Nuclear fuel cycles: interfaces with the environment

The waste materials generated in the nuclear fuel cycle are very varied,ranging from the tailings arising from mining and processing uranium ore, depleted uranium in a range of chemical forms, to a range of process wastes of differing activities and properties. Indeed, the wastes generated are intimately linked to the options selected in operating the nuclear fuel cycle, most obviously to the management of spent fuel. An open fuel cycle implies the disposal of highly radioactive spent fuel, whereas a closed fuel cycle generates a complex array of waste streams. On the other hand, a closed fuel cycle offers options for waste management, for example reduction in highly active waste volume, decreased radiotoxicity, and removal of fissile material. Many technological options have been proposed or explored, and each brings its own particular mix of wastes and environmental challenges.

Solution Combustion Synthesis of Nanosized Copper Chromite and Its Use as a Burn Rate Modifier in Solid Propellants

Nano sized copper chromite, which is used as a burn rate accelerator for solid propellants, was synthesized by the solution combustion process using citric acid and glycine as fuel. Pure spinel phase copper chromite (CuCr2O4) was synthesized, and the effect of different ratios of Cu-Cr ions in the initial reactant and various calcination temperatures on the final properties of the material were examined. The reaction time for the synthesis with glycine was lower compared to that with citric acid. The synthesized samples from both fuel cycles were characterized by X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS), BET surface area analysis, and scanning electron microscope (SEM). Commercial copper chromite that is currently used in solid propellant formulation was also characterized by the same techniques. XRD analysis shows that the pure spinel phase compound is formed by calcination at 700 degrees C for glycine fuel cycle and between 750 and 800 degrees C for citric acid cycle. XPS results indicate the variation of the oxidation state of copper in the final compound with a change in the Cu-Cr mole ratio. SEM images confirm the formation of nano size spherical shape particles. The variation of BET surface area with calcination temperature was studied for the solution combusted catalyst. Burn rate evaluation of synthesized catalyst was carried out and compared with the commercial catalyst. The comparison between BET surface area and the burn rate depicts that surface area difference caused the variation in burn rate between samples. The reason behind the reduction in surface area and the required modifications in the process are also described.

O Protocolo de Kyoto e o mercado de crédito de carbono: estudo de caso das emissões de gases de efeito estufa da fábrica de combustível nuclear pertencente às Indústrias Nucleares do Brasil no ano de 2008

A presente dissertação discute as questões relacionadas à intensificação das mudanças climáticas por causas antrópicas conforme a evolução no uso dos recursos naturais, inovações nos processos produtivos, transformações econômicas, sociais, culturais, políticas e, especialmente ambientais. Aborda a comercialização dos créditos de carbono através de projetos de Mecanismo de Desenvolvimento Limpo (MDL), um dos mecanismos de flexibilização criados pelo Protocolo de Kyoto. No contexto de mudanças climáticas, uma matriz energética que utilize fontes de energia que não emitam gases causadores do efeito estufa (GEE) se mostra uma importante estratégia de desenvolvimento sustentável. Sob essa perspectiva, a energia nucleoelétrica é apresentada como uma alternativa viável aos combustíveis fósseis, considerando que esta é uma energia limpa e compatível com a perspectiva de desenvolvimento sustentável. A Fábrica de Combustível Nuclear (FCN), localizada em Resende (Rio de Janeiro), pertencente às Indústrias Nucleares do Brasil (INB), é um conjunto de sofisticadas fábricas nas quais se processam etapas importantes do ciclo do combustível nuclear. Na FCN, o Centro Zoobotânico realiza a gestão das atividades voltadas para a conservação da natureza tais como o Programa de Recuperação de Mata Ciliar, Reflorestamento e Fauna. O Relatório de inventário das emissões diretas e indiretas de GEE da FCN, elaborado pela INB para o ano de 2008, permite a auto-avaliação da empresa, retratando a preocupação corporativa com as questões relativas às mudanças climáticas. Segundo este Relatório, o total de emissões de GEE quantificado corresponde a 12,14% da capacidade total de sequestro de dióxido de carbono, no período de Janeiro a Dezembro de 2008. A proteção de florestas e a plantação de árvores são componentes essenciais de qualquer estratégia global para mitigação da mudança climática, e a participação da INB no mercado de crédito de carbono pode proporcionar externalidades positivas, tais como ganhos de imagem, adequação a padrões ambientais e melhoria do relacionamento com a sociedade.

