Fourier analysis of energy transfer data obtained by simulating a 14-MeV α-particle in water

Data from Monte Carlo transport codes are used to model radiobiological effects. We previously reported the Fourier analysis of ionization data generated by simulating a 500-keV proton traversing water. Here, we extend Fourier analysis to energy transfer data of another radiation type, a 14-MeV α-particle. A radiobiological model based on this frequency-domain analysis views cell as an information processing system [8]. It lends itself naturally to traditional engineering analyses. One engineering principle—the output response of a linear system to random signal—is applied here to explain the fact that there is measurable difference in the magnitude of the biological effectiveness when a given biological system is irradiated with two different radiation types of the same Linear Energy Transfer (LET).     

Research of CdZnTe detector based portable energy dispersive spectrometer

A kind of excellent CdZnTe crystal has been grown in Yinnel Tech, Inc. in recent years. Based on these CdZnTe crystals and some new techniques, a portable energy-dispersive spectrometer has been constructed which has yielded good results. CdZnTe detector has a 3% relative resolution in high-energy field and can detect gamma rays at room temperature. An integrated circuit based on preamplifier and shaping amplifier chips is connected to the detector. Voltage pulses are transformed into digital signals in MCA （multichannel analyzer） and are then transmitted to computer via USB bus. Data process algorithms are improved in this spectrometer. Fast Fourier transform （FFT） and numerical differentiation （ND） are used in energy peak＇s searching program. Sampling-based correction technique is used in X-ray energy calibration. Modified Gaussian-Newton algorithm is a classical method to solve nonlinear curve fitting problems, and it is used to compute absolute intensity of each detected characteristic line.     

Systematic study of anomalous fragment anisotropies in subbarrier complete fusion—fission reactions

The complete fusion-fission is separated from the transter-induced-fission with the fragment folding angle technique.The cross sections and fragment angular distributions for the complete fusion-fission reactions of 11B 238U(237Np),237NP,16O 232Th(238U) and 19F 232Th at near-and sub-barrier energies have been measured.The present fusion and fission standard models can reproduces both the excitation functions and the fragment anisotropies for the systems of 11B 238U(237Np)and 12C 237Np;but fail to explain both the experimental data for the other 3 systems simultaneously,The evidence of the entrance-channel dependence of fission-fragment anisotropies is revealsed by comparison of the 11B 237NP and ^16O 232Th data.Based on the observations a new version model of preequilibrium fission is put forward to explain the anomaly.     

Development of a dose-adjustable α-particle irradiation facility for radiobiological studies

Bystander effects induced by low-dose or low dose-rate radiation have put a great challenge to the traditional model for radiation protection. In order to get a better assessment of the adverse effects of radiation, especially the low-dose radiation risk in environment, a radioactive irradiation facility, by which the dose and dose-rate to the biological targets can be controlled by rotating the sample bracket and changing the space between the radioactive source and the targets, was developed. The energy of our α-radioactive source (^241Am) is measured averagely 3.5 MeV at the position of the irradiated sample. The dose rate was measured by using a silicon surface barrier detector and a CR39 particle track detector. A dose rate ranging from 0.045 cGy/s to 1.07 cGy/s can be obtained by changing the space length from the radioactive source to the sample dish.     

Application of ion transport simulation to the backscattering in α-particle sources

The standardization of α-particle emitting sources with 2π detectors requires corrections for backscattering from the backing material. The Monte Carlo computer code SRIM, developed to simulate the transport of ions in matter, was applied to the study of the backscattering of α-particles in different materials. Simulations of 5.3 MeV α-particles from ²¹⁰Po were performed for various backing materials with atomic numbers ranging from 13 to 79. In addition, some experiments using a 2π detector and a well-defined low solid angle chamber with a silicon detector were also performed. The experimental backscattering coefficients for ²⁴¹Am α-particle sources in Pt and Ag supports and for ²³⁹Pu αs in stainless steel and quartz backing materials were determined. There was reasonable agreement between calculations and experimental results (carried out by us and by other laboratories). This work shows that Monte Carlo simulation of the transport of ions can be a very useful tool to evaluate the backscattering corrections in the standardization of α-particle sources.     

Calculation of the spectra of fission neutrons emitted at angles of 0, 45, and 90° to the direction of flight of the fragments

The evaporation model is used to solve the problem of the spectrum of neutrons emitted by fission Fragments at angIes of 0, 45, and 90° to the direction of flight of the fragments. It is assumed that the neutrons are emitted isotropically in a coordinate system associated with the fragment. The spectra calculated are in good agreement with the experimental data.Translated from Atomnaya Énergiya, Vol. 13, No. 6, pp. 530–533, December, 1962     

Neutron spectra emitted in the fission of U235 at 0, 45, and 90 ° to the direction of motion of the fragments

Neutron spectra emitted at 0, 45, and 90 ° to the direction of motion of the fragments in the fission of U²³⁵ by 14.3 Mev neutrons have been measured. The angular distribution of the fission neutrons was found to be N(0 °): N(45 °): N(90 °) = (3.23 ± 0.12): (1.75 ± 0.07): 1.00. The experimental data are compared with spectra calculated on the assumption of an isotropic Maxwellianneutron distribution in the center-of-mass system of the fragment.     

