Controlling Energy Flow and Loss in p<Superscript>6</Superscript>Li Fusion Plasma

The proton-Lithium-6 (p⁶Li) fusion reaction is significant because it produces energy through charged particles. By selecting this reaction, the problems of tritium processes and 14 MeV neutron fluxes will be reduced. One of the main concerns for p⁶Li plasma is the control of energy flow and loss that occur in fusion reactor. The calculations of energy balance are essential for investigating the energy flow and loss in p⁶Li plasma. It has a fundamental role for describing the material conditions in this plasma. Energy production must compete with inevitable losses in plasma. The losses perform a principal role in determining the operating temperature of thermonuclear plasma. Some losses of energy can be minimized by the suitable selection of designing parameters while some are intrinsic in reactant system. Calculations of energy flow and loss suggest an operating point at 800 keV for p⁶Li plasma. The effect of electron temperature on ion–electron energy transfer and the bremsstrahlung losses is reviewed. It is indicated that the bremsstrahlung radiation losses resulting from large mean electron energies are a serious difficulty for p⁶Li fusion reactor. It would be highly desirable to reduce the electron temperature below their normal equilibrium values. If the ion–electron energy transfer be reduced from the classical value, the electron temperature and thus bremsstrahlung radiation losses would be reduced substantially and as a result the performance of a p⁶Li fusion reactor would be improved significantly. Meanwhile, the bremsstrahlung radiation losses can be minimized with suitable mixture for p⁶Li plasma in a fusion reactor.     

<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.     

Parametric tests and measurements after shimming of a beryllium reflector in a miniature neutron source reactor (MNSR)

After 10 years operation of Pakistan research reactor-2 (PARR-2), a miniature neutron source reactor (MNSR), a beryllium reflector was added to compensate the loss of reactivity due to burn up of fuel. Beryllium shim plates have been placed at the top of the core in a tray provided for this purpose. The control rod was dismantled and withdrawn from the core and the reactor was made subcritical with cadmium shimming. To monitor the neutron population during this experiment, two additional neutron monitoring channels based on BF₃ were installed around the core. Measurement of important Parameters such as effective delayed neutron fraction, decay constant, excess reactivity, control rod worth, temperature coefficient of reactivity, thermal neutron flux, cadmium ratio was done after the addition of Be reflector. Increase in reactivity worth due to addition of Be shim was 1.0 mk.     

Influence of the uncertainty of the microscopic cross sections of <Superscript>238</Superscript>U and plutonium isotopes on the neutron multiplication coefficient in VVÉR

The influence of the uncertainties in the microscopic cross sections of ²³⁸U and ^239–241Pu on the neutron multiplication coefficient in VVéR-1000 is studied. Data on the uncertainty of the cross sections were obtained by analyzing the ENDF/B files of the JENDL 3.3 system using the ERRORJ program. The characteristics of the reactor are calculated using the TRIFON 2.1 and SHERHAN programs. An approach associated with the calculation of the coefficients of sensitivity of the fuel assemblies and the reactor to standard samples in a small-group approximation is used. __________ Translated from Atomnaya énergiya, Vol. 103, No. 4, pp. 247–254, October, 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.     

Measurement of neutron capture and fission cross sections of <Superscript>236</Superscript>U

The cross sections for radiative capture and fission for ²³⁶U in the neutron energy ranges 1–2000 eV and 1–1000 eV, respectively, have been measured. The measurements were performed on a flight baseline 5.2 m of the FAKEL linear electron accelerator at the National Research Center Kurchatov Institute using a method based on measuring the distribution of the number of particles emitted by an excited nucleus with simultaneous measurement of the energy of each particle. The spectrum of the neutrons incident on a sample was measured by detecting the reaction ¹⁰B(n, α)⁷Li. The cross sections were normalized with respect to the resonance in ²³⁶U at 5.45 eV. The data obtained are compared with the results of other experiments and evaluated data.     

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.     

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.     

