Neutron Spectrum Research Articles

Safeguards and Security for High-Burnup TRISO Pebble Bed Spent Fuel and Reactors

Several high-temperature thermal neutron–spectrum pebble bed reactors are being commercialized. China has started up two helium-cooled pebble bed high-temperature reactors. In the United States, the X-Energy helium-cooled and the Kairos Power salt-cooled pebble bed high-temperature reactors will produce spent nuclear fuel (SNF) with burnups exceeding 150 000 MWd per tonne. The reactor fuel in each case consists of small spherical graphite pebbles (4 to 6 cm in diameter) containing thousands of small TRISO (microspheric tri-structural isotropic) fuel particles embedded in the fuel of zone these pebbles. The unique isotopic, chemical, and physical characteristics of this high-burnup SNF create a technical case to eliminate safeguards based on the low risk for use in nuclear weapons, while maintaining safeguards in terms of risk for use in radiological weapons. These safeguards could be reduced to the simple counting and monitoring of pebbles in storage. Alternatively, there is the option to create a special category with reduced requirements for this SNF in storage, transport, and disposal. No safeguards would be required for a repository with only this type of SNF. Reactor safeguards are required for fresh fuel, partly burnt fuel, and to identify unconventional pebbles with depleted uranium or other materials that might be used to create weapons-useable materials.

Measurements of AmBe and PuC spectra using EJ309 and stilbene scintillators

Two organic scintillators (EJ309 and stilbene) with their different detection chains and elaborate software were used to measure neutron spectra of AmBe and PuC sources. Method of shadow cone ensures minimization of scattered neutron influence. To separate neutrons and γ, a modern method of pulse shape discrimination was employed using fast digitizers. Calibration was performed by γ sources and three different methods for the identification of the Compton edge. Spectrometric systems also used different light output functions and unfolding methods. The same spectra were calculated using the MCNP6 code and the JENDL-5 library. The measured and calculated results were mutually compared, and also with the published results available in the literature. Generally, the results agree on the main peaks of neutron spectra but differ in their abundance. The lower limit for the ability to distinguish between neutrons and γ in spectrometry with organic scintillators was found at about 1.5 MeV.

Neutron spectra from photonuclear reactions: Performance testing of Monte-Carlo particle transport simulation codes

The production of neutrons in photon-induced nuclear reactions in the giant-dipole-resonance energy domain remains a topic of high interest for various applications, including the activation and decommissioning of electron accelerator facilities, the detection of illicit materials for homeland security, and the evaluation of neutron dose received by patients during radiotherapy treatments. General-purpose Monte-Carlo (MC) simulation codes for particle transport are intensively used to account for photoneutron production in these applications. However, due to the current scarcity of measured photoneutron energy spectra in the literature, experimental validation of MC-simulated photoneutron energy distributions is not always feasible. Therefore, a critical benchmark among simulation results from various MC codes presently appears as the only option to systematically assess their capabilities in accurately simulating photoneutron production for nuclear reactions of interest. In this work, neutron energy spectra from several targets under irradiation by 20 MeV photons are simulated, employing various state-of-the-art MC codes (FLUKA, Geant4, MCNP6, and PHITS) in their default or generally employed settings. A detailed analysis of the simulated neutron spectra allows one to not only assess the performance of various MC codes in applications such as those mentioned above, but also to partially gauge the incurred systematic uncertainty, and to highlight the present need for more comprehensive evaluated nuclear data in this domain. Ultimately, this work suggests that more prudence is required when using MC codes for applications where photonuclear reactions play a dominant role and where not only the production rate but also the energy spectrum of the emitted neutrons matters.

Neutronic optimization of thorium-based fuel configurations for minimizing slightly used nuclear fuel and radiotoxicity in small modular reactors

Effective management of slightly used nuclear fuel (SUNF) is crucial for both technical and public acceptance reasons. SUNF management, radiotoxicity risk, and associated financial investment and technological capabilities are major concerns in nuclear power production. Reducing the volume of SUNF can simplify its management, and one possible solution is utilizing small modular reactors (SMR) and advanced fuel designs like those with thorium. This research focuses on studying the neutronic performance and radionuclide inventory of three different thorium fuel configurations. The mass of fissile material in thorium-based fuel significantly impacts Kinf, burn-up, and neutron energy spectrum. Compared to uranium, thorium as a fuel produces far fewer transuranic elements and less long-lived fission products (LLFPs) at the end of the core cycle (EOC). However, certain fission product elements produced from thorium-based fuel exhibit higher radioactivity at the beginning of the core cycle (BOC). Physical separation of thorium and uranium in the fuel block, like seed-and-blanket units (SBU) and duplex fuel designs, generate less radioactive waste with lower radioactivity and longer cycle lengths than homogeneous or mixed thorium-uranium fuel. Furthermore, the SBU and duplex feel designs exhibit comparable neutron spectra, leading to negligible differences in SUNF production between the two.

