Neutron Flux Research Articles

Comparison of Neutronic Analysis for 250 MWth Molten Salt Reactor using Monte Carlo OpenMC Code with Different Nuclear Data Libraries of JENDL 5.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF 3.3

Abstract Nuclear data libraries playian important role in the accuracy of neutronic aspect calculations, which determine various factors in a nuclear reactor design. Likewise, deterministic or Monte Carlo methods significantly affect the calculation results. Thus, the combination of nuclear data library and calculation methods needs to be studied carefully. This research will conduct neutronic analysis for the Molten Salt Reactor (MSR) design, one of the Generation IV reactor types with advanced features for future energy resources, using the Monte Carlo-based program OpenMC and different nuclear data libraries. MSR will be designed for a 250 MWth power capacity, operable for five years. The fuel used will be LiF-BeF2 as a coolant and ThF4-UF4 as the fuel. The compositions for each fuel will be optimized to achieve critical conditions during five years of operation. This research aims to evaluate the use of differentinuclear data libraries on reactor criticality conditions, using the nuclear data libraries ENDF/VIII-B, ENDF/VII.0, JEFF 3.3, and JENDL 5.0. To perform the calculation and analysis, MSR chain depletion modification has to be done for different nuclear data libraries. Then, evaluation will be conducted on effective multiplicationifactor (k-eff), neutron flux spectrum, and cross-section. The results of this research show that the use of JENDL 5.0 provides more optimal critical conditions. This is due to the completeness of JENDL 5.0 nuclear data covering 795 nuclide data and higher fission cross-section values. While ENDF/VIII-B has 557 nuclide data evaluations compared to ENDF/VII.0’s 423 nuclide data, more stable k-eff values are shown for the use of ENDF nuclear data libraries. This research is expected to significantly contribute to meeting Indonesia’s energy needs by providing efficient solutions in MSR design in neutronic aspects using the most accurate nuclear data library.

Full-scale modeling and experimental study of a gas neutron tube with a Penning ion source

This paper studies full-scale gas neutron tube modeling, including the following processes: gas discharge combustion in a Penning ion source, particle motion in an ion optical system, and modeling the in-target processes, such as sputtering, diffusion, thermal desorption of hydrogen isotopes, and nuclear reactions. Plasma modeling in quadrupole electric and axial magnetic fields was based on the electrostatic particle-in-cell method with molecular kinetic processes. The TechX Vsim software package was used. The neutron tube element sputtering by ions was simulated using SRIM/TRIM software based on Monte–Carlo methods. The OpenFOAM, as an open integrated platform for numerical simulation in continuum mechanics, was used to calculate the hydrogen thermal desorption activated by ion irradiation. The time-dependent neutron yield modeling was performed using the Geant4 software based on Monte–Carlo methods with CHIPS-TPT VNIIA-developed library. In addition, an experimental study of a gas neutron tube with a Penning ion source was conducted here as well. Details are given on the experiment and measurement technique used in this study. The operating characteristics for the gas neutron tube, including amplitude-time characteristics of current flashes (discharge and extraction currents), were determined. The neutron flux dependencies on the discharge current at various accelerating voltages were also obtained. Finally, a comparison between the experimental and calculated results is presented.

Retrospective analysis of treatment-positioning accuracy and dose error in boron neutron capture therapy using a sitting-position treatment system for head and neck cancer

The neutron beam in boron neutron capture therapy (BNCT) exhibits poor directionality and significantly decreasing neutron flux with increasing distance. Therefore, the treatment site must be close to the irradiation aperture. Some patients with head and neck cancer may benefit from a sitting-position setup. The study aim was to evaluate the treatment-positioning accuracy and dose error in sitting patients receiving BNCT. Thirty-two patients with head and neck cancer who underwent sitting-position BNCT at Southern Tohoku BNCT Research Center were included in the study. Horizontal (ΔX) and vertical (ΔY) errors were defined as the displacement between the treatment planning system (TPS) digital reconstructed radiograph and the pre-treatment X-ray image. Using in-house software, image matching was performed. The beam-axial directional (ΔZ) error was compared with the parameters entered into the TPS and the actual pre-treatment measured values. The translational-position error was reflected in the TPS’s patient coordinate system with respect to the reference plan. Re-dose calculations were performed to evaluate the effect of positional error on tumor and normal-tissue doses. The [ΔX, ΔY, ΔZ] DRR-CR mean ± 1SD were − 0.40 ± 2.0, 0.30 ± 2.3, and − 1.4 ± 1.5 mm, respectively. The Dmean and D98% tumor-dose errors were 1.22 % ± 1.44 % and 0.99 % ± 1.63 %, respectively. The D2% pharyngeal and oral mucosal-dose errors were 0.98 % ± 1.91 % and 1.21 % ± 1.78 %, respectively. The tumor- and normal-tissue dose errors were typically < 5 %. High-precision treatment was feasible in sitting-positioned BNCT.

