Reactor Neutron Research Articles

Burnup chain compression method preserving neutronics, decay photon source term and decay heat calculation results

A burnup chain compression method, preserving neutronics, decay photon source term and decay heat calculation results, is developed based on the generalized perturbation theory (GPT), to reduce calculation burden for the practical reactor physics simulations. In this method, a certain number of target nuclides which have significant effect on calculation accuracy of the three applications are first selected according to their contribution to the neutron reaction rate, decay heat and decay photon intensity. Then, nuclides important for the accuracy of the nuclide number densities of the target nuclides are selected according to the contribution function calculated based on the GPT. A fine burnup chain containing 1547 nuclides based on ENDF/B-VII.0 is compressed to a burnup chain including 557 nuclides. The compressed burnup chain is adopted to calculate a couple of typical fuel assembly and pin cell problems of PWR. The results are compared with those based on the fine burnup chain. The numerical results show that the compressed burnup chain achieves high accuracy for neutronics, decay photon source term and decay heat calculation, simultaneously.

First principles calculation for generating new thermal neutron scattering data for Zirconium hydride

Lattice parameters of materials have the same magnitude as the energy of thermal neutrons in reactors, which directly affects the neutron cross section and its energy. While they are thermalized, incident neutrons can lose or gain energy during their interactions with materials components. Since several decades, methods and models were developed in the aim to generate nuclear data sub-libraries required in correcting neutrons interactions cross sections at thermal energies. However, very few experimental works were dedicated to this field. In this paper we focus our efforts on reviewing the theoretical models and their adequacy in describing thermal scattering events in the aim of proposing new formalisms to calculate the density of states (DOS) and phonon responses of zirconium hydride material, which constitutes an important moderator of neutrons in TRIGA reactors fuel elements. Generally the effects of thermal scattering are provided in nuclear data evaluations by a thermal sub-library ENDF file 7. Data in file 7 are described by the known thermal scattering law S(α,β) which is a function of momentum transfer and energy transfer parameters α and β respectively. The thermal scattering law has been used to calculate the double differential cross sections and the corresponding results are presented. Although the comparison with other models shows satisfactory results, no previously personalized use of data may be the raison of its usefulness in some cases and not in others.

Studies and cross-comparisons of severe accident prevention and mitigation capabilities of a SFR and a GFR

CEA has developed Generation IV fast neutron reactors with gas and sodium coolants in the two last tenth of years. Namely these reactor projects were the GFR2400 (gas-cooled fast reactor of 2400 thermal megawatts) and ASTRID (sodium-cooled fast reactor demonstrator of 1500 thermal megawatts). The objective of this paper is to provide a cross-comparison of the severe accident prevention and mitigation capability of these reactor concepts based on the work done during a significant time period and taking benefit of the distance since these studies. This comparison can highlight both generic trends resulting from a common study approach used for both concepts and from very detailed results obtained during their conceptual design studies. Despite their power difference and their fuel element design specificities, the main study results and conclusions are quite generical of the two concepts as explained in the paper. Thus they could be extrapolated to other reactor design providing main reactor features are kept (core materials, coolant, neutron spectrum, etc.). The assessment of core melting prevention relies on the study of the natural behavior of the GFR2400 and of ASTRID when facing the various accident sequence families. Then, the efficiency and the features of the systems to be foreseen for core degradation prevention are presented. As far as mitigation is concerned, all the consequences of core melting are investigated (i.e. the induced loadings in terms of nature and of range) by considering various core degraded states. Based on the magnitude of these loadings, the needs of mitigation means are assessed for each concept. Among other trends, the presented work shows the very good prevention capability of the SFR concept but the necessity to mitigate the fast vaporization and expansion of degraded core materials. Conversely, the limited coolant capability of the GFR concept and its low thermal inertia require a pressurized gas circulation into the core, limiting its prevention capability whereas its core melting should not induce substantial mechanical loadings of the reactor structures. However, for this last concept, thermochemical interactions between the core materials are an issue deeply investigated in order to understand and simulate core degradation.

