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Analysis of ANS LWR physics benchmark problems.

Description: Various Monte Carlo and deterministic solutions to the three PWR Lattice Benchmark Problems recently defined by the ANS Ad Hoc Committee on Reactor Physics Benchmarks are presented. These solutions were obtained using the VIM continuous-energy Monte Carlo code and the DIF3D/WIMS-D4M code package implemented at the Argonne National Laboratory. The code results for the K{sub eff} and relative pin power distribution are compared to measured values. Additionally, code results for the three benchmark-prescribed infinite lattice configurations are also intercompared. The results demonstrate that the codes produce very good estimates of both the K{sub eff} and power distribution for the critical core and the lattice parameters of the infinite lattice configuration.
Date: July 29, 1998
Creator: Taiwo, T. A.
Partner: UNT Libraries Government Documents Department

Neutronic assessment of stringer fuel assembly design for liquid-salt-cooledvery high temperature reactor (LS-VHTR).

Description: Neutronic studies of 18-pin and 36-pin stringer fuel assemblies have been performed to ascertain that core design requirements for the Liquid-Salt Cooled Very High Temperature Reactor (LS-VHTR) can be met. Parametric studies were performed to determine core characteristics required to achieve a target core cycle length of 18 months and fuel discharge burnup greater than 100 GWd/t under the constraint that the uranium enrichment be less than 20% in order to support non-proliferation goals. The studies were done using the WIMS9 lattice code and the linear reactivity model to estimate the core reactivity balance, fuel composition, and discharge burnup. The results show that the design goals can be met using a 1-batch fuel management scheme, uranium enrichment of 15% and a fuel packing fraction of 30% or greater for the 36-pin stringer fuel assembly design.
Date: September 15, 2006
Creator: Szakaly, F. J.; Kim, T. K. & Taiwo, T. A.
Partner: UNT Libraries Government Documents Department

Fuel cycle analysis of once-through nuclear systems.

Description: Once-through fuel cycle systems are commercially used for the generation of nuclear power, with little exception. The bulk of these once-through systems have been water-cooled reactors (light-water and heavy water reactors, LWRs and HWRs). Some gas-cooled reactors are used in the United Kingdom. The commercial power systems that are exceptions use limited recycle (currently one recycle) of transuranic elements, primarily plutonium, as done in Europe and nearing deployment in Japan. For most of these once-through fuel cycles, the ultimate storage of the used (spent) nuclear fuel (UNF, SNF) will be in a geologic repository. Besides the commercial nuclear plants, new once-through concepts are being proposed for various objectives under international advanced nuclear fuel cycle studies and by industrial and venture capital groups. Some of the objectives for these systems include: (1) Long life core for remote use or foreign export and to support proliferation risk reduction goals - In these systems the intent is to achieve very long core-life with no refueling and limited or no access to the fuel. Most of these systems are fast spectrum systems and have been designed with the intent to improve plant economics, minimize nuclear waste, enhance system safety, and reduce proliferation risk. Some of these designs are being developed under Generation IV International Forum activities and have generally not used fuel blankets and have limited the fissile content of the fuel to less than 20% for the purpose on meeting international nonproliferation objectives. In general, the systems attempt to use transuranic elements (TRU) produced in current commercial nuclear power plants as this is seen as a way to minimize the amount of the problematic radio-nuclides that have to be stored in a repository. In this case, however, the reprocessing of the commercial LWR UNF to produce the initial fuel will be necessary. For ...
Date: August 10, 2010
Creator: Kim, T. K.; Taiwo, T. A. & Division, Nuclear Engineering
Partner: UNT Libraries Government Documents Department

Summary of Generation-IV transmutation impacts.

