Neutron Source Spectrum Research Articles

The contribution of the air and room scattered neutrons in the PTB calibration facility

Neutron detectors and dosimeters used for radiation protection purposes are commonly calibrated with radioisotope-based neutron sources in dedicated calibration facilities. Due to interactions of the neutrons emitted by the sources with components of the calibration room, the neutron fluence at the measuring point is a sum of several components. As the results of calibration procedures should ideally be independent of the environment in which the measurements are performed, the effects caused by neutrons which are not directly reaching the device under test need to be considered, i.e. the influence of scattered neutrons needs to be determined and corrections have to be made. The contribution of the neutrons scattered from the air and the room to the fluence and dose must be taken into consideration, especially in the case of small and medium sized calibration rooms. The scattered neutron component can be evaluated numerically provided a detailed model of the calibration facility is available. In this work, a realistic MCNP6 model of the calibration room at the Physikalisch-Technische Bundesanstalt (PTB) has been developed in order to evaluate carefully the contribution of scattered neutrons from the air, wall, and equipment in the irradiation room for 252Cf, 241Am–Be and D2O moderated 252Cf (with and without Cd-shielding) neutron sources at different reference points and with and without shadow objects. The calculation also considers new fission spectral data for the neutron reference field and recent cross section data libraries. The results show that the contribution of air out-scattered neutrons depends on the detector position and on the neutron source spectrum. This effect is particularly significant for the moderated 252Cf source. More than a third of the total scored neutrons come from the walls and ceiling and the total contributions of the scattered neutrons to the total fluence at the reference position 170 cm away from the source are 50.6 %, 46.6 %, 55.0 % and 52.5 % for 252Cf, 241Am–Be, D2O-252Cf–Cd and D2O-252Cf sources respectively.

Simulation study of a rhodium self-powered neutron detector for irradiation of nuclear fuel and material reactor

To accurately judge the irradiation test cycle of nuclear fuel and material in-pile irradiation, and provide precise design inputs for the post-irradiation performance evaluation of fuel and material, the NIT-Rh self-powered neutron detector has been designed and developed. The sensitivity characteristics of the detector have been simulated and verified. This study introduces the basic structure and working principle of the rhodium self-powered neutron detector for irradiation of nuclear fuel and materials in the reactor. Additionally, a new computational model has been constructed by the Monte Carlo method, and the current sensitivity of different sizes of emitters and insulators has been analyzed. The model considers a more realistic neutron source and β-electron energy spectrum, accounts for the space charge effect on the signal in the insulator, and compares the simulated sensitivity with the reactor test sensitivity. The results show that the new rhodium self-powered neutron detector calculation model is more accurate, and the relative deviation between the simulated and reactor test sensitivities is only 3.06 %. It can be used for the in-heap irradiation test and provides a method for the application of the irradiation of nuclear fuel and material reactors.

A feasibility study of a H2O-moderated 252Cf source for metrology

The new edition of ISO 8529 series reduced from 4 to 3 standard neutron reference field. Recent studies have highlighted the increasing importance of characterizing neutron source spectra at different energies to obtain appropriate fluence-to-dose-equivalent conversion coefficient. This study presents the determination of neutron energy fluency from a free and moderated 252Cf (120 µg) source placed within spheres of different diameters containing distilled water. The spectrum determination is based on both Monte Carlo simulations and experimental measurements with Bonner Sphere Spectrometer (BSS) of multiple diameters: 5.08 cm (2"), 7.62 cm (3”), 12.70 cm (5”), 20.32 cm (8”), 25.40 cm (10”), and 30.48 cm (12”). The aim was to characterize the new water-moderated spectra, obtaining reference values for operational quantities that represent realistic radiation fields for testing, calibrating, and irradiation of personal dosimetry and survey meters.

