Background Validation of neutron transport calculations for nuclear criticality safety requires selecting relevant benchmarks and consolidating multiple uncertainty sources to estimate application bias and uncertainty. Existing practices rely on sensitivity/uncertainty (S/U) techniques and heuristic similarity thresholds, but they lack a simple, uncertainty-aware metric for comparing validation outcomes across consolidation methods and benchmark sets. Methods We introduce a concept of score that quantifies validation performance as the posterior residual normalized by its associated uncertainty. The score is evaluated for four consolidation approaches—parametric, nonparametric, Whisper, and generalized least-squares methodology (GLLSM)—using S/U-based indices. The analysis considers keff predictions for representative sets of International Handbook of Evaluated Criticality Safety Benchmark Experiments benchmarks and corresponding applications. Results Score distributions reveal distinct behaviors across methods. Nonparametric and Whisper generally yield positively shifted means, reflecting conservative bias estimation by design. Parametric and GLLSM produce means closer to zero but can exhibit wider score spreads when selection is based solely on conventional similarity thresholds. Incorporating the scaled similarity index reduces spread and mean shift for the parametric method by accounting for sensitivity magnitude. For GLLSM, acknowledging additional prior sources of uncertainty not captured by nuclear data alone brings the empirical score spread closer to unity across relevance levels, indicating improved statistical consistency. Conclusions The score provides uncertainty-aware lens for comparing consolidation methods and guiding benchmark selection. Results support using scaled similarity with the parametric method and accounting for unmodeled prior uncertainty within GLLSM. The framework helps analysts select benchmarks and consolidation strategies that yield statistically consistent posterior keff predictions while maintaining defensible safety margins.
Nuclear engineering. Atomic power, Medical physics. Medical radiology. Nuclear medicine
BackgroundThe very small-angle neutron scattering (VSANS) spectrometer is currently being constructed at the China spallation neutron source (CSNS). The neutron detector plays a key role in meeting the detection requirements of the spectrometer.PurposeThis study aims to achieve accurate measurement of neutron diffraction in the small angle scattering mode which requires the position resolution of the neutron detector to be ≤2 mm, and the detection efficiency to be ≥60%@0.4 nm.MethodsA large-area position-sensitive neutron scintillator detector mainly composed of 6LiF/ZnS(Ag) scintillation screen, wavelength shift fiber (WLSF), and a silicon photomultiplier (SiPM) was designed and developed to achieve high efficiency and high-resolution real-time detection of thermal neutrons. The optical transmission performance of the 0.5-mm-diameter wavelength shift fiber was investigated in detail. The gain and the thermal noise characteristics of various SiPMs were examined. Finally, a 300 mm×300 mm engineering prototype of the detector was developed and tested by experiments. The detection efficiency was calculated with comparison of the incident neutron counts of the stander 3He tube whilst the detector's position resolution was tested using a boron-doped aluminum plate with a narrow slit bearing the label 'CSNS'.ResultsThe neutron beam test results show that the position resolution of the detector is 1.2 mm×1.2 mm and that the neutron detection efficiency is (61.8±0.2)%@0.4 nm.ConclusionsThe high-resolution neutron scintillator detector developed in this study satisfies the engineering design target, and meets the neutron scintillator detector meets the neutron diffraction measurement needs for the VSANS spectrometer of the CSNS.
