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Large eddy simulation of flow in LWR fuel bundles.

Description: Advances in computational fluid dynamics (CFD), turbulence modeling, and parallel computing have made feasible the development of codes that can simulate 3-D flows and heat transfer in realistic LWR fuel bundle geometries. Although no single existing RANS (Reynolds averaging of the Navier Stokes equations) turbulence model predicts a sufficiently wide range of flows with accuracy adequate for engineering needs, at this time for most flows the k-{epsilon} models seem to be the best choice. In Ref. 1, it was shown that in LWR fuel-bundle flows the predictions of these models for turbulence intensity are in significant disagreement with experimental measurements. The objective of this work was to assess the predictive power of the constant-coefficient Smagorinsky Large Eddy Simulation (LES) model, the simplest of the LES models, in a typical single-phase LWR fuel-bundle flow.
Date: August 17, 2001
Creator: Tzanos, C. P.
Partner: UNT Libraries Government Documents Department

Natural convection in a uniformly heated pool

Description: In the event of a core meltdown accident, to prevent reactor vessel failure from molten corium relocation to the reactor vessel lower head, the establishment of a coolable configuration has been proposed by flooding with water the reactor cavity. In Reference 3, it was shown that for the heavy-water new production reactor (NPW-HWR) design, this strategy, e.g., the rejection of decay heat to a containment decay heat removal system by boiling of water in the reactor cavity, could keep the reactor vessel temperature below failure limits. The analysis of Ref. 3 was performed with the computer code COMMIX-1AR/P, and showed that natural convection in the molten-corium pool was the dominant mechanism of heat transfer from the pool to the wall of the reactor vessel lower head. To determine whether COMMIX adequately predicts natural convection in a pool heated by a uniform heat source, in Ref. 4, the experiments of free convection in a semicircular cavity of Jahn and Reineke were analyzed with COMMIX. It was found that the Nusselt (Nu) number predicted by COMMIX was within the spread of the experimental measurements. In the COMMIX analysis of Ref. 4, the semicircular cavity was treated as symmetric. The objective of the work presented in this paper was to extend the COMMIX validation analysis of Ref. 4 by removing the assumption of symmetry and expanding the analysis up to the highest Rayleigh (Ra) number that leads to a steady state. In conclusion, this work shows that the numerical predictions of natural convection in an internally heated pool bounded by a curved bottom are in reasonably good agreement with experimental measurements.
Date: May 1, 1996
Creator: Tzanos, C.P.
Partner: UNT Libraries Government Documents Department

Decay heat removal by natural convection - the RVACS system.

Description: In conclusion, this work shows that for sodium coolant the reactor vessel auxiliary cooling system (RVACS) is an effective passive heat removal system if the reactor power does not exceed about 1600 MW(th). Its effectiveness is limited by the effective radiative heat transfer coefficient in the inner gap. In a lead cooled system, economic considerations may impose a lower limit.
Date: August 17, 1999
Creator: Tzanos, C. P.
Partner: UNT Libraries Government Documents Department

Simulation of dynamic processes with adaptive neural networks.

Description: Many industrial processes are highly non-linear and complex. Their simulation with first-principle or conventional input-output correlation models is not satisfactory, either because the process physics is not well understood, or it is so complex that direct simulation is either not adequately accurate, or it requires excessive computation time, especially for on-line applications. Artificial intelligence techniques (neural networks, expert systems, fuzzy logic) or their combination with simple process-physics models can be effectively used for the simulation of such processes. Feedforward (static) neural networks (FNNs) can be used effectively to model steady-state processes. They have also been used to model dynamic (time-varying) processes by adding to the network input layer input nodes that represent values of input variables at previous time steps. The number of previous time steps is problem dependent and, in general, can be determined after extensive testing. This work demonstrates that for dynamic processes that do not vary fast with respect to the retraining time of the neural network, an adaptive feedforward neural network can be an effective simulator that is free of the complexities introduced by the use of input values at previous time steps.
Date: February 3, 1998
Creator: Tzanos, C. P.
Partner: UNT Libraries Government Documents Department

Numerical simulation of two-phase flow with front-capturing

Description: Because of the complexity of two-phase flow phenomena, two-phase flow codes rely heavily on empirical correlations. This approach has a number of serious shortcomings. Advances in parallel computing and continuing improvements in computer speed and memory have stimulated the development of numerical simulation tools that rely less on empirical correlations and more on fundamental physics. The objective of this work is to take advantage of developments in massively parallel computing, single-phase computational fluid dynamics of complex systems, and numerical methods for front capturing in two-phase flows to develop a computer code for direct numerical simulation of two-phase flow. This includes bubble/droplet transport, interface deformation and topology change, bubble/droplet interactions, interface mass, momentum and energy transfer.
Date: February 8, 2000
Creator: Tzanos, C. P. & Weber, D. P.
Partner: UNT Libraries Government Documents Department

Computational fluid dynamics analyses of lateral heat conduction, coolant azimuthal mixing and heat transfer predictions in a BR2 fuel assembly geometry.

