Relative Neutron Flux Research Articles

Solusi Persamaan Difusi Neutron Pada PWR (Pressurized Water Reactor) Berbentuk Heksagonal dengan Bahan Bakar Uranium Daur Ulang

The Research on solution of the neutron diffusion equation with a PWR reactor using recycled uranium fuel at 1⁄6 section of the reactor core with a hexagonal IGT-6 geometry. The purpose of this research is to determine the distribution of the neutron flux in the PWR of recycled uranium fuel. The solution is done by computational simulation using the Dev-C++ programming. The parameters used in this study determine the specifications of the reactor core, determine the volume fraction, determine the atomic density, calculate the macroscopic cross-section with the PIJ module, calculate the neutron diffusion equation, calculate ϕ (x,y) using the Gauss Seidel method. The results obtained in this study are the neutron diffusion equation without a source obtaining the highest relative neutron flux value in group 1 of 4,5729×〖10〗^(-2), with a fission source obtaining the highest relative neutron flux value in group 3 of 7,3327×〖10〗^(-4), with fission and scattering sources obtaining the highest relative neutron flux value found in group 2 of 1,5157×〖10〗^(-3), and 3,200 MW of power is added to the source fission, the value of the neutron flux does not change. This is because the addition of power does not affect the value of the neutron flux.

A Neutron Flow Monitor with a 14 MeV Energy of the Pulse Channel of the NG-12I Neutron Generator

A 14 MeV neutron flux monitor has been designed to configure the NG-12I neutron generator in pulsed mode and measure the relative 14-MeV neutron flux. The principle of operation of the pulsed channel of the neutron generator NG-12I is considered. The structural diagram of the monitor is given and the principle of its operation is described. The monitor allows real-time monitoring of the relative flux of 14-MeV neutrons and the background component due to the action of the neutron source in the time intervals between pulses. The results of the monitor readings are given.

A prototype system for dynamically polarized neutron protein crystallography

Abstract The sensitivity of Neutron Macromolecular Crystallography to the presence of hydrogen makes it a powerful tool to complement X-ray crystallographic studies using protein crystals. The power of this technique is currently limited by the relative low neutron flux provided by even the most powerful neutron sources . The strong polarization dependence of the neutron scattering cross section of hydrogen will allow us to use Dynamic Nuclear Polarization to dramatically improve the signal to noise ratio of neutron diffraction data, delivering order of magnitude gains in performance, and enabling measurements of radically smaller crystals of larger protein systems than are possible today. We present a prototype frozen spin system, built at Oak Ridge National Laboratory to polarize single protein crystals on the IMAGINE beamline at the High Flux Isotope Reactor (HFIR). Details of the design and construction will be described, as will the performance of the system offline and during preliminary tests at HFIR.

Calculation on the Distribution of Ex-Core Neutron Flux during Reactor Start-Up

The distribution of neutron flux is simulated by MCNP code from reactor start-up to criticality when the control rods are drawn different length from reactor core on ex-core detector area of the WWER reactor of TianWan. Because physical model built is very large, in order to save calculation time, the moving process of control rod is simplified. The results of calculation show that The neutron mainly distributed in the range of 0-400cm outside the pressure vessel. The value of the relative neutron flux ex-core is maximum in the range of 110cm to 180cm, so detection effect is better when the detector is set in this region.

Calculation of thermal neutron self-shielding correction factors for aqueous bulk sample prompt gamma neutron activation analysis using the MCNP code

In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF 3 detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample’s surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required.

Determination of neutron generation time in miniature neutron source reactor by measurement of neutronics transfer function

The prompt neutron generation time Λ and the total effective fraction of delayed neutrons (including the effect of photoneutrons) β have been experimentally determined for the miniature neutron source reactor (MNSR) of Syria. The neutron generation time was found by taking measurements of the reactor open-loop transfer function using newly devised reactivity-step- ejection method by the reactor pneumatic rabbit system. Small reactivity perturbations i.e. step changes of reactivity starting from steady state, were introduced into the reactor during operation at low power level i.e. zero-power. Relative neutron flux and reactivity versus time were obtained. Using transfer function analysis as well as least square fitting techniques and measuring the delayed neutrons fraction, the neutron generation time was determined to be 74.6±1.57 μs. Using the prompt jump approximation of neutron flux, the total effective fraction of delayed neutrons was measured and found to be 0.00783±0.00017. Measured values of Λ and β were found to be very consistent with calculated ones reported in the Safety Analysis Report.

Neutron reflectometry from solid-solid and solid-liquid interfaces at the TAS8 spectrometer at risø

Abstract In the last decade, neutron reflectometry has been established as a well-suited technique for probing the surface structure of stratified media along the surface normal. The large penetration depth for most materials and the large difference in scattering length of hydrogen and deuterium make neutrons ideally suited for investigations of soft condensed matter. The low absorption of neutrons in, e.g., single crystal quartz and silicon makes it possible to have liquid samples in thermostated enclosures and to study solid-liquid interfaces. The neutrons are able to transverse quartz and silicon crystals and can be reflected from the surface between the crystal and the liquid. One of the limitations in neutron reflectivity measurements is the relative low neutron flux, which has the consequence that a reasonable flux can only be obtained by relaxing the resolution of the reflectometer. For most samples this is not a serious limitation as the reflectivity curve of, e.g., polymers at a solid-liquid int...

The role of reactor characterization in NAA quality assurance

As a result of the NBSR upgrade to 20 MW, an extensive program of recharacterization of the irradiation facilities was undertaken. These measurements were made to (1) evaluate the neutron flux density and its variability over the irradiation volume, and (2) help establish safe irradiation conditions for samples in the NBSR. Techniques used to measure the following parameters are described: thermal neutron flux density; flux variations within an irradiation volume and between irradiations; relative fast neutron flux density; maximum temperatures generated by different sample matrices in various irradiation facilities; and pressures generated in sealed containers by radiolytic decomposition.

A collimated vanadium bath system for the relative measurement of keV neutron flux

A description is given of a collimated vanadium bath system which has been used to measure relative neutron flux in the keV energy region. The variation in the detection efficiency of the system with neutron energy has been studied using Monte Carlo methods. In particular the effective ‘nec’ size of the collimator and its position relative to the neutron source have been calculated as a function of neutron energy. The results of these calculations are compared with experimental measurements. The system has been used to measure the relative 0° yield from the 7 Li(p,n) 7 Be reaction and the results are compared with measurements of the yield made with a long counter. The development of these detectors is important for the accurate measurement of neutron flux.

A test and research reactor excursion radiation detection system using an on-line computer

A pulse reactor radiation detection system is described. The system uses an on-line Scientific Data Systems (SDS) 920 computer, coupled through an analog to digital converting interface, to record and retain the response data for various radiation detectors. Results of typical measurements are presented. An overall range of sensitivity of 10 6 is possible with a precision to within about 5% (exclusive of the detectors) for determining relative reactor power, neutron and gamma ray flux densities (if the spectra remain constant), and neutron and gamma ray tissue kerma rates or the time integral of these quantities. This range is achieved using combinations of radiation sensors.

The neutron capture cross section of gold in the keV energy region

The capture cross-section shape of gold has been measured in the energy range 25–500 keV and normalized at 30 keV using a ‘best fit’-value. The shape measurement was performed using a large liquid scintillator as a capture γ-ray detector and by means of a grey neutron detector which has a flat efficiency curve to measure the relative neutron flux. The results agree with recent measurements of Harris et al. but disagree in the energy range above 130 keV with all measurements performed relative to the fission cross section of 235U.

Absolute Radiative Capture Cross Section for Fast Neutrons in 238U

The capture cross section of 238U has been measured absolutely at a neutron energy of 30 keV using kinematically collimated neutrons from the 7Li(p, n) 7Be reaction near threshold. Activation techniques were used to determine both the number of capture events and the number of neutrons that occurred during the irradiation. The result of the 238U capture cross section measurement is 479 ± 14 mb at 30 keV. In addition, the shape of the 238 U capture cross section has been measured for neutron energies from 25 to 500 keV using neutrons from the 7Li(p, n)7 Be reaction. The capture reactions in the 238 U target were detected using a large liquid scintillator tank and time-of-flight techniques. The relative neutron flux was measured using a flat response neutron detector. The cross-section shape measurement was normalized to the present absolute measurement at 30 keV. The present measurement has been compared with several measured values, theoretical calculations, and compiled values of the 238U capture cross section as given by other authors.

A “grey” neutron detector for the intermediate energy region

A neutron detector for the accurate measurement of relative neutron flux intensities in the keV energy region is described. The detector consists, in its simples arrangement, of a paraffin pile and a NaI(Tl) detector to detect the 2.2 MeV capture γ-rays. The efficiency curve is relatively flat in the considered energy range. The efficiency curve depends only slightly on a change in the parameters used for its calculation. The application of the detector in neutron cross section experiments is described.

The Use of Nuclear Emulsions to Estimate Stray Neutron Spectra Near Particle Accelerators

The stray neutron field near a particle accelerator generates proton recoil tracks, of various lengths and orientations in photographic emulsions. The lengths are directly convertible to proton kinetic energies by use of five formulae. Three of these are least-square, straight-line fits of the experimental data of LATTES, FOWLER and CUER(1) while the other two are the empirical equations of BRADNER, SMITH, BARKAS and BISHOP(2) and W. H. BARKAS.(3) If the proton recoils are counted over all directions, then an analysis will show that the probability that a neutron of energy, En, will cause a proton to recoil with an energy, E < En, is σ/En. Therefore, the relative neutron flux density per unit energy interval per hydrogen atom, N(E), is related to the number of recoil protons per gram of absorbing material per unit energy interval, P(E), by: N(E) = −E/σ × dP(E)/dE where E is the incident neutron energy in MeV and σ is the n-p cross section for hydrogen at energy E. Because the proton track can only be counted if it both begins and ends in the emulsion, it is necessary to correct for the probability, increasing with energy, that a proton track will escape the emulsion. A correction based on the geometry of the emulsion has been derived.(4) Having established the proton distribution, as shown in Fig. 1, by measurement and correction for protons escaping the emulsion, the Gregory-Newton forward difference formula was used to determine dP/dE.(5) The initial use of this technique was to measure the stray neutron spectrum with Ilford G-5 emulsions, near the Brookhaven National Laboratory Cosmotron. The resulting neutron spectrum is exhibited as Fig. 2.

Calibration of manganese and gold foils for relative neutron flux measurements

The usual method of intercalibrating foils for relative flux measurements, made by comparing the β- activities of the foils after exposure to the same neutron flux, is compared with a method dependent on weighing the foils. The mass intercalibrations agree to within ± 1% of the activity intercalibrations for gold and manganese foils in the range 90 to 100 mg/cm2 thick, if a theoretical correction estimating the β- absorption in the foil is applied to the measured mass values.

Energy Dependence of Fast-Neutron Activation Cross Sections

Fast-neutron capture cross sections of 24 nuclides ranging from $A=51$ to $A=197$ have been measured by an activation method, in the neutron energy region from 0.15 to 6.2 Mev. The neutron energy spreads were of the order of 0.1 Mev so that cross sections averaged over many energy levels of the compound nucleus were measured. Activities induced in samples by fast and thermal neutrons were compared. The relative neutron flux in the fast- and thermal-neutron activations was determined with a ${\mathrm{U}}^{235}$ fission counter. A knowledge of the energy dependence of the ${\mathrm{U}}^{235}$ fission cross section and of the thermal-neutron activation cross sections allows calculation of the fast-neutron activation cross sections.