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The optimization of a nuclear energy system’s performance requires an integrated consideration of multiple design goals - sustainability, safety and reliability (S&R), proliferation resistance and ...

An evaluation of the Compact Nuclear Power Source (CNPS) experiments conducted at Los Alamos National Laboratory in the 1980s has been done using information available in the open literature. The MCNP4C Monte Carlo results for critical test configurations are in good agreement with the experimental values; the k{sub eff} values are generally within 0.5% of the experimental values. The calculated total and differential rod worths and material worths were also found generally close to experimental values. These good results motivated the utilization of the experimental test data for the specification of two- and three-dimensional numerical benchmark cases that could be used for the verification and validation of core physics codes developed for Very High Temperature Reactor (VHTR) analysis, particularly the deterministic lattice and whole-core physics codes. To define the benchmark cases, the irregular arrangement of channels in the actual CNPS core was simplified to a regular Cartesian geometry arrangement in the benchmark cases, while preserving the important neutronics characteristics of the CNPS. The results of deterministic calculations using the HELIOS/DIF3D code package were compared to MCNP4C results to show the usefulness of the numerical benchmark cases.

A systematic study on long-lived fission products (LLFPs) transmutation has been performed with the aim of devising an optimal strategy for their transmutation in critical or subcritical reactor sy...

The Accelerator Transmutation of Waste (ATW) concept has been proposed as a transuranics (TRU) (and long-lived fission product) incinerator for processing the 87,000 metric tonnes of Light Water Reactor (LWR) used fuel which will have been generated by the time the currently deployed fleet of commercial reactors in the US reach the end of their licensed lifetime. The ATW is proposed to separate the uranium from the transuranics and fission products in the LWR used fuel, to fission the transuranics, to send the LWR and ATW generated fission products to the geologic repository and to send the uranium to either a low level waste disposal site or to save it for future use. The heat liberated in fissioning the transuranics would be converted to electricity and sold to partially offset the cost of ATW construction and operations. Options for incineration of long-lived fission products are under evaluation. A six-year science-based program of ATW trade and system studies was initiated in the US FY 2000 to achieve two main purposes: (1) “to evaluate ATW within the framework of nonproliferation, waste management, and economic considerations,” and (2) “to evaluate the efficacy of the numerous technical options for ATW system configuration.” This paper summarizes the results from neutronics and thermal/hydraulics trade studies which were completed at Argonne National Laboratory during the first year of the program. Core designs were developed for Pb-Bi eutectic (LBE) cooled and Na cooled 840 MWth fast spectrum transmuter designs employing recycle. Additionally, neutronics analyses were performed at Argonne for a He cooled 600 MWth hybrid thermal and fast core design proposed by General Atomics Co. which runs critical for 34 and subcritical for 14 of its four year once-thru burn cycle. The mass flows and the ultimate loss of transuranic isotopes to the waste stream per unit of heat generated during transmutation have been calculated on a consistent basis and are compared. (Long-lived fission product incineration has not been considered in the studies reported here.)

The results of blanket design studies for a lead-bismuth eutectic (LBE)-cooled accelerator transmutation of waste system are presented. These studies focused primarily on achieving two important an...

A numerical method for directly computing the mathematical adjoint flux moments and partial currents for the hexagonal-Z geometry interface current nodal formulation in the DIF3D code is described. The new scheme is developed as an alternative to an existing scheme that employs a similarity transformation of the physical adjoint solution to compute the mathematical adjoint. Whereas the existing scheme is rigorous only when the flat transverse-leakage approximation is employed, this new scheme is exact for all leakage approximations in the DIF3D nodal method. in the new scheme, adjoint nodal equations whose form is very similar to that of the forward nodal equations are derived by employing linear combinations of the adjoint partial currents as computational unknowns in the adjoint equations. This enables the use of the forward solution algorithm with only minor modifications for solving the mathematical adjoint equations. By using the new scheme as a reference method, it is shown numerically that while the results computed with the existing scheme are approximate, they are sufficiently accurate for calculations of global and local reactivity changes resulting from coolant voiding in a liquid metal reactor.

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope of the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended tomore » be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.« less

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope of the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended tomore » be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.« less

A reconstruction method is developed for recovering pin burnup characteristics from fuel cycle calculations performed in hexagonal-z geometry using the nodal diffusion option of the DIF3D/REBUS-3 c...

The uncertainty in the burnup reactivity swing {sigma}k{sub b} attributable to nuclear data uncertainties is analyzed using depletion-dependent sensitivity coefficients for single- and multicycle equilibrium depletion. Four systems are analyzed with design features that encompass many of the design options considered for current U.S. advanced liquid-metal reactor cores. These systems, while characterized by very different {sigma}k{sub b} values in the range from {minus}0.22 to 3.87% {Delta}k, exhibit much smaller differences in their {sigma}k{sub b} uncertainties, which range from 0.18 to 0.33% {Delta}k. The {sigma}k{sub b} uncertainties depend primarily on the design choices of core size and fissile fuel type, as well as whether the analysis represents multicycle effects. For all reactors analyzed, the burnup swing uncertainty is dominated by the {sup 238}U capture reaction. In this paper the potential for reducing uncertainties by a factor of 3 by use of available integral experiment results is demonstrated.

Systematic analyses of alternative methods for reducing the sodium void worth for plutonium-fueled liquid-metal reactors (LMRs) have been performed. The focus is on core designs of recent interest ...

Nuclear data needs can play an important role for innovative nuclear systems. However, in order to establish priority items, a systematic sensitivity/uncertainty analysis must be performed. Same selected examples will be discussed in this paper.

A diffusion theory method is developed for synthetic acceleration of nodal Sn calculations in multidimensional Cartesian geometries. The diffusion model is derived from the spatially continuous dif...

A consistently formulated differencing approach is applied to the diffusion-synthetic acceleration of discrete ordinates calculations based on various spatial differencing schemes. The diffusion ''coupling'' equations derived for each scheme are contrasted to conventional coupling relations and are shown to permit derivation of either point- or box-centered diffusion difference equations. The resulting difference equations are shown to be mathematically equivalent, in slab geometry, to equations derived by applying Larsen's four-step procedure to the S/sub 2/ equations. Fourier stability analysis of the acceleration method applied to slab model problems is used to demonstrate that, for any S/sub n/ differencing scheme (a) the upper bound on the spectral radius of the method occurs in the fine-mesh limit and equals that of the spatially continuous case (0.22466), and (b) the spectral radius decreases with increasing mesh size to an asymptotic value