The mechanical integrity of fuel pin cladding in a pulsed-beam accelerator driven subcritical reactor

The Accelerator Driven Subcritical Reactor (ADSR) is one of the reactor designs proposed for future nuclear energy production. Interest in the ADSR arises from its enhanced and intrinsic safety characteristics, as well as its potential ability to utilize the large global reserves of thorium and to burn legacy actinide waste from other reactors and decommissioned nuclear weapons. The ADSR concept is based on the coupling of a particle accelerator and a subcritical core by means of a neutron spallation target interface. One of the candidate accelerator technologies receiving increasing attention, the Fixed Field Alternating Gradient (FFAG) accelerator, generates a pulsed proton beam. This paper investigates the impact of pulsed proton beam operation on the mechanical integrity of the fuel pin cladding. A pulsed beam induces repetitive temperature changes in the reactor core which lead to cyclic thermal stresses in the cladding. To perform the thermal analysis aspects of this study a code that couples the neutron kinetics of a subcritical core to a cylindrical geometry heat transfer model was developed. This code, named PTS-ADS, enables temperature variations in the cladding to be calculated. These results are then used to perform thermal fatigue analysis and to predict the stress-life behaviour of the cladding. © 2011 Elsevier Ltd. All rights reserved.

Safety implications of reactivity variations in fast thorium ADSRs

Nuclear power generation offers a reliable, low-impact and large-scale alternative to fossil fuels. However, concerns exist over the safety and sustainability of this method of power production, and it remains unpopular with some governments and pressure groups throughout the world. Fast thorium fuelled accelerator-driven sub-critical reactors (ADSRs) offer a possible route to providing further re-assurance regarding these concerns on account of their properties of enhanced safety through sub-critical operation combined with reduced actinide waste production from the thorium fuel source. The appropriate sub-critical margin at which these reactors should operate is the subject of continued debate. Commercial interests favour a small sub-critical margin in order to minimise the size of the accelerator needed for a given power output, whilst enhanced safety would be better satisfied through larger sub-critical margins to further minimise the possibility of a criticality excursion. Against this background, this paper examines some of the issues affecting reactor safety inherent within thorium fuel sources resulting from the essential Th90232→Th90233→Pa91233→U92233 breeding chain. Differences in the decay half-lives and fission and capture cross-sections of 233Pa and 233U can result in significant changes in the reactivity of the fuel following changes in the reactor power. Reactor operation is represented using a homogeneous lumped fast reactor model that can simulate the evolution of actinides and reactivity variations to first-order accuracy. The reactivity of the fuel is shown to increase significantly following a loss of power to the accelerator. Where the sub-critical operating margins are small this can result in a criticality excursion unless some form of additional intervention is made, for example through the insertion of control rods. © 2012 Elsