ORNL-TM-7207 - the Molten Salt Energy Technologies Web Site

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20 Cell material Energy Core zone Inner salt Mod era tor Interstitial salt Fast A 1.14 0.96 1.14 Resonance A 0,97 1.01 0- 97 The mal A 0.94 1.03 0*8% Fast R 1.2% Q,98 1.12 Resonance w 0.97 1.00 0.98 The ma P R 0.88 1.01 0.93 ~ ~~ a. Average flux in material divided by average cell flux. imbedded sell calculation, an unusual code system clearly would be necessary to provide fd%y satisfying level of detail to this problem, Ob- viously, a true two-dimensional spatial treatment of the flattened core would be appropriate, but imbedding such a calculation inside a depletion loop is expensive. 3.1 4.1 Fueling policy 3.1.4 Once-through system considerations For purposes of nuclear calculations, the fueling policy for the once-through BMSR is as fohlows. E. Thorium is added to an initial loading of salt in a specified concentration. During operation, the concentration is allowed to decline via burnup. Near the end of plant life, small amounts are removed as requ-ired to keep the total actinide content below the startup value. 2. Uranium is added at the maximum allowable enrichment in the ZLIIIQU~~ necessary to maintain criticality.
h 3. Additional 238Y is added as required to maintain the denaturing inequality * 4. Removal of certain fission products Ps accomplished according to Table 8. 21 Table 8. Removal times for fission products in once-through cycles Fission-product group Element Removal time Noble gases Kr, Xe 50 s Seminoble and Zn, Ga, Ge, As, Nb, 2.4 h noble metals Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sn, Sb 3.1.4.2 Fission-product buildup A study of 30-year fission-product buildup was made as a f ~rtcti~n of various continuous removal rates for those products not listed in Table 8. The reactivity effect may be satisfactorily represented by where p = fission-product reactivity effect (%>, t = time (year), Y = yield (0.93%/year), X = burnout rate (6.8 year)-', R = removal rate (year-1 >. "Other nonfissile uranium wuclides further dilute the 233U.; dilu- tion to 12% 23% may require additional 23%.
Page 1: i *w . . I
Page 5 and 6: P iii CONTENTS ....................
Page 7 and 8: CONCEPTUAL DESIGN CHARACTERISTICS O
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: :.:..y ,. . . w, E9 The code calcul
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
Page 69 and 70: 63 n e results of a program of solu
Page 71 and 72: 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
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73 Contamination of fuel by seconda
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*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
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..... 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,
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. . . :.w . . . . , 4. ALTERNATIVE
Page 105 and 106:
.. . ... . . . . . . .. . . "W 3. T
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.,.. w .." by removing some uranium
Page 109 and 110:
, . . . . . . , -c%w probably still
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and the ~ dvent (or carrier) salt a
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BO 7 TPle fuel makeup requirements
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5. CQPIPIERCIWLPZATION CONSIDERATIO
Page 117 and 118:
11% At the close of MSRE operation,
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12YS 75 22”” 955 30” 1””
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115
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aa 7 some components in common with
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Ae eo un t No * Table 3%. Capital c
Page 127 and 128:
- ,.,.,.,., p .. ..... . G is based
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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
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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
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139 The authors want to acknowledge
Page 147 and 148:
. ... . .,.,.,.,.,. . , i.rc- q....
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.:. _.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
Page 155:
... . . . . . . . , . ,.Y&fl 149 ap
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Account No. -0- Three- digit digit
Page 160 and 161:
.... ..
Page 162:
156 77. He 6. MacPherson, Institute
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