Review of Molten Salt Reactor Physics Calculations [Disc 2]

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14 3. ao6S SECTIONS For the most part the cross sections used both in this review and in previous molten salt breeder reactor design studies are the Bane. The cross section data used have been accumulated over a peabd oln about five years and regularly used for reactor evaluation studies. In this regard they have proven to be reasonably accurate in canparison with experiments and calculations by others. A summary of the basic thermal neutron cross section data used in this review and their experimental sources are given in Table 3.1. Table 3.2 lists the resonance fission and absorption integrals and the data sources for the same nuclides. There are some small differences between the thermal neutron cross sections for sane nuclides in this review and in the previous design studies. These differences are primarily a result of the issuance of Supplement I1 of -325 (Ref. 1) which recammends some renormalization of previously accepted cross section values. The differences are' shown 23% 235, 234, in Table 3.3 Of these differences only those In 3 are signiflcant in the MSTlR and lead to a slightly higher breeding ratio. The use of 2200 dsec V of 2.503 rather than 2.500 for 23% is con- sistent with the BML-325 (Ref. 1) recommended value for the prompt V of 2.497 0.008 and a delayed neutron fraction of 0.00264, The q of 2.295 at 2200 dsec for 23% is within the uncertainty range for this nuclide although not necessarily more accurate than the value of 2.292 used in the previous calculations. The 235U 2200 dsec data have been renomlized as recommended in -325, the primary result being a slightly higher a (0.175 VI 0.174), 8 higher t hedl V (2.442 vs 2.430), and 8 resulting slightly higher 11 (2.078 v8 2.070). The 10 barn difference in the ?34U thermal cross section makes a significant difference in the equilibrium concentrations of 234U and 235U. The Brown St. John2 heavy-gas model was used for the scattering kernel for all nuclides except carbon both in this review and in the previous design studies. For carbon the crystalline model as developed by Parks3 was employed. For the review calculations all kennels were cosrputed for an average temperature of 9oO°K-; This temperature was based 't
"'d n d Tab1 Nomalizatian a Sources of the Thermal Cross Sections Used in the MSER Studies Basis for the Energy Dependence ofthe Cross Section 23% 2-73 Assumed l/v throughout thermal energy range. 23$, 6.0 BNL-325 (Ref. 1) Four lowest energy resonances and a computed negative energy resonance 235, Fission cross section based on recommended curve in RlIL-325, 2nd Ed., Supp: 2, Vol. 1x1. Capture cross sec ion based on recent a(E) measurements s' B and of Westcn Bsmmsure et al.5 234u 233u 233Pa ed from two lowest positive energy reso:- P nances, and a computed negative energy resonance ~1 Macklin The tilevel resonance parameters of Moore and in and Reich- 9 and the a(E) data of BmG325 (Ref. 9). V = 2.503 Magnussonlnaad fram data q = 2.2946 of Block et slmBwith us a = 0.0908 taken to W 13.0 b. based on measurements of 0leksa.- 5 (Ref. 1) The re lved resonance parameters of Simpi3on: et til. E! 7.56 2=Th The eight lowest ene ..resonance parameters as reported by mOrdhed- and a computed negative energy Eesonance . chromium 3 RlIL-325 (Ref. 1) Assumed l/v in thermal range. [.og(2200) = 13.0 b.1 Iron 2.62 -325 (Ref. 9) Assumed l/v in thermal range. [~,(2200) = 11.0 b.1 T
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 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 and 52: References. 1. D. RI Vondy, T. B. F
Page 53 and 54: 53 Table 7.1 Neutron Balance by Nuc
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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