ML16127A483
Text
x A B C 0 E F G w-H K L M N 0 P R ,-1 6 1 3 5 5 3 7 8 7 8 7 3 5 4 4 5 3 1 8 6 2 6 8 I 5 4 2 2 4 5 6 7 2 7 2 7 6 5 4 2 2 4 5 1 8 6 2 6 8 I 3 5 4 4 5 3 7 8 7 8 7 3 5 5 3 I 6 I I 2 3 4 5 6 7 8 9 10II 12 13 14 15-y No.of I"'ods FuncTloo Z WORTH-HZP (/.k/k)BOC EOC----SEE TABLE 3.2-7 D Group Number Gr'OUP I 8 2 8 3 8 4 8 5 12 6 8 7 (TRANSIENT) 9 8 8 TOTAL#69 Sofety Sofety Sofety Sofety ControL Contl"'oL ControL APSRs.N......TM! I Update-12 3/94 p.3.FIG-1 Control Rod Locatton and Group O**tgnatton.For TMI-I.Curr.nt Cycl*.CAD FILE:SIA,SKM.OO,0343,OOO-
,0001 Ftg.3.2-1 x A B C D E F G W-_H__K L M N o p R*
15-y RODS IN I.GROUPS 5-7.Or.WD 2.GROUP 8 AT HFP NOMINAL POSITION.*BOC AND EOC EJECTED ROD Z HZP WORTH OF EJECTED ROD (" A k/k)SEE TABLE 3.2-4.N......TMI Untt I Upaote-12 3/94 p.3.FIG-2 E,ected Rod Location BOC and ECC Co4 File SIA.SIOI.OO.DIR.GOO-.OOOI F 1,.3.2-2 TMI-1 UFSAR RODS IN HZP WORTH OF STUCK ROD (%k / k) Groups 1-7, 0% WD See Table 3.2-4
- BOC Maximum Worth St uck Rod** EOC Maximum Worth St uck Rod p. 3.FIG-3 A B C D E F G H K L M **N
- O P R 1 2 3 4 5 6 7 8 9 101112131415 TMI- 1 UPDATED FINAL SAFETY ANALYSIS REPORT Stuck Rod Location BOC and EOC FIGURE 3.2-3 Rev. 2, 04/1 r*2.0 r*1.0 rfa 1200 I 1&I I 10 loa l..rr.-liiO IlOIUSSI SLOPf (II P)SLOPE (EIPEIIIENT)
Ie.)1-)(I 10,2)(110.2)25.7143 I 00 O.1101.25 50.3921 o 83 o 110&I0 19&1 o 71 0.&1.05 2.0.0910 o&1 0.1305__liP----HEllSTllNO IlO.IEI'.HOINEI 02.IM.0&.01.10.12.14.16.11.20---0....--0....-..-*
..r TMI Unit 1 Update*5 7/86 p.3.FIG-4 Fractions of Change in the Reasonance Integral as Function of Vf: v'BJ for UO 2 Rod (T in Degrees K)Fig.3.2-4 0.00-0.02-0.04-0.0&-0.08..-0.10..........0fit..-0.12-0.14-0.1&-0.18-0.20""'"
PPM'"','"'\"'\\10 20 30 , VOid, 40 50 60
..r TMI Unit 1 Update*5 7/86 p.3.FIG-5 Uniform Void Coefficient for 177 Assembly Core Fig.3.2-5 110 100-wU...J 80HH.60*w ffid=->..0:JH 20....C)tjI-0\\\\\\\.....-V-2.4%6k/k(wtth Stuck Rod}\\\......_-----5.40/.6 kl1l.J......----I o 2..e Ttme.sec 6 7 8-p.3.FIG-6 I£a]!J Nuclear TMI Unit 1 Percent Neutron Power Versus Time Following Trip, BOC Update*9 7/90 Fig.3.2-6 TMI-1 UFSAR Core Loading Diagram for TMI-1 Cycle 21 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 A21C3G11 21F2D05 21E2H02 21F2D1121C3G05 B 22BO11 21DP12 23F F 23G F 22EO04 23G F 23F F 21DP04 22BO05 C 21F2N04 23F F 23H F 23E F 22A2O10 23B F 22A2O06 23E F 23H F 23F F 21F2D04 D 22BM13 23F F 22EO12 22DK02 22B F07 23B F 22E L14 23B F 22B F09 22DK14 22EN03 23F F 22BM03 E 21DN14 23H F 22DB09 22BK08 23A F 22CK04 22EB06 22CK12 23A F 22BH07 22DB07 23H F 21DN02 F21C3M07 23F F 23E F 22BG06 23A F 22CC08 23C F 22A2E06 23C F 22CH13 23A F 22BG10 23E F 23F F 21C3M09 G 21F2E04 23G F 22A2 L13 23B F 22CD09 23C F 21C1 F11 19 23D F 21C1M10 19 23C F 22CD07 23B F 22A2 L03 23G F 21F2E12 H 21E2P08 22EN13 23B F 22EB10 22E L02 22A2 L05 23D F 20A2H15 19 23D F 22A2 F11 22E F14 22EP06 23B F 22ED03 21E2B08 K 21F2M04 23G F 22A2 F13 23B F 22CN09 23C F 21C1E06 19 23D F 21C1 L05 19 23C F 22CN07 23B F 22A2 F03 23G F 21F2M12 L21C3E07 23F F 23E F 22BK06 23A F 22CH03 23C F 22A2M10 23C F 22CO08 23A F 22BK10 23E F 23F F 21C3E09 M 21DD14 23H F 22DP09 22BH09 23A F 22CG04 22EP10 22CG12 23A F 22BG08 22DP07 23H F 21DD02 N 22BE13 23F F 22ED13 22DG02 22B L07 23B F 22E F02 23B F 22B L09 22DG14 22EC04 23F F 22BE03 O 21F2N12 23F F 23H F 23E F 22A2C10 23B F 22A2C06 23E F 23H F 23F F 21F2D12 P 22BC11 21DB12 23F F 23G F 22EC12 23G F 23F F 21DB04 22BC05 R21C3K11 21F2N05 21E2H14 21F2N1121C3K05 Key xxxyyy zzzBatch ID Previous Cycle Location Previous Cycle Number (N/A for Feed, Cycle 20if blank)Note: F denotes Fresh Fuelp. 3.FIG-7 ECR TM 14-00484, Rev. 0 A ttachment 8 Page 21 of 30 TMI-1UPDATEDFINALSAETYANALYSISREPORT
.CoreLoadingDiagramTMI-1CurrentCycle
.FIGURE3.2-7Rev23,04/16 TMI UFSAR
Figure 3.2-8 Deleted
-.-.2.7 4:1......,",,---*........2.8 2.5 2.4.-.**:I......, 2.3 0.-..,GI: CD Z CI 2.2 2.1 2.0 1.9 1.8 10 90 110 130 150 Distance fro.bottaM ot active lenlth.in.0:im Nuclear 1MI Unit 1 Update*6 7/87 p.3.FIG-9 Typical DNB Ratios (BAW-2)in the Hot Unit Cell (Cycle 5)Fig.3.2-9
""'lIIIIIiI"-."\."-"-"""'Ill" ,....1," 1..1.-*1.21u*....--I..e:*.-.-*u*::ID.II....,*4.....I D.Io....0.40 0.21 D.*a 10 20 30 40 5.10 70 10 10 1 DO'.rclnta***f fu.1 I'ds with Hilhlr'.akinl factlrs Thin'lint ValUls.I 0:iElNuclear TMI Unit 1 Update*5 7/86 p.3.FIG-IO Distribution of Fuel Rod Peaking (Initial Cycle)Fig.3.2*10 Alial Peaks 2.0 1.7 1..8**G 1.4-;c I.21 I..0 J 0..8:J0..6 0..4 0..2 0.0 20 40 60 80 100 120 140 Active fuei lenath.Inches Bil!l Nuclear 1MI Unit 1 Update*5 7/86 p.3.FIG-l1 Maximum Allowable Axial Power Distributions for a Radial-Local Peaking Factor of 1.78 Fig.3.2-11 A.I.I.ILS"I 1.4
