Fast Neutron Energy Research Articles

Fast neutron energy based modelling of biological effectiveness with implications for proton and ion beams

A practical neutron energy dependent RBE model has been developed, based on the relationship between a mono-energetic neutron energy and its likely recoil proton energy. Essentially, the linear energy transfer (LET) values of the most appropriate recoil proton energies are then used to modify the linear quadratic model radiosensitivities (α and β) from their reference LET radiation values to provide the RBE estimates. Experimental neutron studies published by Hall (including some mono-energetic beams ranging from 0.2 to 15 MeV), Broerse, Berry, and data from the Clatterbridge and Detroit clinical neutron beams, which all contain some data from a spectrum of neutron energies, are used to derive single effective neutron energies (NEeff) for each spectral beam. These energies yield a recoil proton spectrum, but with an effective mean proton energy (being around 50% of NEeff). The fractional increase in LET is given by the recoil proton LET divided by the proton (LETU) value which provides the highest RBE. This ratio is then used to determine the change in the linear-quadratic model α and β parameters, from those of the reference radiation, to estimate the RBE. The predicted proton recoil RBE is then reasonably close to the experimental neutron RBE values found when taking into account the variation inherent in biological experiments. The work has some important consequences. The data of Hall et al (1975 Radiat. Res. 64 245–55) shows that the highest RBE values are found with neutron energies around 0.3–0.4 MeV, but this energy cannot possibly generate recoil proton energies which are higher, as necessary for a 0.68 MeV proton with a 30.5 keV μm−1 LETU (the LET value which provides the maximum obtainable RBE for a specified ion). For 0.4 MeV neutrons with proton recoil energies of around 0.2 MeV, the latter have a LET of around 62.88 keV μm−1. This could have an impact on proton beam RBE modelling. However, this is compensated by finding that the maximum radiosensitivity for mono-energetic neutrons was around 1.7 times larger than previously suggested from experimental ion beam studies, probably due to the necessary spreading out of Bragg peaks for ion beam experimental purposes, sampling errors and particle range considerations. This semi-empirical model can be used with minimal computer support and could have applications in ionic beams and in radioprotection.

Development of a Fast-Spectrum Self-Powered Neutron Detector for Molten Salt Experiments in the Versatile Test Reactor

The self-powered neutron detector (SPND) is a widely used flux monitor in thermal nuclear reactors. Although this is a mature technology, the current state of the art is tuned for a thermal neutron spectrum, so many of the devices currently in use lack sensitivity to fast neutrons. Because current in SPNDs is produced through nuclear reactions with the neutron flux inside a reactor, sensitivity in SPNDs is determined by the neutron cross section of the neutron-sensitive portion of the detector, termed the emitter. This neutron cross section drops by orders of magnitude between thermal and fast neutron energies for many emitters in currently used SPNDs, with a corresponding drop in current from the detector. This paper discusses efforts to develop a fast-spectrum self-powered neutron detector (FS-SPND) that is sensitive to neutrons with energies ranging from 0.025 eV up to 1 MeV. An in-depth analysis of Evaluated Nuclear Data File (ENDF)/B-VII.1 neutron-capture cross sections was performed, and four new materials were identified that are suitable emitter candidates for use in measuring fast neutrons. All four materials are stable mid-shell nuclei in the region between doubly magic132Sn and208Pb. Each candidate was simulated with the Geant4 Monte Carlo simulation toolkit to optimize overall detector efficiency.

Unfolding simulation of single-energy and continuous fast neutrons spectrum based on micro-pattern gas detector

