Pre-equilibrium Emission Spectra Calculations of <Superscript>232</Superscript>Th(n,xn) Using New Evaluated Method

The knowledge of the thorium (²³²Th) cycle potentialities are required for the design of a fusion-fission (hybrid) reactor. Pre-equilibrium nuclear reactions have been used to investigate the effect of initial exciton numbers on the nucleon emission spectra. In this study, the initial exciton numbers for the target nucleus of ²³²Th were calculated through a method of offered by Tel et al. and then were used to obtain the effect cross section of the neutron emission spectra. Using this new method, a different way from the literature, the initial exciton numbers calculated with the theoretical neutron and proton densities have been obtained with SKM* on the ²³²Th(n,xn) reaction at 14.1 and 18.0 MeV incident neutron energies. The results were analyzed by comparing the empirical results in the literature.     

Excitation Functions Calculations of the Charged Particle-Induced Nuclear Reactions on the <Superscript>9</Superscript>Be Target

In this study, calculation of charged particles induced nuclear reactions of beryllium (⁹Be) target nuclei have been investigated in the incident proton and alpha at energy range from threshold to 50 MeV. The excitation functions for ⁹Be target nuclei reactions have been calculated by using PCROSS nuclear reaction calculation code. Weisskopf–Ewing model for equilibrium, calculations and the full exciton and cascade exciton models for pre-equilibrium calculations were used. Also, the semi-empirical calculations for (p,α) reactions have been done by using cross section formula updated with new coefficient developed by Tel et al. (Pramana Indian Acad Sci 74(6):931, 2010). The calculated results were compared with the experimental data for different energy levels ranging from 0.00 to 9.65 MeV taken from the literature.     

Determination of impurity radionuclides in <Superscript>131</Superscript>Cs solutions

The radiochemical purity of a ¹³¹Cs solution used in brachytherapy is studied. After separating ¹³¹Cs from the neutron irradiated targets BaO, Ba(NO₃)₂, and BaCO₃, the contribution of impurities was evaluated: 0.015% for ¹²⁴Sb and 0.012% for ¹³²Cs. The contribution of the parent ¹³¹Ba to cesium solutions was, on average, 0.0067% for BaO, 0.01% for Ba(NO₃)₂, and 0.011% for BaCO₃.     

Calculation of Equilibrium and Pre-equilibrium <Superscript>4</Superscript>He-Emission Spectra at 62 MeV Proton Incident Energy

In this study, the ⁴He-emission cross sections for ²⁷Al, ^54,56Fe, ¹⁹⁷Au, ²⁰⁸Pb and ²⁰⁹Bi target nuclei have been calculated at 62 MeV proton energy. In these calculations, the pre-equilibrium effects have been investigated. The pre-equilibrium calculations involve the geometry dependent hybrid model and hybrid model. A comparison with theoretical calculations (ALICE/ASH) has been performed and the empirical formulas for the ⁴He-emission cross sections produced at 62 MeV proton energy have been derived.     

Use of γ Radiation from Short-Lived Fission Products for Detecting <Superscript>235</Superscript>U and <Superscript>239</Superscript>Pu

It is suggested that γ radiation with E _γ > 4900 keV from short-lived fission products produced by thermal neutrons be used to detect ²³⁵U and ²³⁹Pu in samples. A time regime is substantiated: 120 sec irradiation, 60 sec holding time, and 120 sec measurement time. The contribution of the reaction (n, p) on fast neutrons is studied.__________Translated from Atomnaya Energiya, Vol. 98, No. 5, pp. 365–370, May 2005.     