.., 2.'1.1 e.*1.41.2..1.1-I***J I.':I-'.4'.2'.0 0 20 41 80 10 101 120 148'ct***fUll Llnat**IlehlS BimNuciear 1MI Unit 1 Update*5 7/16 p.3.FIG-12 Equivalent Axial Power Distribution for a Radical Local Pelklng Factor 01 1.65 (Initial Cycle)Fig.3.2-12 2.0 ,,-......1.8 0-nJu c CD 1.6 z Q 99%Confidence Basi s 0-'0 aJ...., U'01.4 aJ...a."'-"-aJ I c c::oJ: u1.2---'en aJ0%c::-1.0 0Design Overpower nJ ex CD (114")z Q 0.8 100 110 120 130 140 150 REFERENCE DESIGI IIOWER (2568.1 t).%p.3.FIG-13 BimNuclear 1MI Unit 1 DNB Ratios (W-3)in Hot Unit Cell Versus Reactor Power (Initial Cycle)Update*5 7/86 Fig.3.2-13 20 II 16 14 12 10.....B--...CI 8 4 2 0-2-4/V 2120 pS11L V I V/I/A21.5 pSle.I V/,//v/I""/vV'//QUill ty V I SuDcooled r-oOSII" Ove rpower I 15 17'" c:..90.-*..-..93-'"..c::L:::a CI Q N.c:-U N.c:-*..--99*.c:...-GI u c::a.-103...-t-.-0-...109 c U....C....c:.-U*c: 116o--t**..o_.......e:t 127-.c:...GI C..144-t 100 110 120 130 140 150 160 REFEREIKE DESIGII flOWER (2561.1 t)., Bi!JNuclear TMI Unit 1 Maximum Hot Channel Exit Quality Versus Reactor Power (Initial Cycle)p.3.FIG-14 Update*5 7/86 Fig.3.2-14 an N C N g C atI*ca.ca.Ba:::sa CDCoD C an ca N--A.Z\:J M C CIt.........a'" u z'"'" lAo'"a..DNIR In Hot Channel BimNuclea, TMI Unit 1 Update*5 7/86 p.3.FIG-15 Hot Channel DNB Ratio (W-3)Versus Power for Partial Pump Operation (Initial Cycle)Fig.3.2-15
&It CL EA-S CL a aN I CL 0 E eftI A-N-CD 1ft N.W-i*..5w a w 0 u*.....w*w"-w*o.2 o.'" o.2 o.'" o o Coolant Quality At Point of liniMU.DNIR In Hot Channel Bil!J Nuclear 1MI Unit 1 Update*5 7/86 p.3.FIG-16 Hot Channel Quality at Point 01 Minimum ONBR Versus Power lor Partial Pump Opel'ltion (Initial Cycle)Fig.3.2-16
..*"...---....---.-----...--
....._,....----
.....*"OJ II tts I I altl'ISld On laD.I£A'-4124(I k de*II-/C8)I I*a..._N...0-=>>......M._c:**.-Q..U-..=-'".-I 2**t-----+-........
...u...3.**t---_-......
..-I w-IT a..*50**.00.:l***2.1 ,..1.........................._...1*
..r 1MI Unit 1 Update*5 7/86 Thermal Conductivity of U0 2 p.3.FIG-17 Fig.3.2-17 1.8 1.6 Gaussian Distribution 1.4 1.21.0 70&0 lit SOc:-0 Go-40 0...u.a 30 Iiz 20 10 0 0.6 0.8
..r TMI Unit 1 Update*5 7/86 Number of Data Points vs.<t>E/<t>C p.3.FIG-18 Fig.3.2-18 1.6 1.5 1.4 1.3 Finite SampleConfidence
1.2 Finite
Sample 99%Confidence
1.1 Infinite
Sample 100S Confidence 50..................L-..L....J 1.0 60 90 100 DNS Ratio (SAI-2)0]2]Nuclear TMI Unit 1 Update*5 7/86 p.3.FIG-19 DNB Ratio (BAW-2)vs.Population for Various Confidence Levels Fig.3.2-19 L 025 l.020 I.015 I.010""1.005...Q-U ftI w.I.000 60 70 80 90 100-GJ C C l'Q 0.995.s=Co)-Q z 0.990 0.985 FA (Interior Alllmbly Channel)--0.980 o 975 0.970FA (Wall A_mbly Channln 0.965 0.960 Population Protected.
,.-p.3.FIG-20 BiI!l Nuclear 1MI Unit 1 Hot Channel Factors vs.Percent Population Protected Update*5 7/86 Fig.3.2-20 16 14 12 10 8 6-f?<J 4>--CQ 2c 0*2*4*6*8*i th 5'Yo F low Factory/Dis tri bu t i on'"/r.,//" V/....v<-:.No Flo.u V V DIs t r I bu t I on-Factor/I.V V I I//
/SuDcooled/,//'7'DeSign I 100 11 0 120 130 140 150 p.3.FIG-21 Rated Power (2568 MWt>.\>B:il!l Nuclear TMI Unit 1 Hot Channel and Nominal Channel Exit Qualities Versus Reactor Power (Without Engineering Hot Channel Factors)(Initial Cycle)Update*5 7/86 Fig: 3.2-21
- Bundle Burnout rest Conditions Where Stable Operations Were ObservedMaxiMuM Design Conditions, Power*MaxiMum Design Conditions, 13010 Power tt Most Pr.obable Conditions, Power*Most ProbablePower\3.0 1o-----4-----+-----+------ir---+---+--------I o 5 10 15 20 Qual i t1 (I It lit),25 30 Ia:ENuclear 1MI Unit 1 Update*5 7/86 Flow Regime Map for the Hot Unit Cell p.3.FIG-22 Fig.3.2-22 3.0 2.5 w I 0)(N., 2.0I.I:-4-...., u 0..1.5>--..s 1.0.5+Bundle Burnout Test Conditions
'here Stable Operations lere Observed*MUlmum Deslin Conditions.
114'Power*Maximum Deslin Conditions, 130'Power\*Most Probable Conditions, 114'Power*Most Probable Conditions.
130'Power*...*f'++*+***to*.+...,....*+ft*...**-*-\+.....+r+**t t..'t'+...t.ubbl.To**........+*..Annular..(**k.r)+++...!'t+....+++++.+*....4A+,*...+........+....***..to)
ToSI ug (.'k.r)'9'L/-5 o 5 10 15 20 Quality (Ib vapor/total Itt).Ia:il!l Nuclear 1MI Unit 1 25 Update*5 7/86 p.3.FIG-23 Flow Regime Map for the Hot Control Rod Cell Fig.3.2-23 3.0 2.5..I 9 III N..2.0...I.I:......u 0..1.5>>......s 1.0.5*Bundle Burnout Test Conditions Ihere Stable Operations lere Observed.*MUIIIlum Deslin Conditions.
114\Power*MUlmum Deslin Conditions.
130\Power\4t Most Probable Conditions.
114" Power*Most Probable Conditions.Paler*****..****..***************..\*....**...*****Ie.u....,.To Annu'ar (.ak.r)**..****,.*****,************..**.-...u....,.To j Slug (.ak.r)/-5 o 5 10 15 20 Qua'it, ('" vapor/total Ib),Bi!JNuclear TMI Unit 1 25 30 Update*5 7/86 p.3.FIG-24 Flow Regime Map lor the Hot Wall Cell Fig.3.2-24 3.0 2.5.0*2)(2.0 N...'P*.I:.-*.;.,...u 1.5 0 U........z 1.0.5*
Burnout Test ConditIons Where Stabl!Operations Were Observed*Maximum Des1in Conditions.