Ion optical design of the magnetic proton recoil neutron spectrometer for the SPARC tokamak.

A magnetic proton recoil (MPR) neutron spectrometer is being designed for SPARC, a high magnetic field (BT = 12T), compact (R0 = 1.85m, a = 0.57m) tokamak currently under construction in Devens, MA, USA. MPR neutron spectrometers are versatile tools for making high fidelity abinitio calibrated measurements of fusion neutron flux spectra and have been used to infer fusion power, ion temperature, fuel ion ratio, and suprathermal fuel populations at several high performance fusion experiments. The performance of an MPR neutron spectrometer is in large part determined by the design of the magnetic field, which disperses and focuses recoil protons. This article details the ion optical design of a high-resolution MPR neutron spectrometer, including the amelioration of image aberrations due to nonlinear effects. An optimized design is presented that achieves ion optical energy resolution δE/E < 1% and focal plane properties that enable straightforward integration with the hodoscope detector array.

Design study of gas-cooled fast reactor with natural uranium as fuel employing modified CANDLE shuffling strategy in the axial direction

Abstract This study investigated gas-cooled fast reactor performance utilizing natural uranium as the fuel employs modified constant axial shape of neutron flux, nuclide densities, and power shape during life of energy production (CANDLE) shuffling strategy in the axial direction. The performance has been carried out on a reactor with various power levels in the range of 2,700–3,100 MWt and refueling every 10 years of burnup. The main importance of the modified CANDLE (MCANDLE) is that it can utilize natural uranium as fuel without needs reprocessing or enrichment plant. During this work, the core has been partitioned into 10 regions of similar volume along the axial direction. The fuel was filled into the first region, then transported to the second region after 10 years of burnup, and then to the third region 10 years later. This technique was practiced to all 10 regions, and the fuel in the tenth region was discharged in the reactor core. 10 % of natural thorium, natural uranium and enriched nitride (15 N) were used as fuel input. The calculations were performed employing a standard reactor analysis code (SRAC). The collision probability method (PIJ) module of SRAC was used to calculate cell burnup, and the reactor core design calculations were done with the CITATION module of SRAC, which used JENDL-4.0 as a nuclear data library. The result shows that a high power level causes an increasing rate of burnup level and effective multiplication factor. The highest average discharge burnup level is about 31.3 % HM for case E and the lowest average discharge burnup is about 27.2 % HM for case A.

Development of the neutral gas diagnostic system for neutral pressure measurements and gas analysis on the SPARC tokamak.

A suite of plasma diagnostics will be installed on the SPARC tokamak to allow for real-time plasma control, an investigation of high-field tokamak physics, and to de-risk the design of ARC, a compact fusion power plant with the aim to supply electricity to the grid. Among these diagnostics is the neutral gas diagnostics system (NTGS), a set of pressure sensors and gas analyzers used to monitor neutral pressure and gas composition for plasma control, optimization of wall conditioning, and helium ash removal, among other measurement functions linked to operational and scientific research needs. While reliable measurements of neutral pressure and gas composition have been fielded on existing magnetic-confinement fusion devices, SPARC represents a step increase in challenge due to its larger power density, higher field, high vacuum vessel bake temperatures, and higher neutron flux environment, as well as a step decrease in the accessibility for maintenance of in-vessel sensors. Multiple sensor types will be employed to have defense-in-depth and mitigate common failure modes. The NTGS system is currently progressing through final design, working to close out decisions using prototyping and analysis, and then moving on to procuring sensors for assembly and installation on SPARC. This paper outlines the current status of the system design and the diagnostic requirements that motivate neutral gas measurements on SPARC, as well as highlights the planned prototyping activities.

