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1 Controlled Fission 235 U + n X + Y + (~ 2.4) n Moderation of second generation neutrons Chain reaction. Water, D 2 O or graphite moderator. Ratio of number of neutrons (fissions) in one generation to the preceding k (neutron reproduction or multiplication factor). k 1 Chain reaction. k < 1 subcritical. k = 1 critical system. k > 1 supercritical. For steady release of energy (steady- state operation) we need k =1. Fast second generation neutrons Infinite medium (ignoring leakage at the surface). Chain reacting pile Nuclear Reactors, BAU, 1 st Semester, (Saed Dababneh).

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Controlled Fission Average number of all neutrons released per fission (for thermal neutrons, eV). 233 U : U : Pu : Pu : absorbedReactor is critical ( k = 1): rate of neutrons produced by fission = rate of neutrons absorbed + leaked. Size and composition of the reactor.

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Nuclear Reactors, BAU, 1st Semester, (Saed Dababneh). 3 Probability for a thermal neutron to cause fission on 235 U is Controlled Fission 235 U thermal cross sections fission 584 b. scattering 9 b. radiative capture 97 b. If each fission produces an average of neutrons, then the mean number of fission neutrons produced per thermal neutron = <

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4 Controlled Fission 235 U 238 U Assume natural uranium: % 238 U, % 235 U. f / N = ( )(0) + (0.0072)(584) = 4.20 b. / N = ( )(2.75) + (0.0072)(97) = 3.43 b. Thermal f = 0 b584 b Thermal = 2.75 b97 b Nuclear Reactors, BAU, 1 st Semester, (Saed Dababneh). Using the experimental elastic scattering data the radius of the nucleus can be estimated. Doppler effect?

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Moderation (to compare x-section) 1H1H (n, ) (n,n) 2H2H Resonances?

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6 natural uranium Probability for a thermal neutron to cause fission in natural uranium If each fission produces an average of = 2.4 neutrons, then the mean number of fission neutrons produced per thermal neutron = = 2.4 x This is close to 1. If neutrons are still to be lost, there is a danger of losing criticality. (Heavy water?). enriched uranium For enriched uranium ( 235 U = 3%) = ????? (> 1.3). (Light water?). In this case is further from 1 and allowing for more neutrons to be lost while maintaining criticality. Controlled Fission Nuclear Reactors, BAU, 1 st Semester, (Saed Dababneh).

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7 Controlled Fission Verify Comment on the calculation for thermal neutrons and a mixture of fissile and non-fissile materials, giving an example. Comment for fast neutrons and a mixture of fissionable materials, giving an example. HW 11

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8 Converters: Convert non-thermally-fissionable material to a thermally-fissionable material. f,th = 742 b Nuclear Reactors, BAU, 1 st Semester, (Saed Dababneh). f,th = 530 b Conversion and Breeding

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9 If = 2 Conversion and fission. If > 2 Breeder reactor. 239 Pu : Thermal neutrons ( = 2.1) hard for breeding. Fast neutrons ( = 3) possible breeding fast breeder reactors. After sufficient time of breeding, fissile material can be easily (chemically) separated from fertile material. Compare to separating 235 U from 238 U. Nuclear Reactors, BAU, 1 st Semester, (Saed Dababneh). Conversion and Breeding Delicate neutron economy…!

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10 Controlled Fission thermalhave produced so far N fast neutrons. N thermal neutrons in one generation have produced so far N fast neutrons. fast fission factor Some of these fast neutrons can cause 238 U fission more fast neutrons fast fission factor = (= 1.03 for natural uranium). Now we have N fast neutrons. We need to moderate these fast neutrons use graphite for 2 MeV neutrons we need ??? collisions. How many for 1 MeV neutrons? strong The neutron will pass through the eV region during the moderation process. This energy region has many strong 238 U capture resonances (up to ????? b) Can not mix uranium and graphite as powders. resonance escape probability In graphite, an average distance of 19 cm is needed for thermalization the resonance escape probability p ( 0.9). Nuclear Reactors, BAU, 1 st Semester, (Saed Dababneh).

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11 Controlled Fission Now we have p N thermal neutrons. Graphite must not be too large to capture thermal neutrons; when thermalized, neutrons should have reached the fuel. Graphite thermal cross section = b, but there is a lot of it present. Capture can also occur in the material encapsulating the fuel elements. thermal utilization factor The thermal utilization factor f ( 0.9) gives the fraction of thermal neutrons that are actually available for the fuel. Now we have fp N thermal neutrons Now we have fp N thermal neutrons, could be > or < N thus determining the criticality of the reactor. The four-factor formula. k = fp The four-factor formula. k = fp (1- l fast )(1- l thermal ) Fractions lost at surface Nuclear Reactors, BAU, 1 st Semester, (Saed Dababneh).

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12 x 1.03 Fast fission factor Fast fission factor x 0.9 Resonance escape probability p x 0.9 Thermal utilization factor f x What is: Migration length? Critical size? How does the geometry affect the reproduction factor? Neutron reproduction factor k = Nuclear Reactors, BAU, 1 st Semester, (Saed Dababneh).

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