New method for unfolding neutron source spectra

For solving the Fredholm equation of the first kind, the domain can be discretized in M spheres and N energy bins using a Bonner multisphere spectrometer, in which the neutron flux is to be deduced from the response matrix Rij(E), and the count rate obtained for each sphere Ci. For this study we use an approach that introduces certain neutron energy indeterminations, which are difficult to account for if some absolute calculations are required to be performed. For M different from N, a problem arises after obtaining the deconvolution of this equation, because the extracted solution is not unique, even if a host neutron energy spectrum is used. We propose a different method for extracting neutron energy spectra from the count rate obtained using a Bonner spectrometer. The method consists of partitioning the response matrix into a number of matrices such that M is equal to N, and depends on the response matrix and the number of spheres under consideration. Thus, the overall solution of the Fredholm equation can be divided and applied to N/M segments with M equations, and M unknowns, which can be solved in every energy segment for the neutron energy spectra. With this, the neutron spectra can be recovered by combining successive energy intervals as will be thoroughly described in the paper.

Measuring and unfolding fast neutron spectra using solution-grown trans-stilbene scintillation detector

We propose an overall procedure for measuring and unfolding fast neutron spectra using a trans-stilbene scintillation detector. Detector characterization was described, including the information on energy calibration, detector resolution, and nonproportionality response. The digital charge comparison method was used for the investigation of neutron-gamma Pulse Shape Discrimination (PSD). A pair of values of 600 ns pulse width and 24 ns delay time was found as the optimized conditions for PSD. A fitting technique was introduced to increase the trans-stilbene Proton Response Function (PRF) by 28% based on comparison of the simulated and experimental electron-equivalent distributions by the Cf-252 source. The detector response matrix was constructed by Monte-Carlo simulation and the spectrum unfolding was implemented using the iterative Bayesian method. The unfolding of simulated and measured spectra of Cf-252 and AmBe neutron sources indicates reliable, stable and no-bias results. The unfolding technique was also validated by the measured cosmic-ray induced neutron flux. Our approach is promising for fast neutron detection and spectroscopy.

Analytical evaluation of geometry correction factor and effectiveness parameter for BSS irradiated by a voluminous neutron source

The deviation in reading due to inverse square law is an important issue in the calibration of neutron detection systems. The geometry correction factor of a neutron-measuring device is a unique property of this type of device, and may depend on the dimensions, shape of the source and the detector, the distance between them and the neutron source spectrum. This study deals with the geometric analysis of angular distribution for landing neutrons from a non-point source on a Bonner sphere spectrometer (BSS) and solid angles subtended by detectors from distributed source. Using mathematical relations, an analytical function was introduced to calculate the geometry correction factor and the effectiveness parameter. By comparing the analytical and Monte Carlo values of geometry factors, it was found that the effectiveness parameter (δ′) for BSS depends on the source radius, the BSS radius and the energy of the source (energy itself depends on the distance). The effectiveness parameter in this case was δ′≡δ′[RD,RS,E(l′)].

Radiation shielding analysis of the HI-STORM FW storage cask

This paper presents radiation shielding analysis of the HI-STORM FW storage cask designed by the Holtec International, Inc. for the Nuclear Power Plant Krsko Spent Fuel Dry Storage (SFDS) project. The first part of the analysis is addressing the problem of a particle source generation. For each spent fuel assembly (SFA) neutron and gamma source intensities and spectra were calculated using ORIGEN-S module of the SCALE 6.2.3 code package. The second part of the analysis is considering neutron and gamma dose rates calculation, for which hybrid shielding methodology based on Monte Carlo (MC) code MCNP6.2 and ADVANTG3.0.3 code was used. The dose rates for the reference case included the contributions of fuel neutron, fuel gamma, secondary gammas from (n, γ) reactions, as well as gammas from activated SFA hardware components due to Co-60 presence. Expectedly, the results have shown that the dose rates around the cask have nonuniform behavior due to a specific design of the cask. The heavy concrete body of the cask presents a good attenuator for all types of radiation, however, a considerable streaming through the top and bottom ventilation openings was noticed. It was also observed that an important role was played by the isotope Co-60 which emits two lines of high energy deep penetrating gammas. The convergence of dose rates and associated relative uncertainties during MC calculations have shown that significantly more histories were required for gamma calculations compared to neutron calculations. To investigate the impact of different SFA content in the casks on the dose rates, shielding analysis was made for 16 casks that will be loaded in the first campaign. For comparison, the calculations of a single HI-STORM FW cask were additionally performed using MAVRIC module of the SCALE 6.2.3 and the obtained differences in external dose rates of 20–30% are expected. The calculation cases showed the differences in casks doses due to different casks loading within loading plan criteria, reduction provided by heavy concrete in cask body (more important for gammas), importance of careful design of lower and upper casks openings to keep increase of dose rates, compared to closed configuration, within one order of magnitude, and compromise needed when describing details of intensity and spectrum of SFA sources. The increased calculation mesh density was used to show that selected discretization is appropriate, and the limited influence of storage pad thickness was demonstrated by performing calculation with thick concrete slab corresponding to Dry Storage Building (DSB) floor foundation thickness.

