OAK RIDGE NATIONAL LABORATORY operated by UNION CARBIDE CORPORATION % for the U.S. ATOMIC ENERGY COMMISSION ORNL- TM- 611 COPY NO. - 9’5— DATE - August 27, 1963 MASTER INHERENT NEUTRON SOURCES IN CLEAN MSRE FUEL SALT P. N. Haubenreich ABSTRACT Unirradiated MSRE fuel salt will contain an appreciable neutron source due to spontaneous fission of the uranium, and (@, n) reactions of alpha particles from the uranium with the fluo- rine and berylliium of the salt. The spontaneous fission source in the core (25 f£t2 of salt) is 10® neutrons/sec. or less, mostly from U238, The alpha-n source is much larger, giving sbout 4 x 10° neutrons/sec. in the core. Nearly all of this la’gzer source is caused by alpha particles from yec4, NOTICE This document contains information of o preliminary nature and was prepared primarily for internal use at the Oak Ridge Nationa! Laboratory. It is subject to revision or correction and therefore does not represent o final report. 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CONTENTS Introduction==~wr-roeme e m e e e e e Fuel Composition===-~===---rm-cccmmm e m e e Spontaneous Fission Neutrons-----e----mccmcmcmcccc i e Neutrons from Alpha-n Reacticng-=-----------w-e-cr-cmcmmemnnn-- Alpha Emission by Uraniume--=--cem-e-eereemera e e r e m e e e Alpha-n Yields in Fuel Sglt----=-mcmmmmc e e e e DiscuSSioN==m=mc- e e c o e e e e e e — - Appendix — Calculation of Alpha-n Yields in MSRE Fuel Salt------ Dilution by Non-Productive Constituents of Fuel balt------- Be® (@, n) Yield=-mmrmme e e oo e e F12 (@, n) Yield=-==-=cmc e e e e e 117 (0, n) Yield=mm==m=mmmmmm e oo e e e References - v T wr S M T W e amm Sm e mw e e mw v G e e e e mm v el vl M SN T g M Gl e MDA AN EE ED G B NN B A NN N W RN e el WEoam wm m \J O ¢ 1 O\ 10 12 12 1k 1h 14 16 » Introduction When a reactor is subcritical, the fission rate and the neutron flux depend on the strength of the neutron source in the reactor due to various reactions and the multiplication of these source neutrons by fissions in the core. By supporting the fission rate in the subecritical reactor at sufficiently high levels, a source performs several functions in reactor operations. The source strength required for some functions is higher than for others. In all reactor fuels there is always a source of neutrons due to spontaneous fission, but this source is relatively weak, particularly if the fuel is highly enriched urasnium. In many reactor cores there is also an inherent photoneutron source produced by interaction of gamma rays with deuterium or beryllium in the cdre. This sourée is usually not significant, however, until after fission product gamma sources have been built up by power operation. . Therefore, in nearly all reactors an-extranebus neutron source 1s inserted in or near the core. The MSRE is unusual in that the fuel is a homogeneous fused salt in which alpha-émitting uranium is inti- mately dispersed with large gquantities of fluorine and beryllium,both of which readily undergo alpha-n reactions. Thus thefe is a strong alpha-n source inherent in the MSRE fuel salt even before it has been irradiated. It is conceivable that the source inherent in the MSRE fuel salt, which is certain to be present whenever there is any chance of eriticality, is strong enofigh-to satisfy some, if not all, of the requirements which make an extraneous source necessary in most reactors. In determining whether or not an extraneous source is required, it is