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

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110 5.1.1 current status MSW development has been carried through the design and operation of a proof-of-principle test reactor, the MSRE, which was an 8-Wt reactor that operated at ORNL from 1965 to 1969. This reactor demonstrated the basic reliability of a molten-salt system, stability of the fuel salt, compatibility sf fluoride salts with Ustelloy-N and graphite, reliability of molten-salt pumps and heat exchangers, and maintenance of a radloactive flufd-fueled system by remote methods. The reactor was crit- ical over 17,000 h, circulated fuel salt for nearly 22,000 h, and gener- ated over 100,000 plwh of thermal ene~gy. The MSRE had achieved all the objectives of the reactor test program when it was retired in 1969. After the successful operatfon of the MSRE, the reactor concept appeared ready for commercial development, In preparation for further de- ve%opment, three major reports were prepared: a conceptual design study of an MSBR in 1971 (Ref. $I9 a review of the status of development in 1972 (Ref. loa), and a program plan for development in 1974 (Ref. 21). For reasons other than technological, the government decided not to fund further development of MSRs. T%e program was cancelled in 1973, restarted in 1974, and finally terminated in $976, The development of a proli%eratfsn-resistant DKSR would require basi- cally the same technological development program as was proposed for the MSBR, but the emphasis would be on reliability, ease of c~m~~e~~iali~ati~n, Picensing, and proliferathn resistance rather than on high breeding per- formance. Kith these objectives in mind, the 1972 status-of-devebopment report has been updated, and the program plan for development has been moc~il~iec~ for the DMSR."~ (mile the main outline of BMSR development requirements will be presented i~ this report, the reader I s referred to Ref. 102 for greater detail.) 5.1.2 Technology base for reference DMSR The base technology for MSWs is well established and has been largely "'proven in princ€p%e" by the operation of the 24SWE. While no major un- ~esolved technical issues exist at the present time, a large R&B effort would be required to bring nnol.ten-salt technology to commercialization.
11% At the close of MSRE operation, two major technical issues appeared unresolved. The first was the control of tritium, which is produced in fairly large quantities in a molten-salt system and which is known to dif- fuse through metal walls. Subsequent engineering-scale tests have demonstrated that tritiUll is oxidized in SOd~Wil flUOrobOPate, the pPQpos@d SeC- ondary salt for the DMSR, and appears to be handled readily. However, this process is not yet well understood, and the effects of maintaining an ade- quate c~ncentrati~n of the oxidant on the long-term compatibility of the salt with the structural alloy are unknown. The second issue involved the compatibility of Hastelloy-N with fuel salt. Operation of the MSRE showed that the general corrosion of Hastelloy-M and graphite in an operating MSR was near zero, as expected. However, metal surfaces that had been exposed to fuel salt containing fission products were unexpectedly found to exhibit grain-boundary attack, which was subsequently shown to be caused by reac- tion with the fission product, tellurim. Further work has shorn that tel- lurium attack can be controlled by either a modification sf the Hastelloy-N alloy or by eo~ltroP of the oxidation potential of the fuel salt. The major areas of research required for comerciabization of MSWs would involve improvement of the materials sf construction (Hastelloy-PS and graphite), the design of in-line instrumentation for high-temperature use, and the development of fuel processing (at least for the end of reactor life and possibly also for use on-line). The major areas of development involve the scale-up of reactor components (eogc9 pumps) and the design and development of components that were not present in the MSW (e.g., steam generators and mechanical valves). pate that the design of some components such as the fuel drain system and the reaetor cell with its insulation, heating, and cooling requirements would be extensively modifed to meet currently unspecified licensing requirements. the temperatures and flows in the primary and secondary salt systems and in the steam system to avoid salt freezing and excessive thermal stress. Alternatively, some co~~~ponents might be designed to accommodate such freezing. Still another area of development would be advanced remote maintenance techniques, including the replacement of components using remote pipe cutting and welding. In addition, we a~tki- Another large area of development would be the control sf
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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 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
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.... . . . . , c.y..p c "
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.... :...w . . . . . :, %. , dimens
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,. ::.y ~. . .... , 40 0 43 1 2 3 4
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.,.,. . . ., , .. . . . . . . . %..
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.... ... . ....... . .%.W 3.3.1.1 C
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49 Variation of fuel composition wi
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.. .*,p. . . . . . . . Table 25. Ph
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;: c.... x. . w ..i 53 Table 26, St
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would be expected to have an equili
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.... . . . . :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 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: 5. CQPIPIERCIWLPZATION CONSIDERATIO
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
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