Fast neutron spectroscopy with a volumetrically-sensitive, moderating-type neutron spectrometer in a high-radiation environment

Non-fission nuclear reactions were investigated by NASA as part of efforts to develop long-lasting electric power systems for space applications where solar energy is weak or unavailable. Key to investigation of nuclear reactions is spectroscopy of the neutron energies produced in such reactions. In addition to employing a standard scintillating-type neutron spectrometry system, the NASA team investigated the application of a moderating-type neutron spectrometer in a high-radiation environment as a parallel approach. The unfolded log-scale and linear-scale spectra of well-known calibration neutron sources are presented along with the ISO standard spectra for comparison. This new technology shows promise as a compact instrument capable of neutron spectroscopy in laboratory settings, and in particular, space applications where small size, low data rate and no onboard computation are advantages.

Affine reduced-order model for radiation transport problems in cylindrical coordinates

This article focuses on parametric reduced-order models to decrease the computational cost of large-scale neutron transport problems in the atmosphere. The atmospheric humidity content and neutron source spectrum are considered as uncertain parameters throughout the problem. We first present our full-order model, including an axisymmetric SN transport methodology, a geopotential model for computing the thermophysical properties of the atmosphere, and a support vector regression method for rapidly generating one-group cross sections at every point. Then, reduction of the full-order problem using the POD-Galerkin and least-squares POD-Petrov-Galerkin methods is presented. The reduced-order models implemented introduce mean relative errors between 0.06% and 0.4% with acceleration factors between 3.6×103-8.4×104, respectively.

Sensitivity of the active neutron coincidence collar response during simulated and experimental fresh fuel assay

Verification of the fissile (and fertile) content in fresh nuclear fuel assemblies is conducted by the IAEA to enforce the Nuclear Non-Proliferation Treaty using the UNCL (Uranium Neutron Collar — Light Water Reactor Fuel). The UNCL uses an uncorrelated AmLi neutron source to interrogate the fuel, producing a signal of coincident fission neutrons (Doubles rate) used to assay 235U content of the fuel. The cost of producing calibration assemblies and limited availability of diverse commercial assemblies at any one time historically restricted the ability to explore the full parameter space experimentally. Monte Carlo simulations can overcome this, but introduce additional sources of uncertainty. Here the sensitivity of simulations and measurements to various parameters is assessed for a reference 16×16 PWR assembly of uniform 3.2% enrichment. Uncertainty contributions in this evaluation include: simulated AmLi neutron source spectrum, AmLi neutron emission rate, AmLi anisotropicity, high density polyethylene (HDPE) density, the precise position of the fuel assembly within the detector, and experimentally the statistical uncertainty. The overall total systematic uncertainty estimate for the simulation of the absolute/relative Doubles rates responses are estimated to be approximately 2.0%/1.5%, and for experimental measurements systematic uncertainty reduces to 1.1%. This analysis supports further work using the relative Doubles rates in place of measurements for updating and extending the UNCL analysis methodology as systematic uncertainty is reasonably small.

