Mean Neutron Energy Research Articles

Neutron spectrum measurements near KAMINI reactor south beam vault door using nested neutron spectrometer.

KAlpakkam MINI reactor (KAMINI) is a 233U fuelled research reactor has various neutron irradiation locations for experimental purposes. The pit at the south beam end of KAMINI reactor is being extensively utilised for neutron attenuation experiments in prospective shielding materials as well as for neutron radiography. During reactor operation, it will be closed by a movable shield. A vault door is located above the shield and the movable shield is used to attenuate streaming neutrons and gamma-rays during reactor operation. Even with the shield, there exists significant dose because of streaming neutrons and gamma rays. Its variation depends on the power of the reactor. The neutron and gamma dose rates close to the south beam vault door have recently been found to be 275-300μSv/h and 175-200μSv/h, respectively, when the reactor is operating at 10kW. In order to characterise the streaming neutron spectra of vault door place for the first time, measurements are done using the Nested Neutron Spectrometer. Along with the neutron flux, neutron mean energy and ambient dose-equivalent rate are also measured and compared with earlier measurements carried out inside the south beam pit. It is observed that the presence of paraffin shield reduces the neutron average energy from 370 to 178keV. Apart from energy reduction, 10kW normalised neutron flux of south beam pit is also attenuated by the shield by 25000 times and it is found that the neutron spectrum of the measured location is also more thermalized. Neutron reference data of the location are generated.

Mechanical and Architectural Changes in Animal Bone Following Fast Neutron Irradiation.

Damage to healthy bone following exposure to ionizing radiation has been well documented for at least seven decades. Among the reported effects are a transient increase in stiffness and a reduction in breaking strength. These changes have been linked to a decrease in osteoblast proliferation and differentiation, inducing cell cycle arrest, reducing collagen production, and increasing sensitivity to apoptotic agents. In this work, we analyzed some mechanical and structural changes in compact costal bovine bone (Hereford breed, n = 9) subjected to escalating doses of fast neutrons from a 7 Li(p,n) 7 Be reaction. The mean neutron energy was 233 keV with calculated absorbed doses ranging from 0 to 4.05 ± 10% Gy. Samples were subjected to Young's Modulus (YM) and breaking strength testing with a Universal Testing Machine (UTM). We found an increase in Young's Modulus and a decrease in breaking strength as functions of increasing dose equivalent. Optical coherence tomography (OCT) revealed trabecular displacement into compact bone in an irradiated sample (D = 4.05 ± 10% Gy), with breaching of the endosteal wall. OCT further revealed a "crack-like" structure across the irradiated sample, potentially consistent with damage from a proton track resulting from an elastic (n,p) reaction. No previous report has been found on mechanical changes in large mammalian bones following fast neutron doses, nor of the OCT imaging of such samples.

Neutronics for fast spectrum, molten salt reactors and evolution over burnup

There are a wide variety of fast spectrum molten salt reactor (FSMSR) designs and corresponding fuel cycles that shown promise. These designs differ in features. Therefore, it becomes difficult to determine whether a set of parameters is applicable to all, one, or none of these designs. This paper presents a number of iterations of a FSMSR concept with a sodium chloride carrier salt. Different design characteristics and their effects on the neurotics parameters are shown. The results show that effective delayed neutron fraction, mean neutron generation time, neutron yield, and energy of neutrons causing a fission all follow trends based off the heavy metal concentration and leakage rate. This paper also gives preliminary fuel cycle calculations and the evolution of the neutronic parameters over burnup. A FSMSR initially fueled with low enriched uranium will naturally breed at a certain core volume, without a need for refueling, for decades of full power operation.

