Current Light Water Reactors Research Articles

Analyses of Jet Buoyant Flow in a Multicompartment Containment Using an Open-Source Solver

Hydrogen mixing and distribution in a nuclear power plant containment during a postulated severe accident is a phenomenon with high safety relevance for current and advanced light water reactors. The Organisation for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) SESAR Thermal-Hydraulics (SETH) project was carried out to generate experimental data on basic containment phenomena with the required instrumentation to assess computational fluid dynamics (CFD) and lumped parameter codes. In this paper, we analyze the PANDA SETH Test 16 of the OECD/NEA SETH project using containmentFOAM, which is a tailored open-source CFD code package for containment analysis developed at Forschungszentrum Jülich GmbH based on OpenFOAM.® SETH Test 16 addressed steam-air mixture transport driven by jets and plumes in the multicompartment PANDA facility (two vessels and an interconnecting pipe) at the Paul Scherrer Institut in Switzerland. Steam as a wall plume was injected at 0.040 kg/s and 140°C in vessel 1, which was initially filled with dry air at 108°C. The steam-air mixture was vented out from the top of vessel 2. The system pressure was constant at 1.3 bar for the whole test duration (4000 s). To simulate SETH Test 16, the k-ω shear stress transport turbulence model was adopted with the generalized gradient diffusive hypothesis buoyancy turbulence model. The simulation represented the experimental phenomenology reasonably well. The trends for the calculated temperature distribution were similar to those of the experiments and only yielded a slight discrepancy of about 2.7%. In addition, the distribution of the steam molar fraction was well predicted above the injection tube in vessel 1. However, below the injection tube in vessel 1, some discrepancies between the simulations and the experimental results were observed. These may have been caused by the higher turbulent mass diffusivity in the physical model and the challenge of the mesh resolution in the middle region of vessel 1.

Shutdown dose rate calculations in high-temperature gas-cooled reactors using the MCNP-ORIGEN activation automation tool

Background High-temperature gas-cooled reactors (HTGRs) have many distinct features from the current light water reactor (LWR) fleet, and potential decommissioning strategies should take them into consideration. The proper characterization of the shutdown dose rates can establish a suitable strategy. Methods This article introduces a new shutdown dose rate calculation capability that relies on the MCNP-ORIGEN activation automation tool and the MCNP repeated structures to explicitly model the TRistructural ISOtropic (TRISO) particles as decay radiation sources. Results Three exercises are conducted with this capability, and their results discussed. The first two exercises verify and demonstrate the workflow. The third exercise proposes a decommissioning strategy for a high-temperature gas-cooled microreactor and studies its feasibility from the shutdown dose rate standpoint. The calculations yield a dose rate map above at the reactor citadel after a three month cool-down, showing values above 40 mSv/h. Conclusions The findings imply that the decommissioning strategy based on placing the reactor in a safe shutdown state and allowing it to cool down for three months before removing various components and extracting the fuel assemblies, may not be sufficient to minimize unnecessary exposure of personnel.

Prognostic model and failure mechanisms of steam generators in Sodium-Cooled fast reactors