Photoneutron sources spectra calculation

The issues of increasing the accuracy of determining the average energy and the type of energy distribution (shape of the spectrum) of neutrons from Sb-Be- and Ra-Be-sources are considered. The indicated photoneutron sources are used to calibrate the energy scales of neutron spectrometers and to determine the energy dependence of the sensitivity of radiometers and neutron dosimeters. A method is proposed for calculating the energy spectra of radionuclide neutron sources of the Sb-Be (γ, n) and Ra-Be (γ, n) types, and the calculation results are presented. The dependence of the spectrum of the source on its design is investigated. The correctness of the results of calculating the energy spectra of radionuclide sources has been confirmed experimentally. The data obtained will be used to determine the energy dependence of the sensitivity of neutron spectrometers, radiometers and dosimeters.

Design and performance of the Water-moderator nested spheres neutron spectrometer

In this work, a novel neutron spectrometer was designed and its performance was evaluated. The spectrometer is several nested spheres that are filled with water and in the center is a 3He proportional counter The detector part of the spectrometer consists of five concentric stainless-steel spherical shells of different sizes that are nested with each other to form four interlayers with different thicknesses. By filling the interlayers with water or air, 13 thicknesses of water moderator can be obtained. A spherical 3He proportional counter is placed in the core of the stainless-steel spherical shell to detect thermal neutrons. During measurement, different neutron moderation effects are achieved by changing the water-filled thickness of the interlayer. This work uses the Geant4 Monte Carlo code to calculate the response matrix for 100 log-equidistant energies from 10−3eV to 100 MeV. In addition, a highly-integrated electronic circuit, with a STM32L4 series ARM processor as the core, is designed for signal processing and system control. Finally, the energy spectra of 241Am-Be and 252Cf neutron sources was measured using the prototype of the spectrometer. The experimental results are in good agreement with the standard neutron energy spectra given by the IAEA.

Enhancing neutron detection capabilities of a water Cherenkov detector

In this work we show the neutron detection capabilities of a water Cherenkov detector (WCD) using pure water, as well as an aqueous solution of sodium chloride as the detection volume. The experiments were performed using a WCD with a single photomultiplier tube (PMT), and a 252Cf neutron source. We show that fast neutrons from a 252Cf source directly impinging the detector can be clearly detected and identified over the natural radiation background. We also present results from numerical simulations that describe the response of the WCD to neutrons of different energies. To do this, a detailed model for the WCD and of the neutron source spectra have been implemented. The implemented simulation code takes into account the interaction processes involved in the detection of neutrons using WCD, and support the experimental evidence introduced in this work. Being both the active volumes analyzed in this work (pure water) and the additive (sodium chloride), cheap, non-toxic and easily accessible materials, the results obtained are of interest for the development of large neutron detectors for different applications. Of particular importance are the detection of special nuclear materials, as well as those applications related with space weather phenomena. We conclude that WCD used as neutron detectors could be a replace or a complementary tool for standard neutron monitors based on 3He.

A neutronic experiment to support the design of an Indian TBM shield module for ITER

A shield module is associated with an Indian Test Blanket Module (TBM) in ITER to limit the radiation doses in port inter-space areas. The shield module is made of stainless steel plates and water channels. It is identified as an important component for radiation protection because of its radiation exposure control functionality. The radiation protection classification leads to more assurance of the component design. In order to validate and verify the design of the shield module, a neutronic laboratory-scale experiment is designed and executed. The experiment is planned by considering the irradiation under a neutron source of 14 MeV and yields of 1010 n s−1. The reference neutron spectrum of the ITER TBM shield module has been achieved through optimization of the neutron source spectrum by a combination of steel and lead materials. The neutron spectrum and flux are measured using a multiple foil activation technique and neutron dose-rate meter LB 6411 (He-3 proton recoil counter with polyethylene), respectively. The neutronic design simulation is assessed using MCNP5 and FENDL 2.1 cross-section data. The paper covers neutronic design, irradiation and the outcome of the experiment in detail.

