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

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was estimated to be the fluence at which the graphite structure would contain sufficient cracks to be permeable to xenone Experience has shown that, even at volme changes of about lox, the graphite is not cracked brat is unifsmly dilated. ability will not be of as much eoncern, and the limit probably will be established by the ~ OI-~TE~~~QIK of cracks sufficiently large for salt intrusion. me GLCC" K-364 graphite likely could be used to 3 x 10'~ neutro~s/m2, and improved graphites with a limit of 4 x 10'6 neutrons/m2 could be developed. For nonproliferating devices, xenon perme- The speeific performance requirements for graphite suitable for the reactor design presented in this report are a lifetime fluence capability of 2.7 x BOz6 neutrons/rn2 (E > 50 keV) at a peak temperature of ?'50Q@, Most probably, existing cmmereial graphites will satisfy th%s meed. 3.4,2.3 Uncertainties Although existing commercial graphites likely will meet the needs of the presemt design, graphite samples having the same cross section as the referenee-design moderator elements need to be irradiated. These tests need to be kt%n t o the d€?stPUction Of the graphite t o detePnaiRe the point at which the graphite actually heals. the present concept. Physical properties, particularly thermal esndue- tivity, need to be measured as a function of fluenee. This will define failure in A %onger-range effort to develop improved graphite for future re- actors Should be i'ilitiatecf. Early efforts Show prOlEise that graphitf2.s with improved dimensional stability can be developed, 3.5 Safety Considerations The main feature of the DMSR which sets it apart from solid-fuel reactor types is that the nuclear fuel is in fluid form (molten fluoride salt) and is circulated throughout the primary coolant system, becoming critical only in the graphite-moderated core. Possible problems and en- gineered safety features associated with this type of reactor will be * Great Lakes Carbon Company.
t.;$- ... quite different from those of the present EWR and liquid-metal fast 91 breeder reactor QLMFBR) designs. A detailed safety analysis of the DMSK must await the results of a research and development CUD,> program; however, identifying possible generic problem areas and some of the advantages and disadvantages of this concept is already possible. In the DMSW, the primary system fluid serves the dual role of bein the medim in which heat is generated within the reactor core and the me- dium that transfers heat from the core to the primary heat exchangers. Thus, the entire primary system will be subject to both high temperatures (900°C at the core exit) and high levels of radiation by a fluid containing most of the daughter products of the fission process. Because of the low fuel-salt vapor pressures however, the primary system design pressure W i l l b@ IOW, -3s in an LMPBRe %XI t@KUIS Of level Of ConfinelTlent, the entire reactor primary systen is analogous to the fuel claddi in a s~%id-fuel reactor. Although much larger, it will not be subject to the rapid thermal transients (with melting) associated with LWR and LMFBR accident see- narios. Two additional levels of confinement will be provided in the BMSW, in accord with present p~aetice, Note that the once-through DMSR concept has safety advantages QV~P the break-even DMSR because a large and complex part of the primary containment - the chemical reprocessing plant - is substantially reduced and because less radioactive material is routinely removed from containment. The problem of developing a reactor primary system that will be reliable, maintainable (under remote condi- tions), inspectable, and structurally sound over the plant's SO-year lifetime will probably be the key factor in demonstrating ultimate safety and Iicensabil fty . The breach of the reactor primary system b~~daryv, resulting in a spill of highly radioactive salt into the primary containment, will probably provide the design-basis accident, The analogous event in a solid- fuel reactor would be major cladding failure. Possible initiators o% this accidemt include pipe failure, missiles, and pressure or temperature transients in the primary salt system. Failure of the boundary between the prhary and secondary salt in the primary heat exchangers could be espe- cially damaging. In the event of salt spill, a possibly redundant system of drains would be activated to channel the salt t o the continuously
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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.
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!I \.... . , . ~ .... .. . . / I I
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... .. ...... . . . $i 13 Table 1.
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15 Initial scoping studies showed t
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.. . E 6> I S N 0.45 k- 0.50 B --.
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:.:..y ,. . . w, E9 The code calcul
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h 3. Additional 238Y is added as re
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...z,y .. . c .A, Table 9. Actinide
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';x.w' .... Relative power distribu
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........ . . . :..i 27 a Inner 2.54
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29 Table 16. Effect of neutron spec
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ob E 2 a 4 5 6 7 8 9 10 11 12 13 14
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.. . . . . . . . . . 'a# 0.1 0.2 0.
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35 (Doppler effect), (2) changing t
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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
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: With the different ~Bjectives of ns
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
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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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