@article{mikkelson_doster_2022, title={Investigation of two concrete thermal energy storage system configurations for continuous power production}, volume={51}, ISSN={["2352-152X"]}, DOI={10.1016/j.est.2022.104387}, abstractNote={Two Modelica concrete thermal energy storage (CTES) models are built to analyze potential CTES system designs. The first design is the single-pipe network design wherein a heat transfer fluid (HTF) flows in one direction during heat deposition and the opposite direction during heat removal. All pipes in the network are used for either deposition (charging) or removal (discharging). The second design is a dual-pipe network design in which a network of HTF pipes carries charging fluid and a separate piping network carries discharging HTF. This paper evaluates the operation of these two configurations to produce constant steam. Results indicate that the designs are likely appropriate for different applications: the single-pipe network appears appropriate for batch energy applications, and the dual-pipe network is appropriate for continuous energy applications.}, journal={JOURNAL OF ENERGY STORAGE}, author={Mikkelson, Daniel and Doster, J. Michael}, year={2022}, month={Jul} }
@article{mikkelson_frick_bragg-sitton_doster_2021, title={Phenomenon Identification and Ranking Table Development for Future Application Figure-of-Merit Studies on Thermal Energy Storage Integrations with Light Water Reactors}, ISSN={["1943-7471"]}, DOI={10.1080/00295450.2021.1906473}, abstractNote={Abstract There are no standard prioritization criteria for evaluating thermal energy storage (TES) options for use in integrated energy systems. A framework for proposing, analyzing, and presenting energy storage integration with power producers and users is presented along with a specific figure-of-merit (FOM) study based in this framework. This basis for evaluating storage technologies can provide a structure for the energy industry to analyze and prioritize energy storage in different applications and environments. The phenomena identification and ranking table (PIRT) presents a series of design questions specific to energy storage applications. The FOM study, built in this PIRT framework based on a nuclear-renewable hybrid energy system using TES to produce power and provide process energy for a secondary user, successfully identified specific technologies to use based on the project requirements. Expanding the library of projects using this framework will expand the deployable options for energy storage and increase its potential for energy security.}, journal={NUCLEAR TECHNOLOGY}, author={Mikkelson, Daniel and Frick, Konor and Bragg-Sitton, Shannon and Doster, J. Michael}, year={2021}, month={Aug} }
@article{frick_doster_bragg-sitton_2019, title={Design and Operation of a Sensible Heat Peaking Unit for Small Modular Reactors}, volume={205}, ISSN={["1943-7471"]}, DOI={10.1080/00295450.2018.1491181}, abstractNote={Abstract Approximately 19% of the electricity produced in the United States comes from nuclear power plants. Traditionally, nuclear power plants, as well as larger coal-fired plants, operate in a baseload manner at or near steady state for prolonged periods of time. Smaller, more maneuverable plants, such as gas-fired plants, are dispatched to match electricity supply and demand above the capacity of the baseload plants. However, air quality concerns and CO2 emission standards have made the burning of fossil fuels less desirable, despite the current low cost of natural gas. Wind and solar photovoltaic power generation are attractive options due to their lack of carbon footprint and falling capital costs. Yet, these renewable energy sources suffer from inherent intermittency. This inherent intermittency can strain electric grids, forcing carbon-based and nuclear sources of energy to operate in a load-follow mode. For nuclear reactors, load-follow operation can be undesirable due to the associated thermal and mechanical stresses placed on the fuel and other reactor components. Various methods of thermal energy storage (TES) can be coupled to nuclear (or renewable) power sources to help absorb grid variability caused by daily load demand changes and renewable intermittency. Our previous research has shown that coupling a sensible heat TES system to a small modular reactor allows the reactor to run at effectively nominal full power during periods of variable electric demand by bypassing steam to the TES system during periods of excess capacity. In this paper we demonstrate that this stored thermal energy can be recovered, allowing the TES system to act as a peaking unit during periods of high electric demand or used to produce steam for ancillary applications such as desalination. For both applications the reactor is capable of operating continuously at approximately 100% power.}, number={3}, journal={NUCLEAR TECHNOLOGY}, author={Frick, Konor and Doster, J. Michael and Bragg-Sitton, Shannon}, year={2019}, pages={415–441} }