This paper presents a prognostic model for sodium-cooled fast reactor (SFR) steam generators (SGs). The purpose of the model is to estimate the remaining useful life of SFR SGs and thus to support the decision-making of autonomous control. SFR SGs are of great interest for plant integrity due to their harsh operation environment. They operate at higher temperatures and higher coolant-to-steam pressure differences than those of current light water reactors (LWR). The severity of the SFR SG failure consequences, which include water-sodium contact, is another reason. Understanding its failure mechanisms is important for the development of its prognostic model. Based on our literature review and physics-based analysis, we concluded that creep would be a dominant degradation mode of SFR SGs due to SFR’s high-temperature and high-pressure environment. Thus, creep is the focus of the prognostic model. Various other failure modes were also investigated in this study. The mechanical fatigue due to flow-induced vibration is usually observed in early developmental phases and would not be an important issue during normal operation. The thermal fatigue due to thermal stripping is occasionally observed in other components in SFRs but does not affect SG integrity. Pure water stress corrosive crack, fretting, and so on are commonly observed in LWR SGs but are significantly less important in SFR SGs because of the high temperatures, high pressure differences, and chemical properties of liquid sodium. Based on these investigations, a prognostic model focusing on creep failures was developed. It estimates the failure probability profile by sampling the Larson-Miller parameter (LMP) and temperature and associated uncertainties through Monte Carlo methods. Two case studies were presented. The first one demonstrated the model’s capability to calculate the failure probability for a new specimen within a given time. If the tube is working under temperatures of 500C ± 3 and pressures of 15.2 MPa, which leads to LMP of 20,100 ± 50, the failure probabilities within 50, 70, and 90 years are approximately 0.1 %, 3.2 %, and 17.5 %. The second demonstrated how adjusting the workload can help to protect the integrity of the component. For an old tube reaching 99 % of its lifespan, continuing to run at temperatures of 500C ± 3 and LMP of 20,100 ± 50 leads to a failure probability of about 29.0 % within a year. If the temperature is reduced by 5C, the failure probability can be reduced to 2.5 %, and, if the pressure is also reduced such that the LMP is increased by 100, the failure probability can further be reduced to 0.12 %.

Analytic Error Analysis of the Partial Derivatives Cross-Section Model—I: Derivation

Accurate assessment of uncertainties in cross-section data is crucial for reliable nuclear reactor simulations and safety analyses. In this study, we focus on the interpolation procedure of the partial derivatives (PD) cross-section model used to evaluate nodal parameters from pregenerated multigroup libraries. Our primary objective is to develop a systematic methodology that enables prediction of the incurred errors in the cross-section model, leading to the development of optimal case matrices, more efficient cross-section models, and informed case matrix construction for reactor types lacking substantial data and experience. We make progress toward this objective through a rigorous analytic error analysis enabled by the derivation of error expressions and bounds for the PD model based on the discovery that the method is a form of Lagrange interpolation. Our investigations reveal distinct outcomes depending on the chosen cross-section functionalizations, achieved by identifying the sources of error. These error sources are found to include interpolation error, which is always present, and model form error, which is a property of the supplied case matrix. We show that simply increasing grid refinement through the addition of branches may not always lead to decreased cross-section errors, particularly in cases where the model form error predominantly contributes to the total error. We present numerical results and verification in a companion paper, showing these different error characteristics for various cross-section functionalizations. Although applied to current light water reactor environments, our methodology offers a means for advanced reactor analysts to develop case matrices with quantified error levels, advancing the goal of a general methodology for robust two-step reactor analysis. Future work includes exploring different lattice types and functionalizations, extending reactivity bounds to multilattice problems, and investigating historical effects within the macroscopic depletion model.

Testing of fiber optic based sensors for advanced reactors in the Texas A&M University TRIGA reactor

An Optical Fiber Based Gamma Thermometer has been developed and tested at the Texas A&M University reactor as well as the Ohio State Research Reactor. This sensor technology provides temperature measurement of a thermal mass undergoing gamma heating, from which the gamma flux along the length of the sensor may be calculated. The gamma flux data from an array of OFBGTs could allow the determination of local power density within a nuclear reactor core in three dimensions. The design of the sensor makes use of optical fiber as a truly distributed temperature sensor with a spatial resolution less than 1 mm and temperature accuracy of ± 1 °C. The testing program showed that the system provided the expected temperature data for each location. Overall, this sensor technology shows promise for current Light Water Reactors as well as advanced nuclear reactors by providing an unparalleled level of precision in reactor power measurement.