Design and test of a module of a breathable Electrostatic Shield for the MITICA 1 MV negative ion Beam Source

The electrical insulation of the MITICA Beam Source at 1 MV is a challenging issue, which has not been fully addressed so far on the basis of experimental results and of theoretical models available in literature. Being MITICA the full-size prototype of the Heating Neutral Beam Injector for the ITER fusion experiment, its electrical insulation is constituted just by vacuum gaps and alumina insulators, since other insulating materials such as SF6 gas or fibreglass-reinforced plastic (FRP) would be quickly degraded by the expected neutron flux produced by fusion reaction. Extrapolations based on HV tests on reduced-scale models have recently indicated the risk of electrical breakdowns in the vacuum gap between electrodes nominally operating at -1 MV and the vacuum vessel (at ground potential). The risk of electrical breakdown can be mitigated by introducing an intermediate Electrostatic Shield (ES), which essentially is an equipotential (metallic) enclosure surrounding the HV electrode, so as to divide the vacuum gap in two independent insulating gaps of 400 kV and 600 kV respectively. However, for optimal negative ion production, the ion source shall operate in H2 or D2 at a pressure of ∼ 0.3 Pa and unavoidably produces a flow of gas leaking out in the surrounding vacuum. Thus, the presence of an intermediate shield can substantially increase the background gas pressure in the vacuum gaps, and, due to the large gap length (0.6 m), exacerbate the risk of breakdown when the pressure approaches the conditions of Paschen-type discharges. In addition to this, RF-induced breakdowns were found on the rear side of the ion source during the operation of the prototype source SPIDER, which were somewhat correlated to a relatively high hydrogen pressure in that area. For these reasons, a structure capable of constituting a full equipotential barrier all around the BS and, at the same time, having sufficient gas conductivity (breathability) to allow efficient pumping of background gas, has been designed. In the first part of the paper, the requirements and design optimization of a breathable module of the intermediate ES are described. Then, an experimental campaign for the validation of the electrode implementation the test configurations and the experimental procedure is discussed.

Determination of the elemental composition and changes in thermal and electrical conductivity of “Hanford” and low-ash medium-grained graphite’s grade graphites depending on the fast neutrons fluence

Studying the properties of graphite recovered from operating nuclear reactors is vital for predicting the properties and integrity of graphite as part of assessing the continued operation and life extension of nuclear reactors. The purpose of the study is to determine the electrical conductivity and thermal conductivity of low-ash medium-grained graphite’s grade graphite in the thermal column masonry of the VVR-SM research reactor in the measurement temperature range corresponding to the conditions of normal operation of the reactor to determine the service life. In this work, the change in thermal conductivity and electrical conductivity of GMZ graphite was studied, as well as for comparison of Hanford grade graphite depending on the fluence of fast neutrons and the measurement temperature. The dependence of electrical conductivity and thermal conductivity on dose and temperature has been established. It has been shown that the greater the neutron fluence, the more both the thermal conductivity and electrical conductivity of the material decreases. The service life of the thermal column has been determined.

Валідація пакета програм MCPV з безперервною за енергією бібліотекою мікроскопічних констант

The results of the validation of the code package MCPV for the calculation of radiation transport within VVER-1000 reactor type near-vessel space using a continuous-energy data library are presented. The validation is carried out based on the experimental data obtained in mock-up and full-scale experiments. It is shown that the code package MCPV can be used to simulate neutrons and gamma quanta transport through the complicated heterogeneous environment of a nuclear reactor and to obtain the valid values of the functionals of the neutron and gamma fluxes within VVER-1000 near-vessel space.

