Measurement Of Neutron Spectra Research Articles

Preequilibrium Asymmetries in the ^{239}Pu(n,f) Prompt Fission Neutron Spectrum.

The physical properties of neutrons emitted from neutron-induced fission are fundamental to our understanding of nuclear fission. However, while state-of-the-art fission models still incorporate isotropic fission neutron spectra, it is believed that the preequilibrium prefission component of these spectra is strongly anisotropic. The lack of experimental guidance on this feature has not motivated incorporation of anisotropic neutron spectra in fission models, though any significant anisotropy would impact descriptions of a fissioning system. In the present work, an excess of counts at high energies in the fission neutron spectrum of ^{239}Pu is clearly observed and identified as an excess of the preequilibrium prefission distribution above the postfission neutron spectrum. This excess is separated from the underlying postfission neutron spectrum, and its angular distribution is determined as a function in incident neutron energy and outgoing neutron detection angle. Comparison with neutron scattering models provides the first experimental evidence that the preequilibrium angular distribution is uncorrelated with the fission axis. The results presented here also impact the interpretation of several influential prompt fission neutron spectrum measurements.

Study on unfolding method of neutron spectrum of BSS (Bonner Sphere Spectrometer) based on compressed sensing

The Bonner Sphere Spectrometer (BSS) is widely used for neutron spectrum measurement for its broad energy sensitivity (from thermal to GeV). Information of neutron spectrum cannot be directly obtained and must be unfolded from the measurement readings. In this paper, a novel method of unfolding neutron spectrum from BSS measurement based on compressive sensing (CS) is proposed. Sparse representation based on learning dictionary is employed. The orthogonal matching pursuit (OMP) algorithm and the K-SVD algorithm are applied for sparse coding and dictionary update. The SL0 algorithm is used for reconstruction. Computational experiment is carried out, in which 5 types of the neutron spectrum are tested, including the D–T fusion neutron spectrum, Watt fission spectrum, Maxwell fission spectrum, the evaporation spectrum and 241Am-Be neutron spectrum taken from ISO standard 8529-1. The computational experiment reveals that: firstly, with appropriate sparse representation, neutron spectrum can be unfolded with high accuracy; secondly, the unfolding method based on CS is very suitable for solving the highly under determined problem. For instance, neutron spectrum with 300 energy groups is unfolded with 7 equations.

Neutron spectrum change with thermal moderator temperature in a compact electron accelerator-driven neutron source and its effects on spectroscopic neutron transmission imaging

ABSTRACT Recently, Compact Accelerator-driven Neutron Sources (CANSs) are attracting attention. In CANSs, a simple thermal neutron moderator such as polyethylene is often used from the viewpoints of cost, simplicity and maintainability. In most cases, the temperature of such a moderator has not been controlled although it is natural that the moderator temperature and the neutron spectrum will change with accelerator-operation. Thus, we simultaneously measured neutron spectra and the temperature of a polyethylene moderator at the Hokkaido University Neutron Source (HUNS) driven by a compact electron accelerator to observe the effect of any temperature change on the reliability of spectroscopic transmission measurement. The ratio of the neutron effective temperature and the moderator temperature was constant in HUNS case, although both increased by 4–5 K within one hour after the start of accelerator-operation. This indicated that the neutron effective temperature was well estimated by the moderator temperature. The effect of the temperature change can be easily avoided by excluding data collection before the moderator warms up. These results suggested that the monitoring of moderator temperature is recommended in compact electron accelerator-driven neutron sources with a thermal neutron moderator to guarantee reliability of spectroscopic transmission measurement without sacrifices of cost, simplicity and maintainability.

A device for measurements of neutron radiation spectra

The method of neutron spectrum measurement with the use of a device with measuring unit based on dysprosium activation detectors has been developed. This article contains the diagram and the exterior of the measurement device. Moreover, it includes the calculated response function of the detectors from neutron energy in the range of 0.05 to 14.5 MeV along the axis of the measuring unit, the algorithm for reconstruction of neutron spectrum and the results of method implementation at the distance of 12 m from the reactor PRIZ-M.