114\Power*Maximum Deslin Conditions, 130'\Power\*Most PrObable Conditions, 114\Power*Most Probable Conditions, Power*****..*****-*************-\*..*-.****4**Bubble To Annular (Baker)**..,**********,*..*********--***'ubbl.To\J Slul (.....r)_l/-5 o 5 10 15 20 Qual i ty (I b vapor/total I b).IJ.Bil!INuclear 1MI Unit 1 25 30 Update*5 7/86 p.3.FIG-25 Flow Regime Map for the Hot Corner Cell Fig.3.2-25 150/DNBR (1-3).1.30 I I 90 140 Desia" Flow RatlJ..(131.32 I 1D6 Ib/hr)I130---I----,t.120 IiDlsia" Overpower110(1141 I 2568**t)100/I,I I 2400 2600 2800 3000 3200 3400 3600 Reactor Core Po.er,.It BE Nue.ear 1MI Unit 1 Update*5 7/86 p.3.FIG-26 Reactor Coolant System Flow Versus Power (Initial Cycle)Fig.3.2-26 2.4 2.2 LINE flOI MIXING COEff.1 1101.02 cJ 2 1001.02 u 2.0 3 901.02 II*100S.0&Il 5 1001.01 Il C"")I*1.8 QftI ac CD:z Q 1.6 cu c C ftI.c1.4 0:c 1.30 (1-3)-1.2 1.0 o I 100 110 120 130 140 150 REFERENCE DESIGN IIOWER (2511 Btl.I BE Nuclear 1MI Unit 1 Update*5 7/86 p.3.FIG-27 Hot Channel DNB Ratio (W-3)Versus Power with Reactor System Flow and Energy Mixing as Parameters (Initial Cycle)Fig.3.2-27 5200 4800 4400.;4000'-.-"-a.E lU.-'-3600cuu..3200 2800 2400 O.0095"Clearance uo r Overpower (114',100'Power'--Maximum Design Clearance Nominal Clearance 6 8 10 12 14 16 18 20 22 24 26 28 30 linear Heat Rate, kw.ft BiI!I Nuclear 1MI Unit 1 Update*5 7/86 p.3.FIG-28 Fuel Center Temperature for Beginning-of-Cycle Conditions (Initial Cycle)Fig.3.2-28 5200...3600 cu....c: cu c....)3200 2800.0095" ClearanceDesign Overpower (1141)Power"-Maximum Design Clearance Nominal Clearance 6 810 12 16 18 20 22 28 30 linear Heat Rate, kw/ft 1MI Unit 1 Update*5 7/86 p.3.FIG-29 Fuel Center Temperature lor End-ai-Cycle Conditions (Initial Cycle)Fig.3.2-29 5200--1ft--a-N 4800*..-4400 4000*BOC (100 IIO/ITU)...3&00*.;..-....3200*Q,*..to-*...28002400 2000 1608 1200 IZ!imNuclear TMI Unit 1 Update*6 7/87 p.3.FIG-30 Typical Post-Initial Cycle-Center Line Fuel Temperature vs.Linear Heat Rate Hot Pin (Cycle 5)Fig.3.2-30 5000""" 4800 u...lW...u i: 4600...u...c:: u U u""" 4400 4200 o 8 16 11.63 kwlt tHat Spot (1001 Power)24 32 40 5000 u...lW...: 4000 I.......Uc:: u u 3000 Burnup (l.llrU I 10-3)EOl 110 40.900 iTii 2000 o*12 16 20 24 p.3.FIG-31 llnllr Hilt Rltl, kw/tt TMI Unit 1 Burnup Effect on Fuel Center Temperature (Initial Cycle)Update*5 7/86 Fig.3.2-31 3600 3200 2800 lL 2400&):)-"'&)a.E 2000&)-&):::I.....1600 1200 800\\\"" ,....'"o 20 40 60 80 100 ,-YOIU.'Fraction of Totll Fuel., (at or IDove Fuel T.perature)
BE Nuclear 1MI Unit 1 Update*5 7/86 p.3.FIG-32 Fuel Temperature Versus Total Fuel Volume Fraction for Equilibrium Cycle at End-ol-Cycle Fig.3.2-32 BimNuclea, TMI Unit 1*0.119 NU.'lr erells Iurn.d Update*5 7/86 p.3.FIG-33 Typical Reactor Fuel Assembly Power Distribution at End-ol-Cycle Equilibrium Cycle Conditions lor 1/8 Core Fig.3.2-33 10k wi t t Fue I 1200
......2400 2800 3200....2000 0 lU...6 k wIt t....to...lU/---E 1600 lU to-0.24 0.20 0.16 0.12 0.08 0.04____S80 0 F_TIVIl COOl:::!800 400"-...A..-..........
.......___""___6__""'___
.....0.0 fuel Radius, in BiI!lNuclear TMI Unit 1 Update*5 7/86 Fuel Rod Temperature Profiles at 6 and 10 KW/Ft p.3.FIG-34 Fig.3.2-34
§...I...I...II u-:I.....-§..a.I!N..§......-N U:.C§....--...N..!I 0:-I-§I§..l-**--p-'l-**I..0!iII-i*'".-I--0-......I-.,.N-rb.....'-l-I-4..-'i+U...................
- ut...4U 0-+4-I-l1 4-.--....*1II-.<.J..*\*\.(1)IIif'.r"\..-*""'"""'t'\--...j'"---.......
- .-4 hp 8 888 8 o o-o V'o o fi**ten c**1111.'ld, I TMI Unit 1 Update*5 7/86 p.3.FIG-35 Percent Fission Gas Released as a Function of the Average Temperature of the U02 Fuel Fig.3.2-35 d-of-Life lUll;-3.DI,I-IIVIU 1.10 (partial Rod Insertion)
I IFUll lidplln.Cor.I.ttll I PIP-1.50 (Iodifi.d COlin.)1.4
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--.......+---............t-t.......*20 40&0 10 110 120 14a Dlltlnc.trGl lott..of Acti,.FUll.in.Iml!]Nuclear 1MI Unit 1 Update*5 7/86 Axial Local to Average Burnup and Instantaneous Power Comparisons p.3.FIG-36 Fig.3.2-36 Maximum DeSii" Diametral Gap"0 800u 0-600 lU::;).....400 200 0 0 8 12 16 20 24 linear Heat Rate,.kw/f t 2400 2200 20000 1800 I-I...J600:::::l-CD Noml na I lU 1400 D I a.e tra I Gap u="'T" 1200 u::;)l:'c: 0 u 1000 c.ftI e.,:, 0:im Nuclear 1MI Unit 1 Update*6 7/87 Fuel to Clad Gap Conductance
'or End-O'-Cycle Conditions (Initial Cycle)p.3.FIG-37 Fig.3.2-37 2D I I II I lalimull DesiRR Clearance 1.1 Alial Power and EDl Burnup Snape with Closed Pores 1.7 Ilill Power Ind EDl Bu rnup Shape*i th Op.en'ores 1.5 Ilial POler Ind EDL Burnup Snape with C los ed P Q res 5 1510-1:1 U..u-....o o 2 4 6 I Initial Cold Diametral Clearance.
in I 10 3 10 BimNuclear TMI Unit 1 Update*6 7/87 p.3.FIG-38 Fission Gas Release for 1.5 and 1.7 MaxlAvg.Axial Power Shapes (Initial Cycle)Fig.3.2-38 3500 3000 Desi an l illi t----114" Overpower-----Power 2500"'&It a.c"'Closed u m 2000=.-1.7 AIIaI Po.e r an u c EDt Bu rnup Snape-&It c 1.5 Axial Power ana u EDl Burnup Snape:::;)&It Pores&It 1500 u1.7 Ali al Power and Q,.&It EDl Burnup Snape"'CoD 1000 MIII.U.Desie" 500 o 2 4 6 8 10 Initial Cold Diaetral Clearance.