@article{frick_misenheimer_doster_terry_bragg-sitton_2018, title={Thermal Energy Storage Configurations for Small Modular Reactor Load Shedding}, volume={202}, ISSN={["1943-7471"]}, DOI={10.1080/00295450.2017.1420945}, abstractNote={Abstract The increased penetration of intermittent renewable energy technologies such as wind and solar power can strain electric grids, forcing carbon-based and nuclear sources of energy to operate in a load-follow mode. For nuclear reactors, load-follow operation can be undesirable due to the associated thermal and mechanical stresses placed on the fuel and other reactor components. Various methods of thermal energy storage (TES) can be coupled to nuclear (or renewable) power sources to help absorb grid variability caused by daily load demand changes and renewable intermittency. Two TES techniques are investigated as candidate thermal reservoirs to be used in conjunction with a small modular reactor (SMR): a two-tank sensible heat storage system and a stratified chilled-water storage system. The goal when coupling the two systems to the SMR is to match turbine output and demand and bypass steam to the TES systems to maintain reactor power at approximately 100%. Simulations of integral pressurized water reactor dynamics are run in a high-fidelity FORTRAN model developed at North Carolina State University. Both TES systems are developed as callable FORTRAN subroutines to model the time-varying behavior associated with different configurations of these systems when connected to the SMR simulator. Simulation results reveal the sensible heat storage system is capable of meeting turbine demand and maintaining reactor power constant while providing enough steam to power four absorption chillers for chilled-water production and storage. The stored chilled water is used to supplement cooling loads of an adjacent facility.}, number={1}, journal={NUCLEAR TECHNOLOGY}, author={Frick, Konor and Misenheimer, Corey T. and Doster, J. Michael and Terry, Stephen D. and Bragg-Sitton, Shannon}, year={2018}, pages={53–70} }
@article{o'brien_doster_nortier_olivas_stokely_2018, title={Two-way multi-physics coupling for modeling high power RbCl isotope production targets}, volume={433}, ISSN={["1872-9584"]}, DOI={10.1016/j.nimb.2018.07.022}, abstractNote={This work shows successful first application of two-way multi-physics coupling to model RbCl targets in a three-stacked target configuration used at Los Alamos National Laboratory’s (LANL) Isotope Production Facility (IPF). Targets are known to melt at production level beam currents and as in-beam monitoring of the targets in this configuration is not possible, high-fidelity simulation has been utilized to gain insight into target thermal behavior. Thermal hydraulic modeling was performed with ANSYS CFX and particle transport with the Monte Carlo N-Particle (MCNP) code. Multi-physics coupling of these two codes was employed to fully capture the highly coupled nature of the problem physics. Both transient and equilibrium thermal hydraulic results were obtained using this process. The equilibrium thermal hydraulic results were then employed to predict measured 82Sr yields in molten RbCl targets. This technique demonstrates promise as a tool to investigating, understanding, and enhancing high power targetry behavior and limitations.}, journal={NUCLEAR INSTRUMENTS & METHODS IN PHYSICS RESEARCH SECTION B-BEAM INTERACTIONS WITH MATERIALS AND ATOMS}, author={O'Brien, E. M. and Doster, J. M. and Nortier, F. M. and Olivas, E. R. and Stokely, M. H.}, year={2018}, month={Oct}, pages={15–22} }
@article{peeples_magerl_o'brien_doster_bolotnov_wieland_stokely_2017, title={High Current C-11 Gas Target Design and Optimization Using Multi-Physics Coupling}, volume={1845}, ISSN={["0094-243X"]}, DOI={10.1063/1.4983547}, abstractNote={A high current conical C-11 gas target with a well characterized production yield was designed and optimized using multi-physics coupling simulations. Two target prototypes were deployed on an IBA 18/9 cyclotron, and the experimental results were used to benchmark the predictive simulations.}, journal={WTTC16: PROCEEDINGS OF THE 16TH INTERNATIONAL WORKSHOP ON TARGETRY AND TARGET CHEMISTRY}, author={Peeples, J. L. and Magerl, M. and O'Brien, E. M. and Doster, J. M. and Bolotnov, I. A. and Wieland, B. W. and Stokely, M. H.}, year={2017} }
@article{heo_turinsky_doster_2013, title={Optimization of Thermal-Hydraulic Reactor System for SMRs via Data Assimilation and Uncertainty Quantification}, volume={173}, ISSN={["1943-748X"]}, DOI={10.13182/nse11-113}, abstractNote={Abstract This paper discusses the utilization of an uncertainty quantification methodology for nuclear power plant thermal-hydraulic transient predictions, with a focus on small modular reactors characterized by the integral pressurized water reactor design, to determine the value of completing experiments in reducing uncertainty. To accomplish this via the improvement of the prediction of key system attributes, e.g., minimum departure from nucleate boiling ratio, a thermal-hydraulic simulator is used to complete data assimilation for input parameters to the simulator employing