AbstractThe development of new reactor technologies requires careful assessments of the various sources of epistemic uncertainties. In this work, nuclear data uncertainties featuring the main isotopes of the U/Th molten salt fast reactor (MSFR) design are propagated through Monte Carlo calculations to quantify the final uncertainty on some relevant integral parameters. In the first part of this paper, some best-estimate calculations are performed by selecting different nuclear data libraries, showing the remarkable impact of this choice on the final responses. Then the Generalized Perturbation Theory routine available in Serpent 2 is adopted for a preliminary sensitivity and uncertainty analyses with respect to keff, highlighting a significant discrepancy between the covariance of the JEFF-3.3 and ENDF/B-VIII.0 libraries. After the selection of a few relevant nuclides, namely, 7Li, 19F, 232Th, and 233U, the Total Monte Carlo method and the unscented transform (UT) are then adopted to estimate the uncertainty of other responses of interest like the conversion ratio and some multigroup constants. Some potential issues of the UT are highlighted, and a mitigation strategy is applied. A relevant result of this analysis concerns the need for better data evaluations for the nuclides constituting the circulating salt for an effective deployment of the MSFR technology.

AbstractA comprehensive study of the radiological CNGS (CERN Neutrinos to Gran Sasso Experiment) environment characterization is presented. It comprises the evaluation of the residual dose rates of the most relevant standalone beam line equipment, such as the target and horn, as well as overall dose levels in the cavern before and after dismantling. Furthermore, the radionuclide inventories of the main objects to be dismantled were calculated by the Monte Carlo FLUKA code and ActiWiz. The latter is particularly important for transport and waste management. Moreover, we present benchmarking measurements of residual dose rates in the experimental cavern, staying in good agreement with simulation predictions. Additional measurements, as well as FLUKA and ActiWiz studies, allowed for assessing the concrete composition of the cavern’s walls and floor and the shielding blocks. The resulting refined composition allowed for evaluating more precisely the radionuclide inventories and residual dose rates expected before and after the dismantling in the CNGS target area. This was particularly important for the evaluation of the dismantling cost and the substantial savings due to the reusage of the majority of the concrete blocks. Finally, contamination measurements in the accessible parts of the area also are included. All the results discussed are crucial for determining the requirements, planning, and costs of the CNGS dismantling.

AbstractIn the framework of risk assessment in nuclear accidents, simulation tools are widely used to understand and model physical phenomena. These simulation tools take into account a large number of uncertain input parameters. We often use Monte Carlo–type methods to explore their range of variation: The input space is randomly sampled, and a code run is performed on each sampled point. However, some of these code runs may fail to converge. Analyzing these code failures to understand which of the inputs have the most influence on them leads to a better understanding of how the code works. It also intends to improve the robustness of the simulation software and code computations. For this purpose, we propose two complementary approaches performing a statistical analysis of the code failures. The first approach is based on goodness-of-fit tests and compares conditional probability distributions according to code failures to a reference one. A second approach, based on a dependence measure named the Hilbert-Schmidt Independence Criterion, provides another way to measure the global dependence between the inputs and the code failures. The development of this methodology is carried out in the context of severe nuclear accidents. More especially, the presented methods are applied for the study of the simulation code MC3D, which simulates the fuel-coolant interaction in a severe nuclear accident context.

AbstractGas bremsstrahlung, generated by the interaction between stored electrons and residual gas in electron storage rings, is an important radiation source for the shielding of synchrotron radiation (SR) facilities. In recent SR facilities, hydrogen was found dominant in the residual gas of the vacuum chambers of the electron storage rings, although air has been conventionally assumed as the bremsstrahlung target for the shielding designs of SR beamlines extended from the electron storage ring. To study the effect of residual gas composition on the dose rate outside shields, we calculated the intensity of gas bremsstrahlung based on the gas composition for both the air and the residual gas expected in the recent electron storage rings using an analytical formula and general-purpose Monte Carlo codes for particle transport calculations. The analytical shielding calculation with a realistic gas composition was found to well reproduce the energy spectra of gas bremsstrahlung simulated by the Monte Carlo codes. The correction factors between the air and the realistic gas compositions are applied to the conventional analytical formulas for dose estimation of secondary radiations generated by the interaction between the bremsstrahlung from air and beamline components.

