A. Galperin

The Nonproliferative Light Water Thorium Reactor: A New Approach to Light Water Reactor Core Technology

AbstractThe nonproliferative light water thorium technology, also known as RTF (Radkowsky thorium fuel), provides a new approach to light water reactor core design. An RTF core is completely nonpro...

Thorium fuel for light water reactors-Reducing proliferation potential of nuclear power fuel cycle

The proliferation potential of the light water reactor fuel cycle may be significantly reduced by utilization of thorium as a fertile component of the nuclear fuel. The main challenge of thorium utilization is to design a core and a fuel cycle, which would be proliferation‐resistant and economically feasible. This challenge is met by the Radkowsky Thorium Reactor (RTR) concept presented in this paper. So far the concept has been applied to a Russian design of a 1,000 MWe pressurized water reactor, known as a WER‐1000, and designated as WERT. The following are the main results of the preliminary reference design: The amount of plutonium contained in the RTR spent fuel stockpile is reduced by 80 percent in comparison with a WER of a current design. The isotopic composition of the RTR‐Pu greatly increases the probability of preini‐tiation and yield degradation of a nuclear explosion. An extremely large Pu‐238 content causes correspondingly large heat emission, which would complicate the design of an explosiv...

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.

Efficient Generation of One-Group Cross Sections for Coupled Monte Carlo Depletion Calculations

Abstract Coupled Monte Carlo depletion systems provide a versatile and an accurate tool for analyzing advanced thermal and fast reactor designs for a variety of fuel compositions and geometries. The main drawback of Monte Carlo-based systems is a long calculation time imposing significant restrictions on the complexity and amount of design-oriented calculations. This paper presents an alternative approach to interfacing the Monte Carlo and depletion modules aimed at addressing this problem. The main idea is to calculate the one-group cross sections for all relevant isotopes required by the depletion module in a separate module external to Monte Carlo calculations. Thus, the Monte Carlo module will produce the criticality and neutron spectrum only, without tallying of the individual isotope reaction rates. The one-group cross section for all isotopes will be generated in a separate module by collapsing a universal multigroup (MG) cross-section library using the Monte Carlo calculated flux. Here, the term “universal” means that a single MG cross-section set will be applicable for all reactor systems and is independent of reactor characteristics such as a neutron spectrum; fuel composition; and fuel cell, assembly, and core geometries. This approach was originally proposed by Haeck et al. and implemented in the ALEPH code. Implementation of the proposed approach to Monte Carlo burnup interfacing was carried out through the BGCORE system. One-group cross sections generated by the BGCORE system were compared with those tallied directly by the MCNP code. Analysis of this comparison was carried out and led to the conclusion that in order to achieve the accuracy required for a reliable core and fuel cycle analysis, accounting for the background cross section (σ0) in the unresolved resonance energy region is essential. An extension of the one-group cross-section generation model was implemented and tested by tabulating and interpolating by a simplified σ0 model. A significant improvement of the one-group cross-section accuracy was demonstrated.

Use of Thorium in Light Water Reactors

Abstract Thorium-based fuels can be used to reduce concerns related to the proliferation potential and waste disposal of the conventional light water reactor (LWR) uranium fuel cycle. The main sources of proliferation potential and radiotoxicity are the plutonium and higher actinides generated during the burnup of standard LWR fuel. A significant reduction in the quantity and quality of the generated Pu can be achieved by replacing the 238U fertile component of conventional low-enriched uranium fuel by 232Th. Thorium can also be used as a way to manage the growth of plutonium stockpiles by burning plutonium, or achieving a net-zero transuranic production, sustainable recycle scenario. This paper summarizes some of the results of recent studies of the performance of thorium-based fuels. It is concluded that the use of heterogeneous U-Th fuel provides higher neutronic potential than a homogeneous fuel. However, in the former case, the uranium portion of the fuel operates at a higher power density, and care is needed to meet the thermal margins and address the higher-burnup implications. In macroheterogeneous designs, the U-Th fuel can yield reduced spent-fuel volume, toxicity, and decay heat. The main advantage of Pu-Th oxide over mixed oxide is better void reactivity behavior even for undermoderated designs, and increased burnup of Pu.