Description: To support the analyses related to the conversion of the BR2 core from highly-enriched (HEU) to low-enriched (LEU) fuel, the thermal-hydraulics codes PLTEMP and RELAP-3D are used to evaluate the safety margins during steady-state operation (PLTEMP), as well as after a loss-of-flow, loss-of-pressure, or a loss of coolant event (RELAP). In the 1-D PLTEMP and RELAP simulations, conduction in the azimuthal and axial directions is not accounted. The very good thermal conductivity of the cladding and the fuel meat and significant temperature gradients in the lateral directions (axial and azimuthal directions) could lead to a heat flux distribution that is significantly different than the power distribution. To evaluate the significance of the lateral heat conduction, 3-D computational fluid dynamics (CFD) simulations, using the CFD code STAR-CD, were performed. Safety margin calculations are typically performed for a hot stripe, i.e., an azimuthal region of the fuel plates/coolant channel containing the power peak. In a RELAP model, for example, a channel between two plates could be divided into a number of RELAP channels (stripes) in the azimuthal direction. In a PLTEMP model, the effect of azimuthal power peaking could be taken into account by using engineering factors. However, if the thermal mixing in the azimuthal direction of a coolant channel is significant, a stripping approach could be overly conservative by not taking into account this mixing. STAR-CD simulations were also performed to study the thermal mixing in the coolant. Section II of this document presents the results of the analyses of the lateral heat conduction and azimuthal thermal mixing in a coolant channel. Finally, PLTEMP and RELAP simulations rely on the use of correlations to determine heat transfer coefficients. Previous analyses showed that the Dittus-Boelter correlation gives significantly more conservative (lower) predictions than the correlations of Sieder-Tate and Petukhov. STAR-CD 3-D simulations ...
Date: May 23, 2011
Creator: Tzanos, C. P. & Dionne, B. (Nuclear Engineering Division)
Partner: UNT Libraries Government Documents Department

Estimation of steady-state and transcient power distributions for the RELAP analyses of the 1963 loss-of-flow and loss-of-pressure tests at BR2.

Description: To support the safety analyses required for the conversion of the Belgian Reactor 2 (BR2) from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, the simulation of a number of loss-of-flow tests, with or without loss of pressure, has been undertaken. These tests were performed at BR2 in 1963 and used instrumented fuel assemblies (FAs) with thermocouples (TC) imbedded in the cladding as well as probes to measure the FAs power on the basis of their coolant temperature rise. The availability of experimental data for these tests offers an opportunity to better establish the credibility of the RELAP5-3D model and methodology used in the conversion analysis. In order to support the HEU to LEU conversion safety analyses of the BR2 reactor, RELAP simulations of a number of loss-of-flow/loss-of-pressure tests have been undertaken. Preliminary analyses showed that the conservative power distributions used historically in the BR2 RELAP model resulted in a significant overestimation of the peak cladding temperature during the transient. Therefore, it was concluded that better estimates of the steady-state and decay power distributions were needed to accurately predict the cladding temperatures measured during the tests and establish the credibility of the RELAP model and methodology. The new approach ('best estimate' methodology) uses the MCNP5, ORIGEN-2 and BERYL codes to obtain steady-state and decay power distributions for the BR2 core during the tests A/400/1, C/600/3 and F/400/1. This methodology can be easily extended to simulate any BR2 core configuration. Comparisons with measured peak cladding temperatures showed a much better agreement when power distributions obtained with the new methodology are used.
Date: May 23, 2011
Creator: Dionne, B. & Tzanos, C. P. (Nuclear Engineering Division)
Partner: UNT Libraries Government Documents Department