experimental data generated by the virtual reactor. The mathematical approach that is used to complete this analysis depends upon whether the system responses, i.e., sensor signals, and the system attributes are or are not linearly dependent upon the parameters. For a transient producing mildly nonlinear response sensitivities, a Bayesian-type approach was used to obtain the a posteriori distributions of the parameters assuming Gaussian distributions for the input parameters and responses. For a transient producing highly nonlinear response sensitivities, the Markov chain Monte Carlo method was utilized based upon Bayes’ theorem to estimate the a posteriori distributions of the parameters. To evaluate the value of completing experiments, an optimization problem was formulated and solved. The optimization addressed both the experiments to complete and the modifications to be made to the nuclear power plant made possible by using the increased margins resulting from data assimilation. The decision variables of the experiment optimization problem include the selection of sensor types and locations and experiment type imposing realistic constraints. The decision variables of the nuclear power plant modification optimization problem include various design specifications, e.g., power rating, steam generator size, and reactor coolant pump size, with the objective of minimizing cost as constrained by required margins to accommodate the uncertainty. Since the magnitude of the uncertainty is dependent upon the experiments via data assimilation, the nuclear power plant optimization problem is treated as a suboptimization problem within the experiment optimization problem. The experiment optimization problem objective is to maximize the net savings, defined as the savings in nuclear power plant cost due to the modified design specifications minus the cost of the experiments. Both the experiment and the nuclear power plant optimization problems were solved using the simulated annealing method.}, number={3}, journal={NUCLEAR SCIENCE AND ENGINEERING}, author={Heo, Jaeseok and Turinsky, Paul J. and Doster, J. Michael}, year={2013}, month={Mar}, pages={293–311} }
@article{proctor_doster_2012, title={Reactor loose part damage assessments on steam generator tube sheets}, volume={179}, DOI={10.13182/nt12-a14167}, abstractNote={Local space and energy impact densities of various types of loose parts have been generated within a representative steam generator inlet plenum. This work expands upon previous experimental research to identify important mechanisms that govern accumulated loose part damage to steam generator tube sheets. As a result, a computational model for estimating loose part impact damage, including damage to steam generator tube ends from multiple impacts, was previously created. Damage effects were determined to be local effects that depended only on single impacts and impact overlaps in a small region of interest. It was found that the damage could be directly related to local impact density on the steam generator tube sheet. In this work, three-dimensional flow fields were generated, first for a previously used 1:8 scale experimental inlet plenum and then for a 1:1 scale Westinghouse type D steam generator. Monte Carlo simulations were carried out as a function of coolant temperature, coolant inlet velocity, loose part type, shape, mass, density, initial starting location, and initial kinetic energy. No a priori knowledge was assumed for the initial starting location and initial kinetic energy of the parts. Comparisons were performed between previous scaled experimental results and scaled computational simulation results to assess the validity of predictions from the scaled simulation. Combined, both this work and previous work could allow for the assessment of impact damage rates on steam generator tube sheets via simulation. The most-energetic impacts are not localized to any particular region on the tube sheet. The general progression of the spatial distribution of all impact locations as a function of initial kinetic energy accurately depicts the progression for the highest-energy impacts. As the initial kinetic energy increases and as the starting location moves toward the inlet plenum, there is an increase in the number of higher-energy impacts. The higher-initial kinetic energy impacts lead to higher-energy first impacts on the tube sheet. Beyond the first impact, the energy distribution is invariant to initial kinetic energy and initial start location. The invariance seen in the energy distribution does not hold the same for the spatial distribution. The effects of the initial kinetic energy and initial start location ripple into the second and third impacts. Beyond the third impacts little to no change can be discerned and the invariance due to initial kinetic energy and initial start locations is valid. Ultimately, with these types of analyses, reactor facilities will be able to better judge whether a system necessarily needs to shut down due to safety concerns about loose parts damage before a scheduled outage.}, number={3}, journal={Nuclear Technology}, author={Proctor, W. C. and Doster, J. M.}, year={2012}, pages={339–359} }
@article{peeples_stokely_michael doster_2011, title={Thermal performance of batch boiling water targets for 18F production}, volume={69}, ISSN={0969-8043}, url={http://dx.doi.org/10.1016/j.apradiso.2011.06.015}, DOI={10.1016/j.apradiso.2011.06.015}, abstractNote={Batch boiling targets are commonly used in cyclotrons to produce Fluorine-18 by proton bombardment of Oxygen-18 enriched water. Computational models have been developed to predict the thermal performance of bottom-pressurized batch boiling production targets. The models have been validated with experimental test data from the Duke University Medical Cyclotron and the Wisconsin Medical Cyclotron. Good agreement has been observed between experimental measurements and model predictions of average target vapor fraction as a function of beam current and energy.}, number={10}, journal={Applied Radiation and Isotopes}, publisher={Elsevier BV}, author={Peeples, Johanna L. and Stokely, Matthew H. and Michael Doster, J.}, year={2011}, month={Oct}, pages={1349–1354} }
@article{shen_doster_2009, title={Application of a neural network based feedwater controller to helical steam generators}, volume={239}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2009.02.011}, abstractNote={In current generation pressurized water reactors (PWRs), the control of steam generator level experiences challenges over the full range of plant operating conditions. These challenges can be particularly troublesome in the low power range where the feedwater is highly subcooled and minor changes in the feed flow may cause oscillations in the SG level, potentially leading to reactor trip. The IRIS reactor concept adds additional challenges to the feedwater control problem as a result of a steam generator design where neither level or steam generator mass inventory can be measured directly. Neural networks have demonstrated capabilities to capture a wide range of dynamic signal transformation and non-linear problems. In this paper a detailed engineering simulation of plant response is used to develop and test neural control methods for the IRIS feedwater control problem. The established neural network mass estimator has demonstrated the capability to predict the steam generator mass under transient conditions, especially at low power levels, which is considered the most challenging region for a full range feed water controller.}, number={6}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Shen, Hengliang and Doster, J. Michael}, year={2009}, month={Jun}, pages={1056–1065} }
@article{dong_doster_mayo_2008, title={Steam generator control in nuclear power plants by water mass inventory}, volume={238}, ISSN={["0029-5493"]}, DOI={10.1016/j.nucengdes.2007.09.001}, abstractNote={Control of water mass inventory in Nuclear Steam Generators is important to insure sufficient cooling of the nuclear reactor. Since downcomer water level is measurable, and a reasonable indication of water mass inventory near steady-state, conventional feedwater control system designs attempt to maintain downcomer water level within a relatively narrow operational band. However, downcomer water level can temporarily react in a reverse manner to water mass inventory changes, commonly known as shrink and swell effects. These complications are accentuated during start-up or low power conditions. As a result, automatic or manual control of water level is difficult and can lead to high reactor trip rates. This paper introduces a new feedwater control strategy for Nuclear Steam Generators. The new method directly controls water mass inventory instead of downcomer water level, eliminating complications from shrink and swell all together. However, water mass inventory is not measurable, requiring an online estimator to provide a mass inventory signal based on measurable plant parameters. Since the thermal-hydraulic response of a Steam Generator is highly nonlinear, a linear state-observer is not feasible. In addition, difficulties in obtaining flow regime and density information within the Steam Generator make an estimator based on analytical methods impractical at this time. This work employs a water mass estimator based on feedforward neural networks. By properly choosing and training the neural network, mass signals can be obtained which are suitable for stable, closed-loop water mass inventory control. Theoretical analysis and simulation results show that water mass control can significantly improve the operation and safety of Nuclear Steam Generators.}, number={4}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Dong, Wei and Doster, J. Michael and Mayo, Charles W.}, year={2008}, month={Apr}, pages={859–871} }