High quality and large size bulk single crystals are essential to study the basic properties of perovskite materials and their performance limits of X-ray detection and imaging. In this paper, high quality MAPbI3 and MAPbCl3 single crystals were prepared using the silicone-oil assisted method and the properties and the performance of the single crystals and the corresponding X-ray detectors were characterized. These performances of single crystals grown with the silicone-oil assisted method are better than those prepared with the inverse temperature crystallization method(ITC). The ITC, as one of the most prevailing crystal growth methods, has drawbacks over the control of the nucleation and the growth rate. Silicone-oil assisted method was recently developed to overcome the above drawbacks and grew the perovskite single crystals in a controllable manner. In the silicone-oil assisted method, a capping layer on the precursor is adopted to reduce the solvent evaporation rate to a certain level and thus to manage the precursor in a slightly supersaturated condition. Silicone oil is selected as the capping layer for its smaller density and the insolubility with the precursor. The solvent in precursor will diffuse through the silicone oil capping layer and evaporate into the air at the silicone oil/air interface, driven by the content difference and the saturation vapor pressure, respectively. The silicone-oil assisted method realizes the constant growth of perovskite single crystals, since the driving force of the crystallization relies on the solvent diffusion rate but not the temperature-depended solubility. During the perovskite crystal growth process, the precursor is at a steady condition(constant temperature and constant concentration) in majority of the time. According to the classic growth theory, single crystals grown at a constant rate will have better crystal quality. Indeed, MAPbI3 single crystals grown with the silicone-oil assisted method have a FWHM of rocking curves of (1 1 0) crystal planes of 0.011 2°, the fluorescence lifetime of 1 022 ns, and the lowest defect density of 5.61×109 cm-3. MAPbCl3 single crystals grown with the silicone-oil assisted method have a FWHM of rocking curves of (1 0 0) crystal planes of 0.014 3°, the fluorescence lifetime of 957 ns, and the lowest defect density of 3.57×109 cm-3. The X-ray detector based on MAPbI3 single crystals has a high sensitivity of 7 300 μC/(Gy·cm2). All results indicate that this silicone-oil assisted growth method is suitable for the mass production of high-quality single crystals for radiation detection.
Nuclear engineering. Atomic power, Chemical technology
In the present study, we investigated the velocity distribution, temperature distribution and condensation characteristics of steam jet issuing from four different orifice nozzles with a Reynolds number of approximately 79,000 using the phase Doppler particle analyzer system and a K-type thermocouple. The steam jet discharged from the orifice nozzle has a wider jet width compared to pipe nozzle because of the vena-contracta which can enhance the mixing of steam jet with the ambient air. Therefore, the orifice jet showed less condensation due to its wideness, resulting in small velocity decay rate and large temperature decay rate due to momentum conservation and decreased latent heat release compared to pipe nozzle, respectively. Also, the wider jet width of the orifice jet resulted in larger velocity and temperature spread rate compared to the pipe jet. In addition, the increase in the aspect ratio of the orifice jet led to more condensation and larger velocity spread rate and temperature spread rate due to both the vena-contracta and axis-switching effect, resulting in the increase of jet entrainment.
BackgroundIn a tokamak, when fast electrons are deconfined by the tokamak magnetic field constraint and lost to the vacuum wall or limiter, the device may become damaged and the discharge may be affected.PurposeThis study aims to explore the loss behavior of fast electrons during discharge using a diagnostic system based on a ZnS(Ag) scintillator probe for detecting the loss of fast electrons on the Experimental Advanced Superconducting Tokamak (EAST).MethodsThe Geant4 simulation program was employed to simulate the interaction between electrons in different initial states and the scintillator probe of the diagnostic system. Firstly, the probe model and the filling material model of stainless steel and ZnS(Ag) coating were established in Geant4. Then, the interaction between electron beam and scintillator probe under different incident conditions (incident energy, angle, scintillator thickness, magnetic field size, etc.) were simulated, and the physical processes were recorded. Finally, the recorded data were accessed by MATLAB programming for analysis.ResultsThe results show that the contribution of secondary electrons and initial electrons to the luminescence intensity of scintillators occupies different dominant energy ranges. The luminescence intensity first increases and then decreases with the increase of incident electron energy, with a peak value around 12 MeV, and the number of emitted photons at oblique incidence is greater than that at vertical incidence. When the electron energy is lower than 4.3 MeV, secondary particles dominate the scintillation, and when the electron energy is higher than 4.3 MeV, primary particles dominate. The thickness of the scintillator has no significant effect on the peak position. After, the luminous intensity is considerably affected by the magnetic field angle and electron pitch angle after adding a magnetic field.ConclusionsThe results of this study contribute to the understanding of the fast electron loss signal detected by the scintillator probe in the EAST experiments, providing a basis for further study of fast electron loss.
Fast reactors with heavy liquid metal coolant (lead or eutectic bismuth-lead alloy) are one of the most advanced technologies capable to address the accumulated world nuclear energy issues. This innovative power technology is being developed in Russia, the USA, China and the European Union. Russia is the leader since it has focused on this topic for a number of decades. First concrete started to be poured in June 2021 to form the foundation of the Russian BREST-OD-300 lead cooled reactor scheduled to be started up in 2026. Attention is also given to the development status of the Chinese CLEAR reactor series. A large scope of R&D has been undertaken, and large-scale nonnuclear experimental facilities are under construction. International Euro-US consortiums for the development of the ALFRED, PLFR and MYRRHA reactors do not expect any unsolvable technical issues either and are currently formulating requirements to the test facilities and candidate materials and technologies required for further activities.
BackgroundMany existing studies have shown that the use of suitable surface modification methods can enhance the boiling heat transfer effect of metal components, making it have a broad potential application prospect in the pressurized water reactor. However, for the weak alkaline environment of high temperature and high pressure in the reactor, little literature is reported on whether this enhanced effect can be maintained for a long time.PurposeThis study aims to explore the effect of corrosion on boiling heat transfer characteristics of metal specimens with micro-structure surface.MethodsFirst of all, three micro-structures of micro-groove, micro-porous and micro-columns were processed on the surface of stainless steel plate specimens by laser processing. Then the specimens were placed in the high-temperature and high-pressure environment simulating the actual reactor conditions to carry out corrosion experiments for up to 200 d. Finally, the pool boiling experiment and visualization study of the specimens before and after corrosion were carried out for comparison.ResultsThe results show that the surface critical heat flux (CHF) of the three micro-structured metal specimens increases and then decreases with the increase of corrosion time, among which the micro-pores specimens have the largest bubble generation rate at the beginning of nuclear boiling, and the micro-groove specimens have the highest CHF.ConclusionsThe influence law and mechanism of long-term corrosion in pressurized water reactor on the enhanced heat transfer effect of different micro-structure surfaces are partially revealed by this study.
In this research, for improving the measurement performance of Prompt Gamma-ray Neutron Activation Analysis (PGNAA) set-up, a new optimization method for set-up design was proposed and investigated. At first, the calculation method for Signal-to-Noise Ratio (SNR) was proposed. Since the SNR could be calculated and quantified accurately, the SNR was chosen as the evaluation parameter in the new optimization method. For discussing the feasibility of the SNR optimization method, two kinds of PGNAA set-ups were designed in the MCNP code, based on the SNR optimization method and the previous signal optimization method, respectively. Meanwhile, the single element spectra analysis method was proposed, and the analysis effect of single element spectra as well as element sensitivity were used for comparing the measurement performance. Since the simulation results showed the better measurement performance of set-up designed by SNR optimization method, the experimental set-ups were built for the further testing, finally demonstrating the feasibility of the SNR optimization method for PGNAA set-up design.
Background99Mo has important applications in nuclear medicine field. It can be produced in a 235U fission reactor with high yield and specific activity.PurposeThis study aims to optimize the core parameters of new type molten salt fast reactor, and then analyze the production of 99Mo based on the optimized core.MethodsThe Standardized Computer Analyses for Licensing Evaluation Version 6.1 (SCALE6.1) was used to perform the critical calculations, by analyzing the influence of geometry parameters on keff (effective proliferation factor) and temperature coefficient of reactivity, the geometry parameters of the core were determined, and then the extraction position of fuel target in the production of 99Mo was determined by analyzing the neutron energy spectrum and mean microscopic fission cross sections of 235U. Finally, the output of 99Mo in fuel target was calculated.ResultsThe results show that the keff can be increased while the coolant temperature coefficient be decreased by increasing the radius of fuel element and reducing the radius of fuel cell appropriately. When the fuel element container wall thickness is 0.1 cm, fuel element radius is 3.5 cm, fuel cell radius is 5 cm, active area radius and reflector thickness are 63 cm and 100 cm respectively, the operating life of core reaches 32 months and the temperature coefficient of reactivity is -1.615×10-5 K-1, which ensure the inherent safety of reactor core. Selecting the outermost fuel element as the fuel target for 99Mo production can increase the yield of 99Mo. The 99Mo annual output exceeds 6.25×1016 Bq and specific activity exceeds 2.77×1015 Bq∙g-1 when the extraction cycle of fuel target is 7 days.ConclusionsTherefore, the high yield and purity of 99Mo can be obtained by producing 99Mo in the new type molten salt fast reactor.
Alloy 690 tubing has been shown to be highly resistant to primary water stress corrosion cracking (PWSCC). Nevertheless, predicting the failure by PWSCC in Alloy 690 SG tubes is indispensable. In this work, a Bayesian-based statistical approach is proposed to predict the occurrence of failure by PWSCC in Alloy 690 SG tubing. The prior distributions of the model parameters are developed based on the prior knowledge or information regarding the parameters. Since Alloy 690 is a replacement for Alloy 600, the parameter distributions of Alloy 600 tubing are used to gain prior information about the parameters of Alloy 690 tubing. In addition to estimating the model parameters, analysis of tubing reliability is also performed. Since no PWSCC has been observed in Alloy 690 tubing, only right-censored free-failure life of the tubing are available. Apparently the inference is sensitive to the choice of prior distribution when only right-censored data exist. Thus, one must be careful in choosing the prior distributions for the model parameters. It is found that the use of non-informative prior distribution yields unsatisfactory results, and strongly informative prior distribution will greatly influence the inference, especially when it is considerably optimistic relative to the observed data.
This paper is the first of three that overview the main mechanisms that drive the microstructure evolution in Fe alloys under irradiation. It focuses on pure α-Fe and compiles the parameters that describe quantitatively the mobility and stability of point-defects and especially their clusters, including possible reactions and criteria to decide when they should react. These parameters are the result of several years of calculations and application in microstructure evolution models. They are mainly collected from the literature and the parameter choice tries to reconcile different sets of values that, while being in general qualitatively similar, are often quantitatively not coincident. A few calculation results are presented here for the first time to support specific approximations concerning defect properties or features. Since calculations cannot cover all possible defect configurations, the definition of these parameters often requires educated guesses to fill knowledge gaps. These guesses are here listed and discussed whenever relevant. This is therefore a “hands-on” paper that: (i) collects in a single report most microstructure evolution parameters that are found in the literature for irradiated α-Fe, including a discussion of the most important mechanisms at play based on current knowledge; (ii) selects a ready-to-use set that can be employed in microstructure evolution models, such as those based on object kinetic Monte Carlo (OKMC) methods. This work also identifies parameters that are needed, but not known, hopefully prompting corresponding calculations in the future.
Andrea Sánchez Galindo, Yonatan Zuleta Ochoa, Carolina Osorio Castrillón
et al.
