Nuclear utilities are increasingly facing challenges meeting evolving regulatory requirements of probabilistic structural assessments with existing tools. Probabilistic methodologies are widely used as part of reactor safety analyses, including piping component structural integrity and fitness-for-service assessments. Probabilistic structural integrity assessments for nuclear piping components are typically conducted using special-purpose codes that may only provide limited probabilistic capabilities, may not be used with high-fidelity simulation software, or may not be used on high-performance computing clusters for large simulation campaigns. In the Canadian nuclear industry, probabilistic assessments for fuel channels are constrained by the large population of fuel channels in the case of in-core analyses. Meanwhile, for components such as steam generators, assessments are constrained by interfacing capabilities of the complex simulation codes that would be needed to represent the relevant physical phenomena. A generic and scalable probabilistic framework is needed to enable utilities to conduct large-scale probabilistic structural assessments with representative simulation codes. The present study proposes a novel and scalable framework for developing Monte Carlo analysis workflows to assess the structural integrity of nuclear plant piping components. The methodology is demonstrated for two relevant scenarios: a vibration-induced cracking assessment of a steam generator and a CANDU1 pressure-tube fitness-for-service assessment. The study shows that the proposed methodology provides a convenient means for assessing uncertainty propagation, confidence intervals, and output parameter distributions for probabilistic structural integrity cases.

Comparison of results for global responses predicted by different multiphysics simulations of benchmark problems may fail to reveal potentially significant local modeling issues. An examination of code interactions in coupled simulations can provide more information, which may help identify potential modeling issues that went unnoticed during verification (and validation) of the individual codes, or may call into question approximations otherwise deemed reasonable at the individual code level. We illustrate this challenge for the case of coupled neutronics/thermal-hydraulic transient simulations using one of the problems and contributed results documented in International Atomic Energy Agency TECDOC-1994. This recently published report documents the specifications of four numerical multiphysics pressurized heavy water reactor (PHWR) transient challenge problems and the results contributed by 10 participants. Our work is based on the pump rundown problem, where TECDOC-1994 suggests that differences in modeling and methods employed in thermal-hydraulics may be the dominant factor in the observed differences. We performed a more detailed assessment with two different multiphysics coupled computational frameworks using NESTLE-C/ARIANT and PUMA/RELAP-5. We also studied a pump seizure transient, a more challenging variant of the pump rundown transient. Several aspects were investigated: comparisons of standalone results, sensitivity to gap modeling, selection of boundary conditions at the pressurizer, and an examination of correlations used in ARIANT and RELAP-5. Our assessment goes beyond the results for global parameters and dives into details of predictions at the channel level. This paper briefly describes the PHWR pump rundown transient problem and a pump seizure variant, the computational methods employed, and the areas investigated, and discusses some selected results.

CANDU1 1 reactors are a channel-type design where up to 480 zirconium alloy pressure tubes (PTs) act as the reactor pressure boundary. Fitness-for-service standards such as CSA N285.8 (CSA Group, 2015) have been established to prescribe the requirements of pressure tube evaluations to maintain their integrity throughout their lifetime. Probabilistic leak-before-break (LBB) assessments can be used in evaluations to demonstrate that the likelihood of a flaw growing past the stability point while undetected (or before a safe shutdown can be achieved) is below regulatory limits. The assessments typically require a ranking of model parameters based on their influence. The ranking process traditionally uses expert judgment and local sensitivity analysis methods. There has been recent interest in implementing global methods for LBB assessments as they can manage large numbers of variables and extensive sampling spaces. However, their adoption is hindered by the computational cost of the models involved. The current study proposes a method for performing global sensitivity parameter rankings during probabilistic LBB assessments. The study is the continuation of the CANDU nuclear power plant (NPP) pressure tube (PT) case study in El Bouzidi et al., (2024) where a probabilistic approach was implemented in RAVEN (Rabiti and Alfonsi, 2019) to estimate whether the probability of unstable crack growth in PTs is within acceptable limits. This follow-up work proposes surrogate modeling using a high-dimensional model representation to derive Sobol indices and establish a parameter influence ranking. The methodology is demonstrated in a Monte Carlo simulation campaign of a pressure tube inlet rolled joint. The implemented approach successfully identified and quantified critical parameters of the model while efficiently ranking and quantifying the interactions between input variables.