This paper focuses on the feasibility of fast neutron energy spectrum measurement. The MCNPX and Geant4 are used to simulate two conversion models of stacking neutrons to protons in the triple GEM cathode coupled with multilayer polyethylene, with five kinds of single-energy neutron sources and Am-Be continuous neutron sources taken as research objects. The response function to 160 single energy neutrons and the recoil proton spectrum distribution of the above sources of the detection system are obtained by simulation. Using GRAVEL algorithm and MLEM algorithm and through simulation, the recoil proton spectra of six kinds of fast neutron sources are obtained, and they are further analyzed. The spectrum outcome is compared with the standard input spectrum, showing that they are in good agreement with each other. The relative uncertainty of the unfolding spectrum is around 10%–15%. In this part the relation of gas detector with the precision of unfolding spectrum is also discussed. The result shows that when the energy resolution of micro-pattern gas detection is better than 30%, the accuracy of fast neutron spectrum can meet the needs of practical applications. Furthermore, a new transformation model is proposed based on previous experiments and proves the feasibility of applying micro-pattern gas detector to fast neutron detection of simulation. Moreover, spectrum reconstruction can be achieved by using the obtained recoil proton spectrum combined with a suitable inversion algorithm. The modeling and spectrum analysis of this study can provide a different method of applying the fast neutron detection system composed of micro-pattern gas detectors to the detection of unknown fast neutron sources and also to the source recognition through spectrum reconstruction.

Nonstatistical fluctuations in the Cl35(n,p)S35 reaction cross section at fast-neutron energies from 0.6 to 6 MeV

The lack of experimental data on the $^{35}\mathrm{Cl}(n,p)^{35}\mathrm{S}$ reaction above 100 keV has led to nuclear data evaluations that are relatively unconstrained at fast neutron energies. As a result, efforts to explore, develop, and potentially certify next generation reactor designs that incorporate chloride salts as a coolant material have been hindered. In this paper, we report partial cross section data for the $^{35}\mathrm{Cl}(n,p)^{35}\mathrm{S}$ and $^{35}\mathrm{Cl}(n,\ensuremath{\alpha})^{32}\mathrm{P}$ reactions at incident neutron energies between 0.6 MeV and 6 MeV. The measurement was performed using the pulsed beam of neutrons at the unmoderated WNR spallation neutron source at the Los Alamos Neutron Science Center, with the outgoing charged particles detected by the LENZ experimental setup, consisting of annular silicon detectors. Nonstatistical fluctuations in the $^{35}\mathrm{Cl}(n,{p}_{0})$ cross section were observed up to around 3 MeV, and the magnitude of the cross section was systematically lower than all available data evaluations at energies above 1 MeV. Modifications to the ENDF/B-VIII.0 data evaluation are suggested to better reproduce the energy averaged experimental data.

Investigation of the Pu240(n,f) reaction at the n_TOF/EAR2 facility in the 9 meV–6 MeV range

Background: Nuclear waste management is considered amongst the major challenges in the field of nuclear energy. A possible means of addressing this issue is waste transmutation in advanced nuclear systems, whose operation requires a fast neutron spectrum. In this regard, the accurate knowledge of neutron-induced reaction cross sections of several (minor) actinide isotopes is essential for design optimization and improvement of safety margins of such systems. One such case is Pu240, due to its accumulation in spent nuclear fuel of thermal reactors and its usage in fast reactor fuel. The measurement of the Pu240(n,f) cross section was previously attempted at the CERN n_TOF facility EAR1 measuring station using the time-of-flight technique. Due to the low amount of available material and the given flux at EAR1, the measurement had to last several months to achieve a sufficient statistical accuracy. This long duration led to detector deterioration due to the prolonged exposure to the high α activity of the fission foils, therefore the measurement could not be successfully completed. Purpose: It is aimed to determine whether it is feasible to study neutron-induced fission at n_TOF/EAR2 and provide data on the Pu240(n,f) reaction in energy regions requested for applications. Methods: The study of the Pu240(n,f) reaction was made at a new experimental area (EAR2) with a shorter flight path which delivered on average 30 times higher flux at fast neutron energies. This enabled the measurement to be performed much faster, thus limiting the exposure of the detectors to the intrinsic activity of the fission foils. The experimental setup was based on microbulk Micromegas detectors and the time-of-flight data were analyzed with an optimized pulse-shape analysis algorithm. Special attention was dedicated to the estimation of the non-negligible counting loss corrections with the development of a new methodology, and other corrections were estimated via Monte Carlo simulations of the experimental setup. Results: This new measurement of the Pu240(n,f) cross section yielded data from 9meV up to 6MeV incident neutron energy and fission resonance kernels were extracted up to 10keV. Conclusions: Neutron-induced fission of high activity samples can be successfully studied at the n_TOF/EAR2 facility at CERN covering a wide range of neutron energies, from thermal to a few MeV.

