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
Energy group Energy range Fast 14.918 MeV to 52.475 eV Re so na nee 52.475 ev to 2.3824 eV The rmal '2.3824 eV to 0.88647 eV and uranium nuclides were prepared for use in the depletion and reactiv- ity coefficient studies. These were weighted over the neutron spectra calculated in the cell cal.culatPon. The macrospatial effects were treated using the reduced cross sec- tion set with the APC EI esmputer Separate axial and radial flux proflies were found with mutually consistent flux and leakage resu1ts. Core heterogeneity was treated by transverse flux weighting of the de- tailed geometry. Reaction parameters necessary for burnup were deter- mined from these results, with care taken to combine all reactions rep- resenting a particular nuelear species regardless O€ positions in the cell or the identity of the cell involved. This is consistent with an as- sumption of rapid fuel circulation and mixing. Burnup. A simple burnup code, QUAB, was devised to treat the un- usual requirements of this study. Special features include the follow- ing 1. Sufficient 238U is added at d l times to maintain the denatured con- dition. 2. The %ho%-iuwl cQncentrat%Ql3 can be held COnStaRt by automatic addition, allowed to decline naturally, or adjusted to maintain constant total actinide concentration. 3. Periodic additions of enriched fissile material can be made. 4. Pesiodfc withdrawals of fuel can be made selectivePy by nuclide. 5. fuel can be held until the protactinium decays and then be reinserted selectively by nuclide into the machine. placed with fuel identical to the initial loading. Enriched material can be added QPI demand to maintain a specified re- activity margin. The first removal is re- This . .
:.:..y ,. . . w, E9 The code calculates nuclide concentrations, total inventories, reactivity, and breeding ratio as a function of time. Treating the lengthy transplutonium and fission-product chains in QUAB was not practical; multigroup data were not available far many of the required nuclides and were of dubious reliability for others. In- stead, the BRIGEN code15 was used with a library of cross sections" es- pecially devised for its uses The BRIGEN results were then "patched into" the QUAB calculation directly. The burnup calculation allowed the cross sections of thorium and 2 38~ to vary continuously during the calculation; this was accomplished by in- terpolation. 3.1 e 3.2 Evaluation As desired, the method provided relatively rapid response9 detailed treatment of resonances, and a mulitigr~up spectrum and cell treatment. All details of the denatured fuel cycle were treated. The expedient of treating a range of thorium and 2 3 8 ~ densities remevest the necessity of imbedding the expensive and tedious resenamce treatment inside the Loop for varying densities. Deciding on the applicable range was not difficult after a few initial tries. A system ~oupling the spatial calculation and depletion could be used. Kany such systems are available, although all would require exten- sive modification for MSR use. What of the cell calculation? Table 7 shows the cell factors from our reference case which have been condensed to three energy groups. This is clearly a heterogeneous core. Further, the actinide densities are continually changing, resulting in time- dependent cell factors. Studies beyond these would be required t~ prove that a coupled system could be worthwhile without directly co~pled cell. cal cul a t io n s . The requirement to "umix" the revised nuclear densities after hav- ing them lumped together during a depletion step represents a complica- tion that would thwart most existing codes. However, this complication must be coupled with logic to provide interpolation between cross-section sets representing various self-shielding situations. With or without an
- 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: .. . E 6> I S N 0.45 k- 0.50 B --.
- 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
<strong>Energy</strong> group <strong>Energy</strong> range<br />
Fast 14.918 MeV to 52.475 eV<br />
Re so na nee 52.475 ev to 2.3824 eV<br />
The rmal '2.3824 eV to 0.88647 eV<br />
and uranium nuclides were prepared for use in <strong>the</strong> depletion and reactiv-<br />
ity coefficient studies. These were weighted over <strong>the</strong> neutron spectra<br />
calculated in <strong>the</strong> cell cal.culatPon.<br />
The macrospatial effects were treated using <strong>the</strong> reduced cross sec-<br />
tion set with <strong>the</strong> APC EI esmputer Separate axial and radial flux<br />
proflies were found with mutually consistent flux and leakage resu1ts.<br />
Core heterogeneity was treated by transverse flux weighting of <strong>the</strong> de-<br />
tailed geometry. Reaction parameters necessary for burnup were deter-<br />
mined from <strong>the</strong>se results, with care taken to combine all reactions rep-<br />
resenting a particular nuelear species regardless O€ positions in <strong>the</strong><br />
cell or <strong>the</strong> identity of <strong>the</strong> cell involved. This is consistent with an as-<br />
sumption of rapid fuel circulation and mixing.<br />
Burnup. A simple burnup code, QUAB, was devised to treat <strong>the</strong> un-<br />
usual requirements of this study. Special features include <strong>the</strong> follow-<br />
ing<br />
1. Sufficient 238U is added at d l times to maintain <strong>the</strong> denatured con-<br />
dition.<br />
2. The %ho%-iuwl cQncentrat%Ql3 can be held COnStaRt by automatic addition,<br />
allowed to decline naturally, or adjusted to maintain constant total<br />
actinide concentration.<br />
3. Periodic additions of enriched fissile material can be made.<br />
4. Pesiodfc withdrawals of fuel can be made selectivePy by nuclide.<br />
5.<br />
fuel can be held until <strong>the</strong> protactinium decays and <strong>the</strong>n be reinserted<br />
selectively by nuclide into <strong>the</strong> machine.<br />
placed with fuel identical to <strong>the</strong> initial loading.<br />
Enriched material can be added QPI demand to maintain a specified re-<br />
activity margin.<br />
The first removal is re-<br />
This<br />
. .
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