Pencil beam kernel-based dose calculations on CT data for a mixed neutron-gamma fission field applying tissue correction factors.

Objective.For fast neutron therapy with mixed neutron and gamma radiation at the fission neutron therapy facility MEDAPP at the research reactor FRM II in Garching, no clinical dose calculation software was available in the past. Here, we present a customized solution for research purposes to overcome this lack of three-dimensional dose calculation.Approach.The applied dose calculation method is based on two sets of decomposed pencil beam kernels for neutron and gamma radiation. The decomposition was performed using measured output factors and simulated depth dose curves and beam profiles in water as reference medium. While measurements were performed by applying the two-chamber dosimetry method, simulated data was generated using the Monte Carlo code MCNP. For the calculation of neutron dose deposition on CT data, tissue-specific correction factors were generated for soft tissue, bone, and lung tissue for the MEDAPP neutron spectrum. The pencil beam calculations were evaluated with reference to Monte Carlo calculations regarding accuracy and time efficiency.Main results.In water, dose distributions calculated using the pencil beam approach reproduced the input from Monte Carlo simulations. For heterogeneous media, an assessment of the tissue-specific correction factors with reference to Monte Carlo simulations for different tissue configurations showed promising results. Especially for scenarios where no lung tissue is present, the dose calculation could be highly improved by the applied correction method.Significance.With the presented approach, time-efficient dose calculations on CT data and treatment plan evaluations for research purposes are now available for MEDAPP.

A two-layer Timepix3 stack for improved charged particle tracking and radiation field decomposition

Abstract We characterize a novel instrument designed for radiation field decomposition and particle trajectory reconstruction for application in harsh radiation environments. The device consists of two Timepix3 assemblies with 500 µm thick silicon sensors in a face-to-face geometry. These detectors are interleaved with a set of neutron converters: 6LiF for thermal neutrons, polyethylene (PE) for fast neutrons above 1 MeV, and PE with an additional aluminum recoil proton filter for neutrons above ∼4 MeV. Application of the coincidence and anticoincidence technique together with pattern recognition allows improved separation of charged and neutral particles, their discrimination against γ-rays and assessment of the overall directionality of the fast neutron field. The instrument's charged particle tracking and separation capabilities were studied at the Danish Center for Particle Therapy (DCPT), the Proton Synchrotron, and Super Proton Synchrotron with protons (50–240 MeV), pions (1–10 GeV/c and 180 GeV/c). After developing temporal and spatial coincidence assignment methodology, we determine the relative amount of coincident detections as a function of the impact angle, present the device's impact angle resolving power (both in coincidence and anticoicidence channels). The detector response to neutrons was studied at the Czech Metrology Institute (CMI), at n_ToF and the Los Alamos Neutron Science Center (LANSCE), covering the entire spectrum from thermal up to 600 MeV. The measured tracks were assigned to their corresponding neutron energy by application of the time of flight technique. We present the achieved neutron detection efficiency as a function of neutron kinetic energy and demonstrate how the ratio of events found below the different converters can be used to assess the hardness of the neutron spectrum. As an application, we determine the neutron content within a PMMA phantom just behind the Bragg-peak during clinical irradiation condition with protons of 160 MeV.

Burnup-Dependent Neutron Spectrum Behaviour of a Pressurised Water Reactor Fuel Assembly

Understanding the behaviour of a neutron spectrum with burnup is important for describing various phenomena associated with reactor operation. The quest to understand the neutron spectrum comes with a lot of questions. One question that is usually asked by students is: Does the neutron spectrum harden or soften with burnup? Most textbooks used by students do not provide a definite answer to this question. This paper seeks to answer this question using a 3D model of a standard 17 × 17 pressurised water reactor fuel assembly. Two cases were studied using the Serpent Monte Carlo code: the first considered the fuel assembly with constant boron concentration (traditionally found in many published papers), and the second considered boron iteration (where the boron concentration was reduced with burnup). Neutron spectra for the two cases at beginning of life and end of life were compared for spectral shifts. In addition, thermal spectral indices were used to assess spectrum hardening or softening with burnup. Spectral shifts to lower energies were observed in the thermal region of the neutron spectrum, whereas the fast region experienced no spectral shift. There was an increase in thermal spectral indices indicating that the spectrum became soft with burnup.