Measurement of the energy of fast neutrons in the presence of gamma rays using a NaI(Tl) and a plastic scintillator

In many radiological laboratories, the energies of fast neutrons are very important for radiation diagnostics. In this paper we propose a method based on NaI and plastic scintillation (PS) detectors that measures the fast neutrons. The PS detector is sensitive to both gamma rays and fast neutrons. The NaI(Tl) detector is only sensitive to gamma rays. By subtracting the gamma information from the neutron–gamma mixed energy spectrum, pure neutron radiation information is obtained. Then, applying an unfolding method, the energy distribution of the fast neutron is generated. To verify this method, an experiment measuring the energy of monoenergetic neutrons was performed with the result confirming the effectiveness of this method.

Cross section measurement of 241Am(n,f) reaction at the Experimental Area 2 of the n_TOF facility at CERN: First Results

The condition for the safe design and operation of fast neutron reactors and energy boosters (Generation-IV reactors, ADS systems [1]) is the accuracy of nuclear data. The 241Am isotope (T1/2 = 433 years) is highly present in nuclear waste, accounting for about 1.8% of the actinide mass in PWR UOx nuclear reactors’ waste [2]. In addition, the 241Am isotope is further produced by the β decay of the 241Pu isotope (T1/2 = 14.3 years). Given the high production rate of 241Am isotope, its incineration with concurrent energy production is considered to be of utmost importance for the design and implementation of the recycling of existing nuclear waste. Sensitivity studies of the proposed systems for energy production showed that high-precision measurements of the cross section of the 241Am(n,f) reaction are required. In the present work, the 241Am(n,f) reaction cross section was measured in the Second Experimental Area of the n_TOF facility at CERN, using an array of Micromegas detectors. For the measurement, six targets of 241Am with average activity of 17 MBq per sample were coupled with an equal number of detectors in a common chamber. Additionally two 235U and two 238U samples were coupled with Micromegas detectors utilizing the neutron flux determination. Within this work, an overview of the experimental set-up and the adopted data analysis technique is presented along with preliminary results.

237Np reactions with fast neutrons: a phenomenological study

Neptunium presents various opportunities as nuclear fuel, especially in deep–space mission power generators. As it is part of the nuclear spent fuel in PWR, waste management concerns due to 237Np long α-emitting half-life have attracted some attention recently. The scarcity of experimental data in the fast neutron energy range highlights the necessity to investigate the radiative neutron capture and neutron–induced fission cross sections of this radioisotope. In the present work, statistical modeling of these reactions is performed using TALYS in an extended range of neutron energies between 10 keV and 20 MeV. In total, 72 different combinations of code parameters were selected to study the level density and γ-strength function dependence of the cross section in 238Np. Preequilibrium and compound nucleus formation phenomena are also examined. Theoretical calculations are compared to available experimental total cross section data found in literature in an attempt to investigate any discrepancies between experiment and theory and validate statistical uncertainties.

Potential efficiency improvements in fast-neutron semiconductor sensors from finely detailed structuring: A study by simulation including dead layers

The neutron sensor in an active neutron-sensitive personal dosemeter typically comprises a planar silicon diode topped with a converter layer. Neutrons interact in the converter to produce charged particles that are then detected in the diode. However, in the dosimetrically important energy region of a few keV to a few MeV, the efficiency of such devices is small because of the small cross section for producing the charged particles and their very limited range within the converter. The efficiency can in principle be improved by etching fine structures such as trenches or wells in the silicon, and filling these with the converter, so that a considerably larger amount of converter is within one particle range of the silicon. This study quantifies the potential improvements in efficiency for a selection of structure sizes and fast monoenergetic neutron energies. It also investigates how seriously these improvements are degraded if there is a dead layer at the surface of the silicon.

Fast Neutron Capture Reaction Data Measurement of Minor Actinides for Development of Nuclear Transmutation Systems

A research project entitled "study on accuracy improvement of fast-neutron capture reaction data of long-lived MAs for development of nuclear transmutation systems" is ongoing as a joint collaboration, including Tokyo Tech, Japan Atomic Energy Agency and Kyoto University in Japan. This project focuses on neutron capture reaction of minor actinides237Np,241Am and243Am, in the fast neutron energy region. The final goal of this project is to improve the neutron capture cross sections utilizing a high-intensity neutron beam from a spallation neutron source of the Japan Proton Research Complex. The overview of the project and the current progress are presented.