TRistructural ISOtropic (TRISO) fuels are commonly used in High-Temperature Gas cooled Reactors due to their resistance to high temperatures, irradiation, and oxidation. Reactor physics modelling of a fuel form consisting of millions of layered particles is quite challenging, however. Homogenisation techniques are often used to simplify the fuel form, representing it as a solid material to reduce computational costs. The impact of these approximations on reactor physics parameters was evaluated using OpenMC. Geometries at various scales (i.e. fuel pin, fuel block, full core) were developed based on the Modular High-Temperature Gas-cooled Reactor. It was found that homogenisation results in a significant reduction in simulation time. The Reactivity equivalent Physical Transformation (RPT) method was found to be superior to simple Volume-Weighted Homogenization (VWH). The neutron energy spectra and flux distributions calculated by RPT had insignificant differences to the explicit model. Reactivity coefficients were found to be similar, but diverged at very high temperatures.

In an ideal risk-informed approach to safety analysis, all possible operational states of the system and all sources of uncertainty would be factored. Brute force approaches are cost-prohibitive if high-fidelity multiphysics simulation systems are used. To address this barrier, we present a first attempt to develop emulators for coupled neutronics/thermalhydraulics simulations of large break loss-of-coolant accidents in a generic pressurized heavy water reactor. The emulators are based on a linear mean function coupled with a locally approximate Gaussian process model. The emulators are capable of predicting the power pulse amplitude and the maximum bundle enthalpy within mean absolute relative errors of 0.75 and 0.36%, and mean relative margins of error of 1.9 and 1.0%, respectively. In relation to the simulator, the evaluation time is reduced by a factor of nearly 1800. The paper covers the simulation system, training data generation framework, statistical models and uncertainties, and results.

This paper provides an overview of the capabilities of a new toolset that is being developed at Canadian Nuclear Laboratories for modelling advanced non-water cooled small modular reactor concepts. Throughout the study, a technology agnostic approach was adopted. The integrated toolset includes computational fluid dynamics, neutronics and system thermalhydraulics codes to handle a range of multiphysics capabilities. The toolset has been applied to two different non-water cooled SMR concepts: molten salt and prismatic gas-cooled reactors. A step-wise approach was used in which the models were first tested using standalone codes, followed by execution of coupled-code simulations. This paper demonstrates the application of the toolset to gas-cooled and molten-salt reactor concepts, including transient simulations using the CFD code STAR-CCM+ coupled with the Monte Carlo neutronics code Serpent 2 and with the system thermalhydraulics code RELAP5-3D. The scenarios considered include a step reactivity insertion and a pump trip for the molten salt reactors. Overall, the results obtained so far demonstrates the potential of the developed methodology at CNL in simulating the non-water cooled concepts. Crown Copyright (C) 2021 Published by Elsevier Ltd. All rights reserved.

This paper reports on the development and testing of a comprehensive few-group cross section input uncertainty library for the NESTLE-C nodal diffusion-based nuclear reactor core simulator. This library represents the first milestone of a first-of-a-kind framework for the integrated characterization of uncertainties in steady-state and transient CANDU reactor simulations. The objective of this framework is to propagate, prioritize and devise a mapping capability for uncertainties in support of model validation of best-estimate calculations. A complete framework would factor both input and modeling uncertainty contributions. The scope of the present work is limited to the propagation of multi-group cross-section uncertainties through lattice physics calculations down to the few-group format, representing the input to the NESTLE-C core simulator, and finally to core responses of interest.