Las medidas de salud pública adoptadas para mitigar la propagación de la pandemia del Covid-19 han afectado la ejecución del control regulatorio en lo referente a la periodicidad de las inspecciones de instalaciones radiactivas. Se realiza un estudio de caso para identificar los aspectos del programa de protección radiológica de las instalaciones radiactivas que pueden ser verificados mediante visita remota por medio de herramientas tecnológicas, los beneficios de su realización y posibles limitaciones. Se escogió como instalación piloto la Refinería de Cartagena S.A.S., entidad autorizada en el territorio colombiano para el empleo de medidores nucleares. La visita remota se restringe a las verificaciones que tradicionalmente se basan en procedimientos presenciales como la comprobación del inventario de fuentes radiactivas, la medición de los niveles de radiación, las entrevistas al personal ocupacionalmente expuesto y la verificación de las disposiciones de protección radiológica. En la planificación de la visita remota fue necesario realizar varias mesas de trabajo entre la entidad explotadora y el grupo verificador con el fin de delimitar el alcance del estudio y seleccionar las herramientas tecnológicas y de comunicación. Durante la ejecución de la visita se presentaron inconvenientes de conectividad y restricciones de acceso por causas externas, pero aun así, fue posible realizar las verificaciones acorde al plan de visita y registrar las correspondientes evidencias. La adecuada planificación y la disponibilidad de recursos tecnológicos permiten que las verificaciones remotas pueden constituirse en una alternativa a las inspecciones de control regulatorio. De este modo, se facilita el cumplimiento de la función reguladora ante las medidas de salud pública adoptadas en medio de la pandemia como también optimizar el proceso de inspección, generar material de entrenamiento y fomentar la cultura de la seguridad. Se observa la conveniencia de realizar verificaciones remotas previo a la inspección convencional como una alternativa o complemento de esta.
The exchange behaviors of hydrogen isotopes between protonated lithium metal compounds and deuterated water or tritiated water were investigated. The various protonated lithium metal compounds were prepared by acid treatment of lithium metal compounds with different crystal structures and metal compositions. The protonated lithium metal compounds could more effectively reduce the deuterium concentration in water compared with the corresponding pristine lithium metal compounds. The H+ in the protonated lithium metal compounds was speculated to be more readily exchangeable with hydrons in the aqueous solution compared with Li+ in the pristine lithium metal compounds, and the exchanged heavier isotopes were speculated to be more stably retained in the crystal structure compared with the light protons. When the tritiated water (157.7 kBq/kg) was reacted with the protonated lithium metal compounds, the protonated lithium manganese nickel cobalt oxide was found to adsorb and retain twice as much tritium (163.9 Bq/g) as the protonated lithium manganese oxide (69.9 Bq/g) and the protonated lithium cobalt oxide (75.1 Bq/g) in the equilibrium state.
Khawar Nadeem, Javaid Hussain, Noaman Ul Haq
et al.
Ten gizzard and three flesh samples of the broiler were collected from different locations in Tehsil Gujar Khan District Rawalpindi, Pakistan. The samples were dried, crushed and ground. Pellets were prepared by pressing the powder of the samples and that of the Bovine liver 1577c reference material obtained from NIST, USA. Proton induced X-ray emission (PIXE) installed at National Center for Physics, Islamabad, Pakistan has been used as a reliable and improved technique to determine concentration of various major/trace and toxic elements e.g. S, Cl, K, Ca, Cl, Fe, Cu, Mn, Co, Zn, Ti, Cd, Ga, Cr, V and Ni, in the Gizzard and Flesh samples of the broiler. The concentrations of all the detected elements in the samples are statistically significant. The certified and measured values of the elements in the reference material were in agreement with each other within a deviation of 7%. S, Cl, K and Ca are within tolerable limits and are good for human consumption. Ti, V, Cr, Mn, Fe, Co, Ni, Cu and Zn were more than the acceptable limits of World Health Organization, WHO whereas Ga, As, Sn, Sb and Pb are not detected in most of the samples. Keywords: Major/minor minerals, Trace elements, PIXE, Broiler gizzard, Broiler flesh, Permissible limits
Isabelle Guénot-Delahaie, Jérôme Sercombe, Thomas Helfer
et al.
The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs), power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs). As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on PWR-UO2 fuel rods with advanced claddings such as M5® under “low pressure–low temperature” or “high pressure–high temperature” water coolant conditions.This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on UO2-M5® fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE—starting from base irradiation conditions it itself computes—is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur.Areas of improvement are finally discussed with a view to simulating and analyzing further tests to be performed under prototypical PWR conditions within the CABRI International Program.M5® is a trademark or a registered trademark of AREVA NP in the USA or other countries. Keywords: Pressurized Water Reactor, Nuclear Fuel, UO2, M5®, Reactivity-initiated Accident, ALCYONE Code