Evaluation of neutron induced reactions on 56Fe with CONRAD

The evaluation of the 56Fe neutron induced reactions is currently ongoing at the CEA-Cadarache using the code CONRAD, with the goal to cover the whole energy range from the Resolved Resonance Region to the continuum and estimate the corresponding uncertainties and covariance matrices. Some first results and issues occurred from this work are presented and discussed here, more specifically on the analysed transmission and capture data in the Resolved Resonance Region, as well as the optical and statistical model calculations in the fast neutron energy range.

CoolGAL: a Galinstan bathed Be fast neutron production target at the NEPIR facility

In this contribution, we present the CoolGAL fast neutron production target system, to be used in the initial phase of the NEPIR irradiation beamline at the SPES facility, that will be operational in 2022. Initially, NEPIR will be used for shielding studies against fast neutrons for space applications and to investigate neutron-induced single event effects in microelectronic devices and systems. In CoolGAL, the neutron production component, a thick Be cylinder, is immersed in a static bath of Galinstan, a liquid alloy of Ga, In and Sn, contained by an outer water cooled copper cladding. MCNPX calculations indicate that, by using a 1 μA current of 70 MeV protons, it can produce a fast neutron energy spectrum that is somewhat flat the 30-65 MeV energy range and with a sharp cut-off at the beam energy. At the standard test point, located 2.6 m downstream from the source, the beam spot diameter, defined by the peculiar collimation scheme of the initial phase of NEPIR, is 10 cm and the integral fast neutron flux is Φn(1< En<65 MeV) ~ 3×106 n cm-2s-1. Using proton beams with two different energies, one can calculate, by subtraction, the effects due to the neutrons in the energy interval defined by the two cut-off values. Preliminary results of ANSYS calculations, for a 1 μA proton current of 70 MeV protons (70 W), show a limited regime temperature (28 °C) of the Be component, capable of ensuring the exceptional safety level required for the operation at SPES. The thermal reliability of CoolGAL is very promising for future developments that require higher proton currents.

FALSTAFF, an apparatus to study fission fragment properties First arm results

Nuclear fission is a complex process that still need fundamental studies. New measurements, particularly of correlated observables, could allow to develop more sophisticated theoretical models to eventually have truly predictive capabilities for the physics of fission. Moreover, the next generation reactors concepts are mostly foreseen to operate in the fast-neutron energy domain, requiring new high quality nuclear data. In this context, a new experimental setup, called FALSTAFF, dedicated to the study of fission is under development. The FALSTAFF setup aims to investigate the fission of actinides in the fast-neutron energy domain (from a few hundreds of keV to a few MeV). Once completed, this two-arm spectrometer will detect both fragments in coincidence and allow to measure their time of flight (ToF) and kinetic energy. The average neutron multiplicity as a function of the fission fragment mass can then be assessed. The first arm of the FALSTAFF spectrometer was built. It is composed of two main parts: first, two SED-MWPC (Multi-Wire Proportional Counter) detectors are used to measure the time-of-flight as well as the position of the fragments, thus reconstructing their velocity. Second, an axial ionisation chamber gives their kinetic energy and the energy loss profile. This proceeding will describe the FALSTAFF setup as well as the methods that are used to extract the required observables, leading up to the reconstruction of the neutron multiplicity to study the fission process. Then, the recent results obtained with the first arm of FALSTAFF will be presented, exhibiting kinetic energy, velocity and post-evaporation mass distributions. These observables will be displayed for 252Cf spontaneous fission and some of the improvements recently made will be discussed.

Conceptual design of neutron measurement system for input accountancy in pyroprocessing

Abstract One of the possible options for spent-fuel management in Korea is pyroprocessing, which is a process for electrochemical recycling of spent nuclear fuel. Nuclear material accountancy is considered to be a safeguards measure of fundamental importance, for the purposes of which, the amount of nuclear material in the input and output materials should be measured as accurately as possible by means of chemical analysis and/or non-destructive assay. In the present study, a neutron measurement system based on the fast-neutron energy multiplication (FNEM) and passive neutron albedo reactivity (PNAR) techniques was designed for nuclear material accountancy of a spent-fuel assembly (i.e., the input accountancy of a pyroprocessing facility). Various parameters including inter-detector distance, source-to-detector distance, neutron-reflector material, the structure of a cadmium sleeve around the close detectors, and an air cavity in the moderator were investigated by MCNP6 Monte Carlo simulations in order to maximize its performance. Then, the detector responses with the optimized geometry were estimated for the fresh-fuel assemblies with different 235U enrichments and a spent-fuel assembly. It was found that the measurement technique investigated here has the potential to measure changes in neutron multiplication and, in turn, amount of fissile material.

High alloyed new stainless steel shielding material for gamma and fast neutron radiation

Abstract Stainless steel is used commonly in nuclear applications for shielding radiation, so in this study, three different types of new stainless steel samples were designed and developed. New stainless steel compound ratios were determined by using Monte Carlo Simulation program Geant 4 code. In the sample production, iron (Fe), nickel (Ni), chromium (Cr), silicium (Si), sulphur (S), carbon (C), molybdenum (Mo), manganese (Mn), wolfram (W), rhenium (Re), titanium (Ti) and vanadium (V), powder materials were used with powder metallurgy method. Total macroscopic cross sections, mean free path and transmission number were calculated for the fast neutron radiation shielding by using (Geant 4) code. In addition to neutron shielding, the gamma absorption parameters such as mass attenuation coefficients (MACs) and half value layer (HVL) were calculated using Win-XCOM software. Sulfuric acid abrasion and compressive strength tests were carried out and all samples showed good resistance to acid wear and pressure force. The neutron equivalent dose was measured using an average 4.5 MeV energy fast neutron source. Results were compared to 316LN type stainless steel, which commonly used in shielding radiation. New stainless steel samples were found to absorb neutron better than 316LN stainless steel at both low and high temperatures.

Measurement of the 58Ni(n, p)58Co and 58Ni(n, 2n)57Ni reaction cross-sections for fast neutron energies up to 18 MeV

The 58Ni(n, p) and 58Ni(n, 2n) reaction cross-sections were measured from threshold to 18MeV neutron energies using the activation and off-line $\gamma$ -ray spectroscopic technique. The quasi-monoenergetic neutron beam was generated using the 7Li(p, n) reaction. The results from the present work were compared with those of the literature and with the evaluated data from different libraries, like ENDF-B/VII.1, JENDL-4.0, CENDL-3.1, and JEFF-3.3. Nuclear model codes, like TALYS-1.9 and EMPIRE-3.2.3-Malta, were also used for a better description of the present work and literature data. The uncertainties in the measured cross-section were evaluated using the covariance analysis. The present experimental results were found to be in good agreement with those of the literature and with the evaluated data. The nuclear model code TALYS-1.9 was found successful in reproducing the experimental data for both reactions. However, the EMPIRE-3.2.3-Malta model code was found to be able to reproduce only the trend of the 58Ni(n, p) reaction excitation function. The present work provides a better insight on the comparison of both nuclear model codes. The present work is also essential for the production cross-section and the dose rate estimation of the medical isotope 58Co.

A walk-away safe, Very-Small, Long-LIfe, Modular (VSLLIM) reactor for portable and stationary power

