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CONCEPTUAL DESIGN CHARACTERISTICS OF A DENATURED MOLTEN-SALT REACTOR WITH ONCE-THKOUGB FUELING J. 8. Fngel W. R. Grimes H. F. Rauman PI. E. McCoy J. Re Dearing V. A. Rhoades ABSTRACT A study was wade to examine the conceptual feasibility of a molten-salt power reactor fueled with denatured 235U and operated with a minimum of chemical processing. Because suck a reactor would not have a positive breeding gain, reductions in the fuel conversion ratio were allowed in the design to achieve other potentially favorable characteris- ties for the reactor. A conceptual core design was developed in which the power density was low enough to allow a 30-year life expectancy of the moderator graphite with a fluence limit of 3 x 1026 neutrons/rn* (E > 50 kev). mis reactor could be made critical with about 3450 kg of 20% enriched 235U and operated for 30 years with routine additions of denatured 235U and no chemical processing for removal of fission products. The lifetime requirement of natural UjOg for this once-through fuel cycle would be about 1810 Mg (-2000 short tons) for a 1-GWe plant operated at a 75% capacity factor. If the uranium in the fuel at the end of life were recovered (3160 kg fissile uranium at -10% enrichment), the U3Og requirement could be further re- duced by nearly a factor o€ 2. The lifetime net plutonium pro- duction for this fuel cycle would be only 736 kg for all iso- topes (238, 239, 240, 241, and 242). A review of the chemical considerations associated with the conceptual fuel cycle indicates that no substantial dlfficulties would be expected if the soluble fission products and higher ac- tinides were allowed to remain in the fuel salt for the life sf the plant. Some salt treatment to counteract oxide eontamina- tion and to maintain the oxidation potential of the melt probably would be necessary, but these would require only well-known and demonstrated te~hnobogy. Although substantial technology development would be required, the denatured molten-salt reactor concept apparently could be made commercial in about 30 years; if the costs of intermediate developmental reactors are included, the cost for development is estimated to be $3750 million (1978 dollars). The resulting system would be approximately economically competitive with current-technology light-water reactor systems.
Page 1: i *w . . I
Page 5: P iii CONTENTS ....................
Page 9 and 10: 3 The primary purpose of this study
Page 11 and 12: .. .... . w..w c Y ... . r 5 2. GEN
Page 13 and 14: ii.... :.: ..... .,. \ ..... %& . 3
Page 15 and 16: addition StathQR would be required.
Page 17 and 18: !I \.... . , . ~ .... .. . . / I I
Page 19 and 20: ... .. ...... . . . $i 13 Table 1.
Page 21 and 22: 15 Initial scoping studies showed t
Page 23 and 24: .. . E 6> I S N 0.45 k- 0.50 B --.
Page 25 and 26: :.:..y ,. . . w, E9 The code calcul
Page 27 and 28: h 3. Additional 238Y is added as re
Page 29 and 30: ...z,y .. . c .A, Table 9. Actinide
Page 31 and 32: ';x.w' .... Relative power distribu
Page 33 and 34: ........ . . . :..i 27 a Inner 2.54
Page 35 and 36: 29 Table 16. Effect of neutron spec
Page 37 and 38: ob E 2 a 4 5 6 7 8 9 10 11 12 13 14
Page 39 and 40: .. . . . . . . . . . 'a# 0.1 0.2 0.
Page 41 and 42: 35 (Doppler effect), (2) changing t
Page 43 and 44: where Ci = relative delayed-neutron
Page 45 and 46: .... . . . . , c.y..p c "
Page 47 and 48: .... :...w . . . . . :, %. , dimens
Page 49 and 50: ,. ::.y ~. . .... , 40 0 43 1 2 3 4
Page 51 and 52: .,.,. . . ., , .. . . . . . . . %..
Page 53 and 54: .... ... . ....... . .%.W 3.3.1.1 C
Page 55 and 56: 49 Variation of fuel composition wi
Page 57 and 58:
.. .*,p. . . . . . . . Table 25. Ph
Page 59 and 60:
;: c.... x. . w ..i 53 Table 26, St
Page 61 and 62:
would be expected to have an equili
Page 63 and 64:
.... . . . . :i ,.*,&s 57 Corrosion
Page 65 and 66:
. ;;..,..I . . . . . . "LW 0 9 m -
Page 67 and 68:
of the fissioned atom. ea ~arly ass
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63 n e results of a program of solu
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65 Table 27. Sources and rates Sf p
Page 73 and 74:
operation of the PISRE. 67 analyses
Page 75 and 76:
UP3/UF4 ratio would be possible. DM
Page 77 and 78:
$ .p9 3.3.3.1 Preparation of initia
Page 79 and 80:
73 Contamination of fuel by seconda
Page 81 and 82:
*w- through the walls of the primar
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77 4 will apparently be necessary (
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such contamination. For a demonstra
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8% 3.4 Reactor Materials Although s
Page 89 and 90:
: ....., . .’ -. ..WY used in the
Page 91 and 92:
..... i.. ....., . .Yd 85 QRPIL-DWG
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... w . . .y and to salt containing
Page 95 and 96:
With the different ~Bjectives of ns
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t.;$- ... quite different from thos
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93 during noma1 operation. Minor am
Page 101 and 102:
............. . . .. . .=w as SSm,
Page 103 and 104:
. . . :.w . . . . , 4. ALTERNATIVE
Page 105 and 106:
.. . ... . . . . . . .. . . "W 3. T
Page 107 and 108:
.,.. w .." by removing some uranium
Page 109 and 110:
, . . . . . . , -c%w probably still
Page 111 and 112:
and the ~ dvent (or carrier) salt a
Page 113 and 114:
BO 7 TPle fuel makeup requirements
Page 115 and 116:
5. CQPIPIERCIWLPZATION CONSIDERATIO
Page 117 and 118:
11% At the close of MSRE operation,
Page 119 and 120:
12YS 75 22”” 955 30” 1””
Page 121 and 122:
115
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aa 7 some components in common with
Page 125 and 126:
Ae eo un t No * Table 3%. Capital c
Page 127 and 128:
- ,.,.,.,., p .. ..... . G is based
Page 129 and 130:
123 Table 34. Summary of annual non
Page 131 and 132:
...... . . . . ., x.:w* 4 i.. ... _
Page 133 and 134:
E2 7 There are no detailed estimate
Page 135 and 136:
129 particularly if such activities
Page 137 and 138:
. ~.........* . . . . . :*. ... . -
Page 139 and 140:
. .. . . . . . . . ;.:w b 133 the i
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_... ..&/ ; .; with reactor scram m
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137 includlng $900 million for the
Page 145 and 146:
139 The authors want to acknowledge
Page 147 and 148:
. ... . .,.,.,.,.,. . , i.rc- q....
Page 149 and 150:
.:. _.i .. ..... -.=* 45. A. L. Mat
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...... &;.& 3 14 5 74. W, R. Grimes
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p 103. 6. Re Hudson 11, CQNCEPT-5 U
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... . . . . . . . , . ,.Y&fl 149 ap
Page 158 and 159:
Account No. -0- Three- digit digit
Page 160 and 161:
.... ..
Page 162:
156 77. He 6. MacPherson, Institute
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