doses for the case with the accident-initiated iodine spike are: Thyroid Whole Body Skin Exclusion Area Boundary 5.4 rem 0.073 rem 0.11 rem Low Population Zone 0.69 rem 0.01 rem 0.014 rem Control Room 0.22 rem .046 rem 0.009 rem These doses are well within the limits of 25 rem whole body and 300 rem thyroid defined in 10 CFR 100 and the control room dose limits of 5 rem whole body, 30 rem thyroid, and 30 rem skin defined in Section 6.4 of the Standard Review Plan.

[For information only, the TEDE doses are 0.25 rem at the EAB, 0.032 at the LPZ and 0.053 rem in the control room.] For the case with the pre-existing iodine spike the doses are:

Thyroid Whole Body Skin Exclusion Area Boundary 5.3 rem 0.061 rem 0.09 rem Low Population Zone 0.67 rem 0.008 rem 0.011 rem Control Room 0.20 rem 0.037 rem 0.008 rem These doses are well within the limits of 25 rem whole body and 300 rem thyroid defined in 10 CFR 100 and the control room dose limits of 5 rem whole body, 30 rem thyroid, and 30 rem skin defined in Section 6.4 of the Standard Review Plan. [For information only, the TEDE doses are 0.24 rem at the EAB, 0.03 at the LPZ and 0.043 rem in the control room.]

SS15-5.doc 15.5-22 SQN-25 The parameters that effect the environmental consequences of a steam line break are not adversely affected by the steam generator. Therefore, all acceptance criteria continue to be met on Sequoyah Unit 1 and Unit 2 with the RSGs.

15.5.5 Environmental Consequences of a Postulated Steam Generator Tube Rupture The postulated accidents involving release of steam from the secondary system will not result in a release of radioactivity unless there is leakage from the Reactor Coolant System to the secondary system in the steam generators. A conservative analysis of the postulated steam generator tube rupture assumes the loss of offsite power and hence involves the release of steam from the secondary system. A conservative analysis of the potential offsite doses resulting from this accident is presented assuming steam generator leakage prior to the postulated accident for a time sufficient to establish equilibrium specific activity levels in the secondary system. Parameters used in the analysis are listed in Table 15.5.5-1.

SS15-5.doc 15.5-28 SQN-25 15. Not used.

16. NE Calculation SQN-APS3-067, "Offsite and Control Room Operator Doses Due to a MHA LOCA With Maximum Allowable Annulus Inleakage."

17. SQN-DC-V-21.0, "Environmental Design Criteria."

18. NE Calculation SQN-TI- RPS-158, "Maximum Offsite Doses Due to the Post Accident Sampling Facility." 19. Not used.

20. Not used. 21. NE Calculation SQN-APS2-44, "Annulus Pressure Transient Analysis Following A LOCA." 22. Not used. 23. Document No. 77-5016198, Framatome ANP, Replacement Steam Generator Report for Sequoyah Unit 1."

24. Document No. 77-9142036, AREVA NP, "Replacement Steam Generator Report for TVA Sequoyah Unit 2."

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