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Final report on research activities carried out by the PI (Michele Benzi) with his PhD student Massimilano Lupo Pasini and other collaborators under the grant.
Abstract not provided.
To date, inclusion of welding distortion and residual stresses in determining the strength of marine structures is based on empirical formulations due to welding complexity. In this study, a more feasible numerical model was developed from the perspective of inherent strain theory. The accuracy of this simplified numerical model was confirmed by transient thermo-elastic–plastic finite element analysis using a novel dynamic mesh refining and iterative substructure method which is 50 times faster than conventional finite element method. Welding distortion by the two numerical approaches was validated with experimental measurement. The collapse analyses indicated that the inherent strain model predicted a compressive strength and load-displacement curve with high accuracy, as validated by the transient thermo-mechanical model. The global stiffness and the ultimate strength of the welded structure would be overestimated by about 22% if no welding effects were considered. The proposed computation procedure for determining the ultimate strength of welded structures is accurate, simple, and practical and enables improvement of ultimate strength through optimization of welding joint design to reduce the welding residual stress and distortion.
Battery models using robust fail-safe iteration free approach for solving Differential Algebraic Equations, and associated systems and methods are disclosed. In one embodiment, a method includes generating a model of the rechargeable battery; determining one or more initial conditions for one or more algebraic variables of the model using a solver; holding differential variables of the model static by a switch function while determining the one or more initial conditions; applying the initial conditions to the model by the switch function; and determining one or more parameters for the rechargeable battery by solving the algebraic and differential equations.
This paper builds upon previous work to accelerate the Picard iteration (PI) method typically applied for coupled Monte Carlo-Thermal hydraulic (MC-TH) solutions. Previously, the use of the generalized transfer functions (GTFs) to predict variation in macroscopic cross sections following a perturbation in TH properties was demonstrated for a subset of simple 3D problems. In addition, the reduced-order transport prediction block relied on the first order perturbation (FOP) method, which was shown to have computational overheads. Recent work replaced the FOP block with a 1-group nodal diffusion solver to eliminate these overheads. While the use of diffusion is desirable for large-scale problems, the new solver introduces significant homogenization error. This work aims to address this issue by using the Jacobian-Free Newton Krylov (JFNK) method to generate a set of super homogenization (SPH) factors to improve the accuracy of the diffusion solution. The SPH factors will be used in conjunction with an improved cross section prediction method – the expanded transfer function (ETF) method – to produce a highly accurate flux prediction for an axial 1D boiling water reactor (BWR) pincell following a large perturbation in moderator density. The ETF-corrected diffusion (ETF-CD) block is shown to be highly accurate for the 1D test case. Future work will investigate the accuracy of the method for a realistic 3D pressurized water reactor core.
A simple nonlinear Reduced Order Model to study global, regional and local instabilities in Boiling Water Reactors is described. The ROM consists of three submodels: neutron-kinetic, thermal-hydraulic and heat-transfer models. The neutron-kinetic model allows representing the time evolution of the three first neutron kinetic modes: the fundamental, the first and the second azimuthal modes. The thermal-hydraulic model describes four heated channels in order to correctly simulate out-of-phase behavior. The coupling between the different submodels is performed via both void and Doppler feedback mechanisms. After proper spatial homogenization, the governing equations are discretized in the time-domain. Several modifications, compared to other existing ROMs, have been implemented, and are reported in this paper. One novelty of the ROM is the inclusion of both azimuthal modes, which allows to study combined instabilities (in-phase and out-of-phase), as well as to investigate the corresponding interference effects between them. The second modification concerns the precise estimation of so-called reactivity coefficients or C{sub mn}{sup *V,D} - coefficients by using direct cross-section data from SIMULATE-3 combined with the CORE SIM core simulator in order to calculate Eigenmodes. Furthermore, a non-uniform two-step axial power profile is introduced to simulate the separate heat production in the single and two-phase regions, respectively. An iterative procedure was developed to calculate the solution to the coupled neutron-kinetic/thermal-hydraulic static problem prior to solving the time-dependent problem. Besides, the possibility of taking into account the effect of local instabilities is demonstrated in a simplified manner. The present ROM is applied to the investigation of an actual instability that occurred at the Swedish Forsmark-1 BWR in 1996/1997. The results generated by the ROM are compared with real power plant measurements performed during stability tests and show a good qualitative agreement. The present study provides some insight in a deeper understanding of the physical principles which drive both core-wide and local instabilities. (authors)
Recently as part of the CASL project, a capability to perform 3D whole-core coupled neutron transport and computational fluid dynamics (CFD) calculations was demonstrated. This work uses the 2D/1D transport code DeCART and the commercial CFD code STAR-CCM+. It builds on previous CASL work demonstrating coupling for smaller spatial domains. The coupling methodology is described along with the problem simulated and results are presented for fresh hot full power conditions. An additional comparison is made to an equivalent model that uses lower order T/H feedback to assess the importance and cost of high fidelity feedback to the neutronics problem. A simulation of a quarter core Combustion Engineering (CE) PWR core was performed with the coupled codes using a Fixed Point Gauss-Seidel iteration technique. The total approximate calculation requirements are nearly 10,000 CPU hours and 1 TB of memory. The problem took 6 coupled iterations to converge. The CFD coupled model and low order T/H feedback model compared well for global solution parameters, with a difference in the critical boron concentration and average outlet temperature of 14 ppm B and 0.94 deg. C, respectively. Differences in the power distribution were more significant with maximum relative differences in the core-wide pin peaking factor (Fq) of 5.37% and average relative differences in flat flux region power of 11.54%. Future work will focus on analyzing problems more relevant to CASL using models with less approximations. (authors)
The current tendency in multiphysics calculations applied to reactor physics is the use of already validated computer codes, coupled by means of an iterative approach. In this paper such an approach is explained concerning neutronics and thermal-hydraulics coupled analysis with MCNPX and COBRA-IV codes using a driver program and file exchange between codes. MCNPX provides the neutronic analysis of heterogeneous nuclear systems, both in critical and subcritical states, while COBRA-IV is a subchannel code that can be used for rod bundles or core thermal-hydraulics analysis. In our model, the MCNP temperature dependence of nuclear data is handled via pseudo-material approach, mixing pre-generated cross section data set to obtain the material with the desired cross section temperature. On the other hand, COBRA-IV has been updated to allow for the simulation of liquid metal cooled reactors. The coupled computational tool can be applied to any geometry and coolant, as it is the case of single fuel assembly, at pin-by-pin level, or full core simulation with the average pin of each fuel-assembly. The coupling tool has been applied to the critical core layout of the SCK-CEN MYRRHA concept, an experimental LBE cooled fast reactor presently in engineering design stage. (authors)
Perturbation theory based sensitivity analysis is a vital part of todays' nuclear reactor design. This paper presents an extension of standard techniques to examine coupled criticality problems with mutual feedback between neutronics and an augmenting system (for example thermal-hydraulics). The proposed procedure uses a neutronic and an augmenting adjoint function to efficiently calculate the first order change in responses of interest due to variations of the parameters describing the coupled problem. The effect of the perturbations is considered in two different ways in our study: either a change is allowed in the power level while maintaining criticality (power perturbation) or a change is allowed in the eigenvalue while the power is constrained (eigenvalue perturbation). The calculated response can be the change in the power level, the reactivity worth of the perturbation, or the change in any functional of the flux, the augmenting dependent variables and the input parameters. To obtain power- and criticality-constrained sensitivities power- and k-reset procedures can be applied yielding identical results. Both the theoretical background and an application to a one dimensional slab problem are presented, along with an iterative procedure to compute the necessary adjoint functions using the neutronics and the augmenting codes separately, thus eliminating the need of developing new programs to solve the coupled adjoint problem. (authors)
