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

More documents

Recommendations

Info

60 3.3.1.4 Operational c~nstraints and uncertainties Most of the fuel behavior described or implied above may be considered well authenticated, Several constraints and at least minor uncertainties are obvious. 1. Fuel for a DMSR must be prepared from LiF containing a very high 2. The fuel mixture must be managed and maintained so that an appre- ciable fraction of the uranium is present as UP3. 3. Additional experiments are necessary t o establish exactly what fraction of the uranium may be present as UFcj without deleterious chemical reactions of the UP3 with graphite or possibly with other materials within the primary reactor system. 4. Direct measurement of the physical and heat-transfer properties of the BMSR fuel mixture must be made. 5. Further study of the fundamental thermodynamic properties of so- lutes in the LiF-BeF2-ThF4-UF4 mixture are needed to ensure that basic understanding of the chemical behavior is accurate. 3.3.2 Fission-product behavior Fragments produced on fission of a heavy atom originate in energy states and with i ~lsizati~n levels far from those normally considered in chemical reactions. When the fission occurs in a well-mixed mo%ten-salt liquid medium, these fragments must come to a steady state as comonHy encountered chemical entities because they quicldy lose energy through eolLisions with the medium. The valence states that these chemical species assume are presumably defined by the requirements that (1) cation- anion equivalence be maintained in the molten-salt medium and (2) redox e~uilibria be established between the melt and the surface layers of the container metals 19 9 63 9 64 he fission-product cations must satisfy the fission-product anions plus the fluoride ions released by disappearance * The rapid radioactive decay of many species further complicates an already complex situation. *
of the fissioned atom. ea ~arly assessment" indicated that the cations would prove adequate only if some of them assumed oxidation states cor- rosive to H%%telPQy-N. A FBIOPB %€!Cent eXEU'ilin21tion25 strongly SUppOrted this view; these studies indicated that the summation of the products of fission yield and stable valence for each species wight be as low as three per fission event. Accordingly, fission of UP4 [releasing 4 I?-+ 0.815 (~r- + 1-1 per fission] would^ be intrinsica~~y oxidizing to ~astelloy-~. Maintenance of a small fraction of the uranium in the fuel as UF3 was successfully adopted to preclude corrosion from fission of 2 3 5 ~ ~ in 4 the MSWE.~~ A properly maintained redox potential in the fuel salt apparently will prevent any untoward immediate co~seq~ences sf the fission event and will pernit grow-in of the fission products in valence states defined by the redox potential. 3.3.2.2 Effects of radiation When fission occurs in a molten fluoride solution, both eleetro- magnetic radfatio~~ and particles of very high energy arnd intensity originate within the fluid. Local overheating is almost certainly not important in a DMSR where turbulent flow causes rapid intimate mixing. Moreover, the bonding in molten fluorides is completely ionic. Such a mixture, with neither covabernt bonds to rupture nor a lattice to disruptp should be quite resfstant to radiation. Nevertheless, because there pbau- sibly exists a radiation level sufficiently high to dissociate a molten fluoride into metal and fluorine,+ a number of tests of the possibility were wade. 1 9,2 0 9 6 3 9 65 Nany irradiation tests were conducted prior to 1959 with NaF-ZrP4-UP4 mixtures in Ineonel at temperatures at or above 815OC (Refs. 24 and 25) and at quite high fission power densities from 80 to 1080 ~ ~ / of m 3 fuel. No instability of the fuel system was apparent, and the corrosion did not exceed the considerable anmount expected from laboratory-scale tests. "Fission sf PuF3 probably would be nearly neutral in this regardp and fission of a mixture of UF, and PuF3$ as in the DMSR, would be inter- mediate between these extremes. ?This would occur even though the rate of recombination of LiO and Fo should be extremely rapid. *
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 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: . ;;..,..I . . . . . . "LW 0 9 m -
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
Page 79 and 80: 73 Contamination of fuel by seconda
Page 81 and 82: *w- through the walls of the primar
Page 83 and 84: 77 4 will apparently be necessary (
Page 85 and 86: such contamination. For a demonstra
Page 87 and 88: 8% 3.4 Reactor Materials Although s
Page 89 and 90: : ....., . .’ -. ..WY used in the
Page 91 and 92: ..... i.. ....., . .Yd 85 QRPIL-DWG
Page 93 and 94: ... w . . .y and to salt containing
Page 95 and 96: With the different ~Bjectives of ns
Page 97 and 98: t.;$- ... quite different from thos
Page 99 and 100: 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
Page 123 and 124:
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
Page 141 and 142:
_... ..&/ ; .; with reactor scram m
Page 143 and 144:
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
Page 151 and 152:
...... &;.& 3 14 5 74. W, R. Grimes
Page 153 and 154:
p 103. 6. Re Hudson 11, CQNCEPT-5 U
Page 155:
... . . . . . . . , . ,.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
show all

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

Create successful ePaper yourself

Delete template?

Save as template?