Description: An assessment of the potential role of Generation IV nuclear systems in an advanced fuel cycle has been performed. The Generation IV systems considered are the thermal-spectrum VHTR and SCWR, and the fast-spectrum GFR, LFR, and SFR. This report addresses the impact of each system on advanced fuel cycle goals, particularly related to waste management and resource utilization. The transmutation impact of each system was also assessed, along with variant designs for transuranics (TRU) burning. The base fuel cycle for the thermal reactor concepts (VHTR and SCWR) is a once-through fuel cycle using low-enriched uranium fuels. The higher burnup and thermal efficiency of the VHTR gives an advantage in terms of heavy-metal waste mass and volume, with lower decay heat and radiotoxicity of the spent fuel per electrical energy produced, compared to a PWR. Fuel utilization might, however, be worse compared to the PWR, because of the higher fuel enrichment essential to meeting the VHTR system design requirements. The SCWR concept also featured improved thermal efficiency; however, benefits are reduced by the lower fuel discharge burnup. The base fuel cycle for the fast reactor concepts (SFR, GFR, and LFR) is a closed fuel cycle using recycled TRU and depleted uranium fuels. Waste management gains from complete recycle are substantial, with the final disposition heat load determined by processing losses. The base Generation-IV concepts allow consumption of U-238 significantly extending uranium resources (up to 100 times). For both thermal and fast concepts, recent design studies have pursued the development of dedicated burner designs. Preliminary results suggest that a burnup of 50-60% is possible in a VHTR burner design using non-uranium (transuranics) fuel. However, practical limits related to higher actinide buildup and safety impact may limit the extent of TRU burning in thermal reactors. Fast burner designs have been developed for both ...
Date: August 3, 2005
Creator: Taiwo, T. A. & Hill, R. N.
Partner: UNT Libraries Government Documents Department

Preliminary neutronic studies for the liquid-salt-cooled very hightemperature reactor (LS-VHTR).

Description: Preliminary neutronic studies have been performed in order to provide guidelines to the design of a liquid-salt cooled Very High Temperature Reactor (LS-VHTR) using Li{sub 2}BeF{sub 4} (FLiBe) as coolant and a solid cylindrical core. The studies were done using the lattice codes (WIMS8 and DRAGON) and the linear reactivity model to estimate the core reactivity balance, fuel composition, discharge burnup, and reactivity coefficients. An evaluation of the lattice codes revealed that they give very similar accuracy as the Monte Carlo MCNP4C code for the prediction of the fuel element multiplication factor (kinf) and the double heterogeneity effect of the coated fuel particles in the graphite matrix. The loss of coolant from the LS-VHTR core following coolant voiding was found to result in a positive reactivity addition, due primarily to the removal of the strong neutron absorber Li-6. To mitigate this positive reactivity addition and its impact on reactor design (positive void reactivity coefficient), the lithium in the coolant must be enriched to greater than 99.995% in its Li-7 content. For the reference LS-VHTR considered in this work, it was found that the magnitude of the coolant void reactivity coefficient (CVRC) is quite small (less than $1 for 100% voiding). The coefficient was found to become more negative or less positive with increase in the lithium enrichment (Li-7 content). It was also observed that the coefficient is positive at the beginning of cycle and becomes more negative with increasing burnup, indicating that by using more than one fuel batch, the coefficient could be made negative at the beginning of cycle. It might, however, still be necessary at the beginning of life to design for a negative CVRC value. The study shows that this can be done by using burnable poisons (erbium is a leading candidate) or by changing the reference ...
Date: October 5, 2005
Creator: Kim, T. K.; Taiwo, T. A. & Yang, W. S.
Partner: UNT Libraries Government Documents Department

Impact of spectral transition zone in reference ENIGMA configuration.

Description: The gas-cooled fast reactor (GFR) is one of six advanced nuclear energy systems being studied under the auspices of the Gen IV International Forum (GIF). In a bilateral International Nuclear Energy Research Initiative (I-NERI) project French and U.S. national laboratories, industry, and universities are collaborating on the development of the GFR. This effort is led by the ANL in the U.S. and the CEA in France. Some of the attractions of the GFR include: (1) Hard spectrum and core breeding ratio, BR {approx} 1. These features allow minimal waste production, improved transmutation capability, optimal and flexible use of natural resources, potentially better economy (because of use of higher power density relative to current thermal gas-cooled systems), and improved non-proliferation (no fertile blanket); (2) Temperature resistant fuel and structure elements that are favorable to tight fission product confinement and system operation at high temperature; (3) High temperature and transparent helium (He) gas coolant that allows a high thermodynamic conversion efficiency, other energy applications (e.g., hydrogen production), and ease of in-service inspection and repair; and (4) Possible direct energy conversion cycle leading to a simpler design, increased conversion efficiency, and lower investment costs. The French strategy for advanced systems includes the development of the GFR and sodium-cooled fast reactor (SFR) to levels that allow industries to be able to make an informed choice of the fast spectrum system that would provide a sustainable nuclear energy generation option for the future. Current planning calls for the construction of a small experimental research and technology development reactor (ETDR) around 2009 (first operation in 2015) at CEA-Cadarache, France. This would be followed by the construction of a GFR industrial prototype, around 2025. In support of the GFR development efforts, a new physics experimental program (called ENIGMA, Experimental Neutron Investigation of Gas-cooled reactor at Masurca) is ...
Date: October 5, 2005
Creator: Aliberti, G.; Palmiotti, G.; Taiwo, T. A. & Tommasi, J.
Partner: UNT Libraries Government Documents Department