necessary to pre- dict the strength of the neutron source which 1s inherent in the clean slat before the photoneutron source becomes important. The present report describes this prediction. The question of source reQuirements will be considered later, in a separate report. Fuel Composition The strength of the inherent neutron source depends on the composition of the fuel salt and the isotopic composition of the uranium in it. Three fuel salts, with compositions shown in Table 1, have been considered for the MSRE. | Uranium concentrations shown are for the initial critical experiment. For power operation the uranium concentration will be higher by about 15% (to compensate for control rods, xenon and other poisons). The isotopic compositions shown for the uranium are the values used in the criticality calculations., The U234 and U=3€ fractions are based on typical analyses of uranium enriched in the diffusion plant to the indicated U235 content. The lithium composition is that of lithium actually on hand for fuel salt manufacture. Table 1. Tuel Salt Compositions Fuel Type A B C Sglt comp: LiF® 70 66.8 65 (mole %) BeFo 23.7 29 29.2 ZrF, 5 L 5 ThF, 1 0 0 UF, 0.313 0.189 0.831 U comp: =34 0.3 (atom %) =33 U235 0.3 UESB 5 5 & .L‘_ Density at 1200°F 14k .5 134.5 2.7 (1b/£t£3) 299.9926 % 1i7. Spontaneous Fission Neutrons The rate of neutron production by spontaneous fission is a specific property of the each nuclide. In the clean MSRE fuel, U®®® has the shortest half-life for spontaneous fission.* (See Table 2.) Table 2. Neutron Production by - Spoataneous Fission of ' Uranium Isotopes Isotope Specific Emission Rate- (n/kg — sec) ges4 6.1 U=3s 0.51 [ 5.1 ye3s 15.2 The effective core of the MSRE (the’graphite-containing region plus some of the fuel in the upper and lower heads ) contains o5 ££3 df fuel salt. The amounts of each uranium isotope and the spontaneous fission neutron source in this.Volume are given in Table 3 for each of the three fuels described in Table 1. Table 3. Spontaneous Fission Neutron Source in Core | Fuel A Fuel B Fuel C Isotope M, (kg) S(n/sec) M (kg) S(n/sec) M, (kg) S(n/sec) ye34 0.3 2 0.2 1 0.2 1 Uess 27.0 14 16.5 8 26.4 13 ysss 0.3 2 0.2 1 0.2 1 g=s38 1.5 o0 0.9 13 47.5 722 4o 23 737 - Neutrons From Alpha-n Reactions Energetic alpha particles can produce neutrons by nuclear interactions with several'different nuclides. Threshold energies vary widely, depending upon the nuclide. Three nuclides, Li7, Be® and F12, have a-n thresholds below the maximum energy of alphas from uranium. The neutron yield per alpha particle}is a function of the initial energy of the alpha particle and the composition of the medium in which it is slowing down. Alpha Emission by Uranium Among the uranium isotopes present in fresh MSRE fuel, UZ3% has by far the highest specific alpha emission and also emits the highest-energy alpha particles. The specific alpha sources are summarized in Table TR Table 5 gives the total alpha scurce in the effective core of the MSRE during the initial critical experiment (25 £t of salt, containing the amounts of uranium shown in Table 3)}. Table 4. Alpha BEmission by Uranium Tsotope Half-Life for Decay Rate Ea f Q Source P a-decay (y) (ais/sec-kg) (Mev) (&/100 dis.) (a/sec-kg) U234 2.48 x 103 2,28 x 101 L4.77 72 1.64 x 10%L L.72 28 0.64 x 10%% Uess 7.13 x 108 7.9 x 107 L. 58 10 0.79 x 107 L.y 3 0.24 x 107 4.40 83 6.56 x 107 .20 L 0.32 x 107 yess 2.39 x 107 2,35 x 10° L.50 73 1.72 x 10° h.45 27 