Impact of imposed mode 2 laser drive asymmetry on inertial confinement fusion implosions

Low-mode asymmetries have emerged as one of the primary challenges to achieving high-performing inertial confinement fusion implosions. These asymmetries seed flows in the implosions, which will manifest as modifications to the measured ion temperature (Tion) as inferred from the broadening of primary neutron spectra. The effects are important to understand (i) to learn to control and mitigate low-mode asymmetries and (ii) to experimentally more closely capture thermal Tion used as input in implosion performance metric calculations. In this paper, results from and simulations of a set of experiments with a seeded mode 2 in the laser drive are described. The goal of this intentionally asymmetrically driven experiment was to test our capability to predict and measure the signatures of flows seeded by the low-mode asymmetry. The results from these experiments [first discussed in M. Gatu Johnson et al., Phys. Rev. E 98, 051201(R) (2018)] demonstrate the importance of interplay of flows seeded by various asymmetry seeds. In particular, measured Tion and self-emission x-ray asymmetries are expected to be well captured by interplay between flows seeded by the imposed mode 2 and the capsule stalk mount. Measurements of areal density asymmetry also indicate the importance of the stalk mount as an asymmetry seed in these implosions. The simulations brought to bear on the problem (1D LILAC, 2D xRAGE, 3D ASTER, and 3D Chimera) show how thermal Tion is expected to be significantly lower than Tion as inferred from the broadening of measured neutron spectra. They also show that the electron temperature is not expected to be the same as Tion for these implosions.

Measurement of neutron-production double-differential cross sections of [formula omitted]C, 27Al, [formula omitted]Fe, and [formula omitted]Pb by 20, 34, 48, 63, and 78 MeV protons in the most-forward direction

In this study, the neutron-production double-differential cross sections (DDXs) of natC, 27Al, natFe, and natPb bombarded by 20, 34, 48, 63, and 78 MeV protons were measured in the most-forward direction by the time-of-flight (TOF) method using a liquid organic scintillator. Characteristic peak structures were observed in the measured neutron spectra, especially for light nuclei. The experimental data were compared with the results from the Monte Carlo simulation code “Particle and Heavy Ion Transport code System” (PHITS) using the theoretical model Liège Intranuclear Cascade (INCL) or using the Japanese evaluated nuclear data library (JENDL-4.0/HE). While the results of JENDL-4.0/HE were in agreement with the experimental data better than those of INCL, neither could reproduce the measured peak structures for some combinations of the beam and target. For the energy-integrated cross sections, JENDL-4.0/HE reproduced the experimental data within a factor of two in all cases. On the other hand, INCL gave values up to approximately six times larger than the experimental ones.

Measuring fast ions in fusion plasmas with neutron diagnostics at JET

Fast ions in fusion plasmas often leave characteristic signatures in the neutron emission from the plasma. In this paper, we show how neutron measurements can be used to study fast ions and give examples of physics results obtained on present day tokamaks. The focus is on measurements with dedicated neutron spectrometers and with compact neutron detectors used in each channel of neutron profile monitors. A measured neutron spectrum can be analyzed in several different ways, depending on the physics scenario under consideration. Gross features of a fast ion energy distribution can be studied by applying suitably chosen thresholds to the measured spectrum, thus probing ions with different energies. With this technique it is possible to study the interaction between fast ions and MHD activity, such as toroidal Alfvén eigenmodes (TAEs) and sawtooth instabilities. Quantitative comparisons with modeling can be performed by a direct computation of the neutron emission expected from a given fast ion distribution. Within this framework it is also possible to determine physics parameters, such as the supra-thermal fraction of the neutron emission, by fitting model parameters to the data. A detailed, model-independent estimate of the fast ion distribution can be obtained by analyzing the data in terms of velocity space weight functions. Using this method, fast ion distributions can be resolved in both energy and pitch by combining neutron and gamma-ray measurements obtained along several different sightlines. Fast ion measurements of the type described in this paper will also be possible at ITER, provided that the spectrometers have the dynamic range required to resolve the fast ion spectral features in the presence of the dominating thermonuclear neutron emission. A dedicated high-resolution neutron spectrometer has been designed for this purpose.

Neutron Diagnostics for Tokamak Plasma: From a Plasma Diagnostician Perspective

Neutron measurement is considered to be one of the major diagnostic methods in tokamak plasma. The neutron diagnostics provide key information on plasma physics, machine protection and control issues. In this article, the commonly used techniques of neutron emission diagnostics for tokamak plasma are presented. The requirements and limitations imposed on the diagnostic systems are discussed. Several basic techniques of total neutron yield and neutron spectrum measurements such as neutron counters, neutron activation technique, time-of-flight spectroscopy or magnetic proton recoil spectroscopy are described in a practical way, from a plasma diagnostician point of view. Primary physical quantities that can be derived from those measurements are pointed out and 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.