An improved quasistatic option for the DIF3D nodal kinetics code

Description: An improved quasistatic scheme is formulated for solution of the time-dependent DIF3D nodal equations in hexagonal-z geometry. This scheme has been implemented, along with adiabatic and point kinetics solution options, in the DIF3D hexagonal-z nodal kinetics code. The improved quasistatic method is shown to permit significant reduction in computing time, even for transients involving pronounced changes in flux shape. The achievable computing time reduction, in addition to being problem dependent, has also been found to be larger when greater accuracy is required in the computed results.
Date: January 1, 1991
Creator: Taiwo, T.A. & Khalil, H.S.
Partner: UNT Libraries Government Documents Department

An improved quasistatic option for the DIF3D nodal kinetics code

Description: An improved quasistatic scheme is formulated for solution of the time-dependent DIF3D nodal equations in hexagonal-z geometry. This scheme has been implemented, along with adiabatic and point kinetics solution options, in the DIF3D hexagonal-z nodal kinetics code. The improved quasistatic method is shown to permit significant reduction in computing time, even for transients involving pronounced changes in flux shape. The achievable computing time reduction, in addition to being problem dependent, has also been found to be larger when greater accuracy is required in the computed results.
Date: December 31, 1991
Creator: Taiwo, T. A. & Khalil, H. S.
Partner: UNT Libraries Government Documents Department

Th/U-233 multi-recycle in PWRs.

Description: The use of thorium in current or advanced light water reactors (LWRs) has been of interest in recent years. These interests have been associated with the need to increase nuclear fuel resources and the perceived non-proliferation advantages of the utilization of thorium in the fuel cycle. Various options have been considered for the use of thorium in the LWR fuel cycle including: (1) its use in a once-through fuel cycle to replace non-fissile uranium or to extend fuel burnup due to its attractive fertile material conversion, (2) its use for fissile plutonium burning in limited recycle cores, and (3) its advantage in limiting the transuranic elements to be disposed off in a repository (if only Th/U-233 fuel is used). The possibility for thorium utilization in multirecycle system has also been considered by various researchers, primarily because of the potential for near breeders with Th/U-233 in the thermal energy range. The objective of this project is to evaluate the potential of the Th/U-233 fuel multirecycle in current LWRs, with focus this year on pressurized water reactors (PWRs). In this work, approaches for ensuring a sustainable multirecycle without the need for external source of makeup fissile material have been investigated. The intent is to achieve a design that allows existing PWRs to be used with minimal modifications. In all cases including homogeneous and heterogeneous assembly designs, the assembly pitch is kept consistent with that of the current PWRs (21.5 cm used). Because of design difficulties associated with using the same geometry and dimensions as a PWR core, the potential modifications (other than assembly pitch) that would be needed for PWRs to ensure a sustainable multirecycle system have been investigated and characterized. Additionally, the implications of the use of thorium on the LWR fuel cycle are discussed. In Section 2, background information on ...
Date: September 7, 2010
Creator: Yun, D.; Kim, T. K.; Taiwo, T. A. & Division, Nuclear Engineering
Partner: UNT Libraries Government Documents Department

Reactivity estimation for source-driven systems using first-order perturbation theory.

Description: Applicability of the first-order perturbation (FOP) theory method to reactivity estimation for source-driven systems is examined in this paper. First, the formally exact point kinetics equations have been derived from the space-dependent kinetics equations and the kinetics parameters including the dynamic reactivity have been defined. For the dynamic reactivity, exact and first-order perturbation theory expressions for the reactivity change have been formulated for source-driven systems. It has been also shown that the external source perturbation itself does not change the reactivity if the initial {lambda}-mode adjoint flux is used as the weight function. Using two source-driven benchmark problems, the reactivity change has been estimated with the FOP theory method for various perturbations. By comparing the resulting reactivity changes with the exact dynamic reactivity changes determined from the space-dependent kinetics solutions, it has been shown that the accuracy of the FOP theory method for the accelerator-driven system (ADS) is reasonably good and comparable to that for the critical reactors. The adiabatic assumption has also been shown to be a good approximation for the ADS kinetics analyses.
Date: July 2, 2002
Creator: Kim, Y.; Yang, W. S.; Taiwo, T. A. & Hill, R. N.
Partner: UNT Libraries Government Documents Department

Assessment of General Atomics accelerator transmutation of waste concept based on gas-turbine-modular helium cooled reactor technology.