0.63 x 10° =38 L.51 x 10° 1.23 x 107 L.19 T7 0.95 x 108 h.15 23 0.28 x 106 Note: Ea is the initial energy of the alpha particle and f is the percentage yield of alphas of that energy in the natural alpha decay of the nuclide. Table 5. Alpha Source in MSRE Core Source Strength (a/sec) Isotope E, (Mev) Fuel A Fuel B Fuel C 234 b7 L.7h x 10© 2.90 x 101° 3.71 x 10%° L.72 1.85 x 10© 1.13 x 101 1.45 x 101° =35 4 .58 2.13 x 108 1.30 x 108 2.09 x 10® L4 0.65 x 108 0.40 x 108 0.63 x 10® 4.40o 17.7 x 108 10.8 x 108 17.3 x 10®8 4.20 0.86 x 108 0.53 x 10%8 0.85 x 108 Usse L .50 5.01 x 10° 3.06 x 108 3.92 x 108 L.Lh5 1.83 x 108 1.12 x 10%° 1.44 x 108 =38 4.19 1.40 x 108 0.86 x 10° 0.45 x 108 L.15 0.41 x 10° 0.25 x 108 0.13 x 10®8 Alpha-n Yields in Fuel Salt The yields of neutrons from Be®, F®, and Ii” vary with the energy of the alpha-particle, generally increasing with energy. Yields for 4.77-Mev alpha-particles in thick targets of pure material are 40, 6 and 0.1 neutrons per million alpha-particles in beryllium, fluorine, and lithium-7, respec- tively. In the MSRE fuel salt, the productive nuclides comprise only a fraction of the total, and the yield is affécted by the dilution with other elements. Yields for'alpha-particles of each energy in Table 5, in each of three fuel salts, were calculated by proéedures déécribed in the Appendix. Table 6 illustrates how beryllium, fluorine, and lithium contribute to the total yield for the most numerous and highest-energy group of alpha parti- cles. Table 6. Neutron Yields for L.77-Mev Alpha Particles in MSRE Fuel Salt Yield (n/10%/a) Constituent Fuel A Fuel B Fuel C Be 2.65 3.32 3.20 F - L.36 by I, Lo Li ~ 0.02 0.02 0.02 ~ Total = 7.03 7.78 T.62 Table 7 gives the neutron source in the éffective core of thé MSRE when the uranium concentration is at its Initial, clean, critical value. 10 Table 7. Alpha—n Neutron Scurce in Core Alpha Neutron Spurce Strgngthr(n/sec) E_ (Mev) Source O Fuel A Fuel B Fuel C ya34s b .77 3.33 x 105 2.26 x 105 2.83 x 105 h.72 1.21 x 10° 0.82 x 10° 1.03 x 10° U235 4.58 1.14 x 10 0.78 x 10® 1.23 x 10° b7 0.30 x 10° 0.21 x 10° 0.32 x 103 4. 40 7.54 x 10%- 5.20 x 10° 8.13 x 103 k.20 0.28 x 10° 0.20 x 103 0.31 x 103 U238 L.50 2.45 x 103 1.68 x 103 2.10 x 103 4 .45 0.83 x 103 0.57 x 103 0.72 x 103 y=38 4.19 L.36 3.07 157 ©1.15 1.21 0.85 43 Total L.67 x 10° 3.17 x 107 3.99 x 10° Discussion The calculations indicate that the bulk of the neutron source inherent in the clean MSRE fuel is due to alpha-n reactions, with spontaneous fis- sion contributing relatively little. Furthermore, about 97 percent of the neutron source is caused by alpha particles from a single isotope, U=3%, which comprises a very small fraction of the total uranium. Therefore, the neutron source will be very closely proportional to the UZ3% content of the fuel sait. In natural uranium, the abundance of U3* is only 0.0057%, or 0.0079 of the U235 gbundance. In a gaseous diffusion plant, however, the UZ34/y=35 ratio is increased, so that in uranium containing over 90% U=3% the U234 /U235 ratio is 0.010 or above. The U3* fractions which were used in the calculations are based on typical analyses of enriched uranium, and thus are only estimates of what will appear in uranium which will be used in making up the MSRE fuel salt. The estimate is probably good to within +20% in the case of Fuels A and B, which use highly enriched uranium. In the case of Fuel C it was assumed that the uranium would be taken from the diffusion plant at about 35% U=3°, 11 and that the U234 content would be only 0.30%. It now appears that the uranium may be added to the MSRE fuel salt in two batches: the first of natural or depleted uranium; the second, highly enriched. If this course is followed, the U23* content of Fuel C would probably be higher, perhaps by as much as & factor of 1.4. The neutron source for Fuel C would then be higher by the same factor. 