This paper presents the design process of collaborative tools, based on ICT, aiming at supporting the tasks of the team that manages an outage of an energy production plant for maintenance activities. The design process follows an iterative and multidisciplinary approach, based on a collective tasks modeling of the outage management team in the light of Socio Organizational and Human (SOH) field studies, and on the state of the art of ICT. Field test of the collaborative tools designed plays a great place in this approach, allowing taking into account the operational world but involves also some risks which must be managed. To implement tools on all the production plants, we build an 'operational concept' with a level of description which authorizes the evolution of tools and allows some local adaptations. The field tests provide lessons on the ICT topics. For examples: the status of the remote access tools, the potential of use of a given information input by an actor for several individual and collective purposes, the actors perception of the tools meaning, and the requirements for supporting the implementation of change. (authors)
The behavior of nuclear fuel in the reactor environment is affected by multiple physics, most notably heat conduction and solid mechanics, which can have a strong influence on each other. To provide credible solutions, a fuel performance simulation code must have the ability to obtain solutions for each of the physics, including coupling between them. Solution strategies for solving systems of coupled equations can be categorized as loosely-coupled, where the individual physics are solved separately, keeping the solutions for the other physics fixed at each iteration, or tightly coupled, where the nonlinear solver simultaneously drives down the residual for each physics, taking into account the coupling between the physics in each nonlinear iteration. In this paper, we compare the performance of loosely and tightly coupled solution algorithms for thermomechanical problems involving coupled thermal and mechanical contact, which is a primary source of interdependence between thermal and mechanical solutions in fuel performance models. The results indicate that loosely-coupled simulations require significantly more nonlinear iterations, and may lead to convergence trouble when the thermal conductivity of the gap is too small. We also apply the tightly coupled solution strategy to a nuclear fuel simulation of an experiment in a test reactor. Studying the results from these simulations indicates that perhaps convergence for either approach may be problem dependent, i.e., there may be problems for which a loose coupled approach converges, where tightly coupled won't converge and vice versa. (authors)
The development of a fine-mesh coupled neutronic/thermal-hydraulic solver is touched upon in this paper. The reported work investigates the feasibility of using finite volume techniques to discretize a set of conservation equations modeling neutron transport, fluid dynamics, and heat transfer within a single numerical tool. With the long-term objective of developing fine-mesh computing capabilities for a few selected fuel assemblies in a nuclear core, this preliminary study considers an infinite array of a single fuel assembly having a finite height. Thermal-hydraulic conditions close to the ones existing in PWRs are taken as a first test case. The neutronic modeling relies on the diffusion approximation in a multi-energy group formalism, with cross-sections pre-calculated and tabulated at the sub-pin level using a Monte Carlo technique. The thermal-hydraulics is based on the Navier-Stokes equations, complemented by an energy conservation equation. The non-linear coupling terms between the different conservation equations are fully resolved using classical iteration techniques. Early tests demonstrate that the numerical tool provides an unprecedented level of details of the coupled solution estimated within the same numerical tool and thus avoiding any external data transfer, using fully consistent models between the neutronics and the thermal-hydraulics. (authors)
We define a new indirect measure of the diffusion model error called the diffusion model error source. When this model error source is added to the diffusion equation, the transport solution for the angular-integrated intensity is obtained. This source represents a means by which a transport code can be used to generate information relating to the adequacy of diffusion theory for any given problem without actually solving the diffusion equation. The generation of this source does not relate in any way to acceleration of the iterative convergence of transport solutions. Perhaps the most well-known indirect measure of the diffusion model error is the variable-Eddington tensor. This tensor provides a great deal of information about the angular dependence of the angular intensity solution, but it is not always simple to interpret. In contrast, our diffusion model error source is a scalar that is conceptually easy to understand. In addition to defining the diffusion model error source analytically, we show how to generate this source numerically relative to the S{sub n} radiative transfer equations with linear-discontinuous spatial discretization. This numerical source is computationally tested and shown to reproduce the Sn solution for a Marshak-wave problem. (authors)