Shutdown dose rate calculations in high-temperature gas-cooled reactors using the MCNP-ORIGEN activation automation tool

Background High-temperature gas-cooled reactors (HTGRs) have many distinct features from the current light water reactor (LWR) fleet, and potential decommissioning strategies should take them into consideration. The proper characterization of the shutdown dose rates can establish a suitable strategy. Methods This article introduces a new shutdown dose rate calculation capability that relies on the MCNP-ORIGEN activation automation tool and the MCNP repeated structures to explicitly model the TRistructural ISOtropic (TRISO) particles as decay radiation sources. Results Three exercises are conducted with this capability, and their results discussed. The first two exercises verify and demonstrate the workflow. The third exercise proposes a decommissioning strategy for a high-temperature gas-cooled microreactor and studies its feasibility from the shutdown dose rate standpoint. The calculations yield a dose rate map above at the reactor citadel after a three month cool-down, showing values above 40 mSv/h. Conclusions The findings imply that the decommissioning strategy based on placing the reactor in a safe shutdown state and allowing it to cool down for three months before removing various components and extracting the fuel assemblies, may not be sufficient to minimize unnecessary exposure of personnel.

Incineration of trans-uranium elements using chlorides molten salt fast reactors – A proposal of incineration in three steps

ABSTRACT Recently utilization of nuclear energy has been re-thought as means of realizing carbon neutral and diversity of natural resources. On the other hand, operating current light water reactors results in accumulation of spent nuclear fuels, and they contain trans-uranium elements (plutonium and minor actinides). The spent nuclear fuels must be reprocessed for volume reduction and recovery of energy resources, however, disposal sites of the high level radioactive wastes (HLRW) out of the reprocessing have not been found. In order to resolve these issues, we propose to utilize chlorides molten salt fast reactors.In the current study, a three-step cascades incineration scheme is propsed to effectively incinerate TRU using fuel salt of NaCl-CaCl2 system in fast reactors. The proposed reactor concept could solve the current problems concerning utilization of nuclear energy such as accumulation of spent nuclear fuels and plutonium and reduce burdens to decide disposition site for HLRW. Moreover, it could help to achieve carbon neutral.

Assessment of TRU burning in a molten salt reactor moderated by zirconium hydride rods

Inventory reduction of transuranium (TRU) nuclides can effectively reduce the long-lived and high-level radioactive hazards in the spent nuclear fuel (SNF) from the current light water reactors (LWRs), which is of importance to the long-term development and deployment of nuclear energy. Considering its outstanding features such as excellent neutron spectrum, high fuel utilization and compact core structure, a molten salt reactor moderated by zirconium hydride rods (ZrH-MSR) is considered as one of the potential reactor types for TRU burning. In this work, several fuel salts and molten salt volume fractions (SVFs) are analyzed to assess the TRU burning capability in a ZrH-MSR. The fuel salt of 77.5 mol% LiF-22.5 mol% (ThF4 + TRUF3) is selected since it can meet the requirement of TRU solubility limit during 50 years operation and can achieve a high initial TRU loaded mass in the core to improve TRU burning capacity. The optimized SVF is 0.5, and the TRU burning rate and the Support factor can achieve about 28.76 kg/TWthh and 2.9, respectively. The TRU burning mass, TRU fuel utilization ratio and 233U production after 50 years operation are about 12.6 tons, 70.4 %, and 0.7 tons, respectively. The total TRU radiotoxicity after discharge is about 63.9 % smaller than that without burning. This optimized core is beneficial to get a high TRU burning rate to reduce the storage of TRU and to minimize the long-term radiotoxicity risk in a relatively short reactor life. The total temperature feedback coefficient can remain negative during the whole operation time, which ensures the safety of the ZrH-MSR.