Cumulative fission yield measurements with 14.7 MeV neutrons on 238U* *Supported by the National Natural Science Foundation of China (11975113)

The fission yield data in the 14 MeV energy neutron induced fission of 238U play an important role in decay heat calculations and generation-IV reactor designs. In order to accurately measure fission product yields (FPYs) of 238U induced by 14 MeV neutrons, the cumulative yields of fission products ranging from 92Sr to 147Nd in the 238U(n, f) reaction with a 14.7 MeV neutron were determined using an off-line γ-ray spectrometric technique. The 14.7 MeV quasi-monoenergetic neutron beam was provided by the K-400 D-T neutron generator at China Academy of Engineering Physics (CAEP). Fission products were measured by a low background high purity germanium gamma spectrometer. The neutron flux was obtained from the 93Nb (n, 2n)92mNb reaction, and the mean neutron energy was calculated using the cross-section ratios for the 90Zr(n, 2n)89Zr and 93Nb(n, 2n)92mNb reactions. With a series of corrections, high precision cumulative yields of 20 fission products were obtained. Our FPYs for the 238U(n, f) reaction at 14.7 MeV were compared with the existing experimental nuclear reaction data and evaluated nuclear data, respectively. The results will be helpful in the design of a generation-IV reactor and the construction of evaluated fission yield databases.

An improved theoretical method for determining the mean D-T neutron energies in large solid angle sample

ABSTRACT Neutron energy of deuterium-tritium (D-T) generator plays an important role for the application of nuclear energy and nuclear technology. Based on the deuterium energy losing in tritiated titanium (T-Ti) target and differential cross sections of D-T reaction, an improved theoretical formula for calculating the mean neutron energy irradiating a large solid angle sample was proposed. Besides that, the mean neutron energy was discussed under the actual experimental conditions. The results were compared to the reported ratio values of 90Zr(n,2 n)89m+gZr to 93Nb(n,2 n)92mNb reaction cross sections (‘Zr/Nb’ ratio methods) and published theoretical calculated results, respectively. The relative deviations between present work and the ‘Zr/Nb’ ratio methods are below 0.5% and a higher precision was obtained compared with previous published theoretical method. The present work lays a foundation for the application of D-T neutron source in nuclear technology such as nuclear reaction cross section measurement, calibration of neutron detector, and neutron irradiation tests, etc.

Measurement of 54Fe(n,2n)53Fe reaction cross-sections at 14 MeV neutron energies using activation method and its relevance for fusion-plasma temperature diagnosis

Iron as a main structural material in fusion facilities, its activation reaction cross-sections induced by 14 MeV neutron is of great significance. In order to solve the problem of insufficient accuracy of 54Fe(n,2n)53Fe reaction as temperature monitor in fusion reactors, the 54Fe(n,2n)53Fe reaction cross section was measured based on K-400 D-T neutron generator at China Academy of Engineering Physics (CAEP). Three natural abundance iron samples sandwiched with zirconium and niobium foil were irradiated at three angles for 50 min. Neutron activated product, 53Fe, was measured by off-line gamma ray method with a high purity germanium detector at CAEP. The neutron flux was obtained from 56Fe (n,p)56Mn as the standard monitor reaction, and the mean neutron energy was calculated by Q formula. As a result, the cross sections at 14.51 MeV, 14.67 MeV and 14.83 MeV neutron energies are (5.40 ± 0.61) mb, (6.96 ± 0.72) mb and (10.30 ± 0.94) mb, respectively. The experimental results were compared with the previous experimental data and evaluation database. The present findings show a good agreement with the ENDF/B-VIII.0 and TENDL-2019 and with partial existing experimental results. As a temperature monitor in fusion reactors, a new ion temperature curve with an accuracy of (11–17)% was proposed based on the present activation results of 54Fe (n, 2n)53Fe reaction.

Feasibility of 18-MV grid therapy from radiation protection aspects: unwanted dose and fatal cancer risk caused by photoneutrons and scattered photons