Near-complete transuranic waste incineration in a thorium fuelled pressurised water reactor

The production of long-lived transuranic (TRU) waste is a major disadvantage of fission-based nuclear power. Incineration, and virtual elimination, of waste stockpiles is possible in a thorium (Th) fuelled critical or subcritical fast reactor. Fuel cycles producing a net decrease in TRUs are possible in conventional pressurised water reactors (PWRs). However, minor actinides (MAs) have a detrimental effect on reactivity and stability, ultimately limiting the quality and quantity of waste that can be incinerated. In this paper, we propose using a thorium-retained-actinides fuel cycle in PWRs, where the reactor is fuelled with a mixture of thorium and TRU waste, and after discharge all actinides are reprocessed and returned to the reactor. To investigate the feasibility and performance of this fuel cycle an assembly-level analysis for a one-batch reloading strategy was completed over 125 years of operation using WIMS 9. This one-batch analysis was performed for simplicity, but allowed an indicative assessment of the performance of a four-batch fuel management strategy. The build-up of 233U in the reactor allowed continued reactive and stable operation, until all significant actinide populations had reached pseudo-equilibrium in the reactor. It was therefore possible to achieve near-complete transuranic waste incineration, even for fuels with significant MA content. The average incineration rate was initially around 330 kg per GW th year and tended towards 250 kg per GW th year over several decades: a performance comparable to that achieved in a fast reactor. Using multiple batch fuel management, competitive or improved end-of-cycle burn-up appears achievable. The void coefficient (VC), moderator temperature coefficient (MTC) and Doppler coefficient remained negative. The quantity of soluble boron required for a fixed fuel cycle length was comparable to that for enriched uranium fuel, and acceptable amounts can be added without causing a positive VC or MTC. This analysis is limited by the consideration of a single fuel assembly, and it will be necessary to perform a full core coupled neutronic-thermal-hydraulic analysis to determine if the design in its current form is feasible. In particular, the potential for positive VCs if the core is highly or locally voided is a cause for concern. However, these results provide a compelling case for further work on concept feasibility and fuel management, which is in progress. © 2011 Elsevier Ltd. All rights reserved.

Numerical stability of the predictor-corrector method in Monte Carlo burnup calculations of critical reactors

Monte Carlo burnup codes use various schemes to solve the coupled criticality and burnup equations. Previous studies have shown that the simplest methods, such as the beginning-of-step and middle-of-step constant flux approximations, are numerically unstable in fuel cycle calculations of critical reactors. Here we show that even the predictor-corrector methods that are implemented in established Monte Carlo burnup codes can be numerically unstable in cycle calculations of large systems. © 2013 Elsevier Ltd. All rights reserved.

Detailed design of242mAm breeding in pressurized water reactors

There is growing interest in the use of 242mAm as a nuclear fuel. Because of its very high thermal fission cross section and its large number of neutrons released per fission, it can be used for various unique applications, such as space propulsion, medical applications, and compact energy sources. Since the thermal absorption cross section of 242mAm is very high, the best way to obtain 242mAm is by the capture of fast or epithermal neutrons in 241Am. However, fast spectrum reactors are not readily available. In this paper, we explore the possibility of producing 242mAm in existing pressurized water reactors (PWRs) with minimal interference in reactor performance. As suggested in previous studies on the subject, the 242mAm breeding targets are shielded with strong thermal absorbers in order to suppress the thermal neutron flux that causes 242mAm destruction. Since 242mAm enrichment within the Am target mainly depends on the neutron energy distribution, which in turn depends on the Am target thickness and on the neutron filter cutoff energy (thermal absorber type), this unique Am target design was developed. In our study, Cd, Sm, and Gd were considered as thermal neutron filters, as suggested by Cesana et al. The most favorable results were obtained by irradiating Am targets covered either with Gd or Cd. In these cases, up to 8.65% enrichment of 242mAm is obtained after 4.5 yr (three successive PWR fuel cycles) of irradiation. It was also found that significant quantities [up to 1.3 kg/GW (electric)-yr] of 242mAm can be obtained in PWR reactors without notable interference with reactor performance. However, in order to maintain the original fuel cycle length, the enrichment of the driver (UO2) fuel must be increased by ∼1%, raised from the conventional 4.5 to 5.5%, depending on the thermal neutron filter used. The most important reactivity feedback coefficients for fuel assemblies containing the 242mAm breeding targets were evaluated and found to be close to those of a standard PWR. Another product of neutron capture in the 241Am reaction is 238Pu. It was found that in a typical 1000 MW (electric) PWR core with one-third of the fuel assemblies containing 241Am targets, up to 15.1 kg of 238Pu enriched to 80% can be produced per year.