Permeation behavior of gaseous tritium through the assembly composed of Zr and Al2O3 simulating Li rod

• An assembly simulating the Li rod structure for T production in HTGR was used. • The assembly was composed of inner Zr tube - outer Zr tube - Al 2 O 3 tube. • T permeation rate through the assembly from HT was much faster than that from HTO. • The rate-determining step was T permeation through the Al 2 O 3 tube. For starting up a fusion reactor stably, the tritium production by the nuclear reaction of neutrons and lithium in a high-temperature gas-cooled reactor has been studied. In order to suppress the migration of tritium to the whole reactor, it is important to design a Li rod with tritium confinement function and understand tritium behavior. A Li rod structure consisting of Zr - LiAlO 2 - Zr - Al 2 O 3 was proposed, and an assembly simulating proposed Li rod was fabricated. In the previous work using tritium gas containing mainly HTO rather than HT, the present authors succeeded in confining tritium in the assembly at 700 ℃ for 87 h. In this work, tritium gas containing mainly HT rather than HTO was supplied to the assembly, tritium permeation behavior was observed. Unlike the case of HTO, tritium permeated immediately after the introduction of HT. This result indicates that the chemical form of tritium significantly affects tritium confinement performance of the Li rod.

Evaluation of Fuel Reactivity Worth Measurement in the Prototype Fast Reactor Monju

In the prototype fast breeder reactor Monju, fuel reactivity worth was measured at six positions as the reactivity corresponding to the differences of critical control rod positions between cores with and without a dummy fuel subassembly. In this paper, the measurements are evaluated in detail, and their reliability and usefulness as validation data for fast reactor neutronics design methodologies are examined through a comparison with calculations. Calculated-to-experiment values and their uncertainty of fuel reactivity worth were 0.97 to 1.02 and 4% to 6%. Through this study, the consistency and reliability of the measurements and calculations were confirmed.

Design of ITER neutron flux monitor system back-end electronics

The primary aim of the International Thermonuclear Experimental Reactor (ITER) Neutron Flux Monitor (NFM) system is to diagnose the total neutron emission rate and fusion power of the plasma fusion reaction. The neutron flux is measured directly by the system and the fusion power can be calculated according to the neutron flux result. The NFM system is composed of detectors and back-end electronics. This paper focuses on the work of NFM#7 back-end electronics design and development. The back-end electronics consist mainly of four parts: Firstly, the pre-amplifiers are used to charge integrate and amplify the signal of fission chambers. Secondly, the output of pre-amplifiers will be transmitted to Signal Processing Unit (SPU) for analog signal conditioning, data processing, and measurement result uploading. Thirdly, pre-amplifiers and fission chambers have different electricity sources. Pre-amplifiers receive electricity generated from low voltage power whilst fission chambers receive electricity generated from high voltage power. Lastly, the Instrument and Control (I&C) is designed as the interface between NFM hardware and ITER central system and its functions are configuration & control, monitoring, calibration, data communication & archiving and clock synchronization. All the four parts are well integrated and tested in neutron field. The system design also follows the guidelines and standards of ITER.

A method for quickly subtracting the contributions from 182Ta and 94Nb in activity measurement of 93mNb in reactor neutron environment

Niobium is one of the most important neutron dosimeters used outside of the reactor. Owing to the fluorescence effects of 182Ta and 94Nb radionuclides, it is a problem to quickly obtain the activity of 93mNb which produced by the neutron inelastic scattering of Nb. To solve this problem, a fast estimating method based on the Monte Carlo simulation is proposed. All the contributions from fluorescence effects that induced by gamma-rays, beta-rays and internal conversion electrons of the radionuclides in the niobium sample were considered. The comparison results demonstrated that the deviation is less than 8.0% between the calculated and the experimental activities. It shows that the contribution of radionuclide by fluorescence effects in activity measurement of the niobium sample can be subtracted quickly using this method, based just once experimental measurement.