Description: An assessment has been performed for an Accelerator Transmutation of Waste (ATW) concept based on the use of the high temperature gas reactor technology. The concept has been proposed by General Atomics for the ATW system. The assessment was jointly conducted at Argonne National Laboratory (ANL) and Los Alamos national laboratory to assess and to define the potential candidates for the ATW system. This report represents the assessment work performed at ANL. The concept uses recycled light water reactor (LWR)-discharge-transuranic extracted from irradiated oxide fuel in a critical and sub-critical accelerator driven gas-cooled transmuter. In this concept, the transmuter operates at 600 MWt first in the critical mode for three cycles and then operates in a subcritical accelerator-driven mode for a single cycle. The transmuter contains both thermal and fast spectrum transmutation zones. The thermal zone is fueled with the TRU oxide material in the form of coated particles, which are mixed with graphite powder, packed into cylindrical compacts, and loaded in hexagonal graphite blocks with cylindrical channels; the fast zone is fueled with TRU-oxide material in the form of coated particles without the graphite powder and the graphite blocks that has been burned in the thermal region for three critical cycles and one additional accelerator-driven cycle. The fuel loaded into the fast zone is irradiated for four additional cycles. This fuel management scheme is intended to achieve a high Pu isotopes consumption in the thermal spectrum zone, and to consume the minor actinides in the fast-spectrum zone. Monte Carlo and deterministic codes have been used to assess the system performance and to determine the feasibility of achieving high TRU consumption levels. The studies revealed the potential for high consumption of Pu-239 (97%), total Pu (71%) and total TRU (64%) in the system. The analyses confirmed the need for burnable ...
Date: May 8, 2001
Creator: Gohar, Y.; Taiwo, T. A.; Cahalan, J. E. & Finck, P. J.
Partner: UNT Libraries Government Documents Department

Assessment of the teledial gas-cooled transmuter concept

Description: The application of four gas-turbine, modular helium cooled reactors and an accelerator unit (GT/AD-MHR) has been proposed for burning transuranics recycled from LWR waste. The recycled LWR discharged transuranics encapsulated in TRISO coated particles are first loaded into the outer thermal spectrum zone of the GT/AD-MHR for burning in the critical mode for about three years. Previously burned fuel is in a central fast zone. In the fourth year, the same unit is configured as an accelerator-driven system, containing a centrally located spallation target. The three-year, thermal-zone burned fuel and the inner fast-zone fuel from the critical mode operation are used in this subcritical cycle, and remain in their respective zones. At the end of this one-year subcritical irradiation, the outer thermal-zone fuel is reconstituted and used as fast-zone fuel in another critical mode operation. As the fuel in the fast-zone has reached its end of life it is discharged, with very low transuranics content. The critical mode operation is staggered, and each GT/AD-MGR unit undergoes the subcritical burn in one out of four year. The physics performance of the GT/AD-MHR has been evaluated using independent deterministic and Monte Carlo codes and the results of the study are presented in the current paper. A companion paper discussing the verification of the codes is also presented at this meeting. Single-batch and three-batch fuel loading schemes for the GT/AD-MHR have been evaluated using the REBUS-3/DIF3D fuel cycle code, to determine the feasibility of achieving very high burnup without exceeding reactivity and power density limits. The reactor physics of the GT-MHR is complicated by the presence of the low-lying plutonium and Er-167 resonances (0.2--1.1 eV) and by the fact that the neutron spectrum has a low-energy peak about this energy range. This peak can change depending on the core state or material loading. ...
Date: July 24, 2000
Creator: Taiwo, T. A.; Gohar, Y. & Finck, P. J.
Partner: UNT Libraries Government Documents Department

Need for higher order polynomial basis for polynomial nodal methods employed in LWR calculations

Description: The paper evaluates the accuracy and efficiency of sixth order polynomial solutions and the use of one radial node per core assembly for pressurized water reactor (PWR) core power distributions and reactivities. The computer code VARIANT was modified to calculate sixth order polynomial solutions for a hot zero power benchmark problem in which a control assembly along a core axis is assumed to be out of the core. Results are presented for the VARIANT, DIF3D-NODAL, and DIF3D-finite difference codes. The VARIANT results indicate that second order expansion of the within-node source and linear representation of the node surface currents are adequate for this problem. The results also demonstrate the improvement in the VARIANT solution when the order of the polynomial expansion of the within-node flux is increased from fourth to sixth order. There is a substantial saving in computational time for using one radial node per assembly with the sixth order expansion compared to using four or more nodes per assembly and fourth order polynomial solutions. 11 refs., 1 tab.
Date: August 1, 1997
Creator: Taiwo, T.A. & Palmiotti, G.
Partner: UNT Libraries Government Documents Department

Coupled reactor physics and coolant dynamics of heavy liquid metal coolant systems.