Effects of ion irradiation on microstructures and mechanical properties of SiOC nanocomposites

SiOC polymer-derived ceramics (PDCs) are commonly used materials for advanced structural and functional applications. Especially, when in the form of fibers, they are used in the SiCf/SiC composites that is being considered an option for fuel cladding components in the current and next-generation light water reactors. But the radiation effects of the SiOC PDCs have been rarely reported (mostly focused on SiC). Here we prepared SiOC nanocomposites consisting of β-SiC nanocrystals and turbostratic graphite embedded in amorphous SiOC matrix, and irradiated them at room temperature with 4 MeV Kr ions to different damage levels. The irradiation-induced changes in the microstructures, surface morphologies, and mechanical properties were studied. It is indicated much severer irradiation condition leads to higher rearrangement degree of Si-containing tetrahedral units in the amorphous SiOC matrix, higher disordering degree of the turbostratic graphite, but crystallinity of the β-SiC is preserved. No obvious change is detected in the macro properties such as surface morphologies and mechanical properties (hardness and modulus). This study indicates the potential of SiOC PDCs to radiation damage.

Design of a direct-cycle supercritical CO2 nuclear reactor with heavy water moderation

A new reactor concept is described that directly couples a supercritical CO2 (sCO2) power cycle with a CO2-cooled, heavy water moderated pressure tube core. This configuration attains the simplification and economic potential of past direct-cycle sCO2 concepts, while also providing safety and power density benefits by using the moderator as a heat sink for decay heat removal. A 200 MWe design is described that heavily leverages existing commercial nuclear technologies, including reactor and moderator systems from Canadian CANDU reactors and fuels and materials from UK Advanced Gas-cooled Reactors (AGRs). Descriptions are provided of the power cycle, nuclear island systems, reactor core, and safety systems, and the results of safety analyses are shown illustrating the ability of the design to withstand large-break loss of coolant accidents. The resulting design attains high efficiency while employing considerably fewer systems than current light water reactors and advanced reactor technologies, illustrating its economic promise. Prospects for the design are discussed, including the ability to demonstrate its technologies in a small (∼20 MWe) initial system, and avenues for further improvement of the design using advanced technologies.

Heat transfer analysis in sub-channels of rod bundle geometry with supercritical water

Parametric studies of heat transfer and fluid flow are very important research of interest because the design and operation of fluid flow and heat transfer systems are guided by these parametric studies. The safety of the system operation and system optimization can be determined by decreasing or increasing particular fluid flow and heat transfer parameter while keeping other parameters constant. The parameters that can be varied in order to determine safe and optimized system include system pressure, mass flow rate, heat flux and coolant inlet temperature among other parameters. The fluid flow and heat transfer systems can also be enhanced by the presence of or without the presence of particular effects including gravity effect among others. The advanced Generation IV reactors to be deployed for large electricity production, have proven to be more thermally efficient (approximately 45% thermal efficiency) than the current light water reactors with a thermal efficiency of approximately 33 °C. SCWR is one of the Generation IV reactors intended for electricity generation. High Performance Light Water Reactor (HPLWR) is a SCWR type which is under consideration in this study. One-eighth of a proposed fuel assembly design for HPLWR consisting of 7 fuel/rod bundles with 9 coolant sub-channels was the geometry considered in this study to examine the effects of system pressure and mass flow rate on wall and fluid temperatures. Gravity effect on wall and fluid temperatures were also examined on this one-eighth fuel assembly geometry. Computational Fluid Dynamics (CFD) code, STAR-CCM+, was used to obtain the results of the numerical simulations. Based on the parametric analysis carried out, sub-channel 4 performed better in terms of heat transfer because temperatures predicted in sub-channel 9 (corner sub-channel) were higher than the ones obtained in sub-channel 4 (central sub-channel). The influence of system mass flow rate, pressure and gravity seem similar in both sub-channels 4 and 9 with temperature distributions higher in sub-channel 9 than in sub-channel 4. In most of the cases considered, temperature distributions (for both fluid and wall) obtained at 25 MPa are higher than those obtained at 23 MPa, temperature distributions obtained at 601.2 kg/h are higher than those obtained at 561.2 kg/h, and temperature distributions obtained without gravity effect are higher than those obtained with gravity effect. The results show that effects of system pressure, mass flowrate and gravity on fluid flow and heat transfer are significant and therefore parametric studies need to be performed to determine safe and optimum operating conditions of fluid flow and heat transfer systems.

Development and application of a microstructure dependent thermal resistor model for UO2 reactor fuel with high thermal conductivity additives

High thermal conductivity additives are being explored to make next generation accident tolerant fuels for light water reactors (LWRs). The goal is to create composite fuels that will more effectively dissipate heat, thus lowering the high centerline temperatures that develop in current LWR fuel. These fuels can employ various additives and the fuel microstructures vary depending on the manufacturing process. In this work, a thermal resistor model is developed to quickly estimate the effective thermal conductivity of a composite with a high thermal conductivity additive. This model is unique in that it employs a parameter, the continuousness of the secondary constituent, to consider the composite microstructure in its predictions. A genetic algorithm was developed to measure a composite’s continuousness. The model and algorithm are validated using data from literature from four experimental composites composed of UO2 with beryllium oxide and silicon carbide additives. The model predictions have an absolute mean error across a wide temperature range of less than 0.50W/mK in most cases. These tools are then used to create a set of guidelines for microstructure design.

Preliminary modelling of crack nucleation and propagation in SiC/SiC accident-tolerant fuel during routine operational transients using peridynamics

Silicon carbide fibre in silicon carbide matrix composites (SiC/SiC) are a promising cladding for use in accident tolerant fuels (ATF) in current light water reactor (LWR) designs. However, as they are a radically different material from current metal clads, current thermomechanical simulation methods struggle to accurately predict their behaviour, especially regarding the potential development of cracks. Thus, a new peridynamic model for SiC/SiC cladding has been developed in the Abaqus finite element code. The material model was isotropic and considers matrix cracking and fibre pull-out. The thermal expansion, swelling and the degradation of the thermal conductivity are modelled under typical LWR irradiation conditions. The swelling on the outer surface is predicted to be greater than the inner surface due to the lower irradiation temperature, causing a tensile stress on the inside of the cladding; tension being more challenging for a ceramic than a metal. This stress increases during the decrease in power at the start of a typical pressurised water reactor refuelling outage and causes microcracking of the matrix on the cladding inner surface. In models without fibres, cracks would propagate through the cladding. If fibres are modelled, matrix cracking will extend to a depth of around 20% through the cladding from the inner surface, which is unlikely to be an acceptable design. If an inner monolith of SiC is additionally modelled, cracking propagates through the monolith and acts as a stress raiser for matrix cracking in the composite, and therefore does not constitute a design improvement. If an outer SiC monolith is modelled, fibre pull-out strain on the inner surface of the cladding was increased by just under 70%. No cracks are predicted in an outer monolith which may therefore remain gas-tight and thus a more suitable design. These predictions are consistent with experimental findings.

Recent studies on potential accident-tolerant fuel-cladding systems in light water reactors

Accident-tolerant fuel (ATF) has attracted considerable research attention since the 2011 Fukushima nuclear disaster. To improve the accident tolerance of the fuel-cladding systems in the current light-water reactors, it is proposed to develop and deploy (1) an enhanced Zr-based alloy or coated zircaloy for the fuel cladding, (2) alternative cladding materials with better accident tolerance, and (3) alternative fuels with enhanced accident tolerance and/or a higher U density. This review presents the features of the current UO2–zircaloy system. Different techniques and characters to develop coating materials and enhanced Zr-based alloys are summarized. The features of several selected alternative fuels and cladding materials are reviewed and discussed. The neutronic evaluations of alternative fuel-cladding systems are analyzed. It is expected that one or more types of ATF-cladding systems discussed in the present review will be implemented in commercial reactors.

