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e A-10 technique is that a module, de'fined as a separation unit consisting of one set of com- pressors and one set of separation elements, does not as in the classic case, produce only one separation factor of enrichment in one pass but can produce for a constant separative work capacity various degrees of enrichment up to a maximum of several times the separation factor over the element. Full scale modules of this type are nearing the prototype stage. Recent design improvements are expected to result in a nominal capacity of 80 to 90 kg SWU/yrll per separation module. A valuable feature of a plant based on this process is its very low uranium inven- tory, which results in a short cascade equilibrium time, of the order of 16 hours for a commercial plant enriching uranium to 3% 235U. The theoretical lower limit to the specific energy consumption of the separation element can be shown to be about 0.30 MW.h/kg SW. The minimum figure observed by the developers with laboratory separating elements is about 1.80 MW.h/kg SW, based on adiabatic compression and ignoring all system inefficiencies. This difference is a measure of the improvement potential expected by the South Africans. Current and Projected Enrichment Capacity Most of the known installed enrichment capacity is based upon gaseous diffusion technology. Only small increments of centrifuge technology are in operation (1 .e. , URENCO, Japan and U.S.), and one plant utilizing modified nozzle technology (the South African Helikon plant) may be operating. Indicative of the status of other isotope separation methods, all planned additions to the world enrichment capacity are based on either diffusion, centrifuge or nozzle technology. The existing worldwide capacity and planned additions to capacity are shown in Table A-1 by colcntry and technology type. In the table the groups identified as Eurodif and Coredif are multinational organizations building gaseous diffusion plants in France. A.2, New Separation Technologies In addition to the more developed technologies (gaseous diffusion, gas centrifuge, and the Becker nozzle), there are several other separation methods that either have been utilized in the past or are currently being developed. These technologies are listed in Table A-2. L L Id L E. P u L b 6' L
p; ‘U i-- L ,c c Japan A-1 1 Table A-1. Approximate Schedule of World Enrichment Capacitya Caoaci tv World’s Cumulative Year Nation or Group Technology type Increment (MT SWU) Present Status of Increment Capacity (MT SWU) 1977 U.S.’ UK-France Russ i ac China Diffusion Diffusion Diffusion Oi ff us i on 15.400 800-1000 800 Unknown Existing Existing, but dedicated to military use Existing. actual total capacity unknown Existing, mostly military 15.400 16,400 17.200 URENCO Centrifuge 200 Existing 17,400 U.S. S. Africa Centrifuge Helikon-Fixed 50 Unknown Existing Existing pilot plant or in 17,450 wall centrifuge process of coming on-line 1978 US.’ URENCO Japan RussiaC Diffusion Centrifuge Centrifuge Diffusion 3,300 200 20 200 From CIP/CUP olus added power purchase Facilities at Almelo & Capenhurst now in construction Currently under construction 20.750 20.950 20.970 21,170 1979 us.’ RussiaC URENCO Eurodif Diffusion Diffusion Centrifuge Diffusion 2.200 500 400 2,600 From CIP/CUP Under construction 23.370 23.870 24,270 26.870 1980 us.’ URENCO Eurodif RussiaC Diffusion Centrifuge Diffusion Centrifuge Diffusion 1,600 400 3.700 30 500 From CIP/CUP Planned Under construction Under construction 28.470 28,870 32.570 32,600 33.100 1981 1982 us.’ LIREMCO Eurodif Russ i ac U.S.’ IJRENCO Eurodif RussiaC Brazi 1 UREtlCO Coredif us.’ URENCO S. Africa Diffusion Centrifuge Diffusion Diffusion Diffusion Centrifuge Diffusion Diffusion Becker nozzle Centrifuge Diffusion Diffusion Centrifuge Fixed wall centrifuge Diffusion 700 400 2,100 500 300 400 From CIP/CUP Planned Under construction !J 36 ;SO0 T- u i Coredi G; f D 1983 1984 2.400 500 180 1.300 1,800 2,000 1.300 1,600 f 1,800 1985 US.’ Diffusion URENCD Centrifuge S. Africa Fixed wall centrifuge Coredif Diffusion Japan Centrifuge Incr. Power Implementing CUP Planned Under construction Planned Planned Planned Incr. Power Implementing CUP Planned Planned Planned 37,000 37.500 39.900 40.400 40,580 41.880 43,680 45,680 46.980 48,580 50.380 2,000 Incr. Power Implementing CUP 52,380 1.400 Planned 53.780 1.600 Planned 55.380 1,800 Planned 57,180 6,000 Planned, but should be 63,180 considered conditional 1986 US.’ S. Africa URENCO Centrifuge Fixed wall centrifuge Centrifuge 550 1.800 2.000 Planned Planned Planned 63,730 65.530 67,530 1987 U.S.’ URENCO Centrifuge Centrifuge 2.750 2.000 Planned Planned 70,280 72,280 1988 US.’ Centrifuge 3,300 Planned 75,580 1989 U.S.’ Coredif Centrifuge Diffusion 2,200 5.400 Planned Planned, but should be considered conditional 77.780 83,180 ‘Infornation from references 12 and 13. ’Not included in this schedule are possible additions to the U.S. enrichment capacity by private corporations, such as Exxon Nuclear, Garrett and Centar; these may amunt to as much as 10.000 MT SWU by 1990. =For Russia, this is a schedule of growth in enrichment sales avaflability and not necessarily of capacity expansion.
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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 - 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
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! 3-28 Table 3.3-3. Enriched Proauc
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3-30 The high alpha activity of ura
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3- 32 Table 3.3-7. Sumry of Results
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3-34 statistical redundancy through
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introduce time, cost and visibility
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3-38 An additional factor relative
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I b t I t L 1; L t ki 4.0. 4.1. CHA
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L L _ _ { i J _- \ I b B u il c _.
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Y m I I I . aro ID0 10.00 4-5 ORNL-
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t .. -. F 4d -- I ' L 0- 122, L _-
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L L; L t -I I ' b L -- t i b -- I;
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_- I -- L i d .-- L 4-1 1 Reference
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_- id L t I , 4d i, L id L 5 bi t h
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-- - . & I I ' h I ' Y 61 L -- I( L
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u -- x i I*i 1- bi 4-1 7 The use of
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la - i , I br -_ t c c L e L 4-1 9
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4-21 Case B" is a modification of C
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c 4-23 L i a L L 4.2. SPECTRAL-SHIF
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4-25 Initial analyses of spectral-s
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h 1' b t c i--; & I- & 1: i f 4-27
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c- i L L e I b 4-29 . safeguards fo
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L t i- bi L L t i 4-31 a significan
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~ kc'l c" -1 r-"! r! r-'i c Table 4
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F-- I iJ L ii - L c c 4-35 Referenc
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c r- L i] L h L __ f; t 4-37 trons
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Table 4.4-1. Fuel Utilization Chara
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L I 1 I Ld I ie i-- 1 t' 4-41 4.4.2
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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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Page 261 and 262: t a APPENDICES
Page 263 and 264: A-3 Appendix A. ISOTOPE SEPARATION
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Page 267 and 268: id A- 7 The Component Preparation L
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Page 289 and 290: Appendix C. DETAILED RESULTS FROM E
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Page 293 and 294: L c-5 The introduction of the class
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Page 303 and 304: Qulatiw klur W i t y hilt (*I thrmg
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Page 307 and 308: L L L L [i i” Table D -2. Maximum
Page 309 and 310: Fig. D-2 (cont.) 25W I I I I I (I)
Page 311 and 312: [i energy center. It can be seen fr
Page 313 and 314: L L i il L t W 0-9 Table D-4. Summa
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Page 317 and 318: It e L L L t u I' f ' ki Federal Ag
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