Due to the early decommissioning of four Water-Water Energy Reactors (WWER) 440-V230 reactors at the Nuclear Power Plant (NPP) near the city of Kozloduy in Bulgaria, large amounts of low and intermediate radioactive waste will arise much earlier than initially scheduled. In or-der to manage the radioactive waste from the early decommissioning, Bulgaria has intensified its efforts to provide a near surface disposal facility at Radiana with the required capacity. To this end, a project was launched and assigned in international competition to a German-Spanish consortium to provide the complete technical planning including the preparation of the Intermediate Safety Assessment Report. Preliminary results of operational and long-term safety show compliance with the Bulgarian regulatory requirements. The long-term calculations carried out for the Radiana site are also a good example of how analysis of safety assessment results can be used for iterative improvements of the assessment by pointing out uncertainties and areas of future investigations to reduce such uncertainties in regard to the potential radiological impact. The computer model used to estimate the long-term evolution of the future repository at Radiana predicted a maximum total annual dose for members of the critical group, which is carried to approximately 80 % by C-14 for a specific ingestion pathway. Based on this result and the outcome of the sensitivity analysis, existing uncertainties were evaluated and areas for reasonable future investigations to reduce these uncertainties were identified. (authors)
Performance assessments (PAs) have been used for many years for the analysis of post-closure hazards associated with a radioactive waste disposal facility and to provide a reasonable expectation of the ability of the site and facility design to meet objectives for the protection of members of the public and the environment. The use of PA to support decision-making for LLW disposal facilities has been mandated in United States Department of Energy (US DOE) directives governing radioactive waste management since 1988 (currently DOE Order 435.1, Radioactive Waste Management). Prior to that time, PAs were also used in a less formal role. Over the past 20+ years, the US DOE approach to conduct, review and apply PAs has evolved into an efficient, rigorous and mature process that includes specific requirements for continuous improvement and independent reviews. The PA process has evolved through refinement of a graded and iterative approach designed to help focus efforts on those aspects of the problem expected to have the greatest influence on the decision being made. Many of the evolutionary changes to the PA process are linked to the refinement of the PA maintenance concept that has proven to be an important element of US DOE PA requirements in the context of supporting decision-making for safe disposal of LLW. The PA maintenance concept is central to the evolution of the graded and iterative philosophy and has helped to drive the evolution of PAs from a deterministic compliance calculation into a systematic approach that helps to focus on critical aspects of the disposal system in a manner designed to provide a more informed basis for decision-making throughout the life of a disposal facility (e.g., monitoring, research and testing, waste acceptance criteria, design improvements, data collection, model refinements). A significant evolution in PA modeling has been associated with improved use of uncertainty and sensitivity analysis techniques to support efficient implementation of the graded and iterative approach. Rather than attempt to exactly predict the migration of radionuclides in a disposal unit, the best PAs have evolved into tools that provide a range of results to guide decision-makers in planning the most efficient, cost effective, and safe disposal of radionuclides. (authors)
The sheer mass and nature of contaminated materials at DOE and DoD sites, makes it impractical to completely restore these sites to pre-disposal conditions. DOE faces long-term challenges, particularly with developing monitoring and end state approaches for clean-up that are protective of the environment, technically based and documented, sustainable, and most importantly cost effective. Integrated systems-based monitoring approaches (e.g., tools for characterization and monitoring, multi-component strategies, geophysical modeling) could provide novel approaches and a framework to (a) define risk-informed endpoints and/or conditions that constitute completion of cleanup and (b) provide the understanding for implementation of advanced scientific approaches to meet cleanup goals. Multi-component strategies which combine site conceptual models, biological, chemical, and physical remediation strategies, as well as iterative review and optimization have proven successful at several DOE sites. Novel tools such as enzyme probes and quantitative PCR for DNA and RNA, and innovative modeling approaches for complex subsurface environments, have been successful at facilitating the reduced operation or shutdown of pump and treat facilities and transition of clean-up activities into monitored natural attenuation remedies. Integrating novel tools with site conceptual models and other lines of evidence to characterize, optimize, and monitor long term remedial approaches for complex contaminant plumes are critical for transitioning active remediation into cost effective, yet technically defensible endpoint strategies. (authors)