Innovative accident tolerant fuel concept enabled through direct manufacturing technology

Nuclear energy is one of the largest sources of carbon-free electricity in the world. Some countries are looking at new ways to support and revitalize the nuclear sector since the Fukushima disaster. The accident tolerant fuel program is geared toward improving the safety of nuclear energy by investigating materials that can replace or modify the current uranium-dioxide nuclear fuel and zirconium-based cladding. This research program is being supported by all major nuclear countries since 2011. The practical limitations on allowable uranium enrichment has taken the focus away from the most promising fuels with high radioactivity retention such as tristructural isotropic particle fuel. To overcome such enrichment limitation, a new fuel concept is proposed using advanced ceramic direct manufacturing with laser-induced chemical vapor deposition. The fuel-as-fiber concept is an accident-tolerant fuel design that features high thermal conductivity, strong capability of radioactivity retention and most importantly requires reasonable enrichment levels with uranium nitride as the fuel. In this work, the initial fabrication of uranium-based fuel with laser-induced chemical vapor deposition technology is demonstrated. Then an advanced multi-physics guided modeling approach based on finite element analysis codes and informed by the manufacturing capabilities is developed to accelerate the advancement of fuel-as-fiber concept for use in current light-water reactor technology. The detailed thermomechanical analysis showed promising results for viability of the innovative fuel-as-fiber concept. The predicted stresses in the fuel structural materials were similar to the case of tristructural isotropic particle fuel experience base that has shown excellent reliability in retention of fuel radioactivity at high temperatures.

Modeling a fast spectrum molten salt reactor in a systems dynamics fuel cycles code

Molten salt reactor (MSR) concepts are being actively developed by several private companies, with some touting the fuel cycle benefits of this technology. Simulating the deployment of MSRs and their effect on fuel cycle transitions is the role of a system dynamics fuel cycles tool. However, many current fuel cycles tools have difficulty modeling MSR isotopic variations because they were designed for modeling existing reactors based on discrete fuel elements with batch fueling schemes (i.e., charge and discharge). This work presents ORION fuel cycle modeling results using a specific high power density fast spectrum MSR design. For code verification, a single-stage MSR fuel cycle was set up in ORION and results were compared with the results from the SCALE-based reactor physics model (ChemTriton). A transition scenario was set up in ORION from the current light water reactor fleet to a future fleet of MSRs to study the characteristics of an MSR that affect its performance during transition. Several transition scenarios were set with holds and delays in place that are representative of a sodium fast reactor as modeled in Evaluation Group 23 for the Fuel Cycle Options Campaign Evaluation and Screening Study. Several MSR characteristics lend to its high performance and ease in transition from the current fuel cycle: fissile material is only loaded at initial startup; all fissile material is available as it is bred due to online fuel processing; and fissile material recirculating through the core is briefly held in tanks for processing and has very minimal out-of-core time (on the order of a minutes to a few days). These characteristics are in contrast to a solid-fueled reactor, which requires discharge, cooling, and fabrication of fuel all of which could take a few months to a few years and repeated loading of fresh fuel assemblies. A solid-fueled reactor only discharges a fraction of a core every year.

INTRODUCING PASSIVE NUCLEAR SAFETY IN WATER-COOLED REACTORS - NUMERICAL SIMULATION AND VALIDATION OF NATURAL CONVECTION HEAT TRANSFER AND TRANSPORT IN PACKED BEDS OF HEATED MICROSPHERES