<Superscript>239</Superscript>Pu prompt fission neutron spectra

Calculations of ²³⁹Pu(n, F) prompt fission neutron spectra have been performed for neutron energy up to 20 MeV. The exclusive spectra of pre-fission neutron reactions (n, xnf) were calculated on the basis of the Hauser-Feshbach model simultaneously with the cross sections of (n, F) and (n, 2n) reactions. The spectra of neutrons emitted by fission fragments were approximated by a sum of two Watt distributions. The components of the prompt fission neutron spectra due to pre-fission neutrons are manifested in the prompt fission neutron spectra and the average neutron energy. A correlation is established between this effect in the contribution of emissive fission (n, xnf) in the fission cross-section of ²³⁹Pu(n, F) and ²³⁵U(n, F). It is shown that the ²³⁹Pu(n, F) prompt fission neutron spectra used in applied calculations do not correspond to the experimental differential data and the systematic regularities in the spectra and their average energy found for the most carefully studied nuclei ^235,238U and ²³²Th. __________ Translated from Atomnaya énergiya, Vol. 103, No. 2, pp. 119–124, August, 2007.     

<Superscript>212</Superscript>Pb/<Superscript>212</Superscript>Bi Generator for nuclear medicine

A scheme for producing the α-emitters ²¹²Pb and ²¹²Bi, to be used in a promising method of diagnostics and therapy in oncology – radioimmunotherapy, is proposed. The technology is based on two generators operating in tandem: ²²⁸Th/²¹²Pb and ²¹²Pb/²¹²Bi. The first one is based on separation from an initial solution containing thorium isotopes and gaseous ²²⁰Rn, which secures the purity of the desired products ²¹²Pb and ²¹²Bi. For a ²²⁸Th/²¹²Pb model generator, the efficiency of ²²⁰Rn extraction from solution was ~60%. After conditioning, the ²¹²Pb solution from the ²²⁸Th/²¹²Pb generator was used to charge a column, which functioned as a ²¹²Pb/²¹²Bi generator, with a cation exchanger.     

Computations of (α,xn) Reaction Cross-Section for <Superscript>107,109</Superscript>Ag Coated Materials with Possible Application in Accelerators and Nuclear Systems

Silver plated shields are used in components and structures of fusion reactors that may be effected due to the huge radiation. For this reason, we focused on excitation functions of ^107,109Ag on alpha induced reactions. For simplicity to examining the events of nuclear physics, nuclear reaction codes such as ALICE/ASH, TALYS, CEM, EMPIRE, PCROSS, GEANT4 and FLUKA that includes different models have been used for many years. In this study, the cross-sections of (α,xn) reactions, where x = 1–5, for ^107,109Ag by using nuclear reaction codes ALICE/ASH and TALYS 1.6 were computed. Hybrid and Geometry Dependent Hybrid Models were used in ALICE/ASH whereas Two-Component Exciton Model was used in TALYS 1.6 reaction codes. Energy ranges of alpha particles for ¹⁰⁷Ag and ¹⁰⁹Ag were taken into computations as 9.2–96.8 and 10.2–71.2 MeV, respectively. Computed results are also compared each other and with experimental data existing in the literature for each reaction.     

Determination of <Superscript>131</Superscript>I based dose equivalents for emissions of short-lived iodine isotopes

The coefficients required for evaluating the inhalation dose from short-live iodine isotopes entering the human body during a radiation accident associated with a spontaneous chain reaction are validated computationally. Such coefficients make it possible to reduce the effect of a collection of iodine isotopes to that of an equivalent amount of ¹³¹I. The dose inhalation load from an iodine emission aerosol cloud and the site contamination with ¹³¹I are evaluated for a spontaneous chain reaction in systems consisting of enriched metallic uranium with 10¹⁹ fissions with no localizing facilities present. __________ Translated from Atomnaya énergiya, Vol. 102, No. 5, pp. 311–315, May, 2007.     