@article{kim_bourham_doster_2006, title={A wide-beam X-ray source suitable for diffraction enhanced imaging applications}, volume={566}, ISSN={["0168-9002"]}, DOI={10.1016/j.nima.2006.07.041}, abstractNote={Abstract Research in diffraction-enhanced imaging (DEI), using a synchrotron source with an X-ray flux of 1.4×1012 ph/mm2/s, has shown strong potential in obtaining high-resolution images as compared to conventional radiographs. This research investigates the feasibility of developing a large area circular X-ray source with fluxes comparable to a synchrotron source. The source should be capable of integration into a compact system with peak powers not to exceed 200 kW to be feasible for use in a major medical facility, industrial complex or screening facility (such as cargo or airport). A computational study of a circular concentric filament wide-beam area X-ray source has been investigated in this research. The design features are based on generating electrons from three concentric circular filaments to provide an area electron flux, with a 60 kV accelerating potential and a beam current of up to 3 A. The X-ray target is a grounded stationary oxygen-free copper target with a layer of molybdenum. This target feature differs from standard rotating X-ray targets in conventional X-ray systems. Studies of electron trajectories and their distribution on the target were conducted using the SIMION 3D code. Heat loading and thermal management were studied using heat transfer modules from the coupled FEMLAB multi-physics and MATLAB codes. The Monte Carlo code MCNP 5 was used to obtain the X-ray flux and energy distribution for aluminum and beryllium windows. This computational study shows that this target configuration generates X-rays with photon flux comparable to synchrotron source and sufficient for DEI applications. The maximum target temperature rise is 1357 K after 70 s when cooling the back of the target to liquid nitrogen temperature using cold finger contact, and 325 K for an invaded target, in which liquid nitrogen circulates inside the target.}, number={2}, journal={NUCLEAR INSTRUMENTS & METHODS IN PHYSICS RESEARCH SECTION A-ACCELERATORS SPECTROMETERS DETECTORS AND ASSOCIATED EQUIPMENT}, author={Kim, Chang H. and Bourham, Mohamed A. and Doster, J. Michael}, year={2006}, month={Oct}, pages={713–721} }
@article{zaghloul_bourham_doster_2001, title={Semi-analytical modelling and simulation of the evolution and flow of ohmically-heated non-ideal plasmas in electrothermal guns}, volume={34}, ISSN={["0022-3727"]}, DOI={10.1088/0022-3727/34/5/316}, abstractNote={In this paper, the Ohmic power input to the discharge capillary is recovered and used to analyse the basic processes involved in electrothermal (ET) plasma devices operated in an ablation-controlled-arc regime. Such an interplay between theory and experiment is necessary to reduce the number of reasons which might be responsible for the reported discrepancies between theory and experiment, as well as to illuminate the subject of ablation controlled arcs. A consistent methodology for determining detailed composition and thermodynamic functions of the non-ideal plasma generated in such devices is presented and used in the present computations. Different non-ideality effects, due to Debye-Hückel corrections in the Gibbs free energy, which have been ignored in prior publications, have been taken into account. A semi-analytical model for an ET plasma source with non-ideal effects is described and incorporated into a comprehensive computer code to simulate plasma evolution and flow in the discharge capillary. The model is one-dimensional, time dependent and uses the recovered Ohmic power input in the source term of the energy equation. The developed code has been used to investigate the ablation process and has shown the inappropriateness of a widely used ablation model. Code predictions for different plasma parameters are presented, discussed, and compared to available experimental data.}, number={5}, journal={JOURNAL OF PHYSICS D-APPLIED PHYSICS}, author={Zaghloul, MR and Bourham, MA and Doster, JM}, year={2001}, month={Mar}, pages={772–786} }
@article{zaghloul_bourham_doster_2000, title={A simple formulation and solution strategy of the Saha equation for ideal and nonideal plasmas}, volume={33}, ISSN={["0022-3727"]}, DOI={10.1088/0022-3727/33/8/314}, abstractNote={A simple formulation and solution strategy for the Saha equation is introduced. The formulation discriminates between the cases in which either the pressure or the number density of heavy particles is known. This discrimination allows the method to be generalized to include all problems of practical interest, as well as to clarify ambiguities found in other formulations in the literature. The present method overcomes restrictions imposed on other competitive techniques and takes into account all possible formulae for nonideality corrections. In most practical cases the solution of the nonlinear set of the Saha equations is reduced to the simple problem of solving a single transcendental equation.}, number={8}, journal={JOURNAL OF PHYSICS D-APPLIED PHYSICS}, author={Zaghloul, MR and Bourham, MA and Doster, JM}, year={2000}, month={Apr}, pages={977–984} }