Thermal-hydraulic unreliability of passive systems

Description: Advanced light water reactor designs like AP600 and the simplified boiling water reactor (SBWR) use passive safety systems for accident prevention and mitigation. Because these systems rely on natural forces for their operation, their unavailability due to hardware failures and human error is significantly smaller than that of active systems. However, the coolant flows predicted to be delivered by these systems can be subject to significant uncertainties, which in turn can lead to a significant uncertainty in the predicted thermal-hydraulic performance of the plant under accident conditions. Because of these uncertainties, there is a probability that an accident sequence for which a best estimate thermal-hydraulic analysis predicts no core damage (success sequence) may actually lead to core damage. For brevity, this probability will be called thermal-hydraulic unreliability. The assessment of this unreliability for all the success sequences requires very expensive computations. Moreover, the computational cost increases drastically as the required thermal-hydraulic reliability increases. The required computational effort can be greatly reduced if a bounding approach can be used that either eliminates the need to compute thermal-hydraulic unreliabilities, or it leads to the analysis of a few bounding sequences for which the required thermal-hydraulic reliability is relatively small. The objective of this paper is to present such an approach and determine the order of magnitude of the thermal-hydraulic unreliabilities that may have to be computed.
Date: December 31, 1995
Creator: Tzanos, C.P. & Saltos, N.T.
Partner: UNT Libraries Government Documents Department

Status report : guard containment CFD analysis.

Description: Under the auspices of the CEA Cadarache/ANL-US I-NERI project a comprehensive investigation has been made of improvements to the Gen-IV GFR safety case over that of the GCFR safety case twenty five years ago. In particular, it has been concluded and agreed upon [1] that the GFR safety approach for the passive removal of decay heat in a protected depressurization accident with total loss of electric power needs to be different from that taken for the HTRs. The HTR conduction cooldown to the vessel wall boundary mode for an economically attractive core is not feasible in the case of the GFR because the high power densities (100kW/1 compared to 5 kW/1 for pebble bed thermal reactor) require decay heat fluxes well beyond those achievable by the heat conduction and radiation heat transfer mode. A set of alternative novel design options has been evaluated for potential passive safety mechanisms unique to the GFR. In summary, from a technological risk viewpoint and R&D planning, the option which has been identified is the block/plate-based or a pin-based reactor with a secondary guard containment/vessel around the primary vessel to maintain the primary system pressure at a high enough level which would allow primary system natural convection removal of core generated decay heat to be effective. Dedicated emergency decay heat exchangers would have to be connected in a 'failure-proof' configuration to the primary system and have natural convection capability all the way to the ultimate heat sink. What has been collaboratively agreed upon and selected for further development is the natural convection option with a block/plate or pin type derated core and a hybrid passive/active approach.[2] The guard containment will be utilized but it will be sized for an LWR containment range backup pressure (5-7 bars) with an initial pressure of 1 bar. The assessment ...
Date: March 3, 2006
Creator: Tzanos, C. P.
Partner: UNT Libraries Government Documents Department

CFD analysis for the applicability of the natural convection shutdown heat removal test facility (NSTF) for the simulation of the VHTR RCCS. Topical report.

Description: The Very High Temperature gas cooled reactor (VHTR) is one of the GEN IV reactor concepts that have been proposed for thermochemical hydrogen production and other process-heat applications like coal gasification. The USDOE has selected the VHTR for further research and development, aiming to demonstrate emissions-free electricity and hydrogen production at a future time. One of the major safety advantages of the VHTR is the potential for passive decay heat removal by natural circulation of air in a Reactor Cavity Cooling System (RCCS). The air-side of the RCCS is very similar to the Reactor Vessel Auxiliary Cooling System (RVACS) that has been proposed for the PRISM reactor design. The design and safety analysis of the RVACS have been based on extensive analytical and experimental work performed at ANL. The Natural Convective Shutdown Heat Removal Test Facility (NSTF) at ANL that simulates at full scale the air-side of the RVACS was built to provide experimental support for the design and analysis of the PRISM RVACS system. The objective of this work is to demonstrate that the NSTF facility can be used to generate RCCS experimental data: to validate CFD and systems codes for the analysis of the RCCS; and to support the design and safety analysis of the RCCS.
Date: May 16, 2007
Creator: Tzanos, C. P.
Partner: UNT Libraries Government Documents Department

Topical report : CFD analysis for the applicability of the natural convection shutdown heat removal test facility (NSTF) for the simulation of the VHTR RCCS.