Analysis of the characteristics of <Superscript>238</Superscript>U, <Superscript>232</Superscript>Th, <Superscript>237</Superscript>Np,and <Superscript>231</Superscript>Pa radiators for fission ionization chambers for measuring the neutron flux density in the ITER diverter

Computational results, obtained by analyzing possible schemes of nuclear transformations of each of four threshold fission radiators ²³⁸U, ²³²Th, ²³⁷Np, and ²³¹Pa, for fission ionization chambers are presented. The influence of the nuclear reactions (n, ƒ), (n, γ), and (n, 2n) on the characteristics of fission ionization chambers is taken into account in the nuclear transformation schemes for all four radiators. The results are presented in the form of a dependence of the sensitivity of the fission ionization chambers on the neutron fluence in the range 10²¹–10²⁴ cm⁻². The effect of 0.2 and 1 g/cm² thick boron screens is examined. Translated from Atomnaya énergiya, Vol. 106, No. 1, pp. 42–47, January, 2009.     

Investigations of the regularities in the tranport of <Superscript>137</Superscript>Cs, <Superscript>60</Superscript>Co,and <Superscript>24</Superscript>Na photons in iron,lead, bismuth,tungsten, and uranium

The methods and results of benchmark experiments and calculations of the transport of ¹³⁷Cs, ⁶⁰Co, and ²²Na photons in iron, lead, bismuth, tungsten, and uranium are described. The results are photon spectra which served as the basis for determining the photon absorption coefficients in a wide geometry and the accumulation factors. A comprehensive analysis of the experimental and computational results revealed the regularities in the differences between the experimental and computational photon absorption coefficients in a wide geometry.Translated from Atomnaya Ènergiya, Vol. 97, No. 4, pp. 293–299, October, 2004.     

Effect of <Superscript>3</Superscript>He and <Superscript>6</Superscript>Li accumulation in beryllium blocks on the neutron-physical characteristics of the MIR reactor

Reactions resulting in the accumulation of ³He and ⁶Li, whose thermal neutron capture cross-section is large, occur under the action of neutron radiation in the beryllium blocks of the MIR reactor core. When a neutron absorber accumulates in the moderator of a reactor, important physical characteristics change: reactivity access, efficiency of safety and control rods, and reactivity effects; in addition, energy release is redistributed. An algorithm for calculating ³H, ³He, and ⁶Li in each beryllium block of the core has been developed and implemented. This algorithm makes it possible to follow the change in the concentration of these nuclides during reactor operation and shutdown. The ³He and ⁶Li concentrations are used as initial data for calculating the neutron-physical characteristics of the MIR reactor using the MCU and BERCLI programs. The computational results for the effect of the accumulation of the nuclides indicated on the neutron-physical characteristics of the core are presented. __________ Translated from Atomnaya énergiya, Vol. 104, No. 2, pp. 84–88, February, 2008.     

Plasma Focus Device as a Breeder of Proton to Produce Short Lived Radioisotope <Superscript>18</Superscript>F

In a plasma focus device, the nuclear fusion products are created through the thermal and non-thermal (beam-target) mechanisms. The beam target character of the pinched plasma is used to determine the yield of 3.02 Mev protons (when deuterium filling gas is used) at the optimized regime. For this situation, a combination of “moving boiler” model and a shock wave theory are employed. The numerical simulations for the production of the positron emitter nuclide, ¹⁸F (T _1/2 = 110 min; widely used in positron emission tomography), for two Mather type devices (NX2 and PF1000) show that, the rules of the drift velocity as well as the drive parameters have an high impact on the final yields.     

Calculation and analysis of p + Ca reaction cross sections at incident energies from threshold to 250 MeV

All cross sections, angular distributions and energy spectra of neutron, proton, deuteron, triton, helium, alpha particle emission for p + ^{40,42,43,44,46,48,nat}Ca reactions have been calculated and analyzed at incident proton energies from threshold to 250 MeV by nuclear theoretical models. The theoretical calculated results are in good agreement with existing experimental data.     