in I 10 3 BiE]Nuclear TMI Unit 1 Update*6 7/87 p.3.FIG-39 Maximum Gas Release to Pressure Inside the Fuel Clad for Various Axial Burnup and Power Shapes (Initial Cycle)Fig.3.2-39 t_.-.t I I,I Ik In.Fie tor CD HO'UN I'C fll (ntllllp, IIISI Flctor CD"OJ UU CELL CD HO'CO'NlI CELLHOT COhUOl 100 ULL Bil!I Nuclear TMI Unit 1 Update*6 7/87 p.3.FIG-40 Nominal Fuel Rod Power Peaks and Cell Exit Enthalphy Rise Ratios (Initial Cycle)Fig.3.2-40 t I hclllr"1 flCtor Q)NOT UNIT CEll (nUIII"**11'acurMOT tAll CEll OJ MOT CUNEa CEllMOT CONTaOl 100 ell l BE Nuelear TMI Unit 1 Update-6 7/87 p.3.FJG-41 Maximum Fuel Rod Power Peaks and Cell Exit Enthalphy Rise Ratios (Initial Cycle)Fig;3.2-41 1.1 1.5 1.41.3 1.2 1.1 CD 1.0 I CI-..N 0.9...-I...&:::a 0.1...*.0.7 IC::a-...0.&*..z-..0.5 u 00.4 0.3 0.2 011\G*2.21 I 10 1 IIi/tH-f t 2\.'\I\: 1-3 DNa Milt Flu.!\.(DISII" Lillit)'\'\r\.lin IIIUII ON**1.55'\\\\---/V\I/1\V V Calculatld Surflc.\Milt Flu.J'\I o 540 5&0 510600620 140 110&10 100 120 LOCII Enthalpy, Itu/III BimNuclear 1MI Unit 1 Update*5 7/86 p.3.FIG-42 Calculated and Design Limit Local Heat Flux Versus Enthalpy in the Hot Unit Cell at the Most Probable Condition (Initial Cycle)Fig.3.2-42 1.1 1.5 1.4 1.3 1.2 1.1 UI 1..0 I 0-)(0.9 N...-I"-0.8 s:: "-...CD O.7...::I-..-0.6....,%-lQ U Q0.4 0.3 0.2 O.1\,.2.5'I 10'Ib/tH-f t 2I I I I1 r I 11-3 ONI H**t Flu.(D.Slln l'.'t)"\1\\
ON'**1.12\\I..........
V" VI\V"1 Calculated SurfaceJ J....t Flul\, o 560 510 600 620&40 6&0 680 700 720 local Enthalpy, Itu/lb 0ENuclear TMI Unit 1 Update*5 7/86 p.3.FIG-43 Calculated and Design Limit Local Heat Flux Versus Enthalpy in the Hot Unit Cell at the Maximum Design Condition (Initial Cycle)Fig.3.2-43 CONTF,OL ROO ASSEMBLY PL ENUM ASSEMBLY OUTLET NOZ ZLE ILLANCE HOLDE R TUBE LOWEP, GRID FLO W OJS TR IfurOR CONTROL P,OD DRIVES STUDS INTERNALS VENT VALVE CONTROL GUIDE TUBE CORE SUPPORT SHIELD INLET NOZZLE FuEL ASSEMBLY REACTOR VESSEL THERMAL SHIELD GUIDE LUGS INCORE INSTRUMENT GUIDE TUBES________I N CO REI N S T RUM E NTN0ZZLE S Bil!1NucIe.r TMI Unit 1 Update*5 7186 p.3.FIG-44 Reactor VI_lind Intlmlll*Gen.Arnngement Fig.3.2-44 CONTROL ROD ASSEMBLY LDCATION INCORE INSTRUMENT LOCATION REACTOR VESSEL THERMAL SHIELD CORE BARREL SURVEILLANCE SPECIMEN HOLDER TUIE BimNuclea, TMI Unit 1 Update*5 7/86 p.3.FIG-45 Reactor Vessel and Internals-Cross Section Fig.3.2-45 0imNuclear 1MI Unit 1 Core Flooding Arrangement p.3.FIG-46 COlE fL ODD I N&NOZZLE Update*5 7/86 Fig.3.2-46 1-,--+----+++---tr----
t rn I eMf I....'MilL.SICT**I-I Bi!lNucl.r 1MI Unit 1 UI'C..I lie.u".'URl.IU Update*5 7/86 p.3.FIG-47 Internals Vent Valves Fig.3.2-47 BimNuclear 1MI Unit 1 SEE SECTION BELOW Update*5 7/86 p.3.FIG-48 Internals Vent Valve Clearance Gaps Fig.3.2-48 TOP VIEW Update*5 7186 TMI Unit 1 FUll AlUmbly COHTROl.ROO GUIDE TUH , ,'NSTRUMENTATION I TUM I r::: i.'i J.: ,',',r.I'i.i......'\,FUEL ROO ASSEMILY.!iI r:: 1:7'Ir1I n'n I r I , I I , T 1 1 CROSS SECTION I 1 , i I 1\I[i]1*[J, l!I r LOWER END FITTINGIi'-I-'
- r-'---I
'irI INSTRUM*NTATION TUH.-l CONNECTION SNeER GRID U"O ENO trlTTING))Fig.3.2-49 p.3.FIG-49 SPIDER TOP VIIW COUPLINt NEUTRON MATERIAL CONTROl.ROD p.3.FIG-50 ra::imNuclea, TMI Unit 1 Control Rod Assembly Update*5 7/86 Fig.3.2-50 SPIDER TOP VI!W NEUTRON ABSORItNG MATERIAL.CONTROL ROD
..r 1MI Unit 1 Axial Power Shaping Rod Assembly Update*5 7/86 p.3.FIG-51 Fig.3.2-51 SPIOfR----IURNAILI POISON ROD TOP VIIW.URNA.LI POIION MATIRIAL p.3.FIG-52 0iI!JNuclear 1MI Unit 1 Burnable Poison Rod Assembly Update*5 7/86 Fig.3.2-52 TOP VIEW SPIDER-----....