Improvements in Transient Testing Reactor (TREAT) with a Choice of Filter

The safe and reliable operation of nuclear reactors has always been one of the topmost priorities in the nuclear industry. Transient testing allows us to understand the time-dependent behavior of the neutron population in response to either a planned change in the reactor conditions or unplanned circumstances. These unforeseen conditions might occur due to sudden reactivity insertions, feedback, power excursions, instabilities, and accidents. To study such behavior, we need transient testing, which is like car crash testing to estimate the durability and strength of a car design. In nuclear designs, such transient testing can simulate a wide range of accidents due to sudden reactivity insertions and helps study the feasibility and integrity of the fuel used in certain reactor types. This testing involves a high neutron flux environment and real-time imaging technology with advanced instrumentation with appropriate accuracy and resolution to study the fuel slumping behavior. With the aid of transient testing and adequate imaging tools, it is possible to test the safety basis for reactor and fuel designs that serves as a gateway in licensing advanced reactors in the future. To that end, it is crucial to fully understand advanced imaging techniques both analytically and via simulations. This paper presents an innovative method of supporting real-time imaging of fuel pins and other structures during transient testing. The major fuel-motion detection device that is studied in this dissertation is the Hodoscope which requires collimators. This paper provides 1) an MCNP model and simulation of a TREAT core with a central fuel element replaced by a slotted fuel element that provides an open path between test samples and a hodoscope detector, and 2) a choice of good filter to improve image resolution.

Measurement of fast neutron induced (n,γ) reaction cross-section of 68Zn, 96Zr, 121Sb and 123Sb in the energy range of 1 to 2 MeV

The (n,γ) reaction cross-section for the elements 68Zn, 96Zr, 121Sb and 123Sb, present in the reactor structural/shielding materials, was measured by neutron activation technique in the neutron energy region of 1–2 MeV as very limited data is available in this energy range. Further, the neutron spectrum peaks in this energy region for the fast breeder reactors and proposed accelerator driven sub-critical systems. The natural strontium (natSr) element was used as a neutron flux monitor by considering effective combined reaction cross-section for 86Sr(n,γ)87Srm and 87Sr(n,n′)87Srm reactions. The pellets of mixture of sample and monitor were irradiated by a quasi-mono energetic fast neutron beam, generated by 7Li(p,n)7Be reaction at FOTIA, Bhabha Atomic Research Centre, Mumbai, India. The activity of activation products was measured by off-line gamma-ray spectrometry using High Purity Germanium Detector (HPGe). The present data with improved uncertainty and covariance analysis enhance the cross-section data base for better constraining the evaluated data and theoretical models. The theoretical (n,γ) reaction cross-sections were calculated using TALYS 1.96, which could reasonably explain the present data with the Fermi gas level density prescription.

Research on the high-performance computing method for the neutron diffusion spatiotemporal kinetics equation based on the convolutional neural network

Due to the uncertainty of computational results and the lack of interpretability of models in solving physical field equations in current deep learning, this paper designs a convolutional neural network that can be used to solve the neutron diffusion spatiotemporal kinetics equation in polar and cylindrical coordinate systems. This algorithm directly utilizes the macroscopic cross-section of the material without using the lattice homogenization method, replaces the finite volume method with the extended matrices, and solves the extended matrices using the convolutional kernels instead of the iterative algorithms. Taking the simplified Tsinghua High Flux Reactor (THFR) as an example, the feasibility of the algorithm is verified on the PyTorch platform and compared with the calculation results of the source iteration method running on the GPU. The calculation results show that when the number of grids in the radial and axial sections of the simplified THFR model is 804,600 and 3,576,000, respectively, and the algorithm is iterated 3000 times, the normalized power of the convolutional neural network and the source iteration method converges to 10−10, and the maximum point by point error of the neutron flux density of the above two algorithms converges to 10−5. The computational time consumed by the convolutional neural network is approximately 880.64 s and 3729.62 s, which reduces the computational time by 4.66% and 5.05% compared to the GPU parallel accelerated source iteration method, and the former consumes 43.75% less memory compared to the latter. The convolutional neural network is mainly used as the virtual physics engine for the THFR digital twin system, in addition to solving the neutron diffusion spatiotemporal kinetics equation and further improving computational speed. The algorithm directly utilizes the neutron macroscopic cross-section of the material to calculate the neutron flux density distribution without using the lattice homogenization, providing theoretical guidance and algorithm support for developing the high-precision multi-physical field coupling model.

