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

More documents

Recommendations

Info

3,3,2.5 Uncertain features Prom the foregoing discussion, several uncertainties are apparent. Details of behavior of the noble and seminoble fission products are still poorly known. The fractions of each isotope that will appear in the off- gas, deposit on the moderatst graphite, and deposit OR the heat-exchanger metal u% the I>r/n§R can only be crudely estimated. That fraction which appears in the reactor off-gas would seem to cause no insurmountable prsb- lems though that system- at our present state of knowledge -would need to be overdesigned substantially. Their deposition in the heat exchanger is a recognized disadvantage and will be quite insufficient to impede fuel flow, but radioactive decay of the deposited material contributes heat that must be removed during reactor shutdown. Precipztation on the graphite, which appears to be smaller than on heat-exchanger metal, maxi- mizes their opportunities to absorb neutrons. Clearly, a better knowledge of this situation is weeded, While the results obtained in the recent past are highly encouraging, additional data - especially on a larger scale - are needed to establish the redox potential (UF~/UP~ ratio) required to keep the tebburium crack- ing problem to tolerable levels. substantially above 0.85, the probability of precipitation of triflusride solid solutions would be increased. Should the required UF3/UF4 ratio rise Finally, additional infomation abseat the collective solubility behavior of the lanthanide-actinide trffluorides is required. Should they prove appreciably less soluble than wow believed likely, some replacement of fuel might be required late in the life of the essentially unprocessed DMSR e 3.3.3 Fuel. maintenance To achieve fuel maintenance, (1) the fuel must be delivered to and into the reactor in a proper state of purity and homogeneity, (2) the fuel must be sufficiently protected from extraneous impurities, and (3) sound procedures must exist for addition of the required uranium and provision of the required BTF3/UF, ratio,
$ .p9 3.3.3.1 Preparation of initial fuel problems. Initial purification procedures for the DNSR present no formidable Nuclear poisoms (e.g., boron, cadmium, or lanthanides) are not common contaminants of the constituent raw materials. All the pertinent cmpsunds contain at least small amounts of water, and all are readily hyarsliyzea to oxides ana oxyfluorides at elevated temperatures. he COW- pounds LiF and BeF2 generally contain a small quantity of sulfur as SUI- fate ion. Uranium tetrafluoride commonly contains small amonts of U0zp UFg, and U82F2. purification prseeaures63 ,*4 %85 uses to prepare materials for the aircraft reactor experiment (ARE), the MSRE, and many laboratory and en- gineering experiments have treated the mixed materials at. high tempera- ture (usually at 600°C) with gaseous H~-HP mixtures and then with pure H2 in equipment of nickel or copper. me HF-HZ treatment serves to (1) reduce the 6T5* and U6+ to u4+, (2) reduce SU%fat@ ts sulfide and BI€3SlOVe it as H~S, (3) remove CP- as HCP, and ( 4) ~snvert the oxides and oxyfluorides to fluorides. Final treatment with H2 serves to reduce PeF3 and FeF2 to insoluble iron and to remove NiF2 that may have been produced during hy- drofluorination, To date, all preparations have been performed in batch equipment but eontiwuous equipment has been partially developed. 86 8 87 For a DMSR, as for the MSRE,*% purificatfsn of the bulk of the fuel would presumably be conducted on L~F-B~F~-T~FL+-UP~ mixtures containing perhaps 85 to 90% sf the required WL, and on molten Li3UF7 to provide the additional uranium necessary to bring the fuel to the critical and oper- ating Concentration. Such a purification procedure can provide a sufficiently pure and eompletely homogeneous fuel material for initial operation of the reactor. 3.3.3.2 Contamination possibilities Though the fuel material can be supplied and introduced into the reactor in sufficiently pure fom, contamination of the fuel is possible ESOre SeQePal SOUlllC@S e Other reactor materials. a%e moderator graphite can contain a large quantity of 602, C6, and H28 by virtue of its porosity and internal
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:
... .. ...... . . . $i 13 Table 1.
Page 21 and 22:
15 Initial scoping studies showed t
Page 23 and 24:
.. . E 6> I S N 0.45 k- 0.50 B --.
Page 25 and 26: :.:..y ,. . . w, E9 The code calcul
Page 27 and 28: h 3. Additional 238Y is added as re
Page 29 and 30: ...z,y .. . c .A, Table 9. Actinide
Page 31 and 32: ';x.w' .... Relative power distribu
Page 33 and 34: ........ . . . :..i 27 a Inner 2.54
Page 35 and 36: 29 Table 16. Effect of neutron spec
Page 37 and 38: ob E 2 a 4 5 6 7 8 9 10 11 12 13 14
Page 39 and 40: .. . . . . . . . . . 'a# 0.1 0.2 0.
Page 41 and 42: 35 (Doppler effect), (2) changing t
Page 43 and 44: where Ci = relative delayed-neutron
Page 45 and 46: .... . . . . , c.y..p c "
Page 47 and 48: .... :...w . . . . . :, %. , dimens
Page 49 and 50: ,. ::.y ~. . .... , 40 0 43 1 2 3 4
Page 51 and 52: .,.,. . . ., , .. . . . . . . . %..
Page 53 and 54: .... ... . ....... . .%.W 3.3.1.1 C
Page 55 and 56: 49 Variation of fuel composition wi
Page 57 and 58: .. .*,p. . . . . . . . Table 25. Ph
Page 59 and 60: ;: c.... x. . w ..i 53 Table 26, St
Page 61 and 62: would be expected to have an equili
Page 63 and 64: .... . . . . :i ,.*,&s 57 Corrosion
Page 65 and 66: . ;;..,..I . . . . . . "LW 0 9 m -
Page 67 and 68: of the fissioned atom. ea ~arly ass
Page 69 and 70: 63 n e results of a program of solu
Page 71 and 72: 65 Table 27. Sources and rates Sf p
Page 73 and 74: operation of the PISRE. 67 analyses
Page 75: UP3/UF4 ratio would be possible. DM
Page 79 and 80: 73 Contamination of fuel by seconda
Page 81 and 82: *w- through the walls of the primar
Page 83 and 84: 77 4 will apparently be necessary (
Page 85 and 86: such contamination. For a demonstra
Page 87 and 88: 8% 3.4 Reactor Materials Although s
Page 89 and 90: : ....., . .’ -. ..WY used in the
Page 91 and 92: ..... i.. ....., . .Yd 85 QRPIL-DWG
Page 93 and 94: ... w . . .y and to salt containing
Page 95 and 96: With the different ~Bjectives of ns
Page 97 and 98: t.;$- ... 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

You also want an ePaper? Increase the reach of your titles

Delete template?

Save as template?