Description: The Very High Temperature gas cooled reactor (VHTR) is one of the GEN IV reactor concepts that have been proposed for thermochemical hydrogen production and other process-heat applications like coal gasification. The United States Department of Energy has selected the VHTR for further research and development, aiming to demonstrate emissions-free electricity and hydrogen production at a future time. One of the major safety advantages of the VHTR is the potential for passive decay heat removal by natural circulation of air in a Reactor Cavity Cooling System (RCCS). The air-side of the RCCS is very similar to the Reactor Vessel Auxiliary Cooling System (RVACS) that has been proposed for the PRISM reactor design. The design and safety analysis of the RVACS have been based on extensive analytical and experimental work performed at ANL. The Natural Convection Shutdown Heat Removal Test Facility (NSTF) at ANL that simulates at full scale the air-side of the RVACS was built to provide experimental support for the design and analysis of the PRISM RVACS system. The objective of this work is to demonstrate that the NSTF facility can be used to generate RCCS experimental data: to validate CFD and systems codes for the analysis of the RCCS; and to support the design and safety analysis of the RCCS. At this time no reference design is available for the NGNP. The General Atomics (GA) gas turbine - modular helium reactor (GT-MHR) has been used in many analyses as a starting reference design. In the GT-MHR the reactor outlet temperature is 850 C, while the target outlet reactor temperature in VHTR is 1000 C. VHTR scoping studies with a reactor outlet temperature of 1000 C have been performed at GA and INEL. Although the reactor outlet temperature in the VHTR is significantly higher than in the GT-MHR, the ...
Date: May 16, 2007
Creator: Tzanos, C. P. (Nuclear Engineering Division)
Partner: UNT Libraries Government Documents Department

Feasibility study for use of the natural convection shutdown heat removal test facility (NSTF) for VHTR water-cooled RCCS shutdown.

Description: In summary, a scaling analysis of a water-cooled Reactor Cavity Cooling System (RCCS) system was performed based on generic information on the RCCS design of PBMR. The analysis demonstrates that the water-cooled RCCS can be simulated at the ANL NSTF facility at a prototypic scale in the lateral direction and about half scale in the vertical direction. Because, by necessity, the scaling is based on a number of approximations, and because no analytical information is available on the performance of a reference water-cooled RCCS, the scaling analysis presented here needs to be 'validated' by analysis of the steady state and transient performance of a reference water-cooled RCCS design. The analysis of the RCCS performance by CFD and system codes presents a number of challenges including: strong 3-D effects in the cavity and the RCCS tubes; simulation of turbulence in flows characterized by natural circulation, high Rayleigh numbers and low Reynolds numbers; validity of heat transfer correlations for system codes for heat transfer in the cavity and the annulus of the RCCS tubes; the potential of nucleate boiling in the tubes; water flashing in the upper section of the RCCS return line (during limiting transient); and two-phase flow phenomena in the water tanks. The limited simulation of heat transfer in cavities presented in Section 4.0, strongly underscores the need of experimental work to validate CFD codes, and heat transfer correlations for system codes, and to support the analysis and design of the RCCS. Based on the conclusions of the scaling analysis, a schematic that illustrates key attributes of the experiment system is shown in Fig. 4. This system contains the same physical elements as the PBMR RCCS, plus additional equipment to facilitate data gathering to support code validation. In particular, the prototype consists of a series of oval standpipes surrounding the ...
Date: August 31, 2007
Creator: Tzanos, C. P. & Farmer, M. T.
Partner: UNT Libraries Government Documents Department

A coupled Newton-Krylov solver for improved CHAD cache utilization and performance.

Description: CHAD (Computational Hydrodynamics for Advanced Design) is a computer program that has been developed to analyze flows in automotive and defense applications. Extensive performance analysis of the CHAD computer program indicated a need to address cache memory use to increase computational performance. Several strategies have been adopted to achieve this goal: simultaneous solution of the coupled Navier-Stokes equations, data clustering, and data ordering. A coupled Newton-Krylov solver has been incorporated into a version of the CHAD program, resulting in consistent improvement in run times that varies from 50% to 200%. Further work will be required to tune the solver for optimal performance. In addition, experiments with data cluster and reordering indicate a potential for performance improvement.
Date: March 24, 2000
Creator: Canfield, T.R.; Chien, T.H.; Domanus, H.M.; Tentner, A.M.; Tzanos, C.P.; Valentin, R.A. et al.
Partner: UNT Libraries Government Documents Department

Topical report: Natural convection shutdown heat removal test facility (NSTF) evaluation for generating additional reactor cavity cooling system (RCCS) data.

