Uranium fission cross-section

B. R. Leonard, S. M. Hauser, E. J. Seppi, "The Ratio of Plutonium-239 to Uranium-235 Fission Cross Sections from 0.020 to l.,0 Electron Volts, HW-30128, December 1, 1953. 

The discovery of Pu has been described in detail by Seaborg in his Plutonium Story (chapter 1 of the book The Transuranium Elements 1958). First, the separation of Pu from Th caused some difficulties, because both elements were in the oxidation state 4-4. After oxidation of Pu(IV) by persulfate to Pu(VI), separation became possible. Pu is produced in appreciable amounts in nuclear reactors (section 14.1), but it has not immediately been detected, due to its low specific activity caused by its long half-life. After the discovery of Pu, plutonium gained great practical importance, because of the high fission cross section of Pu by thermal neutrons. Very small amounts of Pu are present in uranium ores, due to (n, y) reaction of neutrons from cosmic radiation with The ratio Pu/ U is of the order of 10 In 1971, the longest-lived isotope of plutonium, Pu (ri/2 = 8.00 lO y) was found by Hoffman in the Ce-rich rare-earth mineral bastnaesite, in concentrations of the order of 10 gAg-... 

Consider, for example, a PWR containing 80 te of uranium oxide enriched to 3%, with an average moderator temperature of 310 C. The effective specific volume of uranium oxide pellets will be 1.11 x 10 m /kg. The fission cross-section at a moderator temperature of 20°C/293 K, t/(293), is 582 x 10 m for U-235, while the corresponding figure for plutonium-239 is 743 x 10 m. Calculate NfOf at the start of life and at the end of life, and hence deduce the variation in the constant of proportionality, a. 

In a PWR operating with plutonium recycle the thermal-neutron flux is lower than for uranium fueling because of the hi er fission cross section for plutonium. As a result, less C is produced by thermal-neutron activation within the fuel, as drown in Table 8.11. 

The important point for Dunning, the reason for his passion, was that if U235 was responsible for slow-neutron fission, then its fission cross section must be 139 times as large as the slow-neutron fission cross section of natural uranium, since it was present in the natural substance to the extent of only one part in 140. By separating the 235 isotope, Herbert Anderson emphasizes in a memoir, it would be much easier to obtain the chain reaction. More than this, with the separated isotope the prospect for a bomb with unprecedented explosive power would be very great. ... 

What followed thus made the cross section intuitively obvious it would be more or less the same as the familiar cross section that expressed the odds of hitting the uranium nucleus with a neutron at all—the geometric cross section, 10 square centimeters, an entire order of magnitude larger than the fission cross sections previously estimated for natiural lura-nium that were small multiples of 10 . ... 

In these and the above equations, the a are cross sections per imit volume, the a in (8) is scattering cross section, the average loss in r per collision. The are used because the material may contain different types of atoms. The (Ta is the thermal absorption cross section r(r) the resonance absorption cross section per unit volume. The = qef is the multiplication constant divided by the resonance escape probability. The product of thermal utilization / and (Ta is the effective cross section of uranium per unit volume, i.e., its cross section per unit volume multiplied by the thermal neutron density in it and divided by the average thermal neutron density. One can write, therefore, (Tu for f(Ta- If one multiplies this with rj the result is the same as crfU where fission cross section for thermal neutrons per unit volume, p the number of fast neutrons per fission. As a result, the third term in (7) can be written also as e is the multiplication by fast effect)... 

In the center of the original core and then in the core with neptunium there were measured by several methods, the ratios of average fission cross-sections for 16 isotopes, including minor actinides, and of capture cross-sections in aurum, neptunium, uranium-238, the central reactivity coefficients with the use of samples, the sodium void effect of reactivity, the efficiency of a mock-up of the central control rod with enriched and natural boron carbide, ad well as the fission reaction rate distributions with height. 

There have been carried out investigations concerning uranium-thorium cycle at four critical assemblies of the COBRA facility. The central inserts of these assemblies contained U-235, thorium and hydrogen. Hydrogen/U-235 ratio of nuclear density varied from 0.0 to 70.0. These experiments allow to estimate an influence of thorium on neutron spectrum and an accuracy of nuclear data in a wide energy range (from Mev to tens of Kev). In experiment there have been obtained Keff values of various uranium-thorium compositions, cross-section ratios of some nuclear reactions including thorium capture-to-fission ratio, fission cross-section of some TRU. On completion of the experiments the... 

The a-emitting product was identified as a new element fi-om the study of chemical behavior of this isotope. It was distinctly different from both uranium and neptunium in its redox properties the 3+ and 4+ valence states were more stable. A second isotope of element 94, Pu, with a half-life of24,000 years was synthesized immediately as a daughter of P decay of Np, which confirmed the presence of element 94. The isotope Pu produced in appreciable amounts in nuclear reactors is of major importance, because of its large fission cross section with thermal neutrons. It was named after the planet Pluto in analogy to uranium and neptunium. 

With short reactor irradiation times, relatively pure Pu is formed from natural or low-enriched uranium. The cross section for thermal neutron capture by Pu is about 1,020 barns, and a sizable fraction produces Pu rather than fission ... 