ORIFICE ROO p.3.FIG-'53 Bil!l Nuclear TMI Unit 1 Orifice Rod Assembly Update*5 7/86 Fig.3.2-53 tOJSING ASSY LOAD ARM ASSDfSl.Y p.3.FIG-54 0:W!l Nuclear TMI Unit 1 Side View of BPRA Retainer Update*5 7/86 Fig.3.2-54 UPPER CORE PLAT!ASSY PAD TYP p.3.FIG-55
..r TMI Unit 1 Top View of BPRA Retainer During Operation Update*5 7/86 Fig.3.2-55 POsmoN INDICATQIIt AlIDIILY STATOR.IIIMLV p.3.FIG-56 I I CCUIL.NI AlIDIILV i i BimNuclear TMI Unit 1 Control Rod Drive-Gen.Arrangement Update*5 7/86 Fig.3.2-56
\)SECII**Z*Z SECII**I-I SIC"'I ,.,/E//.c:...-WlI'VALl(ASSEI.U/IIT.I"Il lUI SUI'II'USEI.U---------------.-
_....---------L.l,l, Slell..1*1 SECII**1*1 lUI SCll'**.*l**S*.I.-r L lUI SCIII II"'"'U.sml......" r
/
l Cllllll....""1lI"" CII"
L S SEC"" 1111 SEC 110.,., SEC"" S-S Bil!INucI**r TMI Unit 1 UpdItI.5 7111 p.3.FIG-57 Fig.3.2*57 x A B C 0 E F G w-H K L M N 0 P R ,-1 6 1 3 5 5 3 7 8 7 8 7 3 5 4 4 5 3 1 8 6 2 6 8 I 5 4 2 2 4 5 6 7 2 7 2 7 6 5 4 2 2 4 5 1 8 6 2 6 8 I 3 5 4 4 5 3 7 8 7 8 7 3 5 5 3 I 6 I I 2 3 4 5 6 7 8 9 10II 12 13 14 15-y No.of I"'ods FuncTloo Z WORTH-HZP (/.k/k)BOC EOC----SEE TABLE 3.2-7 D Group Number Gr'OUP I 8 2 8 3 8 4 8 5 12 6 8 7 (TRANSIENT) 9 8 8 TOTAL#69 Sofety Sofety Sofety Sofety ControL Contl"'oL ControL APSRs.N......TM! I Update-12 3/94 p.3.FIG-1 Control Rod Locatton and Group O**tgnatton.For TMI-I.Curr.nt Cycl*.CAD FILE:SIA,SKM.OO,0343,OOO-
,0001 Ftg.3.2-1 x A B C D E F G W-_H__K L M N o p R*
15-y RODS IN I.GROUPS 5-7.Or.WD 2.GROUP 8 AT HFP NOMINAL POSITION.*BOC AND EOC EJECTED ROD Z HZP WORTH OF EJECTED ROD (" A k/k)SEE TABLE 3.2-4.N......TMI Untt I Upaote-12 3/94 p.3.FIG-2 E,ected Rod Location BOC and ECC Co4 File SIA.SIOI.OO.DIR.GOO-.OOOI F 1,.3.2-2 TMI-1 UFSAR RODS IN HZP WORTH OF STUCK ROD (%k / k) Groups 1-7, 0% WD See Table 3.2-4
- BOC Maximum Worth St uck Rod** EOC Maximum Worth St uck Rod p. 3.FIG-3 A B C D E F G H K L M **N
- O P R 1 2 3 4 5 6 7 8 9 101112131415 TMI- 1 UPDATED FINAL SAFETY ANALYSIS REPORT Stuck Rod Location BOC and EOC FIGURE 3.2-3 Rev. 2, 04/1 r*2.0 r*1.0 rfa 1200 I 1&I I 10 loa l..rr.-liiO IlOIUSSI SLOPf (II P)SLOPE (EIPEIIIENT)
Ie.)1-)(I 10,2)(110.2)25.7143 I 00 O.1101.25 50.3921 o 83 o 110&I0 19&1 o 71 0.&1.05 2.0.0910 o&1 0.1305__liP----HEllSTllNO IlO.IEI'.HOINEI 02.IM.0&.01.10.12.14.16.11.20---0....--0....-..-*
..r TMI Unit 1 Update*5 7/86 p.3.FIG-4 Fractions of Change in the Reasonance Integral as Function of Vf: v'BJ for UO 2 Rod (T in Degrees K)Fig.3.2-4 0.00-0.02-0.04-0.0&-0.08..-0.10..........0fit..-0.12-0.14-0.1&-0.18-0.20""'"
PPM'"','"'\"'\\10 20 30 , VOid, 40 50 60
..r TMI Unit 1 Update*5 7/86 p.3.FIG-5 Uniform Void Coefficient for 177 Assembly Core Fig.3.2-5 110 100-wU...J 80HH.60*w ffid=->..0:JH 20....C)tjI-0\\\\\\\.....-V-2.4%6k/k(wtth Stuck Rod}\\\......_-----5.40/.6 kl1l.J......----I o 2..e Ttme.sec 6 7 8-p.3.FIG-6 I£a]!J Nuclear TMI Unit 1 Percent Neutron Power Versus Time Following Trip, BOC Update*9 7/90 Fig.3.2-6 TMI-1 UFSAR Core Loading Diagram for TMI-1 Cycle 21 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 A21C3G11 21F2D05 21E2H02 21F2D1121C3G05 B 22BO11 21DP12 23F F 23G F 22EO04 23G F 23F F 21DP04 22BO05 C 21F2N04 23F F 23H F 23E F 22A2O10 23B F 22A2O06 23E F 23H F 23F F 21F2D04 D 22BM13 23F F 22EO12 22DK02 22B F07 23B F 22E L14 23B F 22B F09 22DK14 22EN03 23F F 22BM03 E 21DN14 23H F 22DB09 22BK08 23A F 22CK04 22EB06 22CK12 23A F 22BH07 22DB07 23H F 21DN02 F21C3M07 23F F 23E F 22BG06 23A F 22CC08 23C F 22A2E06 23C F 22CH13 23A F 22BG10 23E F 23F F 21C3M09 G 21F2E04 23G F 22A2 L13 23B F 22CD09 23C F 21C1 F11 19 23D F 21C1M10 19 23C F 22CD07 23B F 22A2 L03 23G F 21F2E12 H 21E2P08 22EN13 23B F 22EB10 22E L02 22A2 L05 23D F 20A2H15 19 23D F 22A2 F11 22E F14 22EP06 23B F 22ED03 21E2B08 K 21F2M04 23G F 22A2 F13 23B F 22CN09 23C F 21C1E06 19 23D F 21C1 L05 19 23C F 22CN07 23B F 22A2 F03 23G F 21F2M12 L21C3E07 23F F 23E F 22BK06 23A F 22CH03 23C F 22A2M10 23C F 22CO08 23A F 22BK10 23E F 23F F 21C3E09 M 21DD14 23H F 22DP09 22BH09 23A F 22CG04 22EP10 22CG12 23A F 22BG08 22DP07 23H F 21DD02 N 22BE13 23F F 22ED13 22DG02 22B L07 23B F 22E F02 23B F 22B L09 22DG14 22EC04 23F F 22BE03 O 21F2N12 23F F 23H F 23E F 22A2C10 23B F 22A2C06 23E F 23H F 23F F 21F2D12 P 22BC11 21DB12 23F F 23G F 22EC12 23G F 23F F 21DB04 22BC05 R21C3K11 21F2N05 21E2H14 21F2N1121C3K05 Key xxxyyy zzzBatch ID Previous Cycle Location Previous Cycle Number (N/A for Feed, Cycle 20if blank)Note: F denotes Fresh Fuelp. 3.FIG-7 ECR TM 14-00484, Rev. 0 A ttachment 8 Page 21 of 30 TMI-1UPDATEDFINALSAETYANALYSISREPORT
.CoreLoadingDiagramTMI-1CurrentCycle
.FIGURE3.2-7Rev23,04/16 TMI UFSAR
Figure 3.2-8 Deleted
-.-.2.7 4:1......,",,---*........2.8 2.5 2.4.-.**:I......, 2.3 0.-..,GI: CD Z CI 2.2 2.1 2.0 1.9 1.8 10 90 110 130 150 Distance fro.bottaM ot active lenlth.in.0:im Nuclear 1MI Unit 1 Update*6 7/87 p.3.FIG-9 Typical DNB Ratios (BAW-2)in the Hot Unit Cell (Cycle 5)Fig.3.2-9
""'lIIIIIiI"-."\."-"-"""'Ill" ,....1," 1..1.-*1.21u*....--I..e:*.-.-*u*::ID.II....,*4.....I D.Io....0.40 0.21 D.*a 10 20 30 40 5.10 70 10 10 1 DO'.rclnta***f fu.1 I'ds with Hilhlr'.akinl factlrs Thin'lint ValUls.I 0:iElNuclear TMI Unit 1 Update*5 7/86 p.3.FIG-IO Distribution of Fuel Rod Peaking (Initial Cycle)Fig.3.2*10 Alial Peaks 2.0 1.7 1..8**G 1.4-;c I.21 I..0 J 0..8:J0..6 0..4 0..2 0.0 20 40 60 80 100 120 140 Active fuei lenath.Inches Bil!l Nuclear 1MI Unit 1 Update*5 7/86 p.3.FIG-l1 Maximum Allowable Axial Power Distributions for a Radial-Local Peaking Factor of 1.78 Fig.3.2-11 A.I.I.ILS"I 1.4