Improved Mo95 neutron resonance parameters and astrophysical reaction rates

Background: Improved $^{95}\mathrm{Mo}$ neutron resonance parameters and reaction rates are important for nuclear astrophysics, testing nuclear models, and nuclear criticality safety. However, despite many previous neutron-capture and total cross-section measurements on this nuclide, there still is much room for improvement as well as several discrepancies. For example, there are very few firm resonance spin and parity assignments; average resonance parameters are available only for each parity, the currently recommended astrophysical reaction rate results in disagreements between stellar models and meteoric isotopic anomalies, and there are substantial disagreements in the neutron-capture cross section at low energies important for nuclear criticality safety.Purpose: To obtain an improved set of neutron resonance parameters and astrophysical reaction rates for $^{95}\mathrm{Mo}$.Method: High-resolution neutron-capture and transmission data were measured at the Oak Ridge Electron Linear Accelerator (ORELA) using highly isotopically enriched $^{95}\mathrm{Mo}$ samples. The neutron-capture apparatus, data reduction, and analysis were improved so that information contained in the $\ensuremath{\gamma}$-ray cascade following neutron capture were used to assign resonance ${J}^{\ensuremath{\pi}}$ values. Following this, simultaneous analysis of the new neutron-capture and transmission data was used to obtain resonance energies, gamma widths, and neutron widths and their uncertainties to a maximum energy of 10 keV. Accurate neutron-capture cross sections also were obtained for the unresolved resonance region to a maximum energy of 500 keV and, together with the new resonance parameters, used to calculate the astrophysical reaction rates in the temperature range from 5 to 30 keV.Results: A vastly improved set of $^{95}\mathrm{Mo}$ neutron resonance parameters and an astrophysical reaction rate accurate to about 3% were obtained. Firm ${J}^{\ensuremath{\pi}}$ assignments were determined for 261 of the 314 observed resonances. This is a very large improvement over the previously published 32 firm ${J}^{\ensuremath{\pi}}$ assignments for 108 resonances. Also, the number of resonances having both firm ${J}^{\ensuremath{\pi}}$ assignments and ${\mathrm{\ensuremath{\Gamma}}}_{\ensuremath{\gamma}}$ values was increased by almost a factor of 24---from 11 to 261. Neutron- and total-radiation-width distributions and average resonance spacings, average total radiation widths, and neutron strength functions were obtained for the six different $s$- and $p$-wave possibilities. Parameters for the lowest $s$-wave resonance, which is most important for criticality benchmarks, were obtained with high accuracy.Conclusions: Simple modification of the neutron-capture apparatus and expansion and improvement of data analysis techniques led to a large increase in firm ${J}^{\ensuremath{\pi}}$ assignments for $^{95}\mathrm{Mo}$ neutron resonances. The resulting astrophysical reaction rate is 20%--30% larger than the currently recommended rate at $s$-process temperatures, which should lead to better agreement between stellar models and meteoric isotopic anomalies. The neutron-capture cross section at low energies is substantially larger than recommended in the latest evaluation, which is problematical for criticality benchmarks. The average resonance spacing as a function of spin and parity is significantly different from current models. The total-radiation-width distributions are significantly broader than predicted by theory and show significant departures from the expected Gaussian shapes.

Overview and current perspectives for the use of the GFR reactor in research and electricity generation in Brazil / Panorama e perspectivas atuais para o uso do reator GFR em pesquisa e geração de energia elétrica no Brasil

The Gas Cooled Fast Reactor (GFR) is one of the ambitious projects of Generation IV nuclear reactors, which aims to develop a sustainable, safe, reliable, economical and proliferation-resistant nuclear power generation system. This project is currently in a theoretical phase, being reported in this work a history from its idealization to the present moment. Presently, the project that is in a more advanced development stage, ALLEGRO, is in the final design phase and with the licensing process in progress. In the Brazilian perspective, it is known that an increase in investment is necessary to develop the nuclear area and a greater engagement in research and development groups of fast neutron reactors, aspects which are compatible with a GFR generation IV project.