The production and separation of 161Tb with high specific activity at the University of Utah

Targeted radiotherapy (TRT) is an increasingly prominent area of research in nuclear medicine, particularly in the context of treating cancerous tumors. One radionuclide of considerable interest for TRT is terbium-161 (t1/2 = 6.95 days), which undergoes beta emission and shares similar decay properties as 177Lu (FDA-approved as LUTATHERA® and PLUVICTO®). Besides beta emission, 161Tb also emits a significant number of conversion and Auger electrons further enhancing its therapeutic potential. Terbium-161 can be produced using nuclear reactors through an indirect neutron capture reaction, G64160dn,γG64161d→3.66min,β−T65161b, from 160Gd targets. However, a key challenge in utilizing 161Tb for TRT lies in effectively separating target and product materials to attain high specific activity for radiolabeling. Here, we detail the production of no-carrier added 161Tb using low flux research reactors (mean thermal (<0.625 eV) neutron flux: 1.356×1012n∙cm−2∙s−1) like the University of Utah TRIGA Reactor, using enriched 160Gd2O3 targets (1.5 ± 0.3 μCi of 161Tb per mg of 160Gd target per hour of irradiation). We also developed a separation technique based on cation exchange and extraction chromatography, suitable for mCi level irradiations with targets exceeding 200 mg. In a simulated full-scale irradiation, 161Tb was successfully isolated from large mass targets using cation exchange (AG 50W-X8, with 2-hydroxyisobutyric acid at 70 mM, pH 4.75) and extraction chromatography (LN Resin, 0.5–0.75 M HNO3) methods. This resulted in high apparent molar activities of [161Tb]Tb-DOTA (113 ± 3 MBq/nmol), demonstrating high purity 161Tb relevant for potential future preclinical applications.

Effect of gamma and neutron irradiations on the physical, magnetic and electrical properties of Ni–Cu–Cd nano-ferrites

The physical, magnetic and electrical properties variations for the sol-gel auto combustion synthesized Ni0.3Cu0.2Cd0.5Fe2O4 by gamma (γ) and neutron irradiations dose applications are studied in this research. The irradiation have been done through 2.0 × 1011 and 1.9 × 1012 cm−2s−1 per min fast neutron flux and γ rays from the research reactor of TRIGA MARK-II and 64 KCi cylindrical 60Co source respectively within room temperature. The rate of 1 Mrad and 3 Mrad dose is 1.067 Mrad/h. XRD results show greater intensities for the irradiation doses. Due to the redistribution of ions, changes in the lattice constant have been found by the irradiation through the results of XRD and Rietveld Refinement process. The saturated magnetization obtained from the VSM is found to decrease first by the irradiation and then increase drastically. This analysis is performed by the popular pulse field technique of hysteresis loop for nano ferrites and further analyzed by LAS. The frequency dependent characteristics such as dielectric constant, electric modulus, impedance analysis and ac conductivity are inspected for all samples with both types of irradiation process. The electrical modulus Cole-Cole plots insure the dominant property as electric stiffness for investigated ferrties. Extreme decrease has been found in the dielectric natures which also show good agreement in the higher ac conductivity nature at the higher frequencies. Electrical parametric dependent overlapping two semicircles which are not complete have been found in the plots of Cole-Cole for impedance data. The energy of activation for the process of diffusion decreases by the irradiation.

On the Use of a Chloride or Fluoride Salt Fuel System in Advanced Molten Salt Reactors, Part 3; Radiation Damage