Relative yield of delayed neutrons and half-life of their precursor nuclei from <Superscript>238</Superscript>U fission by 14.2–17.9 MeV neutrons

The energy dependence of the relative yield of delayed neutrons and half-life of their precursor nuclei from ²³⁸U fission by 14.2–17.9 MeV neutrons is measured. In addition, the detector-blocking effect, which is due to the exposure of the detector to intense fluxes of high-energy neutrons, and the effect of the secondary neutron source D(d, n)³He, a reaction which occurs when a T(d, n)⁴He based neutron source is used, are investigated experimentally and taken into account in the data analysis. The data are analyzed in terms of the average half-life < T> of the precursor nuclei of the delayed neutrons. It is shown that the observed increase of <T> with increasing primary-neutron energy in the energy range studied is predominantly due to the opening of emissive nuclear fission channels. The energy dependence of the average half-life of the delayed-neutron precursors can be described on the basis of the model of emissive fission of three nuclei that fission in this energy range − ²³⁷U*, ²³⁸U*, and ²³⁹U*. __________ Translated from Atomnaya énergiya, Vol. 102, No. 2, pp. 124–132, February, 2007.     

Critical neutron heating in the control drums of a dual purpose thermionic space reactor for power and propulsion

The critical neutron heating in the reflector control drums is investigated for a fast incore thermionic space craft reactor for power and nuclear propulsion. The reactor is fueled with uranium carbide (UC) and controlled with the help of rotating B₄C drums imbedded into the beryllium reflector. While the neutron heating in the drums would not require a cooling mechanism in the power phase, the heat generation during the thrust phase obliges cooling for a nuclear thermal thrust around F = 5000 N by a specific impulse of 670 s⁻¹ at an hydrogen exit temperature around 1900°K. With a beryllium reflector without extra cooling measures, thermal thrust must be kept F < 2500 N to relieve the thermal load in the reflector. On the other hand, a reflector made of BeO may withstand a thermal load for a nuclear thermal thrust of F = 5000 N. The neutronic analysis has been conducted in S₁₆-P₃ and S₈-P₃ approximation with the help of one- and two-dimensional neutron transport codes ANISN and DORT, respectively. A reactor control with boronated reflector drums (drum diameter = 14 cm) at the outer periphery of the radial reflector of 16 cm thickness would make possible reactivity changes of Δk_{eff = 13.55%}—amply sufficient for a fast reactor—without a significant distortion of the fission power profile during all phases of the space mission. Calculations are conducted for a reactor with a core radius of 22 cm and core height of 35 cm leading to power levels around 50 kW_el.     

Experimental and computed spectra of neutrons and photons emitted from spherical iron models with a <Superscript>252</Superscript>Cf source at the center

Series of experiments and calculations in which the spectra of neutrons and photons emitted from 20, 40, 50, 60, 70, and 100 cm in diameter spherical models with a ²⁵²Cf source at the center are studied are described. The neutron spectra are measured with spectrometers based on stilbene and a hydrogen counter and the photon spectra are measured with spectrometers based on stilbene and a Si(Li) detector. The calculations are performed using the ROZ-6 computer program and the BNAB-78 28-group neutron constants library. The experimental fluxes of >0.8 MeV neutrons are, for the most part, greater than the computed values, and the fluxes of >0.35 MeV photons are also greater than the computed values. The differences reach a factor of 2–3. The data obtained show that the the cross sections for elastic and inelastic neutron scattering and radiative capture of neutrons need to be changed, leaving the total cross sections for the interaction of neutrons with iron in the constants library tested unchanged.__________Translated from Atomnaya Énergiya, Vol. 98, No. 1, pp. 54–61, January, 2005.     

Feasibility of Fast Fission System Confining Long-Lived Nuclides

The feasibility of fast fission system confining long-lived nuclides without other supporting system as synergetics for fuel sustainment and waste incineration was studied from the aspects of nuclear material balance and neutron economy. The continuous utilization of fast fission system which confines all actinides in the reactor but discharges all FP will lead to huge accumulation of radioactive wastes such as ¹²⁹I, ¹³⁵Cs, ¹⁰⁷Pd, ⁹³Zr, ⁹⁹Tc, ¹²⁶Sn and ⁷⁹Se in the far future. Then we studied the feasibility of the system that these long-lived seven FP are also confined in the reactor with actinides. In this scheme, all the long-lived nuclides to be disposed of were exposed with neutrons in the reactor and removed as different nuclides after nuclear transmutation. As the wastes stored in the repository was composed of only shorter-lived nuclides, total amount of radioactive wastes in the repository was suppressed to be less than a few tons per 3 GWt reactor.     

<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.     

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.     