12 APPENDIX Calculation of Alpha-n Yields in MSRE Fuel Salt Information on alpha-n yields from various nuclides usuvally appears in one of two forms: 1) the microscopic cross section of the nuclide for the a-n reaction as a function of alpha energy, or 2) the yield of neutrons per million alpha particles of a given initial energy emitted in an infi- nite medium of the pure nuclide. If the alpha particles are emitted in a mixture, it is necessary to take into account the dilution of the produc- tive nuclides by others which only slow down the alpha particles. Dilution by Non-Productive Constituents of Fuel Salt The correction for the dilution of a productive nuclide in a mixture is essentially the fraction of the alpha energy loss which is attributable to the productive constituent. Let Doax be the yield of neutrons for alpha particles emitted in an infinite medium consisting entirely of a productive nuclide. ILet n be the yield for that nuclide in a mixture. It has been observed=’> that a fairly good approximation is n Nb Sp T T o (1) max j_Ni Si where S is the 'relative atomic stopping power', N is the number density of a nuclide, and p refers to the productive constituent. The best information on relative stopping powers is still a 1937 article by Livingston and Bethe.* They give S relative to air for 16 elements for 6-Mev alpha-particles and for 6 elements at 7 other energies from 2 to 52 Mev. Table 8 gives values of S for the constituents of the MSRE fuel salt, obtained by interpolation in energy and atomic number of the Livingston - Bethe data. The relative stopping powers in this table are evaluated at 4.5 Mev, because this is approximately the energy of the uranium alpha-particles. 13 Table 8. Relative Stopping Power of Constituents of MSRE Fuel Salt for 4.5-Mev Alpha Particles . NS/),N; S3 - Constituent S - ' ' "Fuel A TFuel B Fuel C S Li 0.57 0.163 0.159 0.149 Be . 0.70 0.068 0.085 0.082 F 1.19 0.692. 0.705 0.699 Zr 2.8 0.057 0.0Lk7 0.056 Th 3.9 0.016 0 0 U ‘ .2 0.005 0.003 0.014 ”?Atomic stopping power relative to air. If the microscopic cross section of & nuclide for the alpha-n reaction is known, then the number of neutrons produced by an alpha particle can be found from® EC) .- N_o(E) n = f(_Ei—:E_:T dbE (2) O d‘X, 1 ag\~t : 5 a;)- as a function of alpha energy for several different substances. For a mixture one may assume that Harris! has presented aE 1 4E T & - Puix Z‘“i("'p'dx : | (3) 1 where w, is the weight fraction of constituent i in the mixture. Table 9 gives'values of % %g for the constituents of the MSRE taken from refer- ence 1, and the products of this quantity and the weight fractions for the three different fuel salts. The sum at the bottom of each column is L i for each salt. p_ .. Ox mix 1k Table 9. Slowing-Down Parameters for Alpha Particles 1 dE\ Mev wi (} 1ahy Mev "5 ) e > &x ), 5/ . p d.X i + Fuel A Fuel B Fuel C 4 Mev 5 Mev 4 Mev 5 Mev 4 Mev 5 Mev 4 Mev 5 Mev Li 885 781 103 91 108 95 96 85 Be 840 741 43 38 57 50 53 L7 F 730 €5 L7k 419 513 153 L89 432 Zr 