Benchmarking DRAGON/PARCS Against KRITZ and FFTF Measurements

Calculations have been conducted for the KRITZ-2 (KRITZ-LWR-RESR-001/002/003) and the Fast Flux Test Facility (FFTF) (FFTF-LMFR-RESR-001) Nuclear Energy Agency benchmark problems using the PARCS reactor simulation code with lattice parameters generated by the DRAGON reactor physics code and with the MCNP6 Monte Carlo code. The benchmark analyses examined the DRAGON cross-section library, PARCS energy group structure, DRAGON fuel assembly modeling, and nuclide self-shielding effect. For KRITZ-2, the PARCS 2-group core calculations with a DRAGON 361-group library based on ENDF/B-VII.1 reproduced the benchmark keff with a root-mean-square (rms) error of 0.19% δk. DRAGON/PARCS also predicted the fission rates within 5%. The MCNP results are consistent with the DRAGON/PARCS results but with a small underestimation when compared to the benchmark value. For FFTF, the PARCS 33-group core calculations underpredicted the benchmark keff by 0.19% δk while the MCNP calculation overpredicted the benchmark keff by 0.23% δk. The neutron spectrum distributions calculated by PARCS and MCNP are consistent with measured data. Since the energy boundary values of the measured neutron spectrum are not available, the calculated spectra could not be directly compared to the measured value. The DRAGON/PARCS solution to a numerical benchmark of a gas-cooled fast reactor (GFR), i.e., the Energy Multiplier Module, predicted the keff and assembly power with 0.46% δk and 3.7% rms error, respectively, when compared to the MCNP simulation. The benchmark calculations of the selected thermal and fast reactors have shown that DRAGON/PARCS simulates small reactor cores with good accuracy.

Investigation on A-FNS neutron spectrum monitor system

We proposed the neutron spectrum measurement system for Advanced Fusion Neutron Source (A-FNS). It is configured with fission chambers, Self-Powered Neutron Detectors, neutron activation systems using water flow, and activation monitors. In this study, we mainly investigated the multiple activation foils which are sensitive to high energy neutrons. We selected 14 main dosimetry threshold reactions based on specific A-FNS usage conditions. The reactions are as follows: 55Mn(n,2n)54Mn, 59Co(n,p)59Fe, 59Co(n,2n)58Co, 59Co(n,3n)57Co, 75As(n,2n)74As, 89Y(n,2n)88Y, 93Nb(n,2n)92mNb, 169Tm(n,2n)168Tm, 169Tm(n,3n)167Tm, 197Au(n,2n)196Au, 197Au(n,3n)195Au, 209Bi(n,3n)207Bi, 209Bi(n,4n)206Bi, and 209Bi(n,5n)205Bi. The dosimetry reaction rates were calculated using Monte Carlo code MCNP-5.15 with McDeLicious-11 and nuclear data libraries FENDL-3.1c, IRDFF-1.05 and EAF-2010. As for neutron-induced cross section data of the reactions in the high energy region, we clarified that there were limited experimental data, and proposed new experiment for their evaluations.

Investigation of the neutron spectrum measurement method for dose evaluation in boron neutron capture therapy

In boron neutron capture therapy, it is important to evaluate the dose administered to a patient's body outside the tumour area. The exposure dose is evaluated by calculation; however, the calculated value must be validated using a measured value. The dose evaluations based on the measured neutron spectrum are investigated. Multi-foil activation, combined with a LiCaAlF6 scintillation detector and an imaging plate, is proposed as a measurement method. The proposed method can measure the neutron spectrum at various points quickly.