Neutronic Analysis and Fuel Cycle Parameters of Transuranic-UO2 Fueled Dual Cooled VVER-1000 Assembly

Abstract The feasibility of transuranic (TRU) fuel rods has been studied for a dual-cooled VVER-1000 assembly along with the conventional UO2 fuel rods. It has been found that the heterogeneous arrangements of TRU and UO2 fuel rods help to improve the multiplication factor, enhance the fuel cycle parameters, and maintain the negative Doppler coefficient of the reference VVER-1000 assembly. Within the different heterogeneous combinations of TRU and UO2 fuel rods, the equal number of UO2 and TRU fuel rods, which is referred to as model 6 increased the multiplication factor of the reference assembly by 7.66% at the beginning of the cycle. Furthermore, the fuel cycle parameter becomes almost double for this model in comparison with the reference assembly. TRU rods make the neutron spectrum of reference assembly slightly harder and increase the fast fission rate. However, the Doppler coefficient becomes less negative, and flux level decreases with increasing the number of TRU rods. By considering the safety parameters and neutronic behavior, model 6 (equal number of UO2 and TRU fuel rods) is to be effective for the annular fueled VVER-1000 reactor.

Comparative Radiation Response of GaN and Ga2O3 Exposed to Ground-Level Neutrons

In this work, the radiation response of bulk GaN and Ga2O3 materials exposed to ground-level neutrons is studied by Geant4 numerical simulation, considering the whole atmospheric neutron spectrum at sea level, from thermal to high energies (GeV). The response of the two materials is compared in terms of the number and type of interactions and the nature of the secondary products produced, particularly in nuclear reactions. Our results highlight the importance of 14N(n,p)14C neutron capture in the radiation response of GaN, leading to large differences in the behavior of the two materials in terms of susceptibility to thermal and intermediate-energy (below 1 MeV) neutrons.

Design and validation of a single cylindrical neutron spectrometer using a gold activation foil

A single cylindrical neutron spectrometer (SCNS) with a gold activation foil was designed for mono-directional high-flux neutron spectrum measurement applications. A total of seven gold activation foils with a diameter of 11 mm and a thickness of 0.2 mm were placed in the moderator as detectors with a depth variation of 15 mm. The moderator was cylindrical with a diameter of 8 cm and a length of 16 cm and made of polyethylene. Outside the moderator is a neutron shield made of 3 mm boron-rubber and 0.5-mm cadmium and covered with an aluminum casing of 5 mm thickness. The response value of each detector was determined by simulation using the Monte Carlo MCNPX 2.7 program. The simulated response detectors was then partially validated by comparing the simulation and the measurement result with an AmBe neutron source, where the difference is around 3%. The obtained results showed that SCNS with a gold activation foil could be used to measure the neutron spectrum in a high-flux neutron field.

Distribution of neutron energy and yield variations with natLi target thickness at proton energies between 8 and 20 MeV

The natLi(p,n) system is a prominent source of mono-energetic neutrons capable of delivering neutrons in a large energy band extending from the Coulomb barrier to few hundreds MeV. These mon-energetic neutrons play a pivotal role in various academic and engineering applications. Present work emphasises on the variation of the emission yields and spectral features of natLi(p,n) neutrons at two different Li-foil thicknesses (8 and 16 mg cm−2). The study indicated a proton energy dependent variation in the emission neutron yields. At higher Li target thickness with increase in the neutron yield was observed ∼10 times for 8 MeV whereas, ∼1.5 time increase was observed with 20 MeV incident protons. The spectral shape also indicated softening of neutron spectra with possible merging of high energy emission neutron peaks with lowering of the average energy for 16 mg cm−2 thick Li target.

Status of the 226Ra nuclear data library and its impact on the production amount of 225Ac via the 226Ra (n,2n) reaction

ABSTRACT 225Ac is a promising alpha emitter for targeted alpha therapy. However, the current capability of 225Ac supply is limited to approximately 63 GBq/y, primarily relying on the natural source of 229Th stocked at a few institutes. Therefore, alternative 225Ac production methods are highly desired. The research and development of 225Ac and its parent nuclide production methods using accelerators and reactors to target 226Ra and 232Th are actively pursued worldwide. Hence, the authors focused on the 226Ra(n,2n)225Ra method using fast neutron spectra as an application of the experimental fast reactor Joyo. This study investigated the status of nuclear data libraries for 226Ra, an essential target for 225Ac production, and evaluated the impact of different nuclear data libraries on the amount of 225Ac produced. Consequently, cross-sections with covariance data were stored in TENDL-2021, JEFF-3.3, and EAF-2010 but not in ENDF or JENDL, the major nuclear data libraries. Furthermore, no consistency occurred among the respective nuclear data, and the 225Ra production amount varied.