PurposePhotoneutron production is a common concern when using 18-MV photon beams in radiation therapy. In Spatially Fractionated Grid Radiation Therapy (SFGRT), the grid block in the collimation system modifies the neutron production, photon scattering, and electron contamination in and out of the radiation field. Such an effect was studied with grids made of different high-Z materials by Monte Carlo simulations. The results were also used to evaluate the lifetime risk of fatal cancers. MethodsMCNPX® code (2.7.0 extensions) was employed to simulate an 18-MV LINAC (Varian 2100 C/D). Three types of grid made of brass, cerrobend, and lead were used to study the neutron and electron fluence. Output factors for each grid with different field sizes were calculated. A revised female MIRD phantom with an 8-cm spherical tumor inside the liver was used to estimate the dose to the tumor and the critical organs. A 20-Gy SFGRT plan with Anterior Posterior (AP) - Posterior Anterior (PA) grid beams was compared with a Conventional Fractionated Radiation Therapy (CFRT) plan which delivered 40-Gy to the tumor by AP-PA open beams. Neutron equivalent dose, photon equivalent dose, as well as lifetime risks of fatal cancer were calculated in the organs at risk. ResultsThe grid blocks reduced the fluence of contaminant electrons inside the treatment field by more than 50%. The neutron fluences per electron-history in SFGRT plans with brass, cerrobend and lead were on average 55%, 31% and 31% less than that of the CFRT plan, respectively. However, when converting to fluences per delivered dose (Gy), the cerrobend and lead grid may incur higher neutron dose for 20 × 20 cm2 field size and above. The changes in neutron mean energy, as well as the correlated radiation weighting factors, were insignificant. The total risk due to the photoneutrons in the SFGRT plans was 87% or lower than that in the CFRT plans. In both SFGRT and CFRT plans, the contribution of the primary and scattered photons to the fatal cancer risk was 2 times or more than the photoneutrons. The total risks from photons in SFGRT with brass, cerrobend, and lead blocks were 1.733, 1.374, and 1.260%, respectively, which were less than 30% of the total photon-risk in CFRT (5.827%). ConclusionIn the brass, cerrobend, and lead grids, the attenuation of photoneutrons outweighs its photoneutron production in 18-MV SFGRT. The total cancer risks from photons and photoneutrons in the SFGRT plans were 30% or less of the risks in the CFRT plans (5.911%). Using 18 MV photon beams with brass, cerrobend, and lead grid blocks is still a feasible option for SFGRT.

Measurements of the cross-section of 111Cd(n,n′)111mCd reaction for 241Am/Be neutrons

The cross-section of the [Formula: see text] reaction was measured with [Formula: see text] neutrons using a natural cadmium target [Formula: see text]. The neutron fluence and mean neutron energy of the source were determined using the ISO 8529-1 neutron spectrum and the known cross-sections of the monitor reaction [Formula: see text]. In order to measure the poor [Formula: see text]-ray activity of the reaction products, an HPGe detector with 70% detection efficiency surrounded by an adequate graded shield was applied. The efficiency calculations for the detector were performed using standard point calibration sources and the EFFTRAN efficiency code. Using the measured values of the neutron flux and the induced [Formula: see text]-ray activity of [Formula: see text], the cross-section of the [Formula: see text] reaction at the average neutron energy of 4.05 MeV was found to be [Formula: see text] mb. An estimation of the contribution to the total cross-section by the accompanied reactions [Formula: see text] and [Formula: see text] was achieved and the related cross-sections were found to be 0.16 mb and 8.99 mb, respectively.

Bragg-curve simulation of carbon-ion beams for particle-therapy applications: A study with the GEANT4 toolkit

We used the GEANT4 Monte Carlo MC Toolkit to simulate carbon ion beams incident on water, tissue, and bone, taking into account nuclear fragmentation reactions. Upon increasing the energy of the primary beam, the position of the Bragg-Peak transfers to a location deeper inside the phantom. For different materials, the peak is located at a shallower depth along the beam direction and becomes sharper with increasing electron density NZ. Subsequently, the generated depth dose of the Bragg curve is then benchmarked with experimental data from GSI in Germany. The results exhibit a reasonable correlation with GSI experimental data with an accuracy of between 0.02 and 0.08 cm, thus establishing the basis to adopt MC in heavy-ion treatment planning. The Kolmogorov-Smirnov K–S test further ascertained from a statistical point of view that the simulation data matched the experimentally measured data very well. The two-dimensional isodose contours at the entrance were compared to those around the peak position and in the tail region beyond the peak, showing that bone produces more dose, in comparison to both water and tissue, due to secondary doses. In the water, the results show that the maximum energy deposited per fragment is mainly attributed to secondary carbon ions, followed by secondary boron and beryllium. Furthermore, the number of protons produced is the highest, thus making the maximum contribution to the total dose deposition in the tail region. Finally, the associated spectra of neutrons and photons were analyzed. The mean neutron energy value was found to be 16.29 MeV, and 1.03 MeV for the secondary gamma. However, the neutron dose was found to be negligible as compared to the total dose due to their longer range.

Essential considerations for accurate evaluation of photoneutron contamination in Radiotherapy

Nowadays, high-energy X-rays produced by medical linear accelerators (LINACs) are widely used in many Radiation Therapy (RT) centers. High-energy photons (> 8 MeV) produce undesired neutrons in the LINAC head which raise concerns about unwanted neutron dose to the patients and RT personnel. Regarding the significance of radiation protection in RT, it is important to evaluate photoneutron contamination inside the RT room. Unfortunately, neutron dosimeters used for this purpose have limitations that can under the best conditions cause to > 10% uncertainty. In addition to this uncertainty, the present Monte Carlo (MC) study introduces another uncertainty in measurements (nearly up to 20%) when neutron ambient dose equivalent (Hn*(10)) is measured at the patient table or inside the maze and the change in neutron energy is ignored. This type of uncertainty can even reach 35% if Hn*(10) is measured by dosimeters covered by a layer of 10B as converter. So, in these cases, neglecting the change in neutron energy can threaten the credibility of measured data and one should attend to this energy change in order to reduce measurement uncertainty to the possible minimum. This study also discusses the change in neutron spectra and Hn*(10) at the patient table caused by removing a typical RT room from MC simulations. Under such conditions, neutron mean energy (Ēn) overestimated by 0.2–0.4 MeV at the patient table. Neutron fluence (φn) at the isocenter (IC) was underestimated by 23–54% for different field sizes that caused Hn*(10) to be miscalculated up to 24%. This finding informs researchers that for accurate evaluation of Hn*(10) at the patient table, simulating the RT room is an effective parameter in MC studies.

Calibration of a neutron time-of-flight detector with a rapid instrument response function for measurements of bulk fluid motion on OMEGA.

A newly developed neutron time-of-flight (nTOF) diagnostic with a fast instrument response function has been fielded on the OMEGA laser in a highly collimated line of sight. By using a small plastic scintillator volume, the detector provides a narrow instrument response of 1.7 ns full width at half maximum while maintaining a large signal-to-noise ratio for neutron yields between 1010 and 1014. The OMEGA hardware timing system is used along with an optical fiducial to provide an absolute nTOF measurement to an accuracy of ∼56 ps. The fast instrument response enables the accurate measurement of the primary deuterium-tritium neutron peak shape, while the optical fiducial allows for an absolute neutron energy measurement. The new detector measures the neutron mean energy with an uncertainty of ∼7 keV, corresponding to a hot-spot velocity projection uncertainty of ∼12 km/s. Evidence of bulk fluid motion in cryogenic targets is presented with measurements of the neutron energy spectrum.

Velocity correction for neutron activation diagnostics at the NIF.

The velocity distribution of the hotspot in an inertial confinement fusion implosion changes the energy spectra of fusion neutrons emitted from the experiment as a function of viewing angle. These velocity-induced spectral changes affect the response of neutron activation diagnostics (NADs) positioned around the experiment and must be accounted for to correctly extract information about areal density (ρR) asymmetry from the data. Three mechanisms through which average hotspot velocity affects NAD activation are addressed: change in activation cross section due to the Doppler shift of the mean neutron energy, kinematic focusing of neutron fluence, and change in the scattering cross section due to the Doppler shift. Using the hotspot velocity inferred from neutron time-of-flight measurements of D-T and D-D fusion neutrons, the hotspot velocity is shown to account for the observed NAD activation asymmetry in a calibration shot with negligible fuel ρR. A robust method to evaluate uncertainties in spherical-harmonic fits to the NAD data due to the velocity correction and detector uncertainty is discussed.

A fused silica Cherenkov radiator for high precision time-of-flight measurement of DT and neutron spectra (invited).

A fused silica Cherenkov radiator has been implemented at the National Ignition Facility to provide a new high precision measurement of the time-of-flight spectrum of 14.1 MeV DT fusion neutrons. This detector enables a high precision (<30 ps) co-registered measurement of both a thresholded γ-ray and a neutron spectrum on a single record. Other methods typically require γ and neutron signals to be co-registered via other diagnostics and/or dedicated timing experiments. Analysis of the co-registered γ and neutron signals allows precise extraction of the mean neutron energy and bulk hot-spot velocity, both of which were not possible with prior scintillator technologies. Initial measurements demonstrate the feasibility of this measurement and indicate that combined detection of neutrons and γ-rays on multiple lines-of-sight should enable the bulk vector velocity of the implosion hot-spot to be determined to ≈5 km/s and reduced uncertainty in the spectral width ≈0.1 keV.

Neutron and gamma flux distributions and their implications for radiation damage in the shielded superconducting core of a fusion power plant

The neutron and gamma ray fluxes within the shielded high-temperature superconducting central columns of proposed spherical tokamak power plants have been studied using the MCNP Monte-Carlo code. The spatial, energy and angular variations of the fluxes over the shield and superconducting core are computed and used to specify experimental studies relevant to radiation damage and activation. The mean neutron and gamma fluxes, averaged over energy and angle, are shown to decay exponentially through the shield and then to remain roughly constant in the core region. The mean energy of neutrons is shown to decay more slowly than the neutron flux through the shield while the gamma energy is almost constant around 2 MeV. The differential neutron and gamma fluxes as a function of energy are examined. The neutron spectrum shows a fusion peak around 1 MeV changing at lower energies into an epithermal E−0.85 variation and at thermal energies to a Maxwellian distribution. The neutron and gamma energy spectra are defined for the outer surface of the superconducting core, relevant to damage studies. The inclusion of tungsten boride in the shield is shown to reduce energy deposition. A series of plasma scenarios with varying plasma major radii between 0.6 and 2.5 m was considered. Neutron and gamma fluxes are shown to decay exponentially with plasma radius, except at low shield thickness. Using the currently known experimental fluence limitations for high temperature superconductors, the continuous running time before the fluence limit is reached has been calculated to be days at 1.4 m major radius increasing to years at 2.2 m. This work helps validate the concept of the spherical tokamak route to fusion power by demonstrating that the neutron shielding required for long lifetime fusion power generation can be accommodated in a compact device.

Few groups neutron spectra, and dosimetric features, of isotopic neutron sources

Using Monte Carlo methods, the neutron spectra in 31 energy groups of several isotopic neutron sources were estimated. For each source the neutron mean energy; the Ambient dose equivalent, the Personal dose equivalent and the Effective dose per unit fluence rate were calculated. A convenient way to produce neutrons is the isotopic neutron source, where the production is through (α, n), (γ, n), and spontaneous fission reactions. Isotopic neutron sources are small, easy to handle, and have a relative low cost. On the other hand the neutron yield is small and mostly of them produces neutrons with a wide energy distribution. In this work, the main features of 24NaBe, 24NaD2O, 116InBe, 140LaBe, 238PuLi, 239PuBe, 241AmB, 241AmBe, 241AmF, 241AmLi, 242CmBe, 210PoBe, 226RaBe, 252Cf and 252Cf/D2O isotopic neutron source are also compiled.

A method for determining d–D neutron energies in a large sample

The energy of monoenergetic neutrons generated by the D(d,n)3He reaction was determined as a function of emergent angle and incidence energy of d+-beam, En(θ,Ed). Based on the geometric size of the experimental sample, position of the sample relative to the Ti–D solid or D2 gas targets, volume distribution of D2 gas targets, theoretical formulas were obtained for calculating the mean neutron energy required to irradiate a large sample. Using these formulas, the mean neutron energies of the Ti–D solid and D2 gas targets irradiating a large sample were calculated under various conditions. The results were compared to those reported in the literature. The formulas obtained in this study were found to be applicable for the determination of mean neutron energy irradiating a large sample for the Ti–D solid and D2 gas targets.

Uncertainty in the Fast Neutron Fluence in Function of the Fission Source and Its Burnup Dependence

The fast neutron fluence is an important parameter for the reactor pressure vessel (RPV) lifetime. The uncertainty estimation of this parameter is crucial to manage the RPV with a suitable safety margin. This work focuses on a facet of the problem that concerns the uncertainty contribution of the spectrum of the fission source for different burnups in a thermal neutron reactor. The main goal of this paper is to assess the effect of a possible uncertainty correlation among the spectra of the fissile nuclei, involved in the fission source, on the response uncertainty. Two main simplifications are assumed to reduce the complexity of the problem. The first simplification concerns the geometry of the transport problem that is chosen to calculate as fast as possible the sensitivities and the different responses. The second simplification is related to the way by which one can take into account the correlations among spectra of different fissile nuclei. Simple ENDF-6 models of the fission spectrum (Maxwell, Watt, and simplified Madland-Nix) are used to define correlations among the fissile spectra through the mean neutron energy of the prompt fission spectrum. Results are given to quantify the effect of these postulated correlations on response uncertainties and are compared to the ones using JENDL-4.0 covariances.

A compact neutron spectrometer for characterizing inertial confinement fusion implosions at OMEGA and the NIF.

A compact spectrometer for measurements of the primary deuterium-tritium neutron spectrum has been designed and implemented on the OMEGA laser facility [T. Boehly et al., Opt. Commun. 133, 495 (1997)]. This instrument uses the recoil spectrometry technique, where neutrons produced in an implosion elastically scatter protons in a plastic foil, which are subsequently detected by a proton spectrometer. This diagnostic is currently capable of measuring the yield to ~±10% accuracy, and mean neutron energy to ~±50 keV precision. As these compact spectrometers can be readily placed at several locations around an implosion, effects of residual fuel bulk flows during burn can be measured. Future improvements to reduce the neutron energy uncertainty to ±15-20 keV are discussed, which will enable measurements of fuel velocities to an accuracy of ~±25-40 km/s.

Generation and anisotropy of neutron emission from a condensed Z-pinch

The paper presents results of measurements of neutron emission generated in the constriction of a fast Z-pinch at the S-300 facility (2 MA, 100 ns). An increased energy concentration was achieved by using a combined load the central part of which was a microporous deuterated polyethylene neck with a mass density of 100 mg/cm3 and diameter of 1–1.5 mm. The neck was placed between two 5-mm-diameter agar-agar cylinders. The characteristics of neutron emission in two axial and two radial directions were measured by the time-of-flight method. The neutron spectrum was recovered from the measured neutron signals by the Monte Carlo method. In all experiments, the spatiotemporal characteristics of plasma in the Z-pinch constriction were measured by means of the diagnostic complex of the S-300 facility, which includes frame photography in the optical, VUV, and soft X-ray (SXR) spectral regions; optical streak imaging; SXR detection; and time-integrated SXR photography. The formation of hot dense plasma in the Z-pinch constriction was accompanied by the generation of hard X-ray (with photon energies E > 30 keV), SXR (with photon energies E > 1 keV and duration of 2–4 ns), and neutron emission. Anisotropy of the neutron energy distribution in the axial direction was revealed. The mean neutron energies measured in four directions at angles of 0° (above the anode), 90°, 180° (under the cathode), and 270° with respect to the load axis were found to be of 2.1 ± 0.1, 2.5 ± 0.1, 2.6 ± 0.2, and 2.4 ± 0.1 MeV, respectively. For a 1-mm-diameter neck, the maximum integral neutron yield was 6 × 109 neutrons. The anisotropy of neutron emission for a Z-pinch with a power-law distribution of high-energy ions is calculated.

32P measurement and dose conversion factor evaluation of activated human hair by criticality accident

In order to conduct dose assessment of victims in criticality accidents, a method of fast neutron capture-activated (32)P measurement of hair in which samples are treated by a chemical and analytical procedure that takes 9 h and measurement is conducted by liquid scintillation counting is presented. To validate this measurement method, hair samples spiked with a (32)P reference source were measured and the results analysed and the optimal sample mass and detection efficiency were determined. To verify the correlation between (32)P-specific activity and absorbed dose for spectra with two neutron mean energies, samples collected from three normal individuals were irradiated at various neutron energies and irradiation times using the MC50 Cyclotron of the Korea Institute of Radiological and Medical Sciences. The (32)P-specific activity trend of the irradiated hair agreed well with the absorbed doses. Based on the results, dose conversion factors, which were 0.67 ± 0.15 and 0.59 ± 0.06 Gy (Bq g(-1))(-1) at neutron mean energies of 2.33 and 5.36 MeV, respectively, were calculated as a guide for medical treatment of criticality accident victims.