Neutronic experiments with fluorine rich compounds at LR-0 reactor

Research on molten salt reactor (MSR) neutronics continues in Research Centre Rez (Czech Republic) with experimental work being conducted using fluoride salt that was originally used in the Molten Salt Reactor Experiment (MSRE). Previous results identified significant variations in the neutron spectrum measured in LiF-NaF salt. These variations could originate from the fluorine description in current nuclear data sets. Subsequent experiments were performed to try to confirm this phenomenon. Therefore, another fluorine-rich compound, Teflon, was used for testing. Critical experiments showed slight discrepancies in C/E-1 for both compounds, Teflon and FLIBE, and systematic overestimation of criticality was observed in calculations. Different nuclear data libraries were used for data set testing. For Teflon, the overestimation is higher when using JENDL-3.3, JENDL-4, and RUSFOND-2010 libraries, all three of which share the same inelastic-to-elastic scattering cross section ratio. Calculations using other libraries (ENDF/B-VII.1, ENDF/B-VII.0, JEFF-3.2, JEFF-3.1, and CENDL-3.1) tend to be closer to the experimental value. Neutron spectrum measurement in both substances revealed structure similar to that seen in previous measurements using LiF-NaF salt, which indicates that the neutron spectrum seems to be strongly shaped by fluorine. Discrepancies between experimental and calculational results seem to be larger in the neutron energy range of 100–1300 keV than in higher energies. In the case of neutron spectrum calculation, none of the tested libraries gives overall better results than the others.

Determination of fast neutron spectrum using recoil-proton track imaging method

High sensitivity and good neutron/gamma discrimination, as well as online determination, are goals that researchers are pursuing for fast neutron spectrum measurements (FNSM). For this purpose, we are attempting a novel approach based on the recoil-proton track imaging (RPTI) method to do FNSM. The RPTI neutron spectrometer has been designed and built, and it mainly consists of a polyethylene foil for converting neutrons to protons, a gas scintillation chamber for generating photons along the recoil proton track, and a scientific camera system for recording the track photons in real time. Proof-of-principle experiments have been performed with the RPTI neutron spectrometer at a DT neutron generator at China Institute of Atomic Energy. As a result, numerous recoil proton track images at the single-particle level were successfully acquired and revealed in real time. Also, the proton tracks can be clearly distinguished from the background in the track images, indicating good signal-to-background ratio of the system. Particularly, both the proton ranges and the Bragg curve peak positions can be extracted online from the track images, which can lead to online generation of DT neutron spectra. These results show promising performance of the novel fast neutron spectrometer, which will bring significant advantages in various applications.

Neutron propagation experiments with a lead test section inserted in the core of the LR-0 reactor

Validation of neutronic parameters for lead-cooled systems, by integral and local experiments, became of primary interest with advances in ALFRED lead cooled reactor demonstrator design. Notably, aiming at advancing the core design from the conceptual to the basic stage, the assessment of uncertainties coming from numerical simulation, with respect to real experiments, was of paramount importance to support the claims of outstanding safety demonstration meant for ALFRED. Among the required evidences, the assessment of the spatial distribution of the neutron flux and power in fuel pins required conceiving ad-hoc experiments, and disposing of state-of-the-art post-irradiation examination capabilities. For all these reasons, these experiments were one of the main focuses of the collaboration between the Research Centre Rez and ENEA. This paper presents the conception phase of the test and discusses some of the main results of the first phase of the experimental campaign, dealing with neutrons propagation, collected during its execution at the LR-0 research reactor (hence before the post-irradiation examination of the experimental pins placed in the lead test rig). The experimental work involved at first neutron spectrum measurements in the energy range from 0.1 to 10 MeV. Additionally, measurement of basic neutronic parameters of lead were performed, such as its reactivity worth, its effect on the neutron spectrum and its slowing-down properties. The comparison of calculations and experimental results shows good agreement. In case of calculation in benchmark model with different nuclear data libraries, the criticality is systematically over-predicted by approximately 150 pcm, which is, however, in the 1σ of the uncertainty interval. Neutron spectrum measurement shows only slight variations being around 10% most of the time.

Neutron energy spectrum measurement using an NE213 scintillator at CHARM

To establish a methodology for neutron spectrum measurement at the CERN High energy AcceleRator Mixed field facility (CHARM), neutron spectra were measured using an NE213 scintillator on top of the CHARM roof shielding where is the CERN Shielding Benchmark Facility (CSBF). The spectra were derived as fluences into the scintillator by the unfolding method using an iterative Bayesian algorithm. The methodology was verified based on the agreement of two spectra measured for different positions and directions of incident neutrons by changing the detector orientation. Since the spectra on the roof-top were obtained within a reasonable beam-time, this methodology is suitable for measuring the spectrum when there is less shielding material. Thus, experimental data for neutron transition can be obtained as a function of shielding thickness using this facility.