AbstractIn nuclear engineering, Monte Carlo (MC) methods are commonly used for reactor analysis and radiation shielding problems. These methods are capable of dealing with both simple and complex system models with accuracy. The application of MC methods experiences challenges when dealing with the deep penetration problems that are typically encountered in radiation shielding cases. It is difficult to produce statistically reliable results due to poor particle sampling in the region of interest. Therefore, such calculations are performed by the Monte Carlo N-Particle Transport (MCNP) code in association with the weight window (WW) variance reduction technique, which increases the particle statistics in the desired tally region. However, for large problems, MCNP’s built-in weight window generator (WWG) produces zero WW parameters for tally regions located far from the source. To address this issue, the recursive Monte Carlo (RMC) method was proposed. This paper focuses on the RMC methodology and its implementation in the Helmholtz-Zentrum Dresden-Rossendorf’s (HZDR’s) in-house code TRAWEI, which is responsible for producing optimal zone weight parameters used for optimizing deep penetration MC calculations. In addition, this paper discusses the verification of the TRAWEI weight generator program to that of an existing MCNP WWG. The performance of TRAWEI-generated weight values is assessed using a handful of test cases involving two shield materials. Globally, the TRAWEI-generated weight values improved not only the statistical variance and computational efficiency of the MC run compared to the analog MCNP simulation but also those of the simulation with WW values generated by the standard MCNP WWG.

AbstractMolten Salt Reactor (MSR) systems can be divided into two basic categories: liquid-fueled MSRs in which the fuel is dissolved in the salt, and solid-fueled systems such as the Fluoride-salt-cooled High-temperature Reactor (FHR). The molten salt provides an impediment to fission product release as actinides and many fission products are soluble in molten salt. Nonetheless, under accident conditions, some radionuclides may escape the salt by vaporization and aerosol formation, which may lead to release into the environment. We present recent enhancements to MELCOR to represent the transport of radionuclides in the salt and releases from the salt. Some soluble but volatile radionuclides may vaporize and subsequently condense to aerosol. Insoluble fission products can deposit on structures. Thermochimica, an open-source Gibbs Energy Minimization (GEM) code, has been integrated into MELCOR. With the appropriate thermochemical database, Thermochimica provides the solubility and vapor pressure of species as a function of temperature, pressure, and composition, which are needed to characterize the vaporization rate and the state of the salt with fission products. Since thermochemical databases are still under active development for molten salt systems, thermodynamic data for fission product solubility and vapor pressure may be user specified. This enables preliminary assessments of fission product transport in molten salt systems. In this paper, we discuss modeling of soluble and insoluble fission product releases in a MSR with Thermochimica incorporated into MELCOR. Separate-effects experiments performed as part of the Molten Salt Reactor Experiment in which radioactive aerosol was released are discussed as needed for determining the source term.

AbstractThe increased complexity of modern system designs and demands for quicker time to market have made safety-related verification and validation of such systems more challenging. Incorporating safety and risk considerations at the early stages of design is one way to acquire a more robust initial design for novel systems. Inductive fault analysis has its significance at final stages of design, e.g., verification and validation. However, to preclude certain system failure states—especially for the systems with high failure consequences, a designer would innately prefer to trace back and remedy the causes of failure, as compared to a more cumbersome activity of identifying the faults individually and sifting the combinations that lead to the failure of interest. The work presented in this paper is aimed at the development of a backward failure propagation methodology for analyzing the origins of functional failures in a conceptual design of systems including but not limited to nuclear, mechanical, aerospace, process, electrical/electronics, telecommunication, automotive, etc. This method allows the designer to achieve a robust early design based on the analyses of the system’s functional dependencies before proceeding to the detailed design and testing stages. The insights provided by the analysis at the conceptual design stage also reduce redesign efforts, testing costs, and project delays. The proposed method is a functional analysis approach that extends the Integrated System Failure Analysis for backward failure propagation. When provided with an abstract system configuration, a system’s functional model, and a system’s behavioral model, it utilizes a known functional state (typically a failure) to acquire system component modes and the states of other functions. The method includes inversion of the functional failure logic and component behavioral rules using propositional logic and deductive analysis to assess valid states of a system that satisfy the given initial conditions. To test the method’s scalability, we applied the proposed method to a simplified representation of the secondary loop of a typical pressurized water reactor.

