[en] 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_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_eff and power distribution for the critical core and the lattice parameters of the infinite lattice configuration

[en] The variational nodal method is formulated such that the angular and spatial approximations maybe examined separately. Spherical harmonic, simplified spherical harmonic, and discrete ordinate approximations are coupled to the primal hybrid finite element treatment of the spatial variables. Within this framework, two classes of spatial trial functions are presented: (1) orthogonal polynomials for the treatment of homogeneous nodes and (2) bilinear finite subelement trial functions for the treatment of fuel assembly sized nodes in which fuel-pin cell cross sections are represented explicitly. Polynomial and subelement trial functions are applied to benchmark water-reactor problems containing MOX fuel using spherical harmonic and simplified spherical harmonic approximations. The resulting accuracy and computing costs are compared

No abstract available

[en] 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 conventional and high TRU content fuel forms. In general, the conversion ratio can be varied within a system by changing the uranium loading. Recent studies indicate a low conversion ratio (0.25) SFR retains the favorable passive characteristics of conventional designs, and the cost is similar

No abstract available

[en] 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. The location of the peak and the direction of the spectral shift greatly affect both the resonance fission and capture rates and dictate the core or element criticality state and the magnitude and sign of reactivity coefficients. For these reasons, 23-energy-group, burnup-dependent microscopic cross sections are employed in the REBUS-3/DIF3D model used for evaluating the system. These cross sections were generated with the DRAGON codes using ENDF/B-VI data

[en] An equilibrium cycle method, embodied in the REBUS-3[1] code system, has generally been used in conventional fast reactor design activities. The equilibrium cycle method provides an efficient approach for modeling reactor system, compared to the more traditional non-equilibrium cycle fuel management calculation approach. Recently, the equilibrium analysis method has been utilized for designing Accelerator Transmutation of Waste (ATW)[2,3,4] cores, in which a scattered-reloading fuel management scheme is used. Compared with the conventional fast reactors, the ATW core is significantly different in several aspects since its main mission is to incinerate the transuranic (TRU) fuels. The high burnup non-fertile fuel has large variations in composition and reactivity during its lifetime. Furthermore, a relatively short cycle length is utilized in the ATW design to limit the potentially large reactivity swing over a cycle, and consequently 7 or 8-batch fuel management is usually assumed for a high fuel burnup. The validity of the equilibrium analysis method for the ATW core, therefore, needed to be verified. The main objective of this paper is to assess the validity of the equilibrium analysis method for a Na-cooled ATW core[4], which is an alternative core design of the ATW system under development

[en] 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 ∼ 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 being planned for Cadarache. This new experiment would provide better understanding of GFR neutronic features and will be the basis for the extension of current neutronics code validation domain (particularly, the ERANOS code system in France) to the analysis of GFRs. Experimental planning and decisions are ongoing for ENIGMA. One of the items that have been evaluated is the feasibility of obtaining different flux spectra in the ENIGMA reference configuration, giving the flexibility of simulating a large series of proposed gas-cooled fast systems with harder or softer spectra. In order to achieve this goal it was proposed to use a spectral transition zone in the center region of the ENIGMA core configuration. Another goal of the study is to evaluate the impact of the graphite cross-sections on the performance characteristics of the MASURCA configurations. The work was supported by ANL, through the residence of one of the authors at CEA-Cadarache in 2005. In this report, the impacts of the transition zone on the core physics parameters of the reference ENIGMA configuration are summarized

[en] An advanced gas reactor fuels development and qualification program is being undertaken by the USDOE to address the need for data in support of the licensing operation of the Next Generation Nuclear Power Plant (NGNP). Reactor physics tasks have been defined and are being conducted to support this program. As part of this effort, a preliminary assessment of deterministic lattice codes (WIMS8 and DRAGON) was performed. It was found that the results from these deterministic codes generally agree well with the reference Monte Carlo results. Based on these results, parametric studies were performed with the WIMS8 code to develop an optimum design of the NGNP fuel block. The design goal is to discharge burnup greater than 100 GWd/t and a cycle length of about 18 to 24 months are targeted. As a first step, the effects of fuel kernel size, particle packing fraction, and uranium enrichment have been investigated, while using the NGNP reference design values for the other design parameters such as the fuel compact size and coating thicknesses

[en] 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