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3- 32 Table 3.3-7. Sumry of Results of Centrifuge Enrichment Survey of Potential Fuel Mixture" ~~ Fissile Content Number of Centrifuges Required 232U Content (wt. Fraction) (wt. Fraction) (0.3 kg SWU/yr Zippes) In 232U In Fissile Of 1st Cascade Of 2nd Cascade Of 2nd Of 2nd Cascard Cascade Annual Feed Stripping Enriching T:$ Initial Product Product Tails Product (kg U/yr) Cascade Cascade Total A 8 C 0 E F G H 0 N A ~ 5.02(-4)' PA 5.02 -4 0.005 5.021-41 0.01 5.02(-4) 0.10 0 0.002 4.15(-3) 0.005 2.7 6 0.01 1.31161 0.005 8.1(-7) 0.005 6.564 -4) 6.5641-4) 6.564(-4) 6.564(-4) NA 0.0065 0.01 0.1 5.626(-3) 2.68(-6) 1.63(-6) 8.5(-7) 0.005 0.01 0.005 0.005 0 NA 0 0.01 1.236 -4) 1.2361-4) 1.236 (-4) 0.001236 0.00235 0.00235 2.445(-4) 2.445(-4) 0.002445 0.003 1.134(-4) 0.003 2.331(-4) 2.331(-4) 0.0023 0.003 2.4!-6) 1.14(-6) 6.67(-7) 0,065 0.06 0.0 2.63(-6) 7.87(-6) 0.. 15 0.005 6.42(-7) 0.005 2.5(-6) 7.44(-7) 0.0715 0.005 0.90 2993 0.90 0.90 0.90 0.90 0.90 0.90 0.90 0.90 0.90 832 31 80 1302 81 7 860 3000 1749 853 468 3000 1210 704 3001 860 664 3000 805 Natural Uranium 0 NA 0 0.002 0.90 17575 "Feed and centrifuges needed to produce 100 kg U/yr of 90% fissile product. bSee Table 3.3-2 for description of fuel types. %cad: 5.02 x 'NA = not applicable. A 3.2 wt % 235U D Natural uranium (0.711 wt % 235U) - 20 wt % 235U with thorium 0.90 0.90 0.90 0.90 0.90 0 0 82410 50600 41653 0 86227 61277 45483 0 25244 15459 9292 33033 11872 8758 321 36 1 1889 0 29220 5468 10880 9981 7257 9191 18302 18802 11277 4991 7002 5921 13635 14398 20982 13033 1241 9 19477 7791.8 29220 5468 93290 60581 48910 9191 104529 80079 56760 4991 32246 21380 22927 47431 32854 21791 44555 31366 Table 3.3-8. Enrichment Resistance of Fuel Mixtures Investigated* Fuel Type Number of Centrifuges Feed Requirements Approximate Number of PWR Fuel Assemblies (kg Ulyr) Needed to Supply Feed 1st generation 233U recycle with thorium No 232U removal With 232U removal 5th generation 233U recycle with thorium No 232U removal With 232U removal 1st generation 235U recycle with thorium (Option 1) With 232U removal 5th generation 235U recycle with thorium (Option 1) with 232U removal 1st generation 235U recycle with thorium (Option 2) With 232U removal 5th generation 235U recycle with thorium (Option 2) With 232U removal 29,220 4,991 77,918 5,469 48,910 9,191 80,079 2,993 468 17,575 832 81 7 860 1,750 77916 22,927 704 6.8 32,854 860 21,791 664 6.6 31 ,366 *Feed and centrifuges needed to produce 100 kg U/yr of 90% fissile product. 805 6.7 4.8 Not Applicable 7.1 6.9 7.0 14.2 7.4 7.0 t; I L* c
L U 1 h i t ! id L id L 1 i 1; t L 1 t 1 bi i' L 3-33 1. Of the fuel mixtures that may be in commerce in a thorium-based fuel cycle, 20% 23% 2. 3. mixed with thorium is the most desirable both in ease of enrichment and because it requires diversion of the fewest fuel assemblies to produce a given quantity of highly enriched uranium. Enrichment of 233U recycle fuels, without 232U removal, is an enrichment task com- parable (with respect to the number of centrifuges) to enriching 20% 235U. product, however, will be highly radioactive. If would-be proliferators must remove the 232U, the 235U makeup fuels are less prolifera- tion resistant than the 233U makeup fuels. 4. The 235U recycle fuels with thorium and 232U removal are equivalent to 3.2 wt% 5. slightly enriched uranium fuels with respect to both the number of centrifuges and the number of fuel assemblies to be diverted. The 233U recycle fuels with thorium and 232U removal are equivalent to natural uranium enrichment with respect to the number of centrifuges. 6. If 232U removal is necessary for ease of weapon manufacture and reliability of delivery, then a diverter would probably prefer to divert either slightly enriched uranium fuel or enrich natural uranium than to enrich either 235U or 233U recycle fuel from thorium cycles. This conclusion results from the fact that for each recycle fuel, the cor- responding slightly enriched or natural uranium fuel enrichment plant requires approximately the same number of centrifuges but has the decided advantage of a nonradioactive facility. Reliability of Centrifuge Enrichment Plants. As a final item, the average centrifuge failure rate and its impact on the maintainability and production rate of a centrifuge enrichment plant must be considered. centrifuges is scarce. The URENCO-CENTEC organization has over the years made claims of very long average operating life and correspondingly low failure rates. Typical examples of these claims can be found in some of their sales brochures.13 These claim an average 10-year operating life and a failure rate of less than 0.5%/year. maintenance (e.g., oil changes and bearing inspection) is required to achieve these low failure rates. The Information on the reliability and operating life of It is not clear how much periodic If these claims are accepted as a goal of a long-term development project, then it can be assumed that in the early part of the development somewhat higher failure rates would occur, perhaps greater by a factor of 10. This factor might be further justified in a highly radioactive plant since periodic maintenance would not be practical. The effect of centrifuge failures on the production rate in a radioactive plant has not been determined; however, some qualitative statements can be made. All centrifuge plants must be designed so that failed units or groups of units can be immediately isolated from the rest of the plant. It should also be possible, for a specific cascade layout, an assumed failure rate, and a specified plant operating life, to provide
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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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Page 40 and 41: 3-1 0 3.2. RADIOLOGICAL HAZARDS OF
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4-43 Table 4.4-4. PBR Coated Partic
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HEU/Th 0.59 Seed i3 Breed 0.58 HEU/
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L i; 1 I i; i 1. 2. 3. 4. 5. 6. 7.
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4-49 Table 4.5-1. Fuel Utilization
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ments. 4-51 The calculations for LM
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1. 2. 3. 4. 5. 6. 7. 8. 4-53 0 77-1
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Table 4.6-1. Comparison of Fuel Uti
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4-57 _I 10 10' 2 2' CASE NUMBER -.
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4-59 Most of the information availa
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- i b k L L -- I , b -- I .- I Li L
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4-63 on only the preliminary data p
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c e E c
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! 5-4 5.1. REACTOR RESEARCH AND DEV
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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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