Current light water reactors (LWR) designs have been optimized over decades, and required improvements cannot be based on mere redesign but will require novel technologies. The introduction of an advance fuel in the form of coated micro-fuel particles under Helium atmosphere in the cladding tubes of light water reactor fuel assemblies is a novel technology proposed in this study. This will ensure retention of radioactive fission products by the fuel making LWR attain a passive safety status and among other advantages, higher fuel enrichment in the proposed nuclear fuel will provide higher burn-up. A slender geometrical model with tube-to-particle diameter ratio N = 2.503 and porosity ε = 0.546 mimicking the proposed nuclear fuel in the cladding was numerically simulated using hydrogen as the working fluid. A commercial code, Star CCM+ is used to investigate the heat transfer characteristics and flow distribution under buoyancy driven force expected in the cladding tube of the proposed nuclear fuel. Random packing of the particles is achieved by Discrete Element Method (DEM) simulation. The temperature contour and velocity vector profile obtained can be said to be good illustration of anticipated heat transfer and transport phenomenon to occur in the proposed design. Simulated results for particle-to-fluid heat transfer coefficient, Nusselt number, and Rayleigh number which are of prime importance when analyzing heat transfer performance in fixed bed reactors were validated. Simulated results show close agreement with results obtained in high temperature reactor HTR such as the PBMR.

Estimation of coping time in pressurized water reactors for near term accident tolerant fuel claddings

Abstract The Fukushima Nuclear Power Plant (NPP) accident in Japan has motivated improving the safety of current light water reactors (LWRs). Accident tolerant fuels (ATF) are being developed to enhance the safety of LWRs by tolerating loss of active cooling in the core for a longer duration compared to standard UO2 and Zirconium-based cladding. In this work, the performances of the FeCrAl and Cr-coated ATF claddings under beyond design basis accidents (BDBAs) are modeled with thermal-hydraulics design basis code TRACE. Two models are used for high-temperature oxidation of FeCrAl: a model based on the experimental results of this work, and a model based on experimental results of ORNL’s work. A 3-loop Pressurized Water Reactor model is created and the following BDBAs are simulated for this study: large break loss of coolant accident without safety injection systems, short-term and long term station blackout (SBO) without any mitigation actions. In this work, we define coping time as the initiation time that assumed mitigation actions keep the clad temperature below its melting point. The results showed that ATF claddings increase the coping time and produce significantly less hydrogen compared to Zircaloy cladding under the considered BDBAs scenarios. However, the gains in coping times were found to be marginal.

Long-Term Corrosion Testing of Inconel Alloy 625 in Molten LiCl-Li2O-Li

Introduction In order to incorporate oxide based nuclear fuels from the current light water reactor fleet into a pyrometallurgical based fuel cycle, it is necessary to reduce the oxide fuel to a metallic form. This process is typically conducted electrolytically in a LiCl-Li2O electrolyte at 650°C. Due to the close reduction potentials of UO2 and Li2O, metallic Li is generated at the cathode and subsequently dissolves into the electrolyte, forming a tertiary LiCl-Li2O-Li solution. This study examines the long term corrosion behavior of Inconel 625 when exposed to LiCl-Li2O-Li solutions for 500 and 1000hr. Experimental Similar to our previous work on SS316, exposure testing was performed in LiCl with 1 or 2 wt%Li2O and 0, 0.3, 0.6 and 1.0wt%Li at 650°C for 500 and 1000 hours. I625 samples were polished to a 0.05μm surface finish prior to exposure. The salt solutions were replaced every 96 hours to make up for evaporation and reaction of Li and/or Li2O to ensure consistent solution chemistry. Following exposure, the samples were rinsed with methanol and analyzed using Raman spectroscopy, X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS), scanning electron microscopy (SEM), energy dispersive X-ray spectroscopy (EDS), and micro-Vickers hardness testing. Results The primary corrosion products detected via Raman spectroscopy, XPS, and EDS on samples exposed to LiCl-Li2O solutions in the absence of Li were lithium chromite, nickel-iron-chromium spinel oxides, and lithium chromate. XRD confirmed the presence of lithium chromite on these samples. In the presence of Li, these surface layers were destabilized and the surface of the samples were free of any significant oxide film, while titanium oxynitride films of nm thickness were observed via XPS. The surface morphology of the samples was observed via SEM to undergo increased degradation with higher concentrations of Li. Acknowledgements: This work was performed under the auspices of the Department of Energy (DOE) under contracts DE-NE0008262 and DE-NE0008236, and the US Nuclear Regulatory Commission (NRC) under contracts NRCHQ-11-G-38-0039 and NRC-HQ-13-G-38-0027. W.P. acknowledges the Fellowship Award from the NRC. Dr. Kenny Osborne serves as the program manager for the DOE award and Ms. Nancy Hebron-Isreal serves as the grants program officer for the NRC awards.