Structural materials in fast reactors with harsh radiation environments due to high energy neutrons—compared to thermal reactors—potentially suffer from a higher degree of radiation damage. This radiation damage can change the thermophysical and mechanical properties of materials and, as a result, alter their performance and effective lifetime, in some cases leading to their disintegration. These phenomena can jeopardize the safety of fast reactors and thus need to be investigated. In this study, the effect of radiation damage on the vessels of molten salt fast reactors (MSFR) was evaluated based on two fundamental radiation damage parameters: displacement per atom (dpa) and primary knock-on atom (pka). Following the previous part of this article (Parts 1 and 2), an iMAGINE reactor core design (University of Liverpool, UK—chloride-based salt fuel system) and an EVOL reactor core design (CNRS, Grenoble, France, fluoride-based salt fuel system) with stainless steel and nickel-based alloy material vessels, respectively, were considered as case studies. The SPECTER and SPECTRA-PKA codes and a PTRAC card of MCNPX, integrated with a module which has been developed in MATLAB, named PTRIM and SRIM-2013 (using binary collision approximation), were employed individually to calculate and compare dpa and PKA (this master module containing all three tools has been appended to the iMAGINE-3BIC package for future use during reactor operations). Additionally, SRIM-2013 was applied in a 3D simulation of a radiation damage map on a small sample of vessels based on the calculated PKA. Our results showed a higher degree of radiation damage in the iMAGINE vessel compared to the EVOL one, which could be expected due to the harder neutron flux spectrum of the iMAGINE core compared to EVOL. In addition, the nickel alloy vessel showed better radiation damage resistance against high energy neutrons compared to the stainless steel one, although more investigations are required on thermal neutrons and alloy corrosion mechanisms to determine the best material for use in MSFR vessels.

Systematic derivation of [formula omitted] equations, its discretization using GTIN method and development of a switchable SP3 to [formula omitted] neutron transport solver

The Generalized Simplified Spherical Harmonics (GSPN) method has drawn recent interest owing to the fact that theoretically, it is equivalent to the PN and does not rely on any assumption other than piecewise homogeneity. Its level wise approach to PN offers an agreeable balance between the accuracy and computational efficiency making it a valuable mathematical tool for analyzing neutron transport problems pertaining to nuclear engineering and reactor physics. The lowest level approximation for N = 3, expressed as GSP3(0), offers reducing the complexity while still capturing the essential physics. In the current work, GSP3(0) equations are derived along with the interface and boundary conditions for linear anisotropic scattering. These equations are then discretized using the generalized transverse integration nodal method for a two-dimensional system of piecewise homogeneous rectangular regions. The discretization is done with an option kept for reducing the GSP3(0) formalism to the conventional SP3. Subsequently, a computer code has been developed to solve these discretized equations in the multigroup structure for vacuum, reflective or albedo boundaries. At present, this switchable SP3/GSP3(0) code has the capability of estimating k-eigenvalue and position dependent scalar neutron flux within a given two-dimensional rectangular geometry. The code has been verified by solving SP3, GSP3(0) and other benchmark problems from available literature. The results obtained from our code demonstrate that GSP3(0) is superior to the conventional SP3 and comparable to the recently proposed SDPN (N = 1,2,3) [M. Nazari, A. Zolfaghari, M. Abbasi, Prog. Nucl. Energy, 2023; 166, 104,933] in terms of accuracy as well as computation time.

Polymer Boron-Containing Composite for Protecting Astronauts of Manned Orbital Stations from Secondary Neutron Radiation

This article considers the prospects of using heat-resistant polyimide boron-containing composites to protect astronauts of manned orbital stations from secondary neutron radiation. Variant calculations are performed regarding neutron and gamma-quanta flux distributions in a polyimide composite material with different boron content used to reduce capture radiation. The dependences of spatial distributions of thermal neutron flux density and the gamma-quanta dose rate in a polyimide composite layer with a boron content of 0 to 5% are obtained. An experimental assessment of the energy distribution of neutron and gamma radiation behind the protective polyimide composite is carried out. The introduction of boron atoms in an amount of 3.0 wt.% shows the absence of bursts of secondary gamma radiation energy in the composite, which is due to the high cross-section of thermal neutron absorption by boron atoms. As a result, with a material layer thickness of 3–10 cm, the gamma-quanta dose rate decreases by 2–3 times. The differential thermal analysis method showed that the upper limit of the working temperature of the polyimide composite is 500 °C. The polyimide matrix filled with boron atoms can find effective application in the development of new radiation-protective polymer materials used in manned orbital stations.