.., 2.'1.1 e.*1.41.2..1.1-I***J I.':I-'.4'.2'.0 0 20 41 80 10 101 120 148'ct***fUll Llnat**IlehlS BimNuciear 1MI Unit 1 Update*5 7/16 p.3.FIG-12 Equivalent Axial Power Distribution for a Radical Local Pelklng Factor 01 1.65 (Initial Cycle)Fig.3.2-12 2.0 ,,-......1.8 0-nJu c CD 1.6 z Q 99%Confidence Basi s 0-'0 aJ...., U'01.4 aJ...a."'-"-aJ I c c::oJ: u1.2---'en aJ0%c::-1.0 0Design Overpower nJ ex CD (114")z Q 0.8 100 110 120 130 140 150 REFERENCE DESIGI IIOWER (2568.1 t).%p.3.FIG-13 BimNuclear 1MI Unit 1 DNB Ratios (W-3)in Hot Unit Cell Versus Reactor Power (Initial Cycle)Update*5 7/86 Fig.3.2-13 20 II 16 14 12 10.....B--...CI 8 4 2 0-2-4/V 2120 pS11L V I V/I/A21.5 pSle.I V/,//v/I""/vV'//QUill ty V I SuDcooled r-oOSII" Ove rpower I 15 17'" c:..90.-*..-..93-'"..c::L:::a CI Q N.c:-U N.c:-*..--99*.c:...-GI u c::a.-103...-t-.-0-...109 c U....C....c:.-U*c: 116o--t**..o_.......e:t 127-.c:...GI C..144-t 100 110 120 130 140 150 160 REFEREIKE DESIGII flOWER (2561.1 t)., Bi!JNuclear TMI Unit 1 Maximum Hot Channel Exit Quality Versus Reactor Power (Initial Cycle)p.3.FIG-14 Update*5 7/86 Fig.3.2-14 an N C N g C atI*ca.ca.Ba:::sa CDCoD C an ca N--A.Z\:J M C CIt.........a'" u z'"'" lAo'"a..DNIR In Hot Channel BimNuclea, TMI Unit 1 Update*5 7/86 p.3.FIG-15 Hot Channel DNB Ratio (W-3)Versus Power for Partial Pump Operation (Initial Cycle)Fig.3.2-15
&It CL EA-S CL a aN I CL 0 E eftI A-N-CD 1ft N.W-i*..5w a w 0 u*.....w*w"-w*o.2 o.'" o.2 o.'" o o Coolant Quality At Point of liniMU.DNIR In Hot Channel Bil!J Nuclear 1MI Unit 1 Update*5 7/86 p.3.FIG-16 Hot Channel Quality at Point 01 Minimum ONBR Versus Power lor Partial Pump Opel'ltion (Initial Cycle)Fig.3.2-16
..*"...---....---.-----...--
....._,....----
.....*"OJ II tts I I altl'ISld On laD.I£A'-4124(I k de*II-/C8)I I*a..._N...0-=>>......M._c:**.-Q..U-..=-'".-I 2**t-----+-........
...u...3.**t---_-......
..-I w-IT a..*50**.00.:l***2.1 ,..1.........................._...1*
..r 1MI Unit 1 Update*5 7/86 Thermal Conductivity of U0 2 p.3.FIG-17 Fig.3.2-17 1.8 1.6 Gaussian Distribution 1.4 1.21.0 70&0 lit SOc:-0 Go-40 0...u.a 30 Iiz 20 10 0 0.6 0.8
..r TMI Unit 1 Update*5 7/86 Number of Data Points vs.<t>E/<t>C p.3.FIG-18 Fig.3.2-18 1.6 1.5 1.4 1.3 Finite SampleConfidence
1.2 Finite
Sample 99%Confidence
1.1 Infinite
Sample 100S Confidence 50..................L-..L....J 1.0 60 90 100 DNS Ratio (SAI-2)0]2]Nuclear TMI Unit 1 Update*5 7/86 p.3.FIG-19 DNB Ratio (BAW-2)vs.Population for Various Confidence Levels Fig.3.2-19 L 025 l.020 I.015 I.010""1.005...Q-U ftI w.I.000 60 70 80 90 100-GJ C C l'Q 0.995.s=Co)-Q z 0.990 0.985 FA (Interior Alllmbly Channel)--0.980 o 975 0.970FA (Wall A_mbly Channln 0.965 0.960 Population Protected.
,.-p.3.FIG-20 BiI!l Nuclear 1MI Unit 1 Hot Channel Factors vs.Percent Population Protected Update*5 7/86 Fig.3.2-20 16 14 12 10 8 6-f?<J 4>--CQ 2c 0*2*4*6*8*i th 5'Yo F low Factory/Dis tri bu t i on'"/r.,//" V/....v<-:.No Flo.u V V DIs t r I bu t I on-Factor/I.V V I I//
/SuDcooled/,//'7'DeSign I 100 11 0 120 130 140 150 p.3.FIG-21 Rated Power (2568 MWt>.\>B:il!l Nuclear TMI Unit 1 Hot Channel and Nominal Channel Exit Qualities Versus Reactor Power (Without Engineering Hot Channel Factors)(Initial Cycle)Update*5 7/86 Fig: 3.2-21
- Bundle Burnout rest Conditions Where Stable Operations Were ObservedMaxiMuM Design Conditions, Power*MaxiMum Design Conditions, 13010 Power tt Most Pr.obable Conditions, Power*Most ProbablePower\3.0 1o-----4-----+-----+------ir---+---+--------I o 5 10 15 20 Qual i t1 (I It lit),25 30 Ia:ENuclear 1MI Unit 1 Update*5 7/86 Flow Regime Map for the Hot Unit Cell p.3.FIG-22 Fig.3.2-22 3.0 2.5 w I 0)(N., 2.0I.I:-4-...., u 0..1.5>--..s 1.0.5+Bundle Burnout Test Conditions
'here Stable Operations lere Observed*MUlmum Deslin Conditions.
114'Power*Maximum Deslin Conditions, 130'Power\*Most Probable Conditions, 114'Power*Most Probable Conditions.
130'Power*...*f'++*+***to*.+...,....*+ft*...**-*-\+.....+r+**t t..'t'+...t.ubbl.To**........+*..Annular..(**k.r)+++...!'t+....+++++.+*....4A+,*...+........+....***..to)
ToSI ug (.'k.r)'9'L/-5 o 5 10 15 20 Quality (Ib vapor/total Itt).Ia:il!l Nuclear 1MI Unit 1 25 Update*5 7/86 p.3.FIG-23 Flow Regime Map for the Hot Control Rod Cell Fig.3.2-23 3.0 2.5..I 9 III N..2.0...I.I:......u 0..1.5>>......s 1.0.5*Bundle Burnout Test Conditions Ihere Stable Operations lere Observed.*MUIIIlum Deslin Conditions.
114\Power*MUlmum Deslin Conditions.
130\Power\4t Most Probable Conditions.
114" Power*Most Probable Conditions.Paler*****..****..***************..\*....**...*****Ie.u....,.To Annu'ar (.ak.r)**..****,.*****,************..**.-...u....,.To j Slug (.ak.r)/-5 o 5 10 15 20 Qua'it, ('" vapor/total Ib),Bi!JNuclear TMI Unit 1 25 30 Update*5 7/86 p.3.FIG-24 Flow Regime Map lor the Hot Wall Cell Fig.3.2-24 3.0 2.5.0*2)(2.0 N...'P*.I:.-*.;.,...u 1.5 0 U........z 1.0.5*
Burnout Test ConditIons Where Stabl!Operations Were Observed*Maximum Des1in Conditions.
