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4-20 Table 4.1-5. Fuel Utilization Characteristics for BWRs Under Various Fuel Cycle Options' Separative Work Case Fuel Type Initial Fissile Inventory (kg/GWe) Fissile Charge (kg/G%yrJ Equilibrium Cycle Fissile Discharge Burnup (kg/GWe-yr) (MWO/kg HM) U30s Requirement (ST/GWe 1 30-yr Initial Total' Requirement (10' kg SWU/GWe) 30-yr Initial Total' A 8 8' B" D E F Gh H LEU. no recycle 220OC Dispersible Resource-Based Fuels MEU/Th. no recycle' 1132 244 233U 31.6 428 23sU 53 Puf LEU/Th mixed lattice, - 854 23511 2:; z; L 28.7 no recyclee LEU/HEU/Th mixed lattice, - no recyclee LEU/MEU/Th mixed lattice, self-generated U recyclee Denatured 233U02/Th02, - U recycle (exogeneous - 2S3U makeup)e 799 23% 235 23% 28.4 496C'd 6051d 235C'd 3493' 150 Puf 138 Puf 917 235U 125 233U 30.0 277 23sU 92 Puf 147 233U 152 233U 30.5 742 235U 245 23sU 98 Puf Dispersible Denatured Fuel 770 233U 452 *))U 31.6 15 235U 17 23sU 55 Puf Energy-Center-Constrained Fuels LEU, recycle of U + self-generated Pu 2200' 28.4 PuO2/UO2. Pu recycle - 71 23511 1178 Puf 38 2% 808 puf 27.7 Pu02/Th02. PU recyclee - 1705 Puf 275 233U 954 Puf 29.8 aAll cases assume 0.2 w/o tails and 75% capacity. factor; blank columns included to show no data corresponding to that given for PWRs (Table 4.1-1) are available. bNo credit taken for end-of-reactor-life fissile inventory. 'Initial cycle is 1.47 yr in length at 75% capacity factor. dFrom ref. 9. Based on three-enrichment-zone initial core, axial blankets and improved refueling patterns which are currently being retrofitted into many BWRs. 30-yr Up, and SWU requirements supplied to INFCE for a reference BWR not employing these improvements are 6443 ST UpalGWe and 3887 x 103 SWU/GWe respectively. 'Analyses performed for equilibrium cycle only. 'Approximated from equilibrium cycle requirements. c gFrom ref. 8. 'From ref. 10; adjusted from 80% capacity factor to 75%. $ails uranium used for plutonium diluent. Case B' is a perturbation to the reference U02 BWR assembly design in that the four U02 corner pins in each fuel assembly are replaced with four pure Tho2 pins. The remaining U02 pins are adjusted in enrichment to obtain a desirable local power distribution and to achieve reactivity lifetime. ments by only 2% relative to the reference design. removing the Tho2 corner pins from the spent fuel assemblies, reassembling them into new assemblies, and reinserting them into the reactor. 0 496' i 0 868d 6201 6852 55035 0 38699 i 0 0 235' i 0 7764 3836 5100f 3895f 0 19809 In the once-through mode this design increases U308 require- achieve increased burnups (and also increased 233U production) without reprocessing. U308 requirements for this scheme (i.e., re-use of the Tho2 rods coupled with UO:! stowaway) are approximately 1.3% higher than for the reference U02 cycle.8 This option could be extended by This would permit the Tho2 pins to i 0 L 1:
4-21 Case B" is a modification of Case B' in that in addition to the four Tho2 corner pins, the other peripheral pins in the assembly are composed of MEU(235)/Th. remainder of the pins contain LEU. requirements by 12% relative to the reference BWR U02 design. The In the once-through mode this design increases U308 Both Case B' and Case B" would offer 'operational benefits to the BWR since they have a less negative dynamic void coefficient than the reference U02 design.8 This is desirable since the sensitivity to pressure transients is reduced. As shown in Table 4.14, in equilibrium conditions a BWR employing the Tho2 corner pin once-through design would discharge 24 kg 233U/Gl.le annually while the BWR employing the peripheral Tho2 mixed lattice design would discharge 125 kg 233U/GWe annually. Use of these ciptions in the once-through mode not only could improve the operational performance of the BWR but also would build up a supply of 233U. This supply would then be available if a denatured 233U cycle (together with reprocessing) were adopted at a later time. Furthermore, use of the mixed lattice designs could be used to acquire experience on the performance of thorium-based fuels in BWRs. in the once-through mode may also be feasible in PWRs. Similar schemes for the use of thorium Although only limited scoping analysis of the safety parameters involved in the use of alternate fuels in BWRs has been performed,8 the BWR thorium fuel designs appear to offer some advantageous trends over U02 designs relative to BWR operations and safety. Uranium/thorium fuels have a less negative steam void reactivity coefficient than the U02 reference design at equilibrium. This effect tends to reduce the severity of overpressurization accidents and improve the reactor stability. reactivity coefficient for the denatured 23%/Th fuel indicates that the core will have a flatter axial. oower shape than the reference U02 design. increase in kW/ft margin and increase the maximum average planar heat generation ratio (MAPLHGR). Alternatively, if current margins are maintained, the flatter axial power shape could be utilized to increase the power density or to allow refueling patterns aimed at improved fuel utilization. References for Section 4.1 The less negative void This could result in an 1. N. L. Shapiro, J. R. Rec, and R. A. Ilatzie (Combustion Engineering), "Assessment of Thorium Fuel Cycles in Pressurized Water Reactors," EPRI NP-359 (Feb. 1977). 2. "Thorium Assessment Study Quarterly Progress Report for Second Quarter Fiscal 1977," ORNL/TEl-5949 (June 1977). 3. R. A. Hatzie, J. R. Rec, and A. N. Terney, "An Evaluation of Denatured Thorium Fuel Cycles in Pressurized Water Reactors,' paper presented at the Annual Meeting of the American Nuclear Society, June 12-16, 1977, New York, Mew Yorl:.
