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j ~ 5-20 The R,D&D costs are highest for the HTGR and HWR, which are yet to be demonstrated on their reference cycles for the large unit size (1000-MWe) postulated in this report. The cost of these demonstration units constitutes the largest fraction of the total estimated R,D&D costs, although substantial costs will also be incurr'ed for large plant design and licensing, which includes component R&D, component design, and licensing and SAR development. the assumption that experience equivalent to that of the Fort St. Vrain HTGR prototype can be obtained from Canadian technology. intermediate between those of the LWR and those of the HTGR because of the heavy reliance of the SSCR on LWR technology. As has been discussed in the text, once these reactors have been demonstrated on their reference cycles, additional R,D&D will be required to convert them to DUTH fuels. The R,D&D requirements for the HTGR and HWR are judged to be similar under The SSCR is viewed as having R,D&D costs Section 5.1 References 1. "The Economics and Utilization of Thorium in Nuclear Power Reactors," Resource Planning Associates, Inc., January 16, 1968 (draft).
1 .. L c L i 5-21 5.2. FUEL RECYCLE RESEARCH AND DEVELOPMENT REQUIREMENTS I. Spiewak Oak Ridge National Laboratory The purpose of this section is to summarize the technical problems that must be ad- dressed by a fuel recycle research and development program before reactor systems producing and using denatured uranium-thorium (DUTH) fuels can be deployed commercially. Preliminary estimates of the schedule and costs for such a program are also included to provide some perspective on the comnitments that will be required with the introduction of reactors operating on denatured fuels. Wide ranges in the estimates reflect the current uncertainties in the program. However, detailed studies of the research and development requirements for the recycle of DUTH fuels are now being conducted by the DOE Nuclear Power Division's Advanced Fuel Cycle Evaluation Program (AFCEP), and when the results from these studies be- come available, the uncertainties in costs and schedules should be reduced. 5.2.1. Technology Status Summary The technological areas in a fuel recycle program cover fuel fabrication/refabrication (fuel material preparation, rod fabrication, element assembly); fuel qualification (irradiation performance testing and evaluation); fuel reprocessing (headend treatment, solvent extraction, product conversion, off-gas treatment) ; and waste treatment (concentration, cal- ci nation, vi tri f i ca ti on, and radioact i ve-gas treatment). Fuel Fabri cation/Refabri cati on and Qual if i cati on In general, the basic technology for the fabrication of uranium oxide pellet fuels is established, with the fabrication of both LWR and HWR uranium fuels being conducted on a commercial scale. In contrast, Pu/U oxide pellet fuels have been fabricated only on a small pilot-plant scale, and a significant amount of research and development is still required. Areas requiring further study include demonstration of: (1) a pelletizing process to ensure uniform product characteristics and performance; (2) methods for verifying and controlling the characteristics of the Pu/U fuels; (3) processes for the recovery of contaminated scrap; (4) a reliable nondestructive assay system for powders, fuel rods, and wastes; (5) the ability to operate a large-scale plant remotely, but with hands-on maintenance (in the case where Pu/U oxides containing high quality plutonium are being fabricated); and (6) satisfactory irradiation performance of Pu/U fuels produced in commercial-scale processes and equipment.
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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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-- - . & 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
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
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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.
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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,
Page 142 and 143: 5-8 the LMFBR on its reference cycl
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
Page 150 and 151: 5-1 6 and should also address metho
Page 152 and 153: I DATA BASE DEMu)pMENT DEMO DESIGN
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Page 158 and 159: 5-24 5.2.2. Research, Development,
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Page 164 and 165: 6-4 persed areas ensured - a fact w
Page 166 and 167: 6-6 6.1.2. Reactor Options The reac
Page 168 and 169: Reactor/Cycl ea LWR-U5(LE)/U-S LUR-
Page 170 and 171: 6-10 6.1.3. Nuclear Policy Options,
Page 172 and 173: Table 6.1-4. Nuclear Policy Options
Page 174 and 175: SWU SWU 6-14 UZJS USILEINS WYI U5IL
Page 176 and 177: 6-16 n nM MEDL nows.1 Option 4: In
Page 178 and 179: 6-18 HM HEDL7OOl.70.3 Option 6: In
Page 180 and 181: SRCIFIED CONSTRUCTION LlMllLD INTRO
Page 182 and 183: 50- 40 6-22 c Y \ VL - :-: F 30- -
Page 184 and 185: 6-24 The potential nuclear contribu
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Page 188 and 189: 141.6 ST - U308 4 7.2 lo3 swu ENRIC
Page 190 and 191: 6-30 reactor. Since this increase i
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Page 194 and 195: 6-34 1HE LWR WITH FISSILE UUNIUM PC
Page 196 and 197: THE LWR WITH PURONIUM MlNUllUTlON A
Page 198 and 199: 6-38 be located in energy centers a
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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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