M. Segev

A knowledge-based system for optimization of fuel reload configurations

The authors discuss a knowledge-based production system developed for generating optimal fuel reload configurations. The system was based on a heuristic search method and implemented in Common Lisp programming language. The knowledge base embodied the reactor physics, reactor operations, and a general approach to fuel management strategy. The data base included a description of the physical system involved, i.e., the core geometry and fuel storage. The fifth cycle of the Three Mile Island Unit 1 pressurized water reactor was chosen as a test case. Application of the system to the test case revealed a self-learning process by which a relatively large number of near-optimal configurations were discovered. Several selected solutions were subjected to detailed analysis and demonstrated excellent performance. To summarize, applicability of the proposed heuristic search method in the domain of nuclear fuel management was proved unequivocally.

A Pressurized Water Reactor Plutonium Incinerator Based on Thorium Fuel and Seed-Blanket Assembly Geometry

Abstract A pressurized water reactor (PWR) fuel cycle is proposed, whose purpose is the elimination and degradation of weapons-grade plutonium. This Radkowsky thorium-fuel Pu incinerator (RTPI) cycle is based on a core and assemblies retrofittable to a Westinghouse-type PWR. The RTPI assembly, however, is a seed-blanket unit. The seed is supercritical, loaded with Pu-Zr alloy as fuel in a high moderator-to-fuel ratio configuration. The blanket is subcritical, loaded mainly with ThO2, generating and burning 233U in situ. Blankets are loaded once every 6 yr. The seed fuel management scheme is based on three batches, with one-third of the seed modules replaced every year. The core generates 1100 MW(electric). Equilibrium conditions are achieved with the second seed loading. For equilibrium conditions, the annual average of disposed (loaded) Pu is 1210 kg, of which 702 kg are completely eliminated, and 508 kg are discharged, but with significantly degraded isotopics (i.e., with a high percentage of even mass isotopes). Spontaneous fissions per second in a gram of this degraded Pu are ~500, resulting in significantly increased proliferation resistance. Every 6 yr the blanket discharge contains 780 kg of 233U (including 233Pa) and 36 kg of 235U. However, the blankets are initially loaded with an amount of natural uranium selected such that these U fissile isotopes constitute only 12% of the total U discharge, a percentage equivalent to 20% 235U enrichment; hence, both the discharged uranium isotopics satisfy proliferation-resistant criteria. The RTPI control variables, namely, the moderator temperature coefficient, the reactivity per ppm boron, and the control rods worth, are about equal to those of a PWR. The RTPI spent-fuel stockpile ingestion toxicity over a period of ten million years is about the same as the counterpart toxicities of a regular, or a mixed-oxide (MOX), PWR. Compared with known PWR MOX variants, the RTPI is, per 1000 MW(electric) and per annum, a significantly more efficient incinerator of weapons-grade plutonium.

A test of main stream pin power reconstruction methods

Preliminary to implementing a pin power reconstruction scheme in the nodal core calculations of the ELCOS system, the main stream methods and elements thereof were tested against fine mesh calculations of a number of benchmark small cores consisting of uranium, controlled uranium, and mixed-oxide assemblies. Overall, the results do not clearly favor one of the methods. However, test details conduce the authors to prefer the 32-term expansion for corner-point fluxes over their determination by the separability assumption, and the 21-term expansion of the intranodal flux over the 13-term expansion. There is little difference whether the factorization of the pin power distribution into global and form factors is imposed on the group fluxes or on the power. Data transfers and matrix inversions connected with the many-term flux expansions slow down the nodal calculation. This condition may be alleviated in some cases by an approximation leading to fewer matrix inversions.

BGUCORE - a computational system for neutronic analysis of a pressurized water reactor core

The independently developed and verified computational system BGUCORE for the neutronic analysis of pressurized water reactor cores is introduced. The basic methodology adopted generates cross-section libaries for each fuel type as functions of burnup and soluble boron concentrations. These cross sections are arranged as a two-dimensional matrix of sets, each set corresponding to a particular burnup/boron pair of coordinates. The two-dimensional diffusion analysis of the reactor core utilizes the pregenerated libraries by interpolating between burnup and boron entry points. The present system is especially designed for the analysis of cores with burnable poisons. Such cores are characterized by strong heterogeneity and selfshielding effects. Detailed benchmark calculations, performed for cycle 1 of the Zion 2 power station, validate the performance of the BGUCORE system. Further development of the system, aimed at creating a comprehensive design and fuel cycle analysis tool, includes a three-dimensional representation of the core and thermohydraulic modules.

A WIMS-based calculational route for pebble-bed fuel

A WIMS-based calculational route for pebble-bed fuel has been established. An outstanding advantage of the WIMS code is its integrated route from basic lattice data to burnup-dependent lattice cros...

Applied Resonance Equivalence

AbstractReactor cores are modeled as double heterogeneous lattices. The geometrical cross section for such a lattice is given bywhereGenerally, β < 1, and the effect of double heterogeneity is typically of the order of 5 mk, in comparison with single heterogeneity.

An estimation of the energy generation potential of proton-driven subcritical ThO2 lattices

Interest in generating energy with thorium fuel has increased lately as a result of the activation of subcritical ThO{sub 2} lattices by accelerated protons. A tight, ThO{sub 2} water-cooled lattice has been proposed to generate 200 MW (thermal) with 1.5 GeV protons in a current of {approximately}7 mA. A tight-latticed core, consisting of a ThO{sub 2}/{sup 233}UO{sub 2} seed embedded in a large ThO{sub 2} blanket, has been proposed to generate 2400 MW (thermal) with 1.0-GeV protons in a current of 20 mA. A consistent detailed analysis of these two energy amplifiers, carried out with the HERMES, MCNP, KORIGEN, WIMS, and BOXER codes, results in performances inferior to those claimed. The net power generated will be one-fourth of that claimed for the former and 1/2.5 of that claimed for the latter.

Analytical method for processing neutron multigroup transfer cross sections

In this paper an approach is summarized for developing a full analytical method for the generation of laboratory (lab) coordinate system multigroup transfer cross sections of elastic and discrete-level inelastic scatterings of neutrons, where the angular distribution data of the scattered neutrons are given as coefficients of truncated Legendre polynomial expansions in the center-of-mass (c.m.) coordinate system. In the kernel form of the multigroup approximation, fluxes, cross sections, and angular data are left outside the integration signs of the transfer cross-section expression. Then, the integrand is a four-index kernel the source and sink energy groups and the Legendre polynomials in the c.m. and lab systems each contributing one index integrated over the source and sink groups. In the method introduced, the double integration on the neutron pre- and postscattering energies, in these two groups, is carried out analytically.