The potential use of 241Am as proliferation resistant burnable poison in PWRs

This paper discusses the use of 241Am as proliferation resistant burnable poison for light water reactors. Homogeneous addition of small (as little as 0.12%) amounts of 241Am to the conventional light water reactor fuel results in significant increase in 238Pu/Pu ratio in the discharged fuel improving its proliferation resistance. Moreover, 241Am, admixed to the fuel, acts as burnable absorber allowing for substantial reduction in conventional reactivity control means without a notable fuel cycle length penalty. This is possible due to favorable characteristics of 241Am transmutation chain. The fuel cycle length penalty of introducing 241Am into the core is evaluated and discussed, as well as the impact of He production in the fuel pins and degradation of reactivity feedback coefficients. Proliferation resistance and reactivity control features related to the use of 241Am are compared to those of using 237Np, which has also been suggested as an additive to the conventional fuel in order to improve its proliferation resistance. It was found that 241Am admixture is more favorable than 237Np admixture because of the smaller fuel cycle length penalty and higher burnable poison savings. Addition of either 237Np or 241Am would provide substantial but not ultimate protection from misuse of Pu originating in the spent fuel from the commercial power reactors. Therefore, the benefits from application of the concept would have to be carefully evaluated against the additional costs and proliferation risks associated with manufacturing of 237Np or 241Am doped fuel. Although this work concerns specifically with PWRs, the conclusions could also be applied to BWRs and, to some extent, to other thermal spectrum reactor types. © 2009 Elsevier Ltd. All rights reserved.

Cross-comparison of fast reactor concepts with various coolants

Four fast reactor concepts using lead (LFR), liquid salt, NaCl-KCl-MgCl2 (LSFR), sodium (SFR), and supercritical CO2 (GFR) coolants are compared. Since economy of scale and power conversion system compactness are the same by virtue of the consistent 2400 MWt rating and use of the S-CO2 power conversion system, the achievable plant thermal efficiency, core power density and core specific powers become the dominant factors. The potential to achieve the highest efficiency among the four reactor concepts can be ranked from highest to lowest as follows: (1) GFR, (2) LFR and LSFR, and (3) SFR. Both the lead- and salt-cooled designs achieve about 30% higher power density than the gas-cooled reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor. Fuel cycle costs are favored for the sodium reactor by virtue of its high specific power of 65 kW/kgHM compared to the lead, salt and gas reactor values of 45, 35, and 21 kW/kgHM, respectively. In terms of safety, all concepts can be designed to accommodate the unprotected limiting accidents through passive means in a self-controllable manner. However, it does not seem to be a preferable option for the GFR where the active or semi-passive approach will likely result in a more economic and reliable plant. Lead coolant with its superior neutronic characteristics and the smallest coolant temperature reactivity coefficient is easiest to design for self-controllability, while the LSFR requires special reactivity devices to overcome its large positive coolant temperature coefficient. The GFR required a special core design using BeO diluent and a supercritical CO2 reflector to achieve negative coolant void worth-one of the conditions necessary for inherent shutdown following large LOCA. Protected accidents need to be given special attention in the LSFR and LFR due to the small margin to freezing of their coolants, and to a lesser extent in the SFR. © 2009 Elsevier B.V. All rights reserved.

The potential use of 241Am as proliferation resistant burnable poison

This paper discusses the use of 141Am as proliferation resistant burnable poison for light water reactors. Homogeneous addition of small (less than 1 %) amounts of 241Am to the conventional LWR fuel results in significant increase in 238Pu/Pu ratio in the discharged fuel improving its proliferation resistance. Moreover, 241Am, admixed to the fuel, acts as burnable absorber allowing for substantial reduction in conventional reactivity control means without notable fuel cycle length penalty. This is possible due to favourable characteristics of 241Am transmutation chain. The fuel cycle length penalty of introducing 241Am into the core is evaluated and discussed, as well as the impact of He production in the fuel pins and degradation of reactivity feedback coefficients. Proliferation resistance and reactivity control features related to the use of 241Am are compared to those of using 237Np, which has also been suggested as an additive to the conventional fuel in order to improve its proliferation resistance. It was found that 241Am admixture is more favourable than 237Np admixture because of the smaller fuel cycle length penalty and higher burnable poison savings.