NEUTRON CHARACTERIZATION OF ENSA-DPT TYPE SPENT FUEL CASK AT TRILLO NUCLEAR POWER PLANT.

The Neutron Standards Laboratory of CIEMAT has conducted the characterization of the independent spent fuel storage installation at the Trillo Nuclear Power Plant. At this facility, the spent fuel assemblies are stored in ENSA-DPT type dual purpose casks. Neutron characterization was performed by dosimetry measurements with a neutron survey meter (LB6411) inside the facility, around an individual cask and between stored casks, and outside the facility. Spectra measurements were also performed with a Bonner sphere system in order to determine the integral quantities and validate the use of the neutron monitor at the different positions. Inside the facility, measured neutron spectra and neutron ambient dose equivalent rate are consistent with the casks spatial distribution and neutron emission rates, and measurements with both instruments are consistent with each other. Outside the facility, measured neutron ambient dose equivalent rates are well below the 0.5 μSv/h limit established by the nuclear regulatory authority.

On similarity of various reactor spectra and 235U prompt fission neutron spectrum

A well-defined neutron spectrum is an essential tool not only for calibration and testing of neutron detectors used in dosimetry and spectroscopy but also for validation and verification of evaluated cross sections. A new evaluation of thermal-neutron induced 235U PFNS was performed by the International Atomic Energy Agency (IAEA) in the CIELO (Collaborative International Evaluated Library Organisation Project) project; new measurements of Spectral Averaged Cross sections averaged in the evaluated spectrum are to be obtained. In general, a neutron spectrum in the core is not identical to the pure fission one because fission neutrons undergo many scattering reactions, but it can be shown that PFNS and reactor spectra become undistinguishable from a certain energy boundary. This limit is important for experiments, because when the studied reaction threshold is over this limit, the spectral averaged cross sections in PFNS can be derived from the measured reactions in the reactor core. The evaluation of the neutron spectrum measurements in three different thermal-reactor cores shows that this lower limit is around the energy of 5.5 – 6 MeV. Above this energy the reactor spectra becomes identical with the 235U PFNS. IAEA CIELO PFNS is within 5% of the measured PFNS from 10 to 14 MeV in a LR-0 reactor, while ENDF/B-VII evaluated PFNS underestimated measured neutron spectra.

Benchmark test of JEFF-3.2, FENDL-3.0 and TENDL-2015 with TOF experiments for graphite, silicon carbide, uranium and multiple-slab samples

In order to validate JEFF-3.2, FENDL-3.0 and TENDL-2015 nuclear data libraries, some integral experiments have been conducted with TOF method on the D-T neutron source facility at China Institute of Atomic Energy (CIAE). The measured neutron spectra for Silicon Carbide, Graphite, Uranium and multi-slab sample assemblies were compared with the leakage neutron spectra simulated using the MCNP-4C code with the evaluated nuclear data libraries. The comparison was made by the spectrum shape and the C/E values in five energy regions. From the results we can conclude that all the calculations reproduce well the measurements except at a few neutron energy ranges. It is worth mentioning that the calculated results with FENDL-3.0 are almost the same as those with TENDL-2015for graphite. And compared to JEFF-3.1.1, the nuclear data of carbon were not revised in JEFF-3.2. An interesting phenomenon was significantly observed from the results of multi-slab sample assemblies. The excessive number of elastically-scattered neutrons in the U layer are moderated by the hydrogen in the last layer to lower energies as the C/E ratio satisfactorily becomes consistent with 1.0.

Investigation of an Online Reactor Neutron Spectrum Measurement Method with Ionization Chambers

In this paper, an online reactor neutron spectrum measurement method is presented. The basic theory of this method is based on the unfolding of few-channel data, in which three miniature ionization chambers are applied. The neutron spectrum can be unfolded with the count rates and response functions of the three detectors through an unfolding program. In order to investigate the feasibility of this method, simulation tests have been performed with the reference neutron spectra and neutron spectra from the China LEAd-based Reactor (CLEAR). The research results show that this method can provide an alternative means for an online neutron spectrum measurement in the reactors. This method is suitable to be applied in fast neutron reactors due to the miniature size of ionization chambers and fission threshold of 238U.