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Interim Assessment of the Denatured
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i L i b I, i L L bi L L Contract No
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1, L L t i Ltr 1 L L V PREFACE AND
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L i Li 1, I b k b id ii t L id b i
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I L bi i; L 7.3.1. Possible Procedu
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i xi I b L ABSTRACT A fuel cycle th
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I b .- i, hd t t CHAPTER 1 INTRODUC
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L .- k id .- I L: _ - I iclr .- i b
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huJ -- AI I ' b --- I bl --- t L L
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L - i' u i Lt L r- L t- b L 2-7 den
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2-9 1. Nuclear power is limited to
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e Isr i L i- b t CHAPTER 3 ISOTOPIC
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232" Fig. 3.0-1. Decay of 232U. ORN
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3-6 3.1. ESTIMATED U CONCENTRATIONS
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3-8 The values in Table 3.1-2 are a
Page 40 and 41: 3-1 0 3.2. RADIOLOGICAL HAZARDS OF
Page 42 and 43: 3-1 2 232u IN RECYCLED HTGR FUEL (p
Page 44 and 45: 3-14 3.2.2 Toxicity of 232Th Given
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Page 48 and 49: 3.18 There are three approaches whi
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Page 52 and 53: The 3-22 3.3.3 Detection and Assay
Page 54 and 55: 3-24 3.3.4. Potential Circumvention
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Page 58 and 59: ! 3-28 Table 3.3-3. Enriched Proauc
Page 60 and 61: 3-30 The high alpha activity of ura
Page 62 and 63: 3- 32 Table 3.3-7. Sumry of Results
Page 64 and 65: 3-34 statistical redundancy through
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Page 68 and 69: 3-38 An additional factor relative
Page 71 and 72: I b t I t L 1; L t ki 4.0. 4.1. CHA
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Page 75 and 76: Y m I I I . aro ID0 10.00 4-5 ORNL-
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Page 81 and 82: _- I -- L i d .-- L 4-1 1 Reference
Page 83 and 84: _- id L t I , 4d i, L id L 5 bi t h
Page 85 and 86: -- - . & I I ' h I ' Y 61 L -- I( L
Page 87 and 88: u -- x i I*i 1- bi 4-1 7 The use of
Page 89: la - i , I br -_ t c c L e L 4-1 9
Page 93 and 94: c 4-23 L i a L L 4.2. SPECTRAL-SHIF
Page 95 and 96: 4-25 Initial analyses of spectral-s
Page 97 and 98: h 1' b t c i--; & I- & 1: i f 4-27
Page 99 and 100: c- i L L e I b 4-29 . safeguards fo
Page 101 and 102: L t i- bi L L t i 4-31 a significan
Page 103 and 104: ~ kc'l c" -1 r-"! r! r-'i c Table 4
Page 105 and 106: F-- I iJ L ii - L c c 4-35 Referenc
Page 107 and 108: c r- L i] L h L __ f; t 4-37 trons
Page 109 and 110: Table 4.4-1. Fuel Utilization Chara
Page 111 and 112: L I 1 I Ld I ie i-- 1 t' 4-41 4.4.2
Page 113 and 114: 4-43 Table 4.4-4. PBR Coated Partic
Page 115 and 116: HEU/Th 0.59 Seed i3 Breed 0.58 HEU/
Page 117 and 118: L i; 1 I i; i 1. 2. 3. 4. 5. 6. 7.