On the transformation of ß-particle energies to electric energy in germanium crystals with p-n junctions

We present the results of an investigation performed in 1955 on the transformation of the energy of fast electrons into electric energy by germanium crystals with p-n junctions, as well as data on the efficiency of this transformation and its dependonce on the absorbed radiation flux and total dose. It is found that the basic factor reducing the efficiency of transformation is the increase of the recombination rate of electrons and holes on recombination levels of structural defects arising as a result of the irradiation. Changes in the equilibrium electric conductivity and carrier mobilities play a secondary role. A method is indicated for reestablishing the initial properties of crystals by heating them. Ionization curves are given for monoencrgetic electron beams in germanium.The authors express their sincere gratitude to P.Ia. Glazunov who took part in the experiments with the artificially accelerated electrons, and to P.N. Kodochigov for his valuable help in the preparation and examination of the Sr⁹⁰–Y⁹⁰ samples.     

Structural features of martensite α-phases in alloys of uranium with transition metals

The structures and formation kinetics are considered for martensite α phases in diffusionless transformation of the γ and β phases of alloys between uranium and transition metals. The general features of the {ie118-1} martensite phase and of the martensite in a steel have been determined by metallography, x-ray structure analysis, and dilatometry, and the same has been done for the martensite {ie118-2} and {ie118-3} phases and martensite in indium-thallium and gold-cadmium alloys. Kinetic features are given for the two-stage diffusionless {ie118-4} transformation in uranium-molybdenum and uranium-niobium alloys, in which the second stage {ie118-5} has the characteristics of a thermoelastic martensite transformation that is realized not only as the temperature varies but also in response to external stress. Bochvar All-Russia New Materials Research Institute, Russian Federation State Scientific Center. Translated from Atomnaya énergiya, Vol. 86, No. 2, pp. 113–120, February, 1999.     

Studies on unfolding energy spectra of neutrons using maximumlikelihood expectation–maximization method

Energy spectra of neutrons are important for identification of unknown neutron sources and for determination of the equivalent dose. Although standard energy spectra of neutrons are available in some situations, e.g.,for some radiotherapy treatment machines, they are unknown in other cases, e.g., for photoneutrons created in radiotherapy rooms and neutrons generated in nuclear reactors. In situations where neutron energy spectra need to be determined, unfolding the required neutron energy spectra using the Bonner sphere spectrometer(BSS) and nested neutron spectrometer(NNS) has been found promising. However, without any prior knowledge on the spectra, the unfolding process has remained a tedious task.In this work, a standalone numerical tool named‘‘NRUunfold' was developed which could satisfactorily unfold neutron spectra for BSS or NNS, or any other systems using similar detection methodology. A generic and versatile algorithm based on maximum-likelihood expectation–maximization method was developed and benchmarked against the widely used STAY'SL algorithm which was based on the least squares method. The present method could output decent results in the absence of precisely calculated initial guess, although it was also remarked that employment of exceptionally bizarre initial spectra could lead to some unreasonable output spectra. The neutron count rates computed using the manufacturer's response functions were used for sensitivity studies. The present NRUunfold code could be useful for neutron energy spectrum unfolding for BSS or NNS applications in the absence of a precisely calculated initial guess.     

Study on the performance of a large-size Cs I detector for high energy γ-rays

The measurement of high-energy γ-rays is an important experimental method to study the giant resonance in a nucleus, c reaction in nuclear astrophysics, and so on. The performance of a large-size Cs I detector for crays detection is studied by comparison between the experimental measurements and GEANT4 simulation. The reliability of the simulated efficiency for low-energy γ-rays is verified by comparing with the experimental data. The efficiency of the Cs I detector for high-energy γ-rays was obtained by the GEANT4 simulation. The simulation shows that the detection efficiency of 20 Me V γ-rays can reach 3.8%.     

Neutronic model of a mirror based fusion–fission hybrid for the incineration of the transuranic elements from spent nuclear fuel and energy amplification

The Georgia Institute of Technology has developed several design concepts of tokamak based fusion–fission hybrids for the incineration of the transuranic elements of spent nuclear fuel from Light-Water-Reactors. The present paper presents a model of a mirror hybrid. Concerning its main operation parameters it is in several aspects analogous to the first tokamak based version of a “fusion transmutation of waste reactor”. It was designed for a criticality k_eff ≤ 0.95 in normal operation state. Results of neutron transport calculations carried out with the MCNP5 code and with the JEFF-3.1 nuclear data library show that the hybrid generates a fission power of 3 GW_th requiring a fusion power between 35 and 75 MW, has a tritium breeding ratio per cycle of TBR_cycle = 1.9 and a first wall lifetime of 12–16 cycles of 311 effective full power days. Its total energy amplification factor was roughly estimated at 2.1. Special calculations showed that the blanket remains in a deep subcritical state in case of accidents causing partial or total voiding of the lead–bismuth eutectic coolant. Aiming at the reduction of the required fusion power, a near-term hybrid option was identified which is operated at higher criticality k_eff ≤ 0.97 and produces less fission power of 1.5 GW_th. Its main performance parameters turn out substantially better.     

Measurement of keV-neutron capture cross-sections and capture γ-ray spectra of Fe and Fe

The neutron capture cross-sections and the capture γ-ray spectra of ⁵⁶Fe and ⁵⁷Fe have been measured in the neutron energy range from 10 to 90 keV. Pulsed keV-neutrons were produced from the ⁷Li(p,n)⁷Be reaction by bombarding a lithium target with a 1.5-ns bunched proton beam from a 3 MV Pelletron accelerator. The incident neutron spectrum on the capture sample was measured using a time-of-flight method with a ⁶Li-glass detector. The capture γ-rays emitted from an iron or standard gold sample were detected with a large anti-Compton NaI(Tl) spectrometer. The capture yield of the iron or gold sample was obtained by applying a pulse-height weighting technique to the corresponding capture γ-ray pulse-height spectrum. The capture cross-sections of ^56,57Fe were derived with errors less than 5% using the standard capture cross-sections of ¹⁹⁷Au. The capture γ-ray spectra were obtained by unfolding the observed capture γ-ray pulse-height spectra. The present results for the capture cross-sections were compared with the previous measurements and the evaluated values of ENDF/B-VII.0 and JENDL-3.3. The Maxwellian-averaged capture cross-sections of ⁵⁶Fe and ⁵⁷Fe at 30 keV are derived as 12.22 ± 2.06 mb and 44.48 ± 7.56 mb, respectively.     

Setup of 4π solid angle β detector and measurement of its performance

This setup has very good timing response.Its rising time is just 8ns.By using β-γ coincident technique,the intensities of γ lines following electron capture from neutrondeficient nuclides are strongly suppressed.but the detecting efficiency to the γ lines following 100%β-decay of neutron-rich mercury nuclides arrives highly to 60%.     

Experimental test of a new neutron threshold detector and its application

The Possibility of using 209Bi as a new threshold detector to measure high-energy neutrons was investigated for the first time.At the same time the experiment measured successfully the emitted neutron fluence rate,energy spectrum and dose equivalent rate distributions in the heavy ion target area using a detector complex including 209Bi,^115In,^27Al,^19Fand ^12C samples.     

Design of detector to monitor the Bragg peak location of carbon ions by means of prompt γ-ray measurements with Geant4

Real-time monitoring of the Bragg peak location of carbon ions is urgently required for the quality control of hadron therapy. In this study, we design an annular detector to monitor the Bragg peak location of carbon ions with Geant4 simulation. This 360° surrounding structure has a high detection efficiency for the small-dose situation. The detector consists of a multilayered collimator system and an Na I scintillator for prompt gamma counting. The multilayered collimator includes a lead layer to prevent unwanted gammas and the paraffin and boron carbide layers to moderate and capture fast neutrons. An inclination of the detector further diminishes the background signal caused by neutrons. The detector, with optimized parameters, is applicable to carbon ions of different energies. In addition, the scintillator is replaced by an improved EJ301 organic liquid scintillator to discriminate gammas and neutrons. Inserting thin Fe slices into the liquid scintillator improves the energy deposition efficiency. The Bragg peak location of 200 Me V/u carbon ions can be monitored by prompt gamma detection with the improved liquid scintillator.     

The dependence of cumulative ^238U（n,f）fission yield on incident-neutron energy

This work is aim at studying the dependence of fission yields on incident neutron energy,so as to produce evaluated yield sets of the energy dependence.Experimental data at different neutron energies for gas fission products 85m,87,88Kr and 138Xe resulting from the 238U(n,f) reaction are processed using codes AVERAGE for weighed average and ZOTT for simultaneous evaluation.Energy dependence of the cumulative fission product yields on the incident neutron is presented.The evaluated curve of product yield is compared with the results calculated by the TALYS-0.64 code.The present evaluation is consistent with other main libraries in error permission.The fit curve of 87,88Kr can be recommended to predict the unmeasured fission yields.Comparisons of the evaluated energy dependence curves with theoretical calculated results show that the predictions using purely theoretical model for the fission process are not sufficiently accurate and reliable for the calculations of the cumulative fission yields for the 238U(n,f).