The “virtual density” theory of neutronics

Sustainable nuclear energy will likely require fast reactors to complement the current light water reactor paradigm. In particular, breed-and-burn sodium fast reactors (SFRs) offer a unique combination of fuel cycle and power density features. Unfortunately, large breed-and-burn SFRs are plagued by positive sodium void worth. In order to mitigate this drawback, one must quantify various sources of negative reactivity feedback, among which geometry distortions (bowing and flowering of fuel assemblies) are often dominant. These distortions arise mainly from three distinct physical phenomena: irradiation swelling, thermal swelling, and seismic events.Distortions are notoriously difficult to model, because they break symmetry and repeating patterns. Currently, no efficient and fully general method exists for computing neutronic effects of distortions. Computing them directly via diffusion would require construction of exotic meshes that seldom exist within the neutronics community. Many deterministic transport methods are geometrically flexible but tightly constrained by assumptions of symmetry and repeating patterns. Monte Carlo offers the only high-fidelity approach to arbitrary geometry, but resolving minute reactivities and flux shift tallies within large heterogeneous cores requires CPU years per case and is thus prohibitively expensive. Currently, the most widely-used methods consist of various approximations involving weighting the uniform radial swelling reactivity coefficient by the power distribution. These approximations agree fairly well with experimental data for flowering in some cores, but they are not fully general and cannot be trusted to work for arbitrary distortions in generic cores. Boundary perturbation theory, developed in the 1980s, is fully general and mathematically rigorous, but it is inaccurate for coarse mesh diffusion and has never been applied in industry.Our solution is the “virtual density” theory of neutronics, which alters material density (isotropically or anisotropically) instead of explicitly changing geometry. While geometry is discretized, material densities occupy a continuous domain; this allows density changes to obviate the greatest computational challenges of geometry changes. Although primitive forms of this theory exist in Soviet literature, they are only applicable to cases in which entire cores swell uniformly. Thus, we conceive a much more general and pragmatic form of virtual density theory to model non-uniform and localized geometry distortions via perturbation theory. In this work, we develop and implement virtual density theory entirely in the context of diffusion.In order to efficiently validate virtual density perturbation theory, we conceive the “virtual mesh” method for diffusion theory. This new method involves constructing a slightly perturbed “fake” mesh that produces correct first-order reactivity and flux shifts due to anisotropic swelling or expansion of individual mesh cells. First order reactivities computed on a virtual mesh agree with continuous energy Monte Carlo to within 1σ uncertainty.We validate virtual density theory via the virtual mesh method in 3-D coarse mesh models of the Fast Flux Test Facility (FFTF) and Jo¯yo¯ benchmarks using the MATLAB-PETSc-SLEPc (MaPS) multigroup finite difference diffusion code, which we developed for this purpose. We model a panoply of non-uniform anisotropic swelling scenarios, including axial swelling of individual assemblies, axial swelling of each mesh cell in proportion to its fission power, and radial core flowering with arbitrary axial dependence. In 3-D coarse mesh Cartesian cores with explicit coolant gaps, we model individual assembly motion, assembly row motion with arbitrary axial dependence, and assembly row “s-shape” bowing. In all cases, we find that virtual density perturbation theory predicts reactivity coefficients that agree with virtual mesh reference cases to within 0.01%. These reactivity coefficients are two to four orders of magnitude more accurate than those computed via boundary perturbation theory. In general, this virtual density perturbation method can precisely predict reactivity coefficients due to anisotropic swelling or expansion of any core region in any direction.