This paper describes a Very-Small, Long-LIfe, and Modular (VSLLIM) reactor design, developed at the University of New Mexico’s Institute for Space and Nuclear Power Studies, which offers passive operation and decay heat removal and redundant control. During nominal operation and after shutdown, the VSLLIM reactor is cooled by natural circulation of in-vessel liquid sodium (Na), that is enabled using in-vessel chimney (1–2 m tall) and helically coiled tubes Na-Na heat exchanger, placed at the top of the downcomer. The reactor can generate 1.0–10 MWth, depending on the height of the chimney and the HEX design, at an average fission power density up to 23.47 MWth/m3. The VSLLIM reactor can potentially operate continuously, without refueling for ∼92 and 5.9 full power years (FPY), respectively, and slightly below atmospheric pressure, owing to the sodium low vapor pressure. The core is loaded with hexagonal assemblies of fuel rods with UN enriched to 13.76%, and has independent systems for emergency shutdown and nominal control. It also has two independent systems for safe passive removal of decay heat after shutdown, and following an unlikely malfunction of in-vessel HEX. These are liquid metal heat pipes (LMHPs) embedded in the primary vessel wall, and natural circulation of ambient air along the outer surface of the guard vessel wall. The LMHPs are thermally coupled to thermoelectric elements for generating 10 s of kW of auxiliary DC power, independent of on-site and off-site sources, both during nominal reactor operation and after shutdown. The sodium coolant enters the fast-neutron energy spectrum core at 610 K and exits at ≤755 K, depending on the reactor thermal power. At these temperatures, Na is compatible with the HT-9 Ferritic-Martensitic steel cladding, core structure and reactor vessel. The reactor would be fabricated, assembled and sealed at the factory, and transported by rail, truck or barge to a permanent site and installed underground, to protect against an airplane or a missile impact, and mounted on seismic isolation bearings, to resist earthquakes. At the site, the VSLLIM module could use a superheated steam Rankine cycle or a supercritical CO2 Brayton cycle for electricity generation at high thermal efficiency. Alternatively, a VSLLIM power module, with an open air Brayton cycle, could be deployed on a portable platform or a truck, to provide both electricity and/or process heat at remote sites, hospitals, data centers, natural disaster areas, arid or desert regions, and advanced military bases.

Performance validation of the first arm of FALSTAFF: 252Cf and 235U fission fragment characterisation

The renewed interest for the study of nuclear fission is mainly motivated by the development of GEN-IV reactor concepts, mostly foreseen to operate in the fast neutron energy domain. To support this development, new high-quality nuclear data are needed. In this context, a new experimental setup, the FALSTAFF spectrometer, dedicated to the study of nuclear fission is under development. Employing the double-velocity (2V) and energy-velocity (EV) methods, the fission fragment mass before and after neutron evaporation will be deduced and the correlation between prompt neutron multiplicity and fragment mass will be determined. The first arm of the spectrometer is achieved. It is composed of two SED-MWPC detectors (a combination of a foil to produce secondary electrons and a Multi-Wire Proportional Chamber to detect them) and an axial ionization chamber. The SED-MWPC give access to the velocity (V) via time-of-flight and position measurements. The ionization chamber measures the fragment kinetic energy (E) and the energy loss profile. Preliminary results for spontaneous fission of 252Cf and from the thermal-neutron induced fission experiment on 235U, performed at the Orphée reactor (CEA-Saclay, France), are presented.

Absolute cross section measurements of 238U(n,f) and 237Np(n,f) in the neutron energy range 1-2.4 MeV

New standard (n,f) cross sections other than 235U are important to study the relevant cross sections for Generation-IV power plants. A specific need for such standards is for performing new experiments with quasimonoenergetic neutron beams, such as those produced by Van de Graaf accelerators. Neutrons down-scattered to low energies in the experimental environment, so called room-return, become relevant for this type of measurements. Hence, a standard (n,f) cross section with a fission threshold is of great interest, in order to suppress the contribution from room-return background. For this reason we have performed two experiments at the VDG of the National Physical Laboratory to measure absolutely the (n,f) cross sections of 235U, 238U and 237Np in the fast neutron energy region. Our preliminary results are in agreement with the most up-to-date evaluations.

Improvement of neutron spectrum unfolding based on three-group approximation using CsI self-activation method for evaluation of neutron dose around medical linacs

We previously evaluated ambient neutron dose equivalent by using the self-activation of a CsI scintillator around a high-energy medical linear accelerator (linac) 128I saturated activities were successfully converted to neutron spectrum and ambient neutron dose equivalent by neutron spectrum unfolding with the “three-group approximation.” The principle of the three-group approximation is based on the assumption of fixed shapes of neutron energy spectra for each of the three energy regions to evaluate the neutron spectrum effectively. However, such a neutron dose evaluation with the unfolding method might be affected by the difference between the actual fast neutron energy spectrum and the assumed spectrum. In the present work, we modified the unfolding method by taking into account the differences in the shapes of fast neutron energy spectra for various medical linacs. We verified the unfolding method using Monte Carlo simulation with several neutron spectra obtained from published research articles. The modified three-group approximation evaluates the neutron doses more accurately than the conventional unfolding method.