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4-38 All of the above fuel cycles are for a 3360-MWt, 1344-MWe HTGR with a core power density of 7.1 Wt/cm3. Table 4.4-1 provides a summary, obtained from the detailed mass balance information in ref. 1, of the conversion ratio, fissile requirements, fissile discharge, and U3O8 and separative work requirements. an equilibrium cycle enrichment of 7.4 w/o and 8.0 w/o, respectively. preferred for no-recycle conditions. Cases 1-a and 1-b involve the use of LEU fuel with Case 1-b would be In Case 2 thorium is used with 20% 235U/U (MEU/Th) for no-recycle conditions. that while the initial U308 and fissile loading requirements are higher for the MEU/Th case than for the LEU cases, due to the larger thermal absorption cross section of thorium and the partial unshielding of the 238U resonances resulting from its reduced density, the cumulative U308 requirements are slightly less for the MEU/Th case. attainable in HTGRs and the res+ltant large amount of bred 23% which is burned in situ. Other converter and advanced converter reactors (LWRs , SSCRs , and HWRs ) typically require less U308 for the LEU case than for the MEU/Th case with no recycle. Note This results from the high burnup Case 3 also uses the MEU/Th feed but with recycle of 233U. The unburned 235U and The bred 233U re- plutonium discharged in the denatured 235U particles is not recycled. covered from the fertile particle, however, is denatured, combined with thorium, and recycled. In the calculations for all cases involving recycle of denatured 233U, GA assumed that an isotopic mix of 15% 233U and 85% 238U provided adequate denaturing. Due to the high burnup and the fact that the thermal-neutron spectrum in HTGRs peaks near the 239Pu and 241Pu resonances, a large amount of the fissile plutonium bred in the denatured fuel is burned in situ, thus resulting in the low fissile plutonium content of the fuel at discharge. siderable 238U self-shielding is obtained by the lumping of the 238U in the coated particle kernels. Studies are currently underway at GA concerning the use of larger diameter fissile particles, thereby lowering the 238U resonance integral and, consequently, the amount of bred plutonium di~charged.~ Case 4 employs a denatured 233U feed and includes uranium recycle. It represents a feasible successor to Case 3 once an exogenous source of 233U is available. Case 5 involves Pu/Th Fuel. Since no 238U is present in the core, no plutonium is bred; only 233U is bred. resulting in enhanced neutron economy. essentially achieves the "Phoenix" fuel cycle effect, i .e., the decrease in 239Pu content is largely compensated for by buildup of 241Pu from 240Pu capture and by buildup of 233U from 232Th capture, resulting in a nearly constant ratio of fissile concentration to 240Pu concentration. Therefore the fuel reactivity is relatively constant over a long burnup period, reducing the need for control poison. This allows the core to be batch loaded; i.e., the entire core is reloaded at approximately 5-yr intervals. minimizes down time for refueling and eliminates problems of power sharing between fuel elements of different ages. This reactor has greatly reduced requirements for control poison, This results from the fact that this Pu/Th HTGR This reload scheme Furthermore, it allows easy conversion to a U/Th HTGR after any cycle. It is important to note that the Pu/Th case presented in Table 4.4-1 is not Con- il L c c L f -i Li I' ' C