Neutron detection capabilities of Water Cherenkov Detectors

Abstract In this work we show the neutron detection capabilities of a water Cherenkov detector (WCD). The experiments presented here were performed using a simple WCD with a single photomultiplier tube (PMT) and a 252 Cf neutron source . We compared the use of pure water and water with non contaminant additive as the detection volume. We show that fast neutrons from the 252Cf source can be detected over the flux of atmospheric particles background. Our first estimation for the neutron detection efficiency is at the level of (19)% for pure water and (44)% for the water with the additive. We also present the simulation of the response of the WCD to neutrons using a simulated 252Cf source. We implemented a detailed model of the WCD and of the neutron source spectra using Geant 4. The results of our simulations show the detailed mechanism for the detection of neutrons using WCD and support the experimental evidences presented. Since both active volumes studied, H 2 O pure and with additive, are cheap, non-toxic and easily accessible materials, the results obtained are of interest for the development of large neutron detectors for different applications. Of special importance are those related with space weather phenomena as well as those for the detection of special nuclear materials. We conclude that WCD used as neutron detectors can be a complementary tool for standard neutron monitors based on 3 He.

Neutron spectroscopy with soft computing: development of a computational code based on Support Vector Machine (SVM) for reconstruction of neutron energy spectrum

This paper presents a developed computational code based on Support Vector Machine (SVM) for reconstruction of energy spectrum of neutron source. To reconstruct unknown energy spectrum using known neutron pulse height distribution, the developed machine is trained by known neutron pulse height distribution of detector and corresponding energy spectrum of neutron source. Validation and testing are the next steps to verify the validity of the calculations done with the developed computational code. The calculated neutron pulse height distributions due to randomly generated energy spectrum using MCNPX-ESUT (MCNPX-Energy engineering of Sharif University of Technology) computational code are used as input data of the developed computational code for reconstruction of energy spectrum of neutron source. The considered energy spectrums of neutron source are output data. In the present study, the simulation of neutron pulse height distribution of 241Am-9Be neutron source in the NE-213 liquid organic scintillator is performed using MCNPX-ESUT computational code. In the constructed matrix form of Fredholm integral equation, the effective response matrix is usually close to singular or badly scaled and its condition number is so high. The solution of inverse problem in which the matrix has high condition number does not give accurate results. The motivation of the present study is the development of computational coded based on soft computing algorithm for reconstruction of energy spectrum of the neutron source with high accuracy. To this end, the computational code based on Support Vector Machine (SVM) is developed, where there is no need to solve the inverse problems. The unfolded energy spectra of 241Am-9Be neutron sources using the developed computational code have an excellent agreement with the reference spectrum provided by ISO. A comparison between the results of present study and results of developed computational code based on Artificial Neural Network (ANN) and Adaptive Neuro-Fuzzy Inference System (ANFIS) is performed.

Verification and validation of radiation source term capabilities in STREAM

This paper presents the verification and validation of the radiation source term calculation capability implemented in the reactor analysis code STREAM for pressurized water reactor (PWR) spent nuclear fuel (SNF) analysis. Activity, decay heat, neutron and gamma source spectra of irradiated PWR fuel assemblies are calculated with STREAM and ORIGEN and the results obtained are compared. The verification is performed with Westinghouse 15 × 15, 17 × 17 and Combustion Engineering 16 × 16 designs with enrichment of 3–5 wt.% 235U, burnup of 30–60 GWd/t and cooling times of 0.3–1000 years. The radiation source term comparison between STREAM and ORIGEN indicates that the implementation of the source term equations is correct. The validation is done by comparing STREAM calculation results against 91 decay heat calorimetric measurements for 52 PWR fuel assemblies performed at the Swedish central storage facility for spent nuclear fuel, CLAB, and United States Hanford Engineering Development Laboratory (HEDL) and General Electric Morris Operations. The measured fuel assemblies have enrichment of 2–4 wt.% 235U, burnup of 28–51 GWd/t and cooling times of 2–27 years. The comparison between calculated and measured decay heat shows good agreement. For most of the measurements analyzed, the calculated decay heat is within the range of the measurement uncertainty. The average, overall C/E over 71 CLAB decay heat measurements is 1.000 ± 0.017, and 0.997 ± 0.013 for 20 US measurements. Overall, this work demonstrates that STREAM can be reliably used to predict PWR SNF radiation source terms.