Description: Cooling of advanced nuclear designs with heavy liquid metals such as lead or lead-bismuth eutectic offers the potential for plant simplifications and higher operating efficiencies compared to previously considered liquid metal coolants such as sodium or NaK. Such applications would however also introduce additional safety concerns and design challenges, therefore necessitating a verifiable computational tool for transient design-basis analysis of heavy liquid metal coolant (HLMC) systems. This capability would enable analysts to compare operational and safety characteristics of design alternatives, and to evaluate relative performance advantages with a consistent, deterministic measure.
Date: July 15, 1999
Creator: Cahalan, J. E.; Dunn, F. E. & Taiwo, T. A.
Partner: UNT Libraries Government Documents Department

Uncertainty and target accuracy studies for the very high temperature reactor(VHTR) physics parameters.

Description: The potential impact of nuclear data uncertainties on a number of performance parameters (core and fuel cycle) of the prismatic block-type Very High Temperature Reactor (VHTR) has been evaluated and results are presented in this report. An uncertainty analysis has been performed, based on sensitivity theory, which underlines what cross-sections, what energy range and what isotopes are responsible for the most significant uncertainties. In order to give guidelines on priorities for new evaluations or validation experiments, required accuracies on specific nuclear data have been derived, accounting for target accuracies on major design parameters. Results of an extensive analysis indicate only a limited number of relevant parameters do not meet the target accuracies assumed in this work; this does not imply that the existing nuclear cross-section data cannot be used for the feasibility and pre-conceptual assessments of the VHTR. However, the results obtained depend on the uncertainty data used, and it is suggested to focus some future evaluation work on the production of consistent, as far as possible complete and user oriented covariance data.
Date: September 16, 2005
Creator: Taiwo, T. A.; Palmiotti, G.; Aliberti, G.; Salvatores, M. & Kim, T.K.
Partner: UNT Libraries Government Documents Department

A feasibility study of reactor-based deep-burn concepts.

Description: A systematic assessment of the General Atomics (GA) proposed Deep-Burn concept based on the Modular Helium-Cooled Reactor design (DB-MHR) has been performed. Preliminary benchmarking of deterministic physics codes was done by comparing code results to those from MONTEBURNS (MCNP-ORIGEN) calculations. Detailed fuel cycle analyses were performed in order to provide an independent evaluation of the physics and transmutation performance of the one-pass and two-pass concepts. Key performance parameters such as transuranic consumption, reactor performance, and spent fuel characteristics were analyzed. This effort has been undertaken in close collaborations with the General Atomics design team and Brookhaven National Laboratory evaluation team. The study was performed primarily for a 600 MWt reference DB-MHR design having a power density of 4.7 MW/m{sup 3}. Based on parametric and sensitivity study, it was determined that the maximum burnup (TRU consumption) can be obtained using optimum values of 200 {micro}m and 20% for the fuel kernel diameter and fuel packing fraction, respectively. These values were retained for most of the one-pass and two-pass design calculations; variation to the packing fraction was necessary for the second stage of the two-pass concept. Using a four-batch fuel management scheme for the one-pass DB-MHR core, it was possible to obtain a TRU consumption of 58% and a cycle length of 286 EFPD. By increasing the core power to 800 MWt and the power density to 6.2 MW/m{sup 3}, it was possible to increase the TRU consumption to 60%, although the cycle length decreased by {approx}64 days. The higher TRU consumption (burnup) is due to the reduction of the in-core decay of fissile Pu-241 to Am-241 relative to fission, arising from the higher power density (specific power), which made the fuel more reactivity over time. It was also found that the TRU consumption can be improved by utilizing axial fuel shuffling ...
Date: September 16, 2005
Creator: Kim, T. K.; Taiwo, T. A.; Hill, R. N. & Yang, W. S.
Partner: UNT Libraries Government Documents Department