Port-Hamiltonian Control of Nuclear Reactors

The port-Hamiltonian form (PHF) of nuclear reactor dynamics is first given with the corresponding Hamiltonian function determined by the shifted-ectropies of neutron kinetics and reactor thermal-hydraulics. Then, a passivity-based control (PBC) is proposed by adding extra damping injections to the PHF, and it is shown by theoretical analysis that satisfactory closed-loop stability can be guaranteed. The newly designed control law is applied to a nuclear heating reactor, and the simulation results not only verify the theoretical analysis but also show the satisfactory reactor power-level control performance.

Test of low-dropout voltage regulators with neutrons and protons

The Muon System of the ATLAS experiment at the CERN LHC will be upgraded for the high-luminosity phase of LHC to cope with higher rates and higher radiation levels. Most of the Muon-System on-detector electronics will be replaced. Commercial low-dropout voltage regulators have been considered as a robust, low-noise and economic solution to power distribution. These components should be selected based on their capability to comply to radiation requirements. We tested 7 different types of CMOS LDOs, monitoring online the output voltage of 10 samples of each type. Irradiations were performed in the Radial Channel 1 of the RSV TAPIRO fast neutron reactor at ENEA Casaccia (Rome), to test resistance to non-ionizing energy loss, and at the PIF 200 MeV proton beam at PSI (Villigen), to test for total ionizing dose and single event effects. The experimental setup and the results are presented and discussed in this paper.

Development of Radiation Porosity in the Material of Fuel-Element Cladding Produced by Different Technologies for Fast Neutron Reactors and Its Influence on the Fuel Assembly Service Life in the BN-600 Reactor

Development of Radiation Porosity in the Material of Fuel-Element Cladding Produced by Different Technologies for Fast Neutron Reactors and Its Influence on the Fuel Assembly Service Life in the BN-600 Reactor

Determination of the deformation deceleration coefficient for calculating the initial effective length of the working part of an annular specimen made of fuel cladding

When studying the plastic properties of metallic materials under tension, the initial calculated length of the sample which affects the final results is of significant importance. In the case of determining the short-term mechanical properties of the materials of fuel cladding pipes using annular samples, calculation of the initial effective length of their working part is rather difficult since the plastic deformation of the annular sample is unevenly distributed along the perimeter. This feature is taken into account using the deformation deceleration coefficient. We present the results of determining the deformation deceleration coefficient for calculation of the initial effective length of the working part of an annular sample made of fuel cladding. An experimental method for determining the coefficient is proposed, taking into account the measurements of the distribution of plastic deformation in certain areas along the perimeter of the sample. The method consists in stretching annular and ovalized samples with marks on the end surface. It is shown that for annular and ovalized specimens from the fuel cladding of a BN-600 reactor, the values of the deformation coefficient and effective length are k = 0.50 ± 0.04 and l 0 = 6.1 ± 0.4 mm, respectively. Analysis of the actual relative deformation of the working parts and comparison with the relative elongation showed that when the ovalized specimens are stretched, the value of the relative plastic deformation is higher than that for annular specimens. The proposed method and the results obtained can be used in determination of the parameters of the materials used in the cladding pipes of the of fast neutron reactors.

The design and optimization of a radiation shield for the Tehran Research Reactor's neutron diffraction facility

In this paper, a novel optimal radiation shield for the Tehran research reactor's neutron powder diffraction facility was designed to reduce undesirable neutron and gamma dose rates. A layer-by-layer methodology was considered based on the design approach to achieve a dose rate of less than 30 μSv/hr. MCNPX was used for this purpose as a Monte Carlo radiation transport code. A combination of neutron and gamma absorbers and moderators was employed to create this shield. The results indicated that the shield's six distinct layers met the design objective, totaling 65 cm in thickness. This newly designed shield comprises 15 cm paraffin, 15 cm Portland iron concrete, another 15 cm paraffin, 5 cm solid boric acid, another 10 cm Portland iron concrete, and 10 cm lead. The neutron and photon dose rates are 9.35 and 12.81 μSv/hr, respectively, which is one-sixteenth and one-third of the current facility shield. Validation of the simulation was accomplished by comparing the results for the current shield to experimental data.