Heavy nuclides are unstable they can be disintegrated not only by a or P decay or, in some cases, by spontaneous fission, but also by neutron-induced fission. The fission cross sections, however, of the various nuclides show great differences. Among the naturally occurring fissile nuclides is the only one that can be used in thermal reactors in addition, the artificially produced nuclides Pu and Pu show fission cross sections and halflives which make them appropriate for use as nuclear fuels. In the currently operating light water reactors, which are exclusively based on uranium as the starting element, only the fissile nuclides 2 Pu and 2 Pu are of real interest. 

Typical curves for the decay of and the buildup of plutonium isotopes in a natural uranium reactor are shown in Fig. 4.3. It will be seen that, beyond an exposure of approximately 7500 MW d/tonne of natural uranium, the concentration of Pu exceeds that of Because the fission cross section of is higher than that of the majority of fissions in the reactor will be taking place in plutonium rather than uranium even before the 7500-MW d/tonne exposure is reached. 

A remark on the fission cross section of uranium is in order here. U compounds are available to many workers and a large number of them has been examined at both reactor and spallation sources. Natural uranium has a fission cross section of 4 b, and,even depleted U contains a measurable quantity of What this means, of course, is that when a beam of slow or thermal neutrons is incident on the sample, the fission process gives rise to fast neutrons. These are difficult to stop, thermalize in a variety of ways, and contribute to the background not only on the instrument being used, but, more embarrassing, often on neighboring instruments as well The experimentalist should be aware of this problem, which becomes worse the lower the energy of the incident neutrons. In our experience care in the form of extra... 

The basis for the non-l/v character of the fission cross section of is found in measurements of natural uranium s 7) temperature coefficient, which leads one to choose... 

The thermal neutron flux to which a Zircaloy-2 specimen was exposed while in-pile was determined by measm ing and comparing the amount of the induced activities, Zr -Nb -, in the specimen and in a control sample of Zircalo,y-2. The latter was irradiated together with a cobalt monitor in a separate experiment in which the specimen and cobalt did not contact solution. Tor some experiments which contained steel specimens, similar measurements were made utilizing the activity. The fission power density in the uranium solution adjacent to a specimen was calculated from the resulting value for the thermal flux together with the value for the fission cross. section of uranium, assuming 200 Mev per fission. 

Uranium-233 is a superior fuel for use in molten fluoride-salt reactors in almost every respect. The fission cross section in the intermediate range of neutron energies is greater than the fission cross sections of U and Pu239 Thus initial critical inventories are less, and less additional fuel is required to override poisons. Also, the parasitic cross section is substantially less, and fewer neutrons are lost to radiative capture. Further, the radiative captures result in the immediate formation of a fertile iso-... 

The threshold reaction contributions to the total fission rate can be assumed small for the AGN-201 reactor, since its moderator-to-uranium volume ratio is appreciable and its fuel is enriched with the isotope. Very fast fission is normally accounted for in the four-factor formula by the factor e, the number of neutrons produced by all fissions divided by the number produced by thermal fission. In the AGN-201, nonthermal fission is predominately resonance fission, since has finite fission cross sections at all energies. The amount of epithermal fission can be determined by a simple cadmium-ratio measurement of AGN-201-type fuel. The fission product activity of a bare and cadmium-covered fuel sample can be counted on a proportional counter after two similar irradiations in the reactor core. Their ratio will yield the amount of nonthermal fission to total fission after proper corrections for differences of sample weight, irradiation times, and, power level have been made. The final expression for power level then becomes, . . f... 

The technologically most important isotope, Pu, has been produced in large quantities since 1944 from natural or partially enriched uranium in production reactors. This isotope is characterized by a high fission reaction cross section and is useful for fission weapons, as trigger for thermonuclear weapons, and as fuel for breeder reactors. A large future source of plutonium may be from fast-neutron breeder reactors. 

The only large-scale use of deuterium in industry is as a moderator, in the form of D2O, for nuclear reactors. Because of its favorable slowing-down properties and its small capture cross section for neutrons, deuterium moderation permits the use of uranium containing the natural abundance of uranium-235, thus avoiding an isotope enrichment step in the preparation of reactor fuel. Heavy water-moderated thermal neutron reactors fueled with uranium-233 and surrounded with a natural thorium blanket offer the prospect of successful fuel breeding, ie, production of greater amounts of (by neutron capture in thorium) than are consumed by nuclear fission in the operation of the reactor. The advantages of heavy water-moderated reactors are difficult to assess. 

Xenon occurs in the atmosphere at trace concentrations. It also occurs in gases from certain mineral springs. Xenon also is a fission product of uranium, plutonium, and thorium isotopes induced by neutron bombardment. The radioactive fission product, xenon-135, has a very high thermal neutron cross-section. The element has been detected in Mars atmosphere. 

Fig. 5. Neutron counting as detection method for spontanous fission events of superheavy nuclei. The recorded neutron rates (points) were found to follow the relative cross sections of cosmic-ray induced spallation reactions (curve) and were, thus, due to background events. The numbers are rates for natural uranium and thorium. From W. Grimm, G. Herrmann and H.-D. Schiissler [40].

A radiochemical study [104] of the element distribution in the 238U+238U reaction at the unilac revealed the expected broad distribution of reaction products. Below uranium, where losses by sequential fission of transfer products are not significant, the observed yields decreased exponentially from Z=92 down to Z= 73. This trend was well reproduced [105] by a theoretical model treating nucleon transfer in the intermediate collision complex as a diffusion process. By extrapolation of the model to Z=70 nuclei about 100 microbam total production cross section resulted, associated with broad distributions of neutron numbers and excitation energies. 

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