114\Power*Maximum Deslin Conditions, 130'\Power\*Most PrObable Conditions, 114\Power*Most Probable Conditions, Power*****..*****-*************-\*..*-.****4**Bubble To Annular (Baker)**..,**********,*..*********--***'ubbl.To\J Slul (.....r)_l/-5 o 5 10 15 20 Qual i ty (I b vapor/total I b).IJ.Bil!INuclear 1MI Unit 1 25 30 Update*5 7/86 p.3.FIG-25 Flow Regime Map for the Hot Corner Cell Fig.3.2-25 150/DNBR (1-3).1.30 I I 90 140 Desia" Flow RatlJ..(131.32 I 1D6 Ib/hr)I130---I----,t.120 IiDlsia" Overpower110(1141 I 2568**t)100/I,I I 2400 2600 2800 3000 3200 3400 3600 Reactor Core Po.er,.It BE Nue.ear 1MI Unit 1 Update*5 7/86 p.3.FIG-26 Reactor Coolant System Flow Versus Power (Initial Cycle)Fig.3.2-26 2.4 2.2 LINE flOI MIXING COEff.1 1101.02 cJ 2 1001.02 u 2.0 3 901.02 II*100S.0&Il 5 1001.01 Il C"")I*1.8 QftI ac CD:z Q 1.6 cu c C ftI.c1.4 0:c 1.30 (1-3)-1.2 1.0 o I 100 110 120 130 140 150 REFERENCE DESIGN IIOWER (2511 Btl.I BE Nuclear 1MI Unit 1 Update*5 7/86 p.3.FIG-27 Hot Channel DNB Ratio (W-3)Versus Power with Reactor System Flow and Energy Mixing as Parameters (Initial Cycle)Fig.3.2-27 5200 4800 4400.;4000'-.-"-a.E lU.-'-3600cuu..3200 2800 2400 O.0095"Clearance uo r Overpower (114',100'Power'--Maximum Design Clearance Nominal Clearance 6 8 10 12 14 16 18 20 22 24 26 28 30 linear Heat Rate, kw.ft BiI!I Nuclear 1MI Unit 1 Update*5 7/86 p.3.FIG-28 Fuel Center Temperature for Beginning-of-Cycle Conditions (Initial Cycle)Fig.3.2-28 5200...3600 cu....c: cu c....)3200 2800.0095" ClearanceDesign Overpower (1141)Power"-Maximum Design Clearance Nominal Clearance 6 810 12 16 18 20 22 28 30 linear Heat Rate, kw/ft 1MI Unit 1 Update*5 7/86 p.3.FIG-29 Fuel Center Temperature lor End-ai-Cycle Conditions (Initial Cycle)Fig.3.2-29 5200--1ft--a-N 4800*..-4400 4000*BOC (100 IIO/ITU)...3&00*.;..-....3200*Q,*..to-*...28002400 2000 1608 1200 IZ!imNuclear TMI Unit 1 Update*6 7/87 p.3.FIG-30 Typical Post-Initial Cycle-Center Line Fuel Temperature vs.Linear Heat Rate Hot Pin (Cycle 5)Fig.3.2-30 5000""" 4800 u...lW...u i: 4600...u...c:: u U u""" 4400 4200 o 8 16 11.63 kwlt tHat Spot (1001 Power)24 32 40 5000 u...lW...: 4000 I.......Uc:: u u 3000 Burnup (l.llrU I 10-3)EOl 110 40.900 iTii 2000 o*12 16 20 24 p.3.FIG-31 llnllr Hilt Rltl, kw/tt TMI Unit 1 Burnup Effect on Fuel Center Temperature (Initial Cycle)Update*5 7/86 Fig.3.2-31 3600 3200 2800 lL 2400&):)-"'&)a.E 2000&)-&):::I.....1600 1200 800\\\"" ,....'"o 20 40 60 80 100 ,-YOIU.'Fraction of Totll Fuel., (at or IDove Fuel T.perature)
BE Nuclear 1MI Unit 1 Update*5 7/86 p.3.FIG-32 Fuel Temperature Versus Total Fuel Volume Fraction for Equilibrium Cycle at End-ol-Cycle Fig.3.2-32 BimNuclea, TMI Unit 1*0.119 NU.'lr erells Iurn.d Update*5 7/86 p.3.FIG-33 Typical Reactor Fuel Assembly Power Distribution at End-ol-Cycle Equilibrium Cycle Conditions lor 1/8 Core Fig.3.2-33 10k wi t t Fue I 1200
......2400 2800 3200....2000 0 lU...6 k wIt t....to...lU/---E 1600 lU to-0.24 0.20 0.16 0.12 0.08 0.04____S80 0 F_TIVIl COOl:::!800 400"-...A..-..........
.......___""___6__""'___
.....0.0 fuel Radius, in BiI!lNuclear TMI Unit 1 Update*5 7/86 Fuel Rod Temperature Profiles at 6 and 10 KW/Ft p.3.FIG-34 Fig.3.2-34
§...I...I...II u-:I.....-§..a.I!N..§......-N U:.C§....--...N..!I 0:-I-§I§..l-**--p-'l-**I..0!iII-i*'".-I--0-......I-.,.N-rb.....'-l-I-4..-'i+U...................
- ut...4U 0-+4-I-l1 4-.--....*1II-.<.J..*\*\.(1)IIif'.r"\..-*""'"""'t'\--...j'"---.......
- .-4 hp 8 888 8 o o-o V'o o fi**ten c**1111.'ld, I TMI Unit 1 Update*5 7/86 p.3.FIG-35 Percent Fission Gas Released as a Function of the Average Temperature of the U02 Fuel Fig.3.2-35 d-of-Life lUll;-3.DI,I-IIVIU 1.10 (partial Rod Insertion)
I IFUll lidplln.Cor.I.ttll I PIP-1.50 (Iodifi.d COlin.)1.4
..............1.1--....-.....-..........
,.-......-.....--.,.......-.....-.-...-..................
I....-._--t--I--/-
...
PIp*1.1\1/\1-I o.0.........
....._..A..-_........._.....
...._.........0.'
......
.....-.....................
I I a.4....--.....
.....
....__t O.2....--......
.....--t---.....
...--tI'L IiiD.'
.....
...........
--.......+---............t-t.......*20 40&0 10 110 120 14a Dlltlnc.trGl lott..of Acti,.FUll.in.Iml!]Nuclear 1MI Unit 1 Update*5 7/86 Axial Local to Average Burnup and Instantaneous Power Comparisons p.3.FIG-36 Fig.3.2-36 Maximum DeSii" Diametral Gap"0 800u 0-600 lU::;).....400 200 0 0 8 12 16 20 24 linear Heat Rate,.kw/f t 2400 2200 20000 1800 I-I...J600:::::l-CD Noml na I lU 1400 D I a.e tra I Gap u="'T" 1200 u::;)l:'c: 0 u 1000 c.ftI e.,:, 0:im Nuclear 1MI Unit 1 Update*6 7/87 Fuel to Clad Gap Conductance
'or End-O'-Cycle Conditions (Initial Cycle)p.3.FIG-37 Fig.3.2-37 2D I I II I lalimull DesiRR Clearance 1.1 Alial Power and EDl Burnup Snape with Closed Pores 1.7 Ilill Power Ind EDl Bu rnup Shape*i th Op.en'ores 1.5 Ilial POler Ind EDL Burnup Snape with C los ed P Q res 5 1510-1:1 U..u-....o o 2 4 6 I Initial Cold Diametral Clearance.