AR reactor neutronics multi-code validation* *This manuscript has been authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 with the US Department of Energy (DOE). The US government retains and the publisher, by accepting the article for publication, acknowledges that the US government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow

The affordable, robust, compact (ARC) reactor is a tokamak fusion reactor concept currently under development by Commonwealth Fusion Systems and Massachusetts Institute of Technology. There are three important neutronics considerations for the operation of the ARC reactor: (1) breeding of enough tritium in the blanket to sustain the D–T reaction in the plasma; (2) ensuring low fluence on the superconducting toroidal field coils; and (3) assessing neutron volumetric heating in structural components. This work aims to perform a validation of the neutronics analysis approach by code-to-code comparison. State-of-the-art software stacks are employed for the neutronics analysis of the ARC reactor, and a computer-aided design (CAD) model is used directly for Monte Carlo (MC) neutron transport calculations. Three software stacks, Attila-MCNP, OpenMC-DAGMC, and Shift-DAGMC, are used to perform neutronic analyses of a 90° sector CAD model of the ARC reactor. Results show that the flux tallies calculated by the three software stacks are very close. Also, the volumetric heating and tritium breeding values have less than 0.6% relative difference between codes.

ОПРЕДЕЛЕНИЕ СОДЕРЖАНИЯ ХИМИЧЕСКИХ ЭЛЕМЕНТОВ В ЗДОРОВОЙ ТКАНИ МОЛОЧНОЙ ЖЕЛЕЗЫ ЧЕЛОВЕКА МЕТОДОМ НЕЙТРОННО-АКТИВАЦИОННОГО АНАЛИЗА

Purpose: To develop the method of chemical element (ChE) content measurement in samples of breast tissue (BT) using the activation by neutrons of nuclear reactor combined with the high-resolution spectrometry of gamma-radiation of short-lived radionuclides (INAA-SLR), and to investigate ChE contents in normal human mammary gland. Material and methods: In the samples of BT taken from female with intact mammary glands (mostly died from trauma, n = 38) the contents of calcium (Ca), chlorine (Cl), iodine (I), potassium (K), magnesium (Mg), manganese (Mn), sodium (Na), and strontium (Sr) were measured. To determine these element contents, the method of INAA-SLR was developed. Results: The accuracy of the developed method and the reliability of the results obtained in the study were confirmed by the measurements of international certified reference material IAEA H-4 Animal Muscles and the good agreement with data of its certificate. The main statistic parameters, including, arithmetic mean, standard deviation, standard error of mean, minimum and maximum values, median, percentiles with 0.025 and 0.975 levels was calculated for ChE contents in the normal BT. It was shown, for example, that means of mass fractions ± standard errors of means (M±SEM, mg/kg dry tissue) in normal human mammary gland were: Ca 128±14, Cl 1014±146, I 0,82±0,11, K 196±20, Mg 80,0±9,4, Mn 0,27±0,04, Na 665±71 и Sr 5,20±0,75. It was found that the Ca, Cl, I, K, Mg, Mn, Na и Sr contents in human mammary gland very differ from the normal level of these elements in whole blood, muscle and adipose tissue, particularly in contents of I and Sr. Conclusion: The method of INAA-SLR allow to obtain the representative data on Ca, Cl, I, K, Mg, Mn, Na и Sr contents in human BT specimen. It was shown, that such trace elements as I and Sr may be directly involved in the mammary gland function.

