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Zni c( Y w E: a 9 B 7-1 8 ORNL-DWG 7R-19Anl pc (a) SEGMENT AB: TRANSMUTER x.A + (l-X)*B SEGMENT BC: TRANSMUTER = X-B + (l-X).C SEGMENT AC: TRANSMUTER X-A + (l-X)*C 0 5 x 5 1 c LMFBR Transrnuters + 12% Denatured LMFBRs .0 0.5 I B ENERGY SUPPORT RATIO REACTOR A: (PU/U)02 + U02 RB C REACTOR 8: (Pu/U)02 + Tho2 RB REACTOR C* fPu/Th)02 + Tho2 RB ORNL-DWG 78-19809 SEGMENT AB: TRANSMUTER = X-A + SEGMENT BC: TRANSMUTER = X*B + A SEGMENT AC: TRANSMUTER = X-A + 05x51 + LMFBR Transrnuters ENERGY SUPPORT RATIO REACTOR A: (Pu/U)02 + U02 RB C REACTOR B: (Fu/U)02 + Tho2 RB REACTOR C: (Pu/lh)02 + Tho2 RB Fig. 7.2-1. Operating Envelopes for Symbiotic Systems Utilizing LMFBR Transmuters. 5 4.S I; L c I: L tl Lz L t:
li i i L Li L & I ' i kr ii L i h i h I' ~~ 7-1 9 7.2.4. Conclusions Since optimization of the various reactors for the particular fuel cycle considered was beyond the scope of this study, the results presented above are subject to several uncer- tainties. Nevertheless, certain general conclusions on the impact of the various fuel cycles on reactor performance are believed to be valid: For once-through throwaway systems, the various systems studied are ranked in order of resource utilization as follows: the HWR on the LEU cycle; the HTGR and HTR-PBR on either the LEU cycle or on the MEU/Th cycle; and the SSCR and PWR on the LEU cycle. On the MEU/Th cycle the SSCR and PWR require more uranium than they do on the LEU cycle and hence do not merit further consideration for once-through operation. 0 For once-through stowaway systems, in which the fissile material in the spent fuel is expected to be recovered at some future date, the relative ranking of the systems would depend on the ultimate destination of the fissile material. If future nuclear power systems are to be thermal recycle systems, then early emphasis should be placed on reactors and fuel cycles that have a high 233U discharge. If the future systems are to be fast recycle systems, then emphasis should be placed on reactors and fuel cycles that will provide a plutonium inventory. 0 For recycle systems utilizing thermal reactors, the preferred basic fissile material is 233U. However, implementation of a 233U fuel cycle will require an exogenous source of the fissile material; therefore, it is likely that the MEU(235)/Th cycle would be implemented first to initiate the production of 233U. Both the unburned 235U and the 233U would be recycled; thus the system would evolve towards the MEU(233)/Th cycle, which is the denatured 233U cycle as defined in this study. However, it is to be emphasized that these reactors will not produce enough 233U to sustain themselves and separate 233U production facilities must be operated. A Pu/Th-fueled thermal reactor has been considered as a 233U production facility. 0 For recycle systems utilizing fast reactors, the preferred basic fissile material is 239Pu. Using 233U as the primary fissile material or placing thorium in the core sharply reduces the breeding performance of fast reactors. However, fast reactors using plutonium fuel and thorium blankets would be efficient 23 3U production facil i ties.
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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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-- 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
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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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Page 184 and 185: 6-24 The potential nuclear contribu
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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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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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