Page 119 and 120: 4-49 Table 4.5-1. Fuel Utilization
Page 121 and 122: ments. 4-51 The calculations for LM
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Page 125 and 126: Table 4.6-1. Comparison of Fuel Uti
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5-6 As has been pointed out above,
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5-8 the LMFBR on its reference cycl
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5-10 The fuel performance program u
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. 5.1.2. Government Research and De
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5-1 4 The first aspect of large pla
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5-1 6 and should also address metho
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I DATA BASE DEMu)pMENT DEMO DESIGN
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j ~ 5-20 The R,D&D costs are highes
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5-22 In the case of metal-clad oxid
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5-24 5.2.2. Research, Development,
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5-26 program, including hot testing
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c c
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6-4 persed areas ensured - a fact w
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6-6 6.1.2. Reactor Options The reac
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Reactor/Cycl ea LWR-U5(LE)/U-S LUR-
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6-10 6.1.3. Nuclear Policy Options,
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Table 6.1-4. Nuclear Policy Options
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SWU SWU 6-14 UZJS USILEINS WYI U5IL
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6-16 n nM MEDL nows.1 Option 4: In
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6-18 HM HEDL7OOl.70.3 Option 6: In
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SRCIFIED CONSTRUCTION LlMllLD INTRO
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50- 40 6-22 c Y \ VL - :-: F 30- -
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6-24 The potential nuclear contribu
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%- f $400 J 2 s 2, 1 THE LWR FOLLOW
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141.6 ST - U308 4 7.2 lo3 swu ENRIC
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6-30 reactor. Since this increase i
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‘9. ,1 ST ‘3O8 6-32 12 Kg HM I
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6-34 1HE LWR WITH FISSILE UUNIUM PC
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THE LWR WITH PURONIUM MlNUllUTlON A
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6-38 be located in energy centers a
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6-40 installed capacity must be han
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lm, 1 I I I I mf RR WITH UGHT rwrmi
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6-44 6.2.7. Converter-Breeder Syste
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25.2 ST swu 6-46 19 Kg fir Pu 13 Kg
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6-48 Table 6.3-2. Summary of Result
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6-50 (4) If all plutonium produced
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I d r-- 1 > w I r- k z r L id r- .
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L 1 L u without benefit of U.S. exp
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L a, - L i -- k; L t IJ L 7-7 While
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7-9 Viewed solely from the plutoniu
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L L i; i; 7-1 1 routinely in LWRs a
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'I 7-1 3 t I: k: The isotopic compo
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U L L L L ld 1 i" * required 233U m
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I I, i 7-1 7 ii I 1 I L L k L i Thr
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li i i L Li L & I ' i kr ii L i h i
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7-21 7.3. PROSPECTS FOR IMPLEMENTAT
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L I L t L i L i 1 L 7-23 7.3.1. Pos
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7-25 7.3.2. Considerations in Comme
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.. I b k; L; i b b 1J I L 7-27 cycl
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7-29 7.4. ADEQUACY OF NUCLEAR POWER
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L i' hd .- - I ' i t L i' b i -' b
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x ibd 1 .. I i, L -- i t Table 7.4-
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7-35 cycled in Pu/Th converters, th
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L 1' b I_- ,- i b -- I ' b L 7-37 T
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..a 4 I- 1 ' u c1 Y u L 7-39 betwee
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L e, I I ' lb t 7-41 Systems that u
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e-. . ! hr L3 z i- b: i- b L c i li
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7-45 j I Table 7.5-1. Integrated As
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- isi. c ‘Id i L c 7-47 Although
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e Other e e e On the e 7-49 technol
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t a APPENDICES
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A-3 Appendix A. ISOTOPE SEPARATION
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I] 1- $i ill 8 LI b' The Gas Centri
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id A- 7 The Component Preparation L
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i ?L 1- U t L fl bi L c u La -- ti
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p; ‘U i-- L ,c c Japan A-1 1 Tabl
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13 A-13 L" efficient of the units d
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L L L t I h h u L; _- t f is throug
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u A-1 Aerodynamic Methods 7 Both th
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i i d .- I . ii hi L i: i t ' Li 1
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la t L I I, lj ir 1 Li c I L L' B-1
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i; 1 t u 1 1 B-3 Table 8.5. Reactor
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Table 6-7. Marginal Costs of U308 a
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LJ L L1 a .E 4' P Y W 5 18 2 16 : 0
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Appendix C. DETAILED RESULTS FROM E
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1 i; Li b L lkd t L L I hd I ' L; i
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L c-5 The introduction of the class
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i ' ibd 1 b i L L i L I Li . . b b
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E , 1 ' b _. I ' L 1 b _- i G i' b
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-, ii I ; b i- ii L b 4 i L L i‘
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_- I ii I _ _ I L I td C i b L i, L
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Qulatiw klur W i t y hilt (*I thrmg
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L c c b - i ii L r D-1 Appendix D.
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L L L L [i i” Table D -2. Maximum
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Fig. D-2 (cont.) 25W I I I I I (I)
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[i energy center. It can be seen fr
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L L i il L t W 0-9 Table D-4. Summa
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-- Lid c i' .- .b) i Internal Distr
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It e L L L t u I' f ' ki Federal Ag
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u 1' 1" Outside Organizations (cont
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