SNF processing electrochemical operations: liquid-metal and salt medium purification

The paper investigates the process of regeneration of a liquid metal medium used in the pyroelectrochemical reprocessing of spent mixed uranium-plutonium nitride fuel produced by a fast neutron reactor. The investigation concerns the interaction of liquid cadmium with sludge formed during the anodic dissolution of ceramic nitride pellets in a 3LiCl-2KCl melt medium as well as the possibility of its purification by filtration from individual metal fission products. Anode sludge is represented by fission products of the platinum group, zirconium, molybdenum and technetium. It was determined by scanning electron microscopy that the metal product is composed of several intergrowth phases. It was found that upon contact of a polymetallic alloy simulating anode sludge with a melt, the liquid metal phase is saturated to 0.025 wt% of Pd, 0.01 wt% of Rh for 50 hours at 500 °C, while zirconium forms an insoluble dispersed intermetallic compound ZrCd3. Powders of molybdenum and technetium, which are not wetted with cadmium, can be completely removed using a filter mesh of plain weaving of the P-200 type. It is also possible to remove zirconium from anodic cadmium by filtration. The filtration efficiency of ruthenium and palladium powders did not exceed 54.3 and 13.1 wt%, respectively, due to partial dissolution and thinning of particles, which will lead to saturation of the liquid metal phase and the need to purify it by alternative methods.

Comparison of the minor actinide transmutation efficiency in models of a fast neutron uranium-thorium fueled reactor

In terms of nuclear raw materials, the issue of involving thorium in the fuel cycle is hardly very relevant. However, in view of the large-scale nuclear power development, the use of thorium seems to be quite natural and reasonable. The substitution of traditional uranium-plutonium fuel for uranium-thorium fuel in fast neutron reactors will significantly reduce the production of minor actinides, which will make it attractive for the transmutation of long-lived radioactive isotopes of americium, curium and neptunium that have already been and are still being accumulated. Due to the absence of uranium-233 in nature, the use of thorium in the nuclear power industry requires a closed fuel cycle. At the initial stage of developing the uranium-thorium cycle, it is proposed to use uranium-235 instead of uranium-233 as nuclear fuel. Studies have been carried out on the transmutation of minor actinides in a fast neutron reactor in which the uranium-thorium cycle is implemented. Several options for the structure of the core of such a reactor have been considered. It has been shown that heterogeneous placement of americium leads to higher rates of its transmutation than homogeneous placement does.

Radiation hardness study using SiPMs with single-cell readout

A dedicated single-cell SiPM structure is designed and measured to investigate the radiation damage effects on the gain and turn-off voltage of SiPMs exposed to a reactor neutron fluence up to Φ = 5e13 cm−2. The cell has a pitch of 15μm. The fluence dependence of gain and turn-off voltage are reported. A reduction of the gain by 19% and an increase of Voff by ≈0.5 V is observed after Φ = 5e13 cm−2.

Utilization of accelerator and reactor based nuclear analytical techniques for chemical characterization of automobile windshield glass samples and potential of statistical analyses using trace elements towards glass forensics

Glass forensics is an important area in forensic crime investigations, wherein glass origin or source finding is necessary mainly through chemical composition. In the present work, Nuclear Analytical Techniques namely external (in air) Particle Induced Gamma-ray Emission (PIGE) and Instrumental Neutron Activation Analysis (INAA) were utilized for complete chemical characterization of twenty-five “as received” windshield glass samples of six car manufactures. Concentrations of four major elements (Si, Na, Mg and Al) by PIGE using proton beam and nineteen elements including sixteen trace elements by INAA using research reactor neutrons were determined. Both the methods were validated by analysing matrix matched glass certified (standard) reference materials. Trace elemental concentrations including rare earth elements (REEs) and ternary plot using concentrations of major, transition elements and REEs were utilized to obtain preliminary grouping of the analyzed glass samples. Statistical tools namely K-mean, Cluster Analysis and Principal Component Analysis (PCA) using trace elemental concentrations were utilized for grouping studies, important for forensic applications. Among these statistical analysis techniques, PCA results confirmed that windshield glasses from six manufactures clearly belong to six different groups.