Utilization of light water reactors for plutonium incineration

Abstract In this work a potential of incineration of excess Pu in LWRs is investigated. In order to maintain the economic viability of the Pu incineration option it should be carried out by the existing power plants without additional investment for plant modifications. Design variations are reduced to the fuel cycle optimization, i.e. fuel composition may be varied to achieve optimal Pu destruction. Fuel mixtures considered in this work were based either on uranium or thorium fertile materials and Pu as a fissile component. The slightly enriched U fuel cycle for a typical pressurized water reactor was considered as a reference case. The Pu content of all fuels was adjusted to assure the identical cycle length and discharged burnup values. An equilibrium cycle was simulated by performing cluster burnup calculations. The material composition data for the whole core was estimated based on the core, fuel and cycle parameters. The annual production of Pu of a standard PWR with 1100 MWe output is about 298 kg. The same core completely loaded with the MOX fuel is estimated to consume 474 kg of Pu, mainly fissile isotopes. The MOX-239 fuel type (pure PU-239) shows a potential to reduce the initial total Pu inventory by 220 kg/year and fissile Pu inventory by 420 kg/year. The following two fuel types: TMOX and TMOX-239 are based on Th-232 as a fertile component of the fuel, instead of U-238. The amount of Pu destroyed per year for both cases is significantly higher than that of U-based fuels. Especially impressive is the reduction in fissile Pu inventory: more than 900 kg/year. The safety related reactivity coefficients were found negative, which indicates that the basic behaviour of a reactor core utilizing Pu and Pu-Th based fuel types will be quite similar to that of a standard PWR core utilizing slightly enriched U fuel. It was also found that the reactivity control of a core based on Pu fuel as a fissile component will be more difficult due to a reduced reactivity worth of the soluble boron and control rod control mechanisms.

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.

Modelling and verification of the PWR burnable poison designs by ELCOS code system

Abstract A PWR core of a current design was used as a test problem for a verification of the ELCOS computational system. The chosen core was relatively complex, including 11 fuel types, differing by fuel enrichment and Burnable Poison (BP) design. Two main BP designs were modelled: WABA and IFBA as well as a combination of both. Results of the calculations were compared with utility data, and demonstrated the adequacy of the ELCOS system. The modelling of the BP designs was carried out by performing a series of calculations with different time-step lengths and poison spatial representation. It was found that relatively simple models are sufficiently accurate in simulating core behaviour with accuracy of about 1 mK. Comparison of the depletion rate of two BP designs indicated potential benefits in using a combination of both designs in a single fuel assembly.

Fertile-free fuels in pressurized water reactors : Design challenges and solutions

This paper investigates the basic feasibility of using reactor-grade Pu in fertile-free fuel (FFF) matrix in pressurized water reactors (PWRs). Several important issues were investigated in this work: the Pu loading required to achieve a specific interrefueling interval, the impact of inert matrix composition on reactivity constrained length of cycle, and the potential of utilizing burnable poisons (BPs) to alleviate degradation of the reactivity control mechanism and temperature coefficients Although the subject was addressed in the past, no systematic approach for assessment of BP utilization in FFF cores was published. In this work, we examine all commercially available BP materials in all geometrical arrangements currently used by the nuclear industry with regards to their potential to alleviate the problems associated with the use of FFF in PWRs. The recently proposed MgO-ZrO2 solid-state solution fuel matrix, which appears to be very promising in terms of thermal properties and radiation damage resistance, was used as a reference matrix material in this work. The neutronic impact of the relative amounts of MgO and ZrO2 in the matrix were also studied. The analysis was performed with a neutron transport and fuel assembly burnup code BOXER. A modified linear reactivity model was applied to the two-dimensional single fuel assembly results to approximate the full core characteristics. Based on the results of the performed analyses, the Pu-loaded FFF core demonstrated potential feasibility to be used in existing PWRs. Major FFF core design problems may be significantly mitigated through the correct choice of BP design. It was found that a combination of BP materials and geometries may be required to meet all FFF design goals. The use of enriched (in most effective isotope) BPs, such as 167Er and 157Gd, may further improve the BP effectiveness and reduce the fuel cycle length penalty associated with their use.

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.