in I 10 3 10 BimNuclear TMI Unit 1 Update*6 7/87 p.3.FIG-38 Fission Gas Release for 1.5 and 1.7 MaxlAvg.Axial Power Shapes (Initial Cycle)Fig.3.2-38 3500 3000 Desi an l illi t----114" Overpower-----Power 2500"'&It a.c"'Closed u m 2000=.-1.7 AIIaI Po.e r an u c EDt Bu rnup Snape-&It c 1.5 Axial Power ana u EDl Burnup Snape:::;)&It Pores&It 1500 u1.7 Ali al Power and Q,.&It EDl Burnup Snape"'CoD 1000 MIII.U.Desie" 500 o 2 4 6 8 10 Initial Cold Diaetral Clearance.
in I 10 3 BiE]Nuclear TMI Unit 1 Update*6 7/87 p.3.FIG-39 Maximum Gas Release to Pressure Inside the Fuel Clad for Various Axial Burnup and Power Shapes (Initial Cycle)Fig.3.2-39 t_.-.t I I,I Ik In.Fie tor CD HO'UN I'C fll (ntllllp, IIISI Flctor CD"OJ UU CELL CD HO'CO'NlI CELLHOT COhUOl 100 ULL Bil!I Nuclear TMI Unit 1 Update*6 7/87 p.3.FIG-40 Nominal Fuel Rod Power Peaks and Cell Exit Enthalphy Rise Ratios (Initial Cycle)Fig.3.2-40 t I hclllr"1 flCtor Q)NOT UNIT CEll (nUIII"**11'acurMOT tAll CEll OJ MOT CUNEa CEllMOT CONTaOl 100 ell l BE Nuelear TMI Unit 1 Update-6 7/87 p.3.FJG-41 Maximum Fuel Rod Power Peaks and Cell Exit Enthalphy Rise Ratios (Initial Cycle)Fig;3.2-41 1.1 1.5 1.41.3 1.2 1.1 CD 1.0 I CI-..N 0.9...-I...&:::a 0.1...*.0.7 IC::a-...0.&*..z-..0.5 u 00.4 0.3 0.2 011\G*2.21 I 10 1 IIi/tH-f t 2\.'\I\: 1-3 DNa Milt Flu.!\.(DISII" Lillit)'\'\r\.lin IIIUII ON**1.55'\\\\---/V\I/1\V V Calculatld Surflc.\Milt Flu.J'\I o 540 5&0 510600620 140 110&10 100 120 LOCII Enthalpy, Itu/III BimNuclear 1MI Unit 1 Update*5 7/86 p.3.FIG-42 Calculated and Design Limit Local Heat Flux Versus Enthalpy in the Hot Unit Cell at the Most Probable Condition (Initial Cycle)Fig.3.2-42 1.1 1.5 1.4 1.3 1.2 1.1 UI 1..0 I 0-)(0.9 N...-I"-0.8 s:: "-...CD O.7...::I-..-0.6....,%-lQ U Q0.4 0.3 0.2 O.1\,.2.5'I 10'Ib/tH-f t 2I I I I1 r I 11-3 ONI H**t Flu.(D.Slln l'.'t)"\1\\
ON'**1.12\\I..........
V" VI\V"1 Calculated SurfaceJ J....t Flul\, o 560 510 600 620&40 6&0 680 700 720 local Enthalpy, Itu/lb 0ENuclear TMI Unit 1 Update*5 7/86 p.3.FIG-43 Calculated and Design Limit Local Heat Flux Versus Enthalpy in the Hot Unit Cell at the Maximum Design Condition (Initial Cycle)Fig.3.2-43 CONTF,OL ROO ASSEMBLY PL ENUM ASSEMBLY OUTLET NOZ ZLE ILLANCE HOLDE R TUBE LOWEP, GRID FLO W OJS TR IfurOR CONTROL P,OD DRIVES STUDS INTERNALS VENT VALVE CONTROL GUIDE TUBE CORE SUPPORT SHIELD INLET NOZZLE FuEL ASSEMBLY REACTOR VESSEL THERMAL SHIELD GUIDE LUGS INCORE INSTRUMENT GUIDE TUBES________I N CO REI N S T RUM E NTN0ZZLE S Bil!1NucIe.r TMI Unit 1 Update*5 7186 p.3.FIG-44 Reactor VI_lind Intlmlll*Gen.Arnngement Fig.3.2-44 CONTROL ROD ASSEMBLY LDCATION INCORE INSTRUMENT LOCATION REACTOR VESSEL THERMAL SHIELD CORE BARREL SURVEILLANCE SPECIMEN HOLDER TUIE BimNuclea, TMI Unit 1 Update*5 7/86 p.3.FIG-45 Reactor Vessel and Internals-Cross Section Fig.3.2-45 0imNuclear 1MI Unit 1 Core Flooding Arrangement p.3.FIG-46 COlE fL ODD I N&NOZZLE Update*5 7/86 Fig.3.2-46 1-,--+----+++---tr----
t rn I eMf I....'MilL.SICT**I-I Bi!lNucl.r 1MI Unit 1 UI'C..I lie.u".'URl.IU Update*5 7/86 p.3.FIG-47 Internals Vent Valves Fig.3.2-47 BimNuclear 1MI Unit 1 SEE SECTION BELOW Update*5 7/86 p.3.FIG-48 Internals Vent Valve Clearance Gaps Fig.3.2-48 TOP VIEW Update*5 7186 TMI Unit 1 FUll AlUmbly COHTROl.ROO GUIDE TUH , ,'NSTRUMENTATION I TUM I r::: i.'i J.: ,',',r.I'i.i......'\,FUEL ROO ASSEMILY.!iI r:: 1:7'Ir1I n'n I r I , I I , T 1 1 CROSS SECTION I 1 , i I 1\I[i]1*[J, l!I r LOWER END FITTINGIi'-I-'
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'irI INSTRUM*NTATION TUH.-l CONNECTION SNeER GRID U"O ENO trlTTING))Fig.3.2-49 p.3.FIG-49 SPIDER TOP VIIW COUPLINt NEUTRON MATERIAL CONTROl.ROD p.3.FIG-50 ra::imNuclea, TMI Unit 1 Control Rod Assembly Update*5 7/86 Fig.3.2-50 SPIDER TOP VI!W NEUTRON ABSORItNG MATERIAL.CONTROL ROD
..r 1MI Unit 1 Axial Power Shaping Rod Assembly Update*5 7/86 p.3.FIG-51 Fig.3.2-51 SPIOfR----IURNAILI POISON ROD TOP VIIW.URNA.LI POIION MATIRIAL p.3.FIG-52 0iI!JNuclear 1MI Unit 1 Burnable Poison Rod Assembly Update*5 7/86 Fig.3.2-52 TOP VIEW SPIDER-----....
ORIFICE ROO p.3.FIG-'53 Bil!l Nuclear TMI Unit 1 Orifice Rod Assembly Update*5 7/86 Fig.3.2-53 tOJSING ASSY LOAD ARM ASSDfSl.Y p.3.FIG-54 0:W!l Nuclear TMI Unit 1 Side View of BPRA Retainer Update*5 7/86 Fig.3.2-54 UPPER CORE PLAT!ASSY PAD TYP p.3.FIG-55
..r TMI Unit 1 Top View of BPRA Retainer During Operation Update*5 7/86 Fig.3.2-55 POsmoN INDICATQIIt AlIDIILY STATOR.IIIMLV p.3.FIG-56 I I CCUIL.NI AlIDIILV i i BimNuclear TMI Unit 1 Control Rod Drive-Gen.Arrangement Update*5 7/86 Fig.3.2-56
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L S SEC"" 1111 SEC 110.,., SEC"" S-S Bil!INucI**r TMI Unit 1 UpdItI.5 7111 p.3.FIG-57 Fig.3.2*57