@article{zaghloul_bourham_doster_2000, title={Energy-averaged electron-ion momentum transport cross section in the Born approximation and Debye-Huckel potential: Comparison with the cut-off theory}, volume={268}, ISSN={["0375-9601"]}, DOI={10.1016/S0375-9601(00)00217-6}, abstractNote={An exact analytical expression for the energy-averaged electron–ion momentum transport cross section in the Born approximation and Debye–Hückel exponentially screened potential has been derived and compared with the formulae given by other authors. A quantitative comparison between cut-off theory and quantum mechanical perturbation theory has been presented. Based on results from the Born approximation and Spitzer's formula, a new approximate formula for the quantum Coulomb logarithm has been derived and shown to be more accurate than previous expressions.}, number={4-6}, journal={PHYSICS LETTERS A}, author={Zaghloul, MR and Bourham, MA and Doster, JM}, year={2000}, month={Apr}, pages={375–381} }
@article{shi_doster_mayo_2000, title={Numerical simulation of accumulated steam generator tube sheet damage and loose part impact distributions}, volume={129}, ISSN={["0029-5450"]}, DOI={10.13182/NT00-A3066}, abstractNote={An experimental research program into the loose part damage process identified important mechanisms that govern accumulated loose part damage to steam generator tube sheets. Relationships were developed to quantify damage due to single and multiple impacts, including such effects as tube end open diameter reduction and tube end contour deformation. These experimental investigations have led to the development of a computational model for estimating loose part impact damage on steam generator tube ends. Comparisons to experimental data show the loose part damage model to be a good approximation of actual loose part impact damage and provide a convenient and quantitative linkage between loose part impact properties and damage. Impact damage effects are local effects that depend only on the single impacts and impact overlaps in a small region of interest. The damage can be directly related to local impact density. Since in general the local impact density on a steam generator tube sheet is unknown, a model developed to simulate loose part impact distributions as a function of operating conditions is described.}, number={3}, journal={NUCLEAR TECHNOLOGY}, author={Shi, L and Doster, JM and Mayo, CW}, year={2000}, month={Mar}, pages={338–355} }
@article{shi_doster_mayo_1999, title={Drag coefficients for reactor loose parts}, volume={127}, ISSN={["0029-5450"]}, DOI={10.13182/NT99-A2981}, abstractNote={To estimate the range of impact velocities of potential reactor loose parts (LPs) requires information on regional flow velocities, LP mass, and LP drag coefficients. Flow velocities and the mass of potential LPs can generally be bounded and therefore are assumed to be known. In this work, drag coefficients for prototype LP shapes, including objects such as bolts, nuts, pins, and hand tools, were measured in the fluid velocity range typical of reactor coolant systems. Unlike drag coefficients measured for stationary objects, or by moving a body through a stagnant fluid, these experiments are performed on objects moving freely in a turbulent flow stream. In general, the measured drag coefficients for all tested LP shapes are shown to be close to the standard drag coefficient for a sphere, especially in the low-Reynolds-number region. However, significant differences exist in the wake transition region, which indicates that the drag coefficient for a freely moving body in turbulent flow is different from the drag coefficient for a confined body under the same flow conditions or for a body moving in a stagnant fluid.}, number={1}, journal={NUCLEAR TECHNOLOGY}, author={Shi, L and Doster, JM and Mayo, CW}, year={1999}, month={Jul}, pages={24–37} }
@article{shi_mayo_doster_1999, title={Loose-part damage}, volume={34}, ISSN={["0149-1970"]}, DOI={10.1016/S0149-1970(98)00009-2}, abstractNote={Experiments and analytical methods were used to develop models that describe loose part impact damage. Single impact damage volume was characterized as a function of impact energy and contact shape. The deformation volume for multiple, overlapping impacts was used to define work hardening. Integrated impact damage experiments were conducted at representative pressurized water reactor steam generator inlet flow velocity using a target plate that represented a section of a steam generator tube sheet. Data from the integrated impact tests were used to develop an empirical relationship for tube end open diameter reduction as a function of loose part energy and accumulated impact density. Monte-Carlo programs were developed to investigate the tube sheet impact density distribution and the penetration of tube end welds. The results provide insight with respect to single and multiple impact damage mechanisms and methods that can be used to predict steam generator tube sheet accumulated impact damage.}, number={3}, journal={PROGRESS IN NUCLEAR ENERGY}, author={Shi, L and Mayo, CW and Doster, JM}, year={1999}, pages={231–261} }