The effect of Ln(III)/An(III) separation on TRU multi-recycling in CONFU assembly

This work examines the basic feasibility of the net-zero-balance TRU multi-recycling concept in which trivalent lanthanide fission products (Ln(III) ) are not separated from trivalent actinides (An(III)). The TRU together with Eu and Gd isotopes are recycled in a standard PWR using Combined Non-Fertile and UO2 (CONFU) assembly design. The assembly assumes a heterogeneous structure where about 20% of U02 fuel pins on the assembly periphery are replaced with Inert Matrix Fuel (IMF) pins hosting TRU, Gd, and Eu generated in the previous cycles. The 2-D neutronic analysis show potential feasibility of Ln / An recycling in PWR using CONFU assembly. Recycling of Ln reduces the fuel cycle length by about 30 effective full power days (EFPD) and TRU destruction efficiency by about 5%. Power peaking factors and reactivity feedback coefficients are close to those of CONFU assembly without Ln recycling.

Efficient generation of one-group cross sections for coupled Monte Carlo depletion calculations

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 onegroup 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.

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.

Optimization of burnable poison design for Pu incineration in fully fertile free PWR core

The design challenges of the fertile-free based fuel (FFF) can be addressed by careful and elaborate use of burnable poisons (BP). Practical fully FFF core design for PWR reactor has been reported in the past [1]. However, the burnable poison option used in the design resulted in significant end of cycle reactivity penalty due to incomplete BP depletion. Consequently, excessive Pu loading were required to maintain the target fuel cycle length, which in turn decreased the Pu burning efficiency. A systematic evaluation of commercially available BP materials in all configurations currently used in PWRs is the main objective of this work. The BP materials considered are Boron, Gd, Er, and Hf. The BP geometries were based on Wet Annular Burnable Absorber (WABA), Integral Fuel Burnable Absorber (IFBA), and Homogeneous poison/fuel mixtures. Several most promising combinations of BP designs were selected for the full core 3D simulation. All major core performance parameters for the analyzed cases are very close to those of a standard PWR with conventional UO2 fuel including possibility of reactivity control, power peaking factors, and cycle length. The MTC of all FFF cores was found at the full power conditions at all times and very close to that of the UO2 core. The Doppler coefficient of the FFF cores is also negative but somewhat lower in magnitude compared to UO2 core. The soluble boron worth of the FFF cores was calculated to be lower than that of the UO2 core by about a factor of two, which still allows the core reactivity control with acceptable soluble boron concentrations. The main conclusion of this work is that judicial application of burnable poisons for fertile free fuel has a potential to produce a core design with performance characteristics close to those of the reference PWR core with conventional UO2 fuel.

Recycle of transuranium elements in light water reactors for reduction of geological storage requirements

This paper presents results of a feasibility study aimed at developing a zero-transuranic-discharge fuel cycle based on the U-Th-TRU ternary cycle. The design objective is to find a fuel composition (mixture of thorium, enriched uranium, and recycled transuranic components) and fuel management strategy resulting in an equilibrium charge-discharge mass flow. In such a fuel cycle scheme, the quantity and isotopic vector of the transuranium (TRU) component is identical at the charge and discharge time points, thus allowing the whole amount of the TRU at the end of the fuel irradiation period to be separated and reloaded into the following cycle. The TRU reprocessing activity losses are the only waste stream that will require permanent geological storage, virtually eliminating the long-term radiological waste of the commercial nuclear fuel cycle. A detailed three-dimensional full pressurized water reactor (PWR) core model was used to analyze the proposed fuel composition and management strategy. The results demonstrate the neutronic feasibility of the fuel cycle with zero-TRU discharge. The amount of TRU and enriched uranium loaded reach equilibrium after about four TRU recycles. The reactivity coefficients were found to be within a range typical for a reference PWR core. The soluble boron worth is reduced by a factor of ∼2 from a typical PWR value. Nevertheless, the results indicate the feasibility of an 18-month fuel cycle design with an acceptable beginning-of-cycle soluble boron concentration even without application of burnable poisons.