Description: As part of the Department of Energy (DOE) Generation IV roadmapping activity, the Very High Temperature gas cooled Reactor (VHTR) has been selected as the principal concept for hydrogen production and other process-heat applications such as district heating and potable water production. On this basis, the DOE has selected the VHTR for additional R&D with the ultimate goal of demonstrating emission-free electricity and hydrogen production with this advanced reactor concept. One of the key passive safety features of the VHTR is the potential for decay heat removal by natural circulation of air in a Reactor Cavity Cooling System (RCCS). The air-cooled RCCS concept is notably similar to the Reactor Vessel Auxiliary Cooling System (RVACS) that was developed for the General Electric PRISM sodium-cooled fast reactor. As part of the DOE R&D program that supported the development of this fast reactor concept, the Natural Convection Shutdown Heat Removal Test Facility (NSTF) was developed at ANL to provide proof-of-concept data for the RVACS under prototypic natural convection flow, temperature, and heat flux conditions. Due to the similarity between RVACS and the RCCS, current VHTR R&D plans call for the utilization of the NSTF to provide RCCS model development and validation data, in addition to supporting design validation and optimization activities. Both air-cooled and water-cooled RCCS designs are to be included. In support of this effort, ANL has been tasked with the development of an engineering plan for mechanical and instrumentation modifications to NSTF to ensure that sufficiently detailed temperature, heat flux, velocity and turbulence profiles are obtained to adequately qualify the codes under the expected range of air-cooled RCCS flow conditions. Next year, similar work will be carried out for the alternative option of a water-cooled RCCS design. Analysis activities carried out in support of this experiment planning task have shown that: ...
Date: September 1, 2005
Creator: Farmer, M. T.; Kilsdonk, D. J.; Tzanos, C.P.; Lomperski, S.; Aeschlimann, R.W.; Pointer, D. et al.
Partner: UNT Libraries Government Documents Department

Topical report : NSTF facilities plan for water-cooled VHTR RCCS : normal operational tests.

Description: As part of the Department of Energy (DOE) Generation IV roadmapping activity, the gas-cooled Very High Temperature Reactor (VHTR) has been selected as the principal concept for hydrogen production and other process-heat applications such as district heating and potable water production. On this basis, the DOE has selected the VHTR for additional R&D with the ultimate goal of demonstrating emission-free electricity and hydrogen production with this advanced reactor concept.
Date: September 1, 2006
Creator: Farmer, M. T.; Kilsdonk, D. J.; Tzanos, C. P.; Lomperski, S.; Aeschlimann, R. W. & Division, Nuclear Engineering
Partner: UNT Libraries Government Documents Department

Interim status report on lead-cooled fast reactor (LFR) research and development.

Description: This report discusses the status of Lead-Cooled Fast Reactor (LFR) research and development carried out during the first half of FY 2008 under the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. Lead-Cooled Fast Reactor research and development has recently been transferred from Generation IV to the Reactor Campaign of the Global Nuclear Energy Partnership (GNEP). Another status report shall be issued at the end of FY 2008 covering all of the LFR activities carried out in FY 2008 for both Generation IV and GNEP. The focus of research and development in FY 2008 is an initial investigation of a concept for a LFR Advanced Recycling Reactor (ARR) Technology Pilot Plant (TPP)/demonstration test reactor (demo) incorporating features and operating conditions of the European Lead-cooled SYstem (ELSY) {approx} 600 MWe lead (Pb)-cooled LFR preconceptual design for the transmutation of waste and central station power generation, and which would enable irradiation testing of advanced fuels and structural materials. Initial scoping core concept development analyses have been carried out for a 100 MWt core composed of sixteen open-lattice 20 by 20 fuel assemblies largely similar to those of the ELSY preconceptual fuel assembly design incorporating fuel pins with mixed oxide (MOX) fuel, central control rods in each fuel assembly, and cooled with Pb coolant. For a cycle length of three years, the core is calculated to have a conversion ratio of 0.79, an average discharge burnup of 108 MWd/kg of heavy metal, and a burnup reactivity swing of about 13 dollars. With a control rod in each fuel assembly, the reactivity worth of an individual rod would need to be significantly greater than one dollar which is undesirable for postulated rod withdrawal reactivity insertion events. A peak neutron fast flux of 2.0 x 10{sup 15} (n/cm{sup 2}-s) is calculated. For comparison, ...
Date: March 31, 2008
Creator: Tzanos, C. P.; Sienicki, J. J.; Moisseytsev, A.; Smith, C. F.; de Caro, M.; Halsey, W. G. et al.
Partner: UNT Libraries Government Documents Department

Gas-Cooled Fast Reactor (GFR) FY05 Annual Report

Description: The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection. Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in ...
Date: September 1, 2005
Creator: Weaver, K. D.; Marshall, T.; Totemeier, T.; Gan, J.; Feldman, E.E.; Hoffman, E.A et al.
Partner: UNT Libraries Government Documents Department