AbstractDue to the presence of sodium, it is a challenging task to achieve the reliable and safe operation of steam generators in a sodium-cooled fast reactor (SFR). Water flow oscillations in a two-phase flow system worsen the tube integrity. An accurate prediction of two-phase pressure drop is essential in designing steam generators to operate in a stable regime. Toward this, experiments have been carried out on an industrial-size 19-tube model sodium-heated steam generator of 5.5-MW capacity to understand two-phase pressure drop characteristics at various operating conditions. The measured data are used to estimate the two-phase frictional pressure drop. The concept of a two-phase friction multiplier has been used in the present study. A significant variation in the two-phase frictional multiplier is seen with steam quality, whereas the variation of the two-phase friction multiplier is insignificant at saturated steam condition. Based on the experiments, complemented by computational model, a correlation has been developed for the two-phase frictional multiplier as a function of steam quality for sodium-heated once-through straight-tube steam generators.

AbstractTRIPOLI-4® is a general-purpose Monte Carlo radiation transport code developed by the Service d’Études des Réacteurs et de Mathématiques Appliquées at CEA-Saclay. It uses continuous-energy nuclear data to simulate neutron, photon, electron, and positron transport in fields like radiation shielding, reactor physics, and nuclear criticality safety. To study radiation protection dosimetry in human tissues and organs, male and female adult computational phantoms from the Medical Internal Radiation Dose–Oak Ridge National Laboratory phantoms family and the International Commission on Radiological Protection (ICRP) publication 110 were recently modeled and calculated using the geometry options of TRIPOLI-4. To easily use the ICRP 110 voxel-based phantoms in different exposure scenarios, a newly developed phantom option is available in TRIPOLI-4 and its display tool T4G. This new phantom option is helpful for modeling one or more phantoms and for improving calculation performance in real irradiation environments. The 2020 published pediatric computational reference phantoms are accessible from ICRP publication 143. Male and female pediatric phantoms are also verified with the new T4G tool and TRIPOLI-4 code.This paper reports on recent works using TRIPOLI-4 on adult and pediatric computational phantoms. The modeling methods of stylized and voxel-based phantoms, the graphic displays of modeled phantoms with T4G, and the verification procedures for single-phantom and two-phantom application cases are presented. Validation for external and internal dosimetry calculations has been performed. Calculation results on organ dose S values for nuclear medicine applications are presented for single female and single male voxel phantoms using 131I and 177Lu radiation sources. Effective dose calculations for two-phantom cases using 99mTc and 18F sources are compared with traditional H*(10) calculations from nuclear medicine patient to patient caregiver.

AbstractThe European Spallation Source (ESS), presently under construction in Lund, Sweden, is a multidisciplinary international laboratory that, once completed at full specifications, will operate the world’s most powerful pulsed neutron source. Supported by a 3 M Euro Research and Innovation Action within the European Union Horizon 2020 program, a design study (HighNESS) is now underway to develop a second neutron source located below the spallation target. Compared to the first source, which is located above the spallation target and designed for high cold and thermal brightness, the new source is being optimized to deliver higher intensity and a shift to longer wavelengths in the spectral regions of cold neutrons (CNs) (2 to 20 Å), very cold neutrons (VCNs) (10 to 120 Å), and ultracold neutrons (UCNs) (>500>500>500 Å). The second source consists of a large liquid deuterium moderator to deliver CNs and serve secondary VCN and UCN sources, for which different options are under study. These new sources will boost several areas of condensed matter research and will provide unique opportunities in fundamental physics. The HighNESS project is now entering its last year, and we are working toward the Conceptual Design Report of the ESS upgrade. In this paper, results obtained in the first 2 years, ongoing developments, and future perspectives are described.

AbstractWithin the maximum entropy principle, we present a general theory able to describe in a dynamical context the transport properties of hot carriers in monolayer graphene under electric fields of arbitrary strength. Therefore, we obtain a closed extended hyperbolic system of hydrodynamic (HD) equations in which all the unknown constitutive functions are completely determined. In particular, we consider the different scattering mechanisms used in the literature in the kinetic approaches. The closed extended HD system is applied to monolayer graphene at 300 K and is validated by comparing numerical calculations with ensemble Monte Carlo simulations.

