Review of Molten Salt Reactor Physics Calculations [Disc 2]

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52 7. DEPIJEPIOI? CUULATIoIdS The previous calculations used the MERC code which obtains equilibrium reaction rates and equllibrik concentrations automatically. trattons are obtained from the ERC portionof.MERC w hich solves directly the coupled equations of the equilibrium condition for a circulating fuel reactor. Since we wished to make an independent check on the validity of the solution, and had no other code available which solves the same problem, we made extensive modifications to an existing code. We took the UllM code, which does a multigroup point-depletion calculation of the equilibrium cycle for 8 solid-fuel reactor, and changed 2t to do a one- group, point-depletion calculation of the fluid fuel cycle. The concen- This required that the code use average stream densities for losses and removals, that it permit specification of separate cycle times for the two streams, and that the nuclide chains be modified to include all of those discussed in section 4 of this report. We used the one-group cross sections (microscopic reaction rates) obtained from the ASSAULT calculation for each nuclide in each stream. The removal of fission products was treated as indicated in section 4. Losses of uranium isotopes were assumed to be 0.001 per reprocessing cycle, loss of 23% was assumed to be O.oooO1 per cycle, and '37m was assumed to be:.removed completely each cycle. Complete transfer of uranium isotopes from the fertile to the fissile stream at the end of each fertile stream cycle was assumed. Table 7.1 gives the resulting neutron balance. The gross (nuclear) breeding ratio is 1.062 compared with 1.054 previously obtained'(0RIm- TM-1467, calculated from Table 8). Previous calculations have made the tacit assumption that the equilibrium fuel cycle can be used to represent the reactor history though it is known that sane molten salt reactors might have to be started up on a uranium fully enriched in 235,. In order to check the validity of the equilibrium assumption, we did a time-dependent depletion calcul8tion for the heavy nuclides over a 30- - 74 23% fuel. year reactor history, starting with a 93$ 235U We held the thorium concentratio9 constant and varied the fissile concentration to
53 Table 7.1 Neutron Balance by Nuclide Absorptions Fissions Productions C 0.0261 .!Ii 0.0172 0.0159 0.0103 0.0205 ‘ F 0.0274 INOR 0.0050 Leakage 0.0010 Delayed neutrons 0.0051 u5Xe 0.0050 Fissile stream '23% . 0.9070 " 0.8047 2.0148 234v 0.0907 o.ooo5 0.0014 235,. '0.0844 0.0682 01664 236u 0.0105 0.0001 0.0002 237,, O*ooo9 0.0063 151, 0.0018 147h 0.0023 . 14% 0.0009 0.0008 Other fission products 0.0086 Fertile stream _. 2%h 0.9825 23%a 0.0078 23?, 234i 6 Li 149& 151, 0.0086 o*oooo 0.0017 0.0001 . o.oooo iMo24 0.0056 0.&3 0.0190 Other fission products 0.0007 '- - TOTAL 2.:227g w355 2.2279
Page 1 and 2: OAK RIDGE NATIONAL LABORATORY opero
Page 3: 3 . CONTENTS Intmduction ..........
Page 6 and 7: 6 The cross sections selected were
Page 8 and 9: 8 2. QF RESULTS AND FECOMMEIPDATION
Page 10 and 11: 10 2. The average '1 of 235U was 1.
Page 12 and 13: The power distributions obtained in
Page 14 and 15: 14 3. ao6S SECTIONS For the most pa
Page 16 and 17: Table 3.1 (cont 'a) mUclide ua(2200
Page 18 and 19: I Table 3.1 (eont'd) Nuclide Ua(220
Page 20 and 21: Table 3.2 20 (cont'd) W i Absorptio
Page 22 and 23: 22 on an average fuel salt temperat
Page 24 and 25: 24 28. 29 9 30 31 32. 33 34 J. Halp
Page 26 and 27: 26 Fig. 4.1 kclide Chains Used in b
Page 28 and 29: ’ 28 Table 4.1 Yield F’ractions
Page 30 and 31: 7.5249"7 30 Table 4.3 Fission Produ
Page 32 and 33: 32 Table 4.5 (cont'd) U kclide .,(Z
Page 34 and 35: I The infinite-medium code M-GAM wa
Page 37 and 38: Table 5.1. Figure 5.1Materials Zone
Page 39 and 40: 39 Table 5.3. (cont'd) Nuclide Pure
Page 41 and 42: * u E of these dips and elastic res
Page 44 and 45: i 44 6. TWO-DmIONAL CAICULDIONS The
Page 46 and 47: 46 !Cable 6.1 Region Composition a
Page 48 and 49: 48 Table 6.3 Region Neutron Balance
Page 51: References. 1. D. RI Vondy, T. B. F
Page 55 and 56: ,- i 55 i !
Page 57 and 58: ~ . -. I . . ... ) ~ ,..I ........
Page 59: A DISTRIBUTIOl 1-5. DTIE, OR 6-15.

Review of Molten Salt Reactor Physics Calculations [Disc 2]

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