<Superscript>63</Superscript>Ni-based β-electric current source

The possibility of developing a β-electric atomic battery with microwatt power based on ⁶³Ni with a lifetime of at least 30 years is analyzed. A method is proposed for large-scale commercial production of ⁶³Ni with concentration 80–90% in the final product. A design of a β-electric transducer based on macroporous silicon with optimized configuration and macropore depth (microchannels) is presented and analyzed. ⁶³Ni is deposited on the inner surface of the microchannels. A laboratory model of an atomic battery based on ⁶³Ni has been fabricated. The working parameters and the current-voltage characteristics of a 500 μW(e) atomic battery based ⁶³Ni (90% concentration) are presented. __________ Translated from Atomnaya énergiya, Vol. 103, No. 6, pp. 353–356, December, 2007.     

Assessment of <Superscript>137</Superscript>Cs decontamination of concrete by the reagent method

The particulars of radioactive contamination of concrete wastes by ¹³⁷Cs were studied. x-Ray phase analysis and chemical analysis show that clayey materials, including Al₂O₃, Fe₂O₃, K₂O, and MgO, on which ¹³⁷Cs sorption is possible, were present in the concrete wastes. The content and form in which ¹³⁷Cs was found in radioactive concrete wastes from nuclear power facilities as well as in model samples were determined. When the wastes were treated with nitric acid the binder dissolved and a polydisperse suspension formed. The suspension consisted of three phases: solution, fine suspension, and rapidly settling precipitate of heavy filler particles. x-Ray phase analysis was performed and the ¹³⁷Cs mass ratio and distribution in the phases were determined. The possibility of decontaminating the concrete by a reagent method was evaluated.     

Stepped decontamination of <Superscript>137</Superscript>Cs-contaminated sandy soil

The possibilities of reagent-based processing of sandy soil contaminated with ¹³⁷Cs are examined. To attain a high decontamination factor, the process is conducted in several steps. First, water-gravity separation is used to separate finely dispersed matter (first step). The main goal of the next (second) reagent-based processing step is to destroy the strongly bound fragments and hydroxide films with which a part of the finely dispersed fraction is bound. At the next step, part of the ¹³⁷Cs is extracted into the reagent solution from the purely sandy fraction; repeated water-gravity separation (third step) makes it possible to completely separate the finely dispersed substance. The subsequent steps of reagent-based processing give ¹³⁷Cs concentration in the sandy fraction of the initial soils that meets the sanitary norms.     

New method of producing <Superscript>99</Superscript>Mo in molten-salt fluoride fuel

The status and the problems of world ⁹⁹Mo production are presented. A comparative analysis is made of reactor methods of ⁹⁹Mo production. It is noted that the currently used technologies and research reactors are not satisfying the growing demand in medicine for this isotope. It is underscored that the role of alternative production technologies has grown. In the development of new ⁹⁹Mo production technologies, the experimental results obtained on the basis of research program conducted on the MSRE reactor with molten-salt fluoride fuel have been analyzed. The analysis revealed a special behavior of certain fission products including ⁹⁹Mo: they leave the melt spontaneously and enter the gas phase. The authors hypothesize that highly volatile fluorides of the indicated products are formed in the melt; this explains the effect indicated. The effect is used as a basis to propose a new reactor method of producing fissionproduced ⁹⁹Mo. Concrete examples of a way to implement the new method of producing fission-produced ⁹⁹Mo using molten-salt fluoride nuclear fuel are presented.     

<Superscript>59</Superscript>Ni Accumulation in Corrosion-Resistant Steel under the Action of Neutron Flux

A γ-ray line with energy E_γ = 11.3 MeV was detected during an experiment, performed on a nuclear reactor, investigating the characteristics of the energy spectrum of γ-rays. The most likely source of this line is radiative capture of thermal neutrons by ⁵⁹Ni nuclei, which accumulated in the corrosion-resistance steel as a result of the more than 20 years of irradiation in the reactor, via the reaction ⁵⁸Ni(n, γ)⁵⁹Ni. It was found that for thermal-neutron fluence 10²¹ cm⁻² the ⁵⁹Ni concentration is 0.47% of the ⁵⁸Ni concentration. __________ Translated from Atomnaya Energiya, Vol. 99, No. 4, pp. 268–272, October, 2005.