The accurate prediction of multiphase flow when particles are differentiated by a set of “internal” variables, such as size or temperature, can pose modelling and numerical challenges. Although Lagrangian particle methods can provide predictive simulations of a wide spectrum of complex multiphase problems, they can become prohibitively expensive as the number of particles becomes large. Alternatively, Eulerian approaches have the potential to improve the computational efficiency of multiphase flows, but classical methods produce modelling artifacts or do not properly treat the local statistical dependence between the particle velocities and internal variables, or between the internal variables themselves. In this paper an extension is proposed of the classical Gaussian ten-moment model from gaskinetic theory to a model for the treatment of a dilute particle phase with an arbitrary number of internal variables based on an entropy-maximization argument. Unlike previous formulations, this new model provides a set of first-order robustly-hyperbolic balance laws that include a direct treatment for the local statistical variance of each variable, as well as the covariance between the internal variables or the internal variables and particle velocity. A study of the wave speeds of the general hyperbolic system is presented. To demonstrate an example application, the model is then specialized for polydisperse flows that are subject to viscous fluid drag as well as gravitational and buoyancy forces. The complete eigenstructure of this fifteen-moment polydisperse Gaussian model (PGM) is presented and the PGM is shown to maintain a physically realizable distribution function for all admissible initial conditions. Finally, several illustrative low-dimensional problems are studied to demonstrate the predictive capabilities of the new model.

Canadian Nuclear Laboratories has developed a toolset of coupled reactor transient simulation codes for pressure tube heavy water reactors. This paper presents the first demonstration of the toolset's capabilities, using a hypothetical 20% reactor inlet header break (critical break) scenario. This scenario induces strong feedback between thermalhydraulic and reactor physics phenomena, owing to a positive coolant voiding coefficient. The scenario was simulated using the advanced thermalhydraulics code ARIANT and the reactor physics code NESTLE-C, coupled together using the SALOME platform. This paper summarizes recent advances in ARIANT, the coupling methodology, and the results of the transient simulation. Crown Copyright (C) 2018 Published by Elsevier Ltd. All rights reserved.

Sensitivity of fast reactor physics coefficients to nuclear data differs from thermal reactors. A conceptual lead-cooled reactor has been modelled and assessed for nuclear data sensitivity. A TSUNAMI sensitivity analysis was performed to estimate uncertainty and sensitivity of the neutron multiplication factor to the nuclear data of all nuclides and reactions relevant to the reactor concept. The results were compared against MCNP6, which also employs an adjoint-based method for sensitivity calculations. Sensitivity of the neutron multiplication factor was found to agree closely between MCNP6 and TSUNAMI, for lead cross sections and other important nuclear data. The TSUNAMI assessment of nuclear data impact was also examined through a brief survey of nuclear data evaluations. The evaluations were found to have differences of interpretation that impact the TSUNAMI calculations. The consequent uncertainties in fuel temperature and coolant-voiding coefficients were also examined in TSUNAMI. Based on the TSUNAMI ranking of important data, a subset of nuclides and reactions were selected for stochastic sampling of cross-section data using NJOY. The sampled cross-section data were then used in SERPENT simulations to assess uncertainties in calculated parameters, which were consistent with the TSUNAMI results. Crown Copyright (C) 2018 Published by Elsevier Ltd. All rights reserved.

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Many nuclear data uncertainty propagation methods are implemented on the basis of first-order perturbation theory. These methods are complex and integral verification using direct perturbation of the nuclear cross section data is difficult due to the structure of nuclear data files. We present a new implementation of the direct perturbation method, which eliminates the need to modify these files. The method was implemented in DPERT, a patched version of MCNP5.Using the DPERT patch, we present an integral verification of TSURFER based on 77 heavy water moderated ZED-2 critical experiments. TSURFER is a module of the SCALE code suite and applies first-order perturbation theory to propagate nuclear data uncertainties. The experiments were modeled using the standard MCNP5 code to establish the a priori k(eff) calculation biases. TSURFER was used to minimize these biases by adjusting the underlying nuclear data. The proposed cross section alterations were then applied to the experiment models, and the DPERT patch was used to verify TSURFER's evaluation of the a apsteriori k(eff). biases. The study confirmed the TSURFER bias reduction prediction, but suggests TSURFER may underestimate the impact of the nuclear data corrections by 1.35 +/- 0.05 mk on average. Crown Copyright (C) 2013 Published by Elsevier Ltd. All rights reserved.