Potential of Radioactive Isotopes Production in DEMO for Commercial Use

There is widespread use of nuclear radiation for medical imagery and treatments. Worldwide, almost 40 million treatments are performed per year. There are also applications of radiation sources in other commercial fields, e.g., for weld inspection or steelmaking processes, in consumer products, in the food industry, and in agriculture. The large number of neutrons generated in a fusion reactor such as DEMO could potentially contribute to the production of the required radioactive isotopes. The associated commercial value of these isotopes could mitigate the capital investments and operating costs of a large fusion plant. The potential of producing various radioactive isotopes was studied from material pieces arranged inside a DEMO equatorial port plug. In this location, they are exposed to an intensive neutron spectrum suitable for a high isotope production rate. For this purpose, the full 3D geometry of one DEMO toroidal sector with an irradiation chamber in the equatorial port plug was modeled with an MCNP code to perform neutron transport simulations. Subsequent activation calculations provide detailed information on the quality and composition of the produced radioactive isotopes. The technical feasibility and the commercial potential of the production of various isotopes in the DEMO port are reported.

Adaptive energy group division in the few-group cross-section generation for full spectrum reactor modeling with deterministic method

Advanced nuclear reactors, especially the newly developed small and micro-reactors have complex neutron spectrum, which makes the deterministic reactor core calculations sensitive to the energy group structure of few-group cross-sections. To avoid significantly increasing the cost of energy discretization in the core calculation, two energy group structures with 31 groups and 33 groups were adopted for typical thermal and fast reactor cores, respectively. Then, an adaptive scheme of group division for reactor cores with a medium neutron spectrum was proposed. The works were based on the full spectrum nuclear reactor analysis code SARAX/TULIP. An equivalent one-dimensional model of the core was proposed to capture the key neutron spectrum features of the reactor core. Such features were used to adaptively determine a few-group structure for the following reactor core calculations. Then, the neutron spectrum in different zones with more details was calculated. With this spectrum, the cross-sections were condensed into the determined energy groups. Three tests based on different neutron spectrum were calculated to verify the schemes. The results show that using the adaptive energy group division scheme, the following core calculation can meet the accuracy requirement of different reactors with different neutron spectra.

Exploration of genetic algorithms to build a balanced neutron spectra dataset useful to train unfolding techniques based on artificial neural networks

Diverse domains need neutrons unfolding technics to assess the incident neutron energy spectrum. Examples are radiation protection, nuclear reactor physics or criticality safety. Traditionally, methods based on the Bayesian approach requires an initial guess of the solution which may significantly impact the unfolding result. This work proposes a novel method for neutron spectrum reconstruction using machine learning (ML) techniques trained on a large dataset. To ensure the ML algorithm to perform on a large domain of application particular attention has been paid to the dataset creation. We propose a comparison of two methods of building large dataset where the most adequate solution is obtained using a dynamic genetic algorithm (GA). This GA targets optimal combinations of 48 parameters to generate a variety of neutron spectra. The resulting dataset is then used to train a new convolutional neural network architecture for unfolding neutron spectra. Obtained performance metrics of the tested architecture show high efficiency and emphasize the added value of the built dataset.

Measurement of secondary neutron spectra induced by 480 MeV proton and 430 MeV/u 4He beams with a thick aluminum target

Knowledge of the characteristics of secondary neutrons produced by the interaction of Galactic Cosmic Radiation with spacecraft shielding materials is becoming increasingly important for predicting and mitigating biological risks of space explorers during deep-space travel. Hadron accelerators for medical applications are well suited to reproduce part of the conditions found in deep-space in terms of ion species and energies. The objectives of this work are to measure the secondary neutron spectra produced by proton and helium ion beams hitting an aluminum target with energies that correspond to the Galactic Cosmic Radiation peak during solar minimal activity and to validate and compare physical models of Monte Carlo simulations. Neutron spectra were measured with the extended-range Bonner sphere system NEMUS at two positions, 0° and 90° relative to the direction of the primary ion beam. The experimental setup consisted of 480 MeV proton and 430 MeV/u 4He beams colliding with a 30×30×63.5 cm3 aluminum target. The experimental neutron spectra were analyzed using the MAXED unfolding code and compared to several Monte Carlo simulation codes. The results show deviations in terms of the shape of the neutron energy distributions ranging between 1% and 14% and of the integral quantities of fluence and ambient dose equivalent ranging between 1% and 5.2%.