385 351 Lo 38 37 33 L2 39 Th 228 208 13 12 0 0 0 0 U 222 202 h Y 3 2 10 9 679 602 718 633 690 612 Be® (o, n) Yield For alpha particles with an initial energy EO, emitted in pure beryl- 2,3 Jium n = 0.152 E 358 neutrons/10% « (4) max o For Fuels A, B, and C, n/nmax is 0.068, 0.085 and 0.082 respectively. (See Table 8). The product is the yield in the fuel salt which is tabu- lated in Table 10, F¥° (o, n) Yield Segre and Wiegand® measured neutron yields for alpha particles of various energies in thick targets of F. The yield, LN begins to be measurable at 3 Mev and rises to 10 neutrons/10® alphas at 5.3 Mev. From Table 8, n/nmaX for Fuels A, B and C are 0.692, 0.705 and 0.699. The pro- duct of this n/nmax and Noox from the data of Segre and Wiegand is given in Table 10. Li” (@, n) Yield This writer knows of no direct measurements of nmax for Li7. The cross-section for the Li” (&, n) B reaction as a function of alpha energy 15 was calculated and reported by Hess.® Above a threshold at 4.36 Mev, the cross-section rises to 8 mb at 4.8 Mev, then decreases to about & mb at higher energies. Hess' cross-section was used to compute yields from Li7 in the fuel salt from equations (2) and (3). The integral in Eq. (2) was evaluated by fepresenting the cross section curve by straight-line segments and‘by‘approximating - % ;“ %ED-l vs. B by linear relations fitted to points at 4 and 5 Mev ginQXin Table 9. Results appear in Table 10. Table 10. Neutron Yields for Alpha Particles in MSRE Fuel Salt (n/108 ) By Fuel & =~ Fuel B Fuel C (Mev) 5o 5 Ii Be F Ii Be F Li h.77 2.65 L4.36 0.019 3.32 L.4h 0.019 3.20 L4.40 0.017 L2 2.60 3.9% 0.016 3.25 L4.02 0.016 3.13 3.98 0.01k L.58 2.31 3.0k 0.008 2.89 3.10 0.008 2.79 3.08 0.007 4.50 2.18 2.70 0.005 2.72 2.75 0.004 2.62 2.73 0.004 L.y o 2.12 2.56 0.003 2.65 2.61 0.003 2.% 2.59 0.003 Lis 2,10 2.42 0.002 2.63 2.47 0.002 2.53 2.45 0.002 L.ho 2.01 2.25 0.00. 2.52 2.29 0.001 2.43 2.27 0.000 h.20 1.70 1.52 0 2.13 1.55 0 2.05 1.54 0 L.19 1.67 1.45 0 2.09 1.48 0 2.02 1.47 0 4,15 1.63 1.31 0 2.04 1.34 0 1.97 1.33 0 16 References 1. D. R. Harris, "Calculation of the Background Neutron Source in New, Uranium-Fueled Reactors,” USAEC Report WAPD-TM-220, Bettis Atomic Power Iaboratory, March 1960. 2. A. O. Hansen, "Radiocactive Neutron Sources," p. 3 in Fast Neutron Physics, Part 1, ed. by J. B. Marion and J. L. Fowler, Interscience, New York, 1960. 3. 0. J. C. Runnalls and R. R. Boucher, "Neutron Yields from Actinide-Beryllium Alloys," Can. J. Phys., 34: 949 (1956). L. M. 8. Livingston and H. S. Bethe, "Nuclear Dynamics, Experimental," Revs. Mod. Phys., 9: 272 (1937). 5. W. N. Hess, "Neutrons from (@, n) Sources,' Annals of Phys., 2: 115233 (1959). 6. E. Segre and C. Wiegand, "Thick-Target Excitation Functions for Alpha Particles," USAEC Report LA-136, Los Alamos Scientific Laboratory, September 194L (also issued as USAEC Report MDDC-185). = H e FwW RO 15-19. 20. 21. 22. 4748, 51. 52, 53-67. O O~ Ovn Fw o 17 Internal Distribution MSRP Director's Office, Rm. 219, 9204-1 . E. Beall . Bender Bettis Boch . Blankenship Claiborne Ditto . Engel . Epler . Fraas . Grimes Guymon . Hanauer Haubenreich Lindauer Lyon MacPherson HybDduwEprpEHGCRDODRPREHEZN NMEWEmndxwgdg o+, 23. ok . 25. 26. 27 . 28. 29. 30. 31. 32. 33. 3. 35. 36. 37-38. 39-40. h1-h2. 4345, L6, . McDonald . McDuffie . Miller . Moore Perry Prince ITEEr4" W H. Shaffer J . Skinner Splewak . Taboada . R. 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