AbstractA new method for directly sampling the resonance upscattering effect is presented. Alternatives have relied on inefficient rejection sampling techniques or large tabular storage of relative velocities. None of these approaches, which require pointwise energy data, are particularly well suited to the windowed multipole cross-section representation. The new method, called multipole analytic resonance scattering, overcomes these limitations by inverse transform sampling from the target relative velocity distribution where the cross section is expressed in the multipole formalism. The closed-form relative speed distribution contains a novel special function we deem the incomplete Faddeeva function, and we present the first results on its efficient numerical evaluation.

AbstractNeutron irradiation experiments are carried out in source-driven subcritical cores with high-energy neutrons generated by spallation reactions between a 100-MeV proton beam and a lead-bismuth target at the Kyoto University Critical Assembly. The main objective of the experiments is to investigate the effect of epithermal and resonance neutrons on the accuracy of capture reaction rates with respect to a subcriticality variation. Activation foils of copper, indium, tantalum, and tungsten are employed to obtain capture reaction rates for epithermal and resonance neutrons by applying the cadmium difference method. Also, the applicability of the foils for the measurement of the reaction rates for epithermal and resonance neutrons is substantiated in the critical irradiation experiments performed prior to the subcritical experiments. The subcritical experiments are conducted with three different subcriticalities by changing the control rod insertion pattern.The measured reaction rates are compared with the calculated values obtained by the Monte Carlo code MVP with JENDL-4.0, and the ratio of the calculation and experiment values of the reaction rates shows equivalent values within the 1σ errors regardless of a difference in the subcriticality. The compared results indicate that the numerical analyses have a consistent accuracy of reaction rates in epithermal and resonance energy regions for a subcriticality variation in source-driven subcritical cores.

AbstractHigh-energy radiotherapy treatments of a medical Linear Accelerator (LinAc) generate secondary neutrons that can produce health damage on the human body as the induction of secondary cancers. The energy spectrum of these neutrons must be determined to estimate the extra dose received by patients inside a radiotherapy room during radiotherapy treatment. To quantify the neutron production, a Ludlum Bonner sphere spectrometer (BSS) is used for measurement at different points of a LinAc bunker at the Hospital Universitari i Politècnic La Fe de València. With the neutron measured data and a set of response detector curves obtained by Monte Carlo simulations with MCNP6.1.1, the Maximum Likelihood Expectation Maximization unfolding method is used to unfold the energy neutron spectrum. Unfolded neutron spectra at different locations were compared to those obtained by Monte Carlo simulation of the same setup, showing the same energetic behavior. The fluence rate decreases with source distance, and the shape changes from a fast neutron peak in the nearest LinAc head location to a prominent thermal neutron peak in the bunker maze region. Moreover, the neutron ambient equivalent dose was obtained from the unfolded spectra and compared to Berthold detector measurements, being consistent.

AbstractPending the availability of an operational long-term spent nuclear fuel (SNF) repository or other disposal methods, SNF will be increasingly stored in interim dry casks. Casks loaded with commercial SNF may contain several significant quantities of plutonium, so appropriate nuclear material safeguards monitoring is in order. An external remote monitoring system (RMS) developed by researchers at Texas A&M University is proposed to further the current dry cask safeguards regime, which is limited to containment and surveillance mechanisms. In this study, neutron measurements of SNF in dry cask storage were performed with the external RMS at a commercial interim spent fuel storage installation. Corresponding neutron transport simulations using MCNP were conducted with two types of detector responses (tallies) and the results were compared with measurements.The objectives of the study were to add dry cask measurement data to the literature, to assess the performance of the external RMS in full-scale dry cask measurements, and to investigate the degree to which measurements could be estimated with high-fidelity radiation transport simulations. The study demonstrated that the external RMS can acquire neutron count rate measurements with a relative error of less than 0.5% in 5 min or less through the shielding of a dry cask lid. Additionally, the developed simulation model matched trends in the measurement data to a degree that exceeds results in current literature, and normalization factors were calculated to better estimate the magnitude of neutron count rates.