@article{doster_kauffman_1999, title={Numerical stability of the mixture drift-flux equations}, volume={132}, ISSN={["0029-5639"]}, DOI={10.13182/NSE99-A2051}, abstractNote={Drift flux models are commonly used to describe two-phase flow systems when explicit representation of the relative phase motion is not required. In these models, relative phase velocity is typically described by flow-regime-dependent, semi-empirical models. Although they are a somewhat simple description of the two-phase conditions that might be expected in nuclear power systems, drift flux models can still be expected to give reasonable results in a significant range of operating conditions and can be useful in incorporating thermal-hydraulic feedback into steady-state and transient neutronics calculations. In this paper, we examine the numerical stability associated with the finite difference solution of the mixture drift flux equations. We assume a standard semi-implicit discretization on a staggered spatial mesh, where the drift flux terms are evaluated purely explicitly.}, number={1}, journal={NUCLEAR SCIENCE AND ENGINEERING}, author={Doster, JM and Kauffman, JM}, year={1999}, month={May}, pages={90–104} }
@article{zaghloul_bourham_doster_powell_1999, title={On the average electron-ion momentum transport cross-section in ideal and non-ideal plasmas}, volume={262}, ISSN={["0375-9601"]}, DOI={10.1016/S0375-9601(99)00560-5}, abstractNote={The energy-averaged electron–ion (e–i) momentum transport cross-section has been derived analytically and computed numerically. The result shows inaccuracy of the computations and fitting formula given by other authors.}, number={1}, journal={PHYSICS LETTERS A}, author={Zaghloul, MR and Bourham, MA and Doster, JM and Powell, JD}, year={1999}, month={Oct}, pages={86–89} }
@article{shi_mayo_doster_1999, title={Reactor loose-part activity}, volume={34}, ISSN={["0149-1970"]}, DOI={10.1016/S0149-1970(98)00011-0}, abstractNote={Measured Loose-Part drag coefficients were used to calculate reactor loose-part activity as a function of mass and reactor coolant flow conditions for prototype loose-part shapes. The maximum loose-part mass that could be levitated in vertical flow and the maximum loose-part mass that could impact a recirculating steam generator tube sheet were calculated as a function of flow velocity for Pressurized Water Reactor (PWR) cold and operating primary coolant water properties. The energy of steam generator tube sheet impacts was calculated as a function of mass at cold and operating conditions for a 1130 MW Pressurized Water Reactor. Substantial decreases in active loose-part mass and impact energy occurred between cold and hot flow conditions due to the decrease in water density and viscosity. Loose-parts with higher surface area to mass ratios had higher maximum levitation masses and impact energy. These calculations provide insight into the range of active loose-part mass and impact energy as a function of flow conditions. The associated range in detected signal amplitude can assist in the screening and evaluation of unknown loose-part signals. The loose-part activity modeling methods can be used to extend the results to other reactor coolant system flow conditions and geometries.}, number={3}, journal={PROGRESS IN NUCLEAR ENERGY}, author={Shi, L and Mayo, CW and Doster, JM}, year={1999}, pages={283–298} }
@article{doster_kendall_1999, title={Stability of one-dimensional natural-circulation flows}, volume={132}, ISSN={["0029-5639"]}, DOI={10.13182/NSE99-A2052}, abstractNote={Natural circulation is important for the long-term cooling of light water reactors in off-normal conditions, and it is therefore important to understand the numerical behavior of reactor safety codes used to simulate flows under those conditions. While the methods and models in these codes have been studied in some detail, the impact of the weight force term on the numerical behavior has been largely ignored. The dynamic and numerical stability of the one-dimensional, single-phase-flow equations are examined for natural-circulation problems. It is shown that the presence of the weight force in the momentum equation results in a minimum value of the frictional loss coefficient for the equations to be stable. It is further shown that the numerical solution is unstable unless this dynamic stability limit is satisfied. The stability limits developed are verified by numerical solution of the single-phase-flow equations under natural-circulation conditions.}, number={1}, journal={NUCLEAR SCIENCE AND ENGINEERING}, author={Doster, JM and Kendall, PK}, year={1999}, month={May}, pages={105–117} }
@inproceedings{bourham_gilligan_doster_orton_tucker_1997, title={Simulation of plasma-surface interactions in electrothermal-chemical devices: progress on the 2-D code TURFIRE}, volume={1}, number={1997}, booktitle={Proc. 34th JANNAF Combustion Meeting, CPIA Publications 662, West Palm Beach, FL, 27-31 October 1997}, publisher={CPIA, Chemical Propulsion Information Agency}, author={Bourham, M. A. and Gilligan, J. G. and Doster, J. M. and Orton, N. P. and Tucker, E. C.}, year={1997}, pages={57–63} }