Investigation of the reflector temperatures of a space craft thermionic reactor

In a fast incore thermionic spacecraft reactor for nuclear propulsion, the temperature rise due to the neutron heating in the reflector control drums is investigated. The reactor is fuelled with (U-Ta)C, consisting of 80UC-20TaC with a sinter density of 80% and controlled with the help of rotating drums imbedded into the beryllium reflector. The control drums contain natural B₄C strips (with 20% ¹⁰B and 80% ¹¹B) and produce nuclear heat via neutron absorption in ¹⁰B. The neutronic analysis has been conducted in S₁₆-P₃ and S₈-P₃ approximation with the help of one- and two-dimensional neutron transport codes ANISN and DORT, respectively. Calculations are conducted for a reactor with a core radius of 22 cm and core height of 35 cm leading to 50 kW_el in power phase. Reflector drums with 100% natural B₄C in form of strips (drum diameter=13.5 cm, strip width=5 mm) at the outer periphery of the radial reflector of 16 cm thickness would make possible reactivity changes of Δk_eff,max=10.7% without a significant distortion of the fission power profile during all phases of the space mission. A reduction of the B₄C in the strips to 20 and 10% would still allow a reactivity change of Δk_eff,max=8.4 and 7.7%, respectively, amply sufficient for an effective control of a fast reactor during all phases of the space mission. By a nuclear thermal thrust around F=5000 N and a specific impulse of 670 s⁻¹ at an hydrogen exit temperature around 1900 K, the maximum temperature in the drums rises to 1023 K, with 100% natural B₄C content in the strips, far below the melting point of beryllium. The maximum drum temperature is depressed to 663 and 519 K, with 20 and 10% natural B₄C content in the strips, respectively.     

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.     

Investigation of the infinite medium integral neutronic data for incident fusion source neutrons

In the present work, the physical behavior of integral data in infinite medium has been evaluated for incident fusion neutrons with the help to the 3-D Monte Carlo code. In a fusion reactor blanket with finite dimension, the integral quantities will be more or less different from the infinitive medium results, depending on the neutron leakage fraction. Design studies foresee the reduction of the neutron leakage out of the blanket as possible in order to prevent the nuclear heating in super conducting fusion magnets and to keep all neutrons primarily in the coolant. The most important materials in fusion technology, namely tritium, beryllium, lead, thorium, and uranium have been investigated in infinitive medium. The main purpose of this work is to calculate the integral tritium breeding ratio, ²³³U breeding rate, ²³⁹Pu breeding rate, heat release, neutron multiplication ratio through (n,x) and fission (when applicable) reactions in those mixtures which are composed when first UO₂ and ThO₂ are mixed with natural lithium (Nat.Li) or ⁶Li for a volume fraction from 0 to 100%. Then the variable UO₂-Nat.Li (UO₂ mixed with Nat.Li) and UO₂-⁶Li (UO₂ mixed with ⁶Li) compositions will be mixed with Beryllium (Be) and Lead (Pb) for a volume fraction from 0 to 100%. However, the variable TO₂-Nat.Li (ThO₂ mixed with Nat.Li) and ThO₂-⁶Li (ThO₂ mixed with ⁶Li) compositions will be mixed with Be and Pb for a volume fraction mentioned above.     

Improved Formula for (<Emphasis Type="Italic">n</Emphasis>,<Superscript>3</Superscript>He) Fusion Reactions Cross Sections Using Optical Model

Non-elastic cross-sections have been calculated by using optical model for (n, ³He) reactions at 14–15 MeV energy. The new empirical formula including optical model non-elastic effects by fitting two parameters for the (n, ³He) reaction cross-sections have been suggested. The excitation function character and reaction Q-values depending on the asymmetry term effect for the (n, ³He) reaction have been investigated. The obtained cross-section formula with new coefficients has been compared with the experimental data and the other fitting formulae existed in the literature and discussed. It has seen that the fit of new formula in this paper is greatly improved with the experimental data.     

<Superscript>243</Superscript>Am neutron fission cross section and resonance parameters

The SVZ-100 lead moderation time neutron spectrometer at the Institute of Nuclear Physics of the Russian Academy of Sciences was used to measure the fission cross section for ²⁴³Am in the neutron energy range E_n = 0.3 eV – 10 keV. The resonance fission integrals and the area and fission width of the resolved resonances were calculated. The properties of the intermediate-structure resonances were evaluated. The results were compared with existing data and recommended evaluations.