AbstractThe Fokker-Planck equation (FPE) is one of the quintessential equations of particle transport theory. Representing small angle scattering characteristics of electron and photon transport by differential scattering indeed is a mathematical/numerical challenge. Here, we address the challenge with the method of response matrix applied to the Sn approximation to arrive at a nearly six-place-precision benchmark. Our approach aligns with the response matrix solution of the radiative transfer equation for anisotropic scattering published previously. We conclude with the comparison of the response matrix benchmark to a classical finite difference approximation.

AbstractThe set of benchmark solutions used in the thermal radiative transfer community suffers some coverage gaps, in particular, nonlinear, nonequilibrium problems. Also, there are no nonequilibrium, optically thick benchmarks. These shortcomings motivated the development of a numerical method free from the requirement of linearity and easily able to converge on smooth optically thick problems, i.e., a moving mesh Discontinuous Galerkin framework that utilizes an uncollided source treatment. Having already proven this method on time-dependent scattering transport problems, we present here solutions to nonequilibrium thermal radiative transfer problems for familiar linearized systems together with more physical nonlinear systems in both optically thin and thick regimes, including both the full transport and the S2/P1 solution. Geometric convergence is observed for smooth sources at all times and some nonsmooth sources at late times when there is local equilibrium. Also, accurate solutions are achieved for step sources when the solution is not smooth.

AbstractStable density stratification in a large enclosure could significantly hamper the effectiveness of natural convection cooling in pool-type liquid metal or gas-cooled advanced reactors. In addition, accurate prediction of stratified front behavior remains to be a challenging task for turbulence modeling. With the rapid growth of high-performance-computing capabilities in recent years, conducting high-fidelity simulations for a large-timescale transient has become more affordable and hence a valuable data source to support turbulence modeling as well as to gain further physical insights. In this work, direct numerical simulation is performed at experiment-consistent conditions to simulate the density stratification transient High-Resolution Jet (HiRJET) facility. Specifically, we focus on the case where an injected aqueous sugar solution has 1.5% density higher than that in the enclosure. In the early stage of the transient, the impingement of the denser jet to the bottom surface of the enclosure promoted turbulent mixing locally. This rendered the establishment of the mixture layer, formation and swift upward propagation of the stratified front, and elevation with (locally) the highest vertical concentration gradient. As the front rose, the diminishing turbulent mass flux slowed down the propagation, and a larger vertical concentration gradient was established. In this stage, both the velocity and the concentration scalar showed large-timescale fluctuation behavior around the stratified front. For the concentration time signal, the characteristic frequency in the power spectral density was found to agree well with the Brunt-Väisällä frequency. The preliminary validation endeavor showed that the stratified front location and the corresponding concentration gradient magnitude in the simulation agreed well with the experiment data. Further validation will mainly revolve around benchmarking against high-resolution density measurement and high-order flow statistics.

AbstractThe pinhole gamma camera is a simple and useful device for determining the radiation distribution in a certain region. Using this device, we developed a method to measure the distribution of 137Cs contamination density on surfaces using the total energy absorption peak count rate of gamma rays, where each camera pixel was projected onto the surface to determine the corresponding measured area and distance to the surface. We applied this method to measure the 137Cs contamination density of the wall, ceiling, and floor of the Unit 2 Operation Floor at the Fukushima Daiichi Nuclear Power Station in 2020 and 2022 and compared the results obtained in 2020 to those of a robot-operated, conventional, high-dose-area smear test. We found a pinhole gamma camera with the proposed method is useful for obtaining contamination density distribution results quickly, without the complexities of using a robot.

AbstractA production route for 99mTc and 101Tc using the (γ,p) reaction was considered. For an electron beam with energy of 40 MeV and power of 10 kW, distributions of produced 99mTc and 101Tc were obtained. For the optimized target (0.5 g) configurations, values of 99mTc specific activities of 9.53 GBq/g (0.26 Ci/g) for a one-stage setup and of 1.77 GBq/g (0.049 Ci/g) for a two-stage setup were obtained. For 101Tc, the corresponding values were 6.51 GBq/g (0.18 Ci/g) and 1.24 GBq/g (0.033 Ci/g).