The sustainability criterion for Gen IV nuclear energy systems inherently presumes the availability of efficient fuel recycling capabilities. One area for research on advanced fuel recycling concerns safeguards aspects of this type of facilities. Since a recycling facility may be considered as sensitive from a non-proliferation perspective, it is important to address these issues early in the design process, according to the principle of Safeguards By Design. Presented in this paper is a mode of procedure, where assessments of the proliferation resistance (PR) of a recycling facility for fast reactor fuel have been performed so as to identify the weakest barriers to proliferation of nuclear material. Two supplementing established methodologies have been applied; TOPS (Technological Opportunities to increase Proliferation resistance of nuclear power Systems) and PR-PP (Proliferation Resistance and Physical Protection evaluation methodology). The chosen fuel recycling facility belongs to a small Gen IV lead-cooled fast reactor system that is under study in Sweden. A schematic design of the recycling facility, where actinides are separated using solvent extraction, has been examined. The PR assessment methodologies make it possible to pinpoint areas in which the facility can be improved in order to reduce the risk of diversion. The initial facility design may then be slightly modified and/or safeguards measures may be introduced to reduce the total identified proliferation risk. After each modification of design and/or safeguards implementation, a new PR assessment of the revised system can then be carried out. This way, each modification can be evaluated and new ways to further enhance the proliferation resistance can be identified. This type of iterative procedure may support Safeguards By Design in the planning of new recycling plants and other nuclear facilities. (authors)
The Former Guterl Specialty Steel Corporation Site (Guterl Site) is located 32 kilometers (20 miles) northeast of Buffalo, New York, in Lockport, Niagara County, New York. Between 1948 and 1952, up to 15,875 metric tons (35 million pounds) of natural uranium metal (U) were processed at the Guterl Site. The resulting milling dust, shavings, thermal scale, and associated on-site land disposal contaminated both the facility and soils. These combined manufacturing and disposal processes promoted the development of a large-scale uranium plume in the groundwater underlying the site. Site soils are anthropogenic fill and re-worked, glacially-derived sediments that vary between 0.6 meters (m) and 3 meters (2.0 to 10.0 feet [ft]) in thickness. These soils blanket a highly permeable weathered and fractured zone in the upper Lockport Dolostone bedrock. Aerial recharge through the thin soils flushes soluble constituents to groundwater, as exemplified by variations in specific conductivity and groundwater levels in the bedrock during precipitation events. Seasonally higher groundwater levels also promote contact with uranium contamination in deeper soil, thus complicating the transport characteristics of the site (e.g., oxidizing metallic uranium in the soil that increases leach rates). The Seasonal Soil (SESOIL) compartment model was used to simulate a series of site soil conditions and predict the leaching of uranium to groundwater. The SESOIL input included physical soil characteristics, uranium profiles common in five soil areas, climate data, and an assumed oxidation-state (or chemical valence) for metallic uranium in the soil. The iterative modeling process evaluated site-specific variables to achieve a final deterministic set of input that was applied to the five soil areas. The leaching simulations estimated time-dependent influx rates for uranium, which were input to a numerical groundwater flow (MODFLOW) and transport (MT3DMS) model. The combined modeling effort produced spatially and temporally variable transport conditions that were simulated over a 1,000-year period. The leaching models predicted a preliminary remediation goal (PRG) for uranium of 11 milligrams per kilogram (mg/kg) that is protective of groundwater. However, this low uranium PRG has field implementation concerns, so a modified direct-exposure PRG of 70 mg/kg was developed and evaluated in the flow and transport model. The primary difference in the PRGs is the longevity and location of the uranium plume, which will affect the remedial costs and potential stakeholder acceptance. (authors)