ORNL-TM-7207 - the Molten Salt Energy Technologies Web Site
ORNL-TM-7207 - the Molten Salt Energy Technologies Web Site ORNL-TM-7207 - the Molten Salt Energy Technologies Web Site
24 Table 10. Nuclide concentrations and neutron utilization after 15 years sf DNSR operation Nus 1 ide 2 56% 1.13 49.0 9.21 25-1 16.2 1.83 474 4.34 2.46 1.84 2.38 e Transplutoniram 8Pu 0.882 Total actinfdes Fluorine Lithium Beryl Piurn Graphi t e Fission products a Concentration Neutron Fission vcrf/aa (x 1024) absorption' fraction 48,000 24,500 5 440 92 9 270 0,2561 0.0018 0 s 2483 8.0120 0,1161 0.0075 0.0047 0.0901 0.0896 0.0324 0.0293 0.0039 0.0014 0.0024 0.8956 0,0079 0 0 0062 0.0012 0.9109 0.0172 0 e 0563 Total 0.9844 Nuclei per cubic meter of salt or moderator. h AbsorptTon per neutron horn; leakage is 0.0156. 0,0017 0*0000 0.5480 0 e 0002 0.2292 0.0001 0 s 0000 0.0017 0.31578 0.0001 0.0628 0,0001 0.9803 a e 0000 3.1.5.2 Flux and power distributions and graphite lifetime ~ ~~ 0.0090 0.0033 2.2427 0,0143 l e 9894 0.0168 0.0182 0.0194 310 7905 0.0032 2. €754 0.0136 0. E245 e he relative fast flux (E > 52.4 ke~) and power-peaking factors are given in Table 11. These factors include the effects of flattening. For comparison, the overall fast flux peaking in an unflattened core would be -2.3; the neutron leakage, however, wsbald be only 0.8% vs 1.56% for this core c
';x.w' .... Relative power distributions (Fig. 4) show no serfous problems. The ... . . . . . . . . w:w . 25 peak occurs in the well-cooled inner zone. A power peak per unit O€ core volume occurs in the gap between the core and the reflector, but the power per unit volume of salt is actually relatively low in that region. CORE A (20% SALT) Table El. Neutron flux and power-peaking factors Fast flux Power Radial 1.32 1.36 Axial 1.15 1.15 Overall 1.52 1.56 CORE B (12 9% SALT) OHNL UWG80 4265 FT3 AXIAL RACIAL- (50% SALT) 05 1 15 2 45 3 35 4 45 DISTANCE FROM CENTER OF CORE (mi Fig. 4. DMSR relative power-density distribution. Axial and radial profiles ate separately and arbitrarily norwa%ized.
- 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: ...z,y .. . c .A, Table 9. Actinide
- 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 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
24<br />
Table 10. Nuclide concentrations and neutron utilization<br />
after 15 years sf DNSR operation<br />
Nus 1 ide<br />
2 56%<br />
1.13<br />
49.0<br />
9.21<br />
25-1<br />
16.2<br />
1.83<br />
474<br />
4.34<br />
2.46<br />
1.84<br />
2.38<br />
e<br />
Transplutoniram<br />
8Pu 0.882<br />
Total actinfdes<br />
Fluorine<br />
Lithium<br />
Beryl Piurn<br />
Graphi t e<br />
Fission products<br />
a<br />
Concentration Neutron Fission vcrf/aa<br />
(x 1024) absorption' fraction<br />
48,000<br />
24,500<br />
5 440<br />
92 9 270<br />
0,2561<br />
0.0018<br />
0 s 2483<br />
8.0120<br />
0,1161<br />
0.0075<br />
0.0047<br />
0.0901<br />
0.0896<br />
0.0324<br />
0.0293<br />
0.0039<br />
0.0014<br />
0.0024<br />
0.8956<br />
0,0079<br />
0 0 0062<br />
0.0012<br />
0.9109<br />
0.0172<br />
0 e 0563<br />
Total 0.9844<br />
Nuclei per cubic meter of salt or moderator.<br />
h<br />
AbsorptTon per neutron horn; leakage is 0.0156.<br />
0,0017<br />
0*0000<br />
0.5480<br />
0 e 0002<br />
0.2292<br />
0.0001<br />
0 s 0000<br />
0.0017<br />
0.31578<br />
0.0001<br />
0.0628<br />
0,0001<br />
0.9803<br />
a e 0000<br />
3.1.5.2 Flux and power distributions and graphite lifetime<br />
~<br />
~~<br />
0.0090<br />
0.0033<br />
2.2427<br />
0,0143<br />
l e 9894<br />
0.0168<br />
0.0182<br />
0.0194<br />
310 7905<br />
0.0032<br />
2. €754<br />
0.0136<br />
0. E245<br />
e he relative fast flux (E > 52.4 ke~) and power-peaking factors are<br />
given in Table 11. These factors include <strong>the</strong> effects of flattening. For<br />
comparison, <strong>the</strong> overall fast flux peaking in an unflattened core would be<br />
-2.3; <strong>the</strong> neutron leakage, however, wsbald be only 0.8% vs 1.56% for this<br />
core<br />
c
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