Godiva IV central cavity neutron environment characterization with threshold neutron detectors

Godiva IV is a cylindrical fast burst reactor comprised of approximately 65 kg of highly enriched uranium that is operated by Los Alamos National Laboratory and sited at the National Criticality Experiments Research Center at the Nevada National Security Site in Nevada in the United States. Godiva IV is typically operated at delayed critical and in the regime spanning from sub-prompt to super-prompt bursts. Godiva IV is used for sample irradiations, criticality safety demonstrations, dosimetry studies, and for studying super-prompt behavior. In preparation for both an upcoming experiment to reduce uncertainties in the prompt fission spectrum for 235U using threshold neutron detectors, and for future research using Godiva IV, it was desired to exercise the process of the selection of threshold neutron detectors/activation foils, radiation metrology, and the subsequent adjustment of the neutron spectrum. For this exercise, nine high purity threshold neutron detectors/activation foils were irradiated in a Godiva IV burst. The foils were then analyzed using a high-purity germanium detector in the NCERC counting laboratory to determine end of irradiation specific activities for available IRDFF-II reactions. This work summarizes the Godiva IV foil irradiation, radiation metrology results, and adjusted neutron spectrum. The results of this exercise ultimately characterized the neutron environment inside the sample irradiation cavity inside Godiva IV to a higher degree than previously performed, informed decisions for the upcoming larger scale experiment, and will inform future neutron spectrum characterizations at NCERC.

Development and testing of WEASEL for least-squares neutron spectrum adjustment

A new least-squares adjustment program called WEASEL has been developed. WEASEL is implemented in Python and uses standard ASCII interfaces for user input and output, and HDF5 interfaces for cross section, covariance data, and weighting function data. WEASEL allows the user to supply a trial neutron spectrum in an arbitrary group structure. The trial neutron spectrum is re-binned into a group structure consistent with the covariance data using a weighting function that can either be read from the data library or supplied as input by the user. WEASEL performs least-squares adjustments of a single irradiation environment. The adjusted neutron spectrum, covariance matrices, and other derived quantities can then be plotted and interrogated with standard software packages.

On radiation situation in vicinity of PET production cyclotron

The medical cyclotrons intended for the production of medical isotopes are relatively widespread. The most common radioisotope product of these facilities is 18F which origins from 18O(p, n)18F reaction. Due to relatively large amounts of produced nuclei, the leakage neutron and gamma flux is a big issue in radiation safety in such cyclotron labs. Due to relatively high energies of accelerated protons, high energy gammas and neutrons are produced with energies above 10 MeV with high penetration properties. Neutron and gamma spectra around cyclotron IBA Cyclone 18/9 intended for production of 18F was measured with stilbene detector. It was compared to simulated neutron spectra with various Monte Carlo codes and with spectral indices calculated from reaction rates in activation detectors. Strong disagreement in simulation of spectra in backward angles was found for MC codes Geant4 and MCNP6.2.

Turbulent induced vibration of a guide tube in experimental reactor

The design of advanced experimental nuclear reactors consists in integrating safety and operational requirements as well as reaching targets in terms of thermal power and neutron spectrum. In order to meet theses constraints, slender structures with little supports and crossing the entire reactor vessel are implemented in the reactor and are subjected to an axial flow that generates flow-induced vibration (FIV). From the industrial point of view, the mastering of the occurrence of FIV and its associated wear in case of contacts is requested. The stationary fluid forces applying on slender tubular structure in response to its motion are of primary interest since they can significantly affect the vibration amplitudes and even the stability of the system, especially in case of confined axial flow with high velocities (typically higher than 10 m/s). In this study it proposed to compare two approaches to estimate the vibration induced by turbulent excitation of an industrial device encountered in a research nuclear reactor. The control-rod guide-tube mock-up of the Jules Horowitz Reactor, previously tested in an hydraulic channel at the Technical Center from Le Creusot in France, is retained for this benchmark. Two models are proposed, one based on derivation of leakage flow theory and the other one was based on potential flow theory with adjusted coefficients given by CFD simulations. The flow induced vibration amplitude is consistent with the experimental data. Also, the calculation and experiment provide similar trends when the boundary conditions are changed.