Bonner sphere measurements of high-energy neutron spectra from a 1 GeV/u 56Fe ion beam on an aluminum target and comparison to spectra obtained by Monte Carlo simulations

The goal of this work is to characterize the secondary neutron spectra produced by 1 GeV/u56Fe beam colliding with a thick cylindric aluminum target and to perform a quantitative comparison with simulated results obtained with Monte Carlo codes. The measurements were performed using extended-range Bonner sphere spectrometers at two positions (15° and 40°) with respect to the beam direction. The secondary radiation field was simulated using four Monte Carlo codes (FLUKA, MCNP6, Geant4 and PHITS) and several physical models of nuclei transport and interaction. Neutron and proton energy distributions were simulated for the experimental measurement positions. The simulated neutron spectra, together with data measured with Bonner sphere spectrometers, after carrying out the correction of the contributions induced by the secondary protons, were used as input for the MAXED spectrum unfolding code to obtain the measured neutron spectra. Unfolded neutron spectra were compared with simulated ones to carry out a quantitative analysis of the performance of the chosen Monte Carlo codes and their corresponding physical models. This comparison showed that, because of experimental uncertainties and physical models, there are no unique solutions for each measurement location, but a range of solutions where the true experimental neutron spectra probably lie. The results showed deviations between 4.23% and 8.42% for some simulated spectra. Regarding the total integral values of neutron fluence and ambient equivalent dose, the unfolded neutron spectra showed deviations lower than 2%.

Dopant, coating, and grating effects in silica optical fibers under extreme neutron irradiation

Radiation effects on fiber Bragg gratings (FBGs) have been studied but data gaps related to the effects of displacement damage resulting from high fast neutron fluence remain. In this work, Type-I and Type-II FBGs inscribed in optical fibers with various core dopants (Ge and F) and fiber coatings (acrylate, polyimide) were monitored in situ during 75 days of neutron irradiation to a peak fast (>0.1 MeV) neutron fluence of 3×1021 nfast/cm2. The reflected intensity of the Type-I FBGs inscribed in a Ge-doped core fiber decreased by >30 dB within 7 h of irradiation (1019 nfast/cm2), whereas Type-II FBGs inscribed in a pure silica core fiber eventually approached >40 dB attenuation after accumulating a fast neutron fluence on the order of 1020 nfast/cm2. Type-II FBGs inscribed in F-doped core fiber improved stability: the attenuation approached an equilibrium value in the range of 10 to 20 dB.

Enhancing fuel breed-burn performance in a sodium-cooled fast reactor using a novel reactivity control method

This study aims to enhance the fuel cycle and fissile breeding performance of a sodium-cooled fast breeder reactor (FBR) by utilizing minor actinides (MAs) as a means of reactivity control alongside partially-inserted control rods. Choosing the PFBR-500 as the reference design, four core models, designated as Cases A, B, C, and D, utilizing various proportions of minor actinides (MAs) were built and simulated using OpenMC. The MA concentrations were optimized to compensate for the withdrawal of control rods and ensure the same initial reactivity for all cores. Burnup analysis over 365 EFPDs revealed a significant increase in cycle length and burnup for the modified cores along with a modest rise in breeding ratio. Notably, Case C, employing 3.45 wt.% MAs in the 88 inner-core fuel subassemblies achieved an extra 62.25 EFPDs cycle length, a 33.74% rise in single-cycle burnup, and a 3.86% increase in breeding gain compared to the reference. Loading MAs into the inner-core region proved to be more effective in enhancing both fertile-to-fissile conversion (thus extending the cycle length and fuel burnup) and transmutation than utilizing MAs throughout the core due to greater neutron flux at the core center. While Case D utilizing 2.2 wt.% MAs both in the inner and outer core fuel subassemblies had the highest overall MA loading, it demonstrated lower increments in cycle length, burnup, and breeding gain compared to Case C. Case C also exhibited the highest overall destruction rate (approximately 24%/y) for Np and Am isotopes, and successfully transmuted around 24.08 kg 237Np, 9.33 kg 241Am and 3.82 kg 243Am over the course of a single year. The addition of MAs also achieved a slight flattening of the axial and radial flux profile and a decrease in the flux peaking factor. However, it slightly lowered the beta-effective, Doppler constant, and control rod assembly worth, and shifted the coolant void reactivity worth to the positive side.