Neutron spectra from Neutron Standards Laboratory (LPN/CIEMAT) sources with two Bonner sphere spectrometers

The Neutron Standards Laboratory (LPN: Laboratorio de Patrones Neutrónicos) of the Ionizing Radiation Metrology Laboratory (LMRI) of the CIEMAT (Centro de Investigaciones Energéticas Medioambientales y Tecnológicas) has two calibration neutron sources, 241Am-Be and 252Cf. Studies to verify the shielding of the facility and to characterize the neutron field in the irradiation room have been carried out using Monte Carlo methods. However, spectral measurements were not performed yet. In this study, we measured the total and direct neutron spectra of both calibration neutron sources. Two different Bonner Sphere Spectrometers (BSS) systems were allocated to 115 cm from the sources, on the calibration bench, with and without the shadow cones (SC). From the count rates obtained with both BSS systems (BSS-6LiI and BSS-3He), the total neutron spectrum and direct neutron spectrum were unfolded. Total and direct spectra obtained for the 241Am-Be and 252Cf sources with the MCNPX and different BSS systems, unfold codes and response matrices presented similar shape and are consistent with each other.

Evaluation of a new neutron energy spectrum unfolding code based on an Adaptive Neuro-Fuzzy Inference System (ANFIS).

The purpose of the present study was to reconstruct the energy spectrum of a poly-energetic neutron source using an algorithm developed based on an Adaptive Neuro-Fuzzy Inference System (ANFIS). ANFIS is a kind of artificial neural network based on the Takagi–Sugeno fuzzy inference system. The ANFIS algorithm uses the advantages of both fuzzy inference systems and artificial neural networks to improve the effectiveness of algorithms in various applications such as modeling, control and classification. The neutron pulse height distributions used as input data in the training procedure for the ANFIS algorithm were obtained from the simulations performed by MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). Taking into account the normalization condition of each energy spectrum, 4300 neutron energy spectra were generated randomly. (The value in each bin was generated randomly, and finally a normalization of each generated energy spectrum was performed). The randomly generated neutron energy spectra were considered as output data of the developed ANFIS computational code in the training step. To calculate the neutron energy spectrum using conventional methods, an inverse problem with an approximately singular response matrix (with the determinant of the matrix close to zero) should be solved. The solution of the inverse problem using the conventional methods unfold neutron energy spectrum with low accuracy. Application of the iterative algorithms in the solution of such a problem, or utilizing the intelligent algorithms (in which there is no need to solve the problem), is usually preferred for unfolding of the energy spectrum. Therefore, the main reason for development of intelligent algorithms like ANFIS for unfolding of neutron energy spectra is to avoid solving the inverse problem. In the present study, the unfolded neutron energy spectra of 252Cf and 241Am-9Be neutron sources using the developed computational code were found to have excellent agreement with the reference data. Also, the unfolded energy spectra of the neutron sources as obtained using ANFIS were more accurate than the results reported from calculations performed using artificial neural networks in previously published papers.

PHOTON SPECTRA IN NPL STANDARD RADIONUCLIDE NEUTRON FIELDS.

A HPGe detector has been used to measure the photon spectra from the majority of radionuclide neutron sources in use at NPL (252Cf, 241Am-Be, 241Am-Li, 241Am-B). The HPGe was characterised then modelled to produce a response matrix. The measured pulse height spectra were then unfolded to produce photon fluence spectra. Changes in the photon spectrum with time from a 252Cf source are evident. Spectra from a 2-year-old and 42-year-old 252Cf source are presented showing the change from a continuum to peaks from long-lived isotopes of Cf. Other radionuclide neutron source spectra are also presented and discussed. The new spectra were used to improve the photon to neutron dose equivalent ratios from some earlier work at NPL with GM tubes and EPDs.