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68 3, The mobility of the pod of noble-metal material suggests that de- posits occur as an accmulatfsn of finely divided well-mixed material rather than as a Suck precipitation within the reactor, though expected, is a dis- adVantag@. Precfpftataon OIE the metal surface (most Sf Which iS in the heat exchanger) will be quite insufficient to impede fuel flow, but radio- active decay SS the deposited material contributes to heat generation Bur- iRg ?X?actor SfiutdoWII. Precipitation OFI the mOdePatOlP graphite, Which ap- peared to be considerably smaller than on the metal, would maximize their opportunities to absorb valuable neutrons. Operation Of the MSaE did prOdUCe One WktOWZ3sd effect Sf fission 2% products, Metal surfaces exposed to the fuel in the MSRE showed grain boundaries that were embrittled to depths of 0-1 to O.? IILH~ (5 to 10 mils). In the heat exchanger, the embrittled boundaries opened to form metalls- graphically visi$%@ cracks; in other regfsns such cracks formed QES%~ when the specimens were deliberately strained. Early studies*0 implicated fission-product tellurim as responsible for this embrittlement, and sub- SeqUE?Ist WQPk has CQnfi9fmed thiS."6a4* However, mOPe Irecent S%Udfes46,81982 strongly suggest that (1) if the molten fuel is made to contain as much as 5% of the uranim 8s U F the ~ ~ te~~urim would be present as ~e2- and (2) in that fom, tellurium is much less aggressive. Much further study will be necessary, but use sf this higher but still moderate UF,/IJP, ratlo apparently will. markedly alleviate, and probably control, the tellurim embrittlemerat problem. 3.302.4 Operational ConSthaintS Avoiding the detrimental effects sf fission-produet tellurium (de- scribed immediately preceding) may wake necessary the operation of the DMSR with as mush as 5% of the uranium fluoride present as UF3" eration of possible reacti~ns of ~ T F to ~ produce uranium carbides (de- C~nsfd- scribed preaiousPy) suggests that operation with a considerably higher A This subject is addressed further in Sect, 3.4, .&!&.,.-
UP3/UF4 ratio would be possible. DMSR operation possibly may result. 69 knm.rer, a more subtle constraint on The lanthanide trifluorides are only moderately soluble in w~lten LiF-BeP2-ThP4-UF4 mixtures; if more than one such trifluoride is present, they crystallize as a solid solution of all the trifluorides on cooling of a saturated melt. Though not definitive, there is evidence that the actinide triflusrides (including W3> might also join in such solid solu- tions. If so, the total (lanthanide plus actinide) trifluorides in the end-of-life reactor might possibly exceed their combined solubility. The solubilities of PuFq (Ref. 29) and CeF3 (Ref. 31) have been cape- fully determined im %~F-B~F~-T~F~ (72-16-12 mole %> and em be eo~~ihfered to be moderately well known in DMSR fuel. According to Baes et al 929 the solubility of PuF3 in the LiF-BeF2-ThFq melt at 565°C (the minimum temperature anticipated within the BMSR fuel circuit) is 1.35 mole %. The solubility of CeP3 umder the same conditions appears to be very slightly smaller ( 1.3 mole %) .29 9 3 1 are not well known. trifluorides ( including CeF3 ) have been shown83 to be considerably smaller and to vary with some more and some less soluble than CeF3. As a reason- able approximation (~bviously, many additional data are needed), the solubility of the lanthanide-actinide trifluoride solid soluti~n may be assumed to be near 1.3 mole 2. Solubilities of the other pertinent fluorides In L12BeF4, the solubilities of several lanthanide From Table 9, the BMSR fuel at end-of-life will contain some 11.404 x 185 moles of uranium isotopes and 3.64 x PO3 moles of transuranium isotopes. The end-of-life inventory of lanthanide plus yttrium isotopes will be near 4.7 x 104 moles. If 5X of the uranium is present as U F and ~ if all transuranic and lanthanide species are assumed to be trivalent, the end-of-life reactor fuel will contain about 5.77 x 104 moles of trifluorides. fore, would contain about 1.1 mole % of trifluorides. The solubility of the combined trifluorides likely would not be exceeded within the reactor circuit, but additional solubility data are needed to make this point certain. 'Fke DMSR system (with about 5.3 x IO6 moles of fluorides), there-
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P iii CONTENTS ....................
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CONCEPTUAL DESIGN CHARACTERISTICS O
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3 The primary purpose of this study
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.. .... . w..w c Y ... . r 5 2. GEN
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ii.... :.: ..... .,. \ ..... %& . 3
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addition StathQR would be required.
Page 17 and 18:
!I \.... . , . ~ .... .. . . / I I
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... .. ...... . . . $i 13 Table 1.
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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
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: operation of the PISRE. 67 analyses
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
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123 Table 34. Summary of annual non
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...... . . . . ., x.:w* 4 i.. ... _
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E2 7 There are no detailed estimate
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129 particularly if such activities
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. ~.........* . . . . . :*. ... . -
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. .. . . . . . . . ;.: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
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. ... . .,.,.,.,.,. . , 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
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... . . . . . . . , . ,.Y&fl 149 ap
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Account No. -0- Three- digit digit
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.... ..
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156 77. He 6. MacPherson, Institute
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