Table 4.4-1. Fuel Utilization Characteristics for HTGRs Under Various Fuel Cycle Options fnitial Core Requirements: U308 Requirementc Esuilibrium Cycleb - (ST/GWe) Separative Work Requiremento (lo3 kg SWU/GWe) Case, Fuel Type - Conversion Ratio . (1st Cy./Eq. Cy.) Fissile Inventory (kg/GWe) HM Loading (MT/GWe) Fissile Makeup (kg/GWe-yr) Discharge of Nonrecycl able Fissile Material (kg/GWe-yr) Initial 30-yr Total for CF of;, 65.9%/75% Initial 30-yr Total for CF os 65.9%/75% 1-a, LEU. no recycle, C/U = 350 1-b, LEU, no recycle, C/U = 400 2, MEU(PO% 235U)/Th. no recycle, C/Th = 650 3 MEU(2O% 235U)/Th,f 233~ recycle, C/Th = 3O6/40Og 4, MEU(l5X 233U)/Th,f U recycle, C/Th = 274/300 5, Pu/Th. C/Th = 650 6. HEU(235U)/Th, no recycle. C/Th = 214/238 7, HEU(233U)/Th, hi/gain. U recycle, C/Th = 150 8. HEU(235U)/Th, hi/gain, U recycle, C/Th = 180/180 0.580/0.553 901 23511 0.557/0.526 819 235U 0.630/0.541 1077 235U 0.682/0.631 1474 235U 0.824/0.764 1168 233U 0.617/0.617 3153 Pufh 0.723/0.668 1358 235U 0.915/0.859 1395 23% 139 235U ~ - Dispersible Resource-Based Fuels 24.6 U 608 235U 113 235U 69 Puf 21.6 U 576 235U 77 23% 52 Puf 5.4 U 551 235U 47 2% 20.2 Th 74 23311 22 Puf 7.4 u 397 235u 27.5 Th _bisperslble Denatured Fuel 65 235U 36 Puf 7.9 U 246 23311 35 Puf 30.7 Th Energy-Center-Constrained Fuel 12.2 Th 630 Puf 102 Puf 97 23% Reference Fuels 1.5 U 508 235U 49 235u 37.2 Th 2.0 u 53.0 Th 120 2330 12 235U 183 233U 1 Puf - : f 25u 217 197 274 371 0 0 345 0 sodBk 4272/4860 4040/4594 3967/4515 3229/3666 0 0 3864/4395 :Initial cycle lasts one calendar year at 60% capacity factor. _Eaullibrium cycle capacity factor is 72%. >sumes 0.2 40 tails. -Value preceding slash is for an average 30-yr capacity factor of 65.9; value following slash Is for a constant capacity factor of 75%. o credit taken for end of life core. 40 23511 frm MEU particle or Pu recycled in Case 3; all U recycled in Case 4, but no Pu recycled. !Initial coreheload sement. :Core is batch loaded; initial load provides fissile material for 45 yr of operation. jReference fuels are considered only as limiting cases. Initial cycle length is 1.6 yr. 'Numbers shown are for a capacity factor of 75%. 0 /2280 142 3319/3781 130 3188/3629 249 3640/4143 340 2933/3361 0 0 0 0 344 3858 f 4387 0 0 50dsk /2278
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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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-- t Li - I' b -. c (u I i ai L i i
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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
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3-1 0 3.2. RADIOLOGICAL HAZARDS OF
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3-1 2 232u IN RECYCLED HTGR FUEL (p
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3-14 3.2.2 Toxicity of 232Th Given
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1. 2. 3. 4. 5. 6. 7. 8. 9. 10. 11.
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3.18 There are three approaches whi
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1 i i 3-20 3.3.2. Fabrication and H
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The 3-22 3.3.3 Detection and Assay
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3-24 3.3.4. Potential Circumvention
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3-26 - either large centrifuge pilo
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
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Page 87 and 88: u -- x i I*i 1- bi 4-1 7 The use of
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Page 91 and 92: 4-21 Case B" is a modification of C
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: c r- L i] L h L __ f; t 4-37 trons
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
Page 127 and 128: 4-57 _I 10 10' 2 2' CASE NUMBER -.
Page 129 and 130: 4-59 Most of the information availa
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Page 138 and 139: ! 5-4 5.1. REACTOR RESEARCH AND DEV
Page 140 and 141: 5-6 As has been pointed out above,
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Page 144 and 145: 5-10 The fuel performance program u
Page 146 and 147: . 5.1.2. Government Research and De
Page 148 and 149: 5-1 4 The first aspect of large pla
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Page 152 and 153: I DATA BASE DEMu)pMENT DEMO DESIGN
Page 154 and 155: j ~ 5-20 The R,D&D costs are highes
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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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