Photostop is an accessible technique capable of producing atoms or molecules at a standstill in the laboratory frame. Starting with a NO2/Xe molecular beam with a mean velocity of 415 m s(-1) and a longitudinal translational temperature of 6K, NO2 molecules are photodissociated to yield NO(X-2 Pi(3/2), v = 0, J = 1.5) fragments with a recoil speed equal to the molecular beam speed. The fraction of NO fragments that recoil opposite to the molecular beam are produced with a 6K longitudinal velocity distribution centred at zero. The NO molecules are allowed to 'evaporate' from the probe volume by waiting for 10 mu s and the molecules left behind are probed with a translational temperature of 1.6K along the molecular beam axis and an estimated density of 10 7 cm(-3) per quantum state. Through the choice of suitable precursors, the photostop technique has the potential to extend the list of atoms and molecules that can be slowed or trapped. It should be possible to accumulate density in a trap through consecutive loading of multiple pulses.

An entry from the Cambridge Structural Database, the world’s repository for small molecule crystal structures. The entry contains experimental data from a crystal diffraction study. The deposited dataset for this entry is freely available from the CCDC and typically includes 3D coordinates, cell parameters, space group, experimental conditions and quality measures.

The H+ +D2(v=0,j=0)-->HD+D + reaction has been theoretically investigated by means of a time independent exact quantum mechanical approach, a quantum wave packet calculation within an adiabatic centrifugal sudden approximation, a statistical quantum model, and a quasiclassical trajectory calculation. Besides reaction probabilities as a function of collision energy at different values of the total angular momentum, J, special emphasis has been made at two specific collision energies, 0.1 and 0.524 eV. The occurrence of distinctive dynamical behavior at these two energies is analyzed in some detail. An extensive comparison with previous experimental measurements on the Rydberg H atom with D2 molecules has been carried out at the higher collision energy. In particular, the present theoretical results have been employed to perform simulations of the experimental kinetic energy spectra.

Photoinduced Fe-to-bpy charge transfer in [{Cp(dppe)Fe}(mu-C[triple bond]CC[triple bond]N){Re(CO)(3)(bpy)}]PF(6) has been observed by ps-TRIR spectroscopy, supported by UV-Vis/IR spectroelectrochemistry and DFT calculations.

After H-2 and water, CO and CO2 are the most common molecules in stellar and interstellar environments. They are often dominant components in icy mantles of comets and interstellar dust grains. These ice mantles are subject to a varying degree of cosmic radiation, leading to chemical alterations. This paper presents a study of the near-IR absorption spectrum (1900 - 4000 cm(-1)) of proton-irradiated CO thin films. The spectra of such films display several features arising in the vicinity of the CO fundamental that are attributed to carbon oxides such as CnO and CnO2 for n less than or equal to 7. The formation mechanisms of carbon chains in interstellar environments are of interest since carbon chains ( or molecules derived from them) have been considered to be carriers of some of the diffuse interstellar bands.

Proton-beam irradiation, using the 3 MV accelerator at Guelph, has been successful in allowing us to study the emission spectra from liquid helium and hydrogen-doped liquid helium despite the fact that the beam penetrates the target a scant 0.5 mm. H-2 emission lines can be observed from liquid samples following doping with small amounts of hydrogen into the helium gas before condensation. The line intensities are a sensitive unction of the sample temperature, becoming strongest as the liquid approaches 4.2 K. The H-2 lines themselves are much broader than those from He-2 because the helium lines actually originate in small bubbles of helium gas trapped in the liquid, which is above the lambda point The interesting fact is that electronic emission spectra from H-2 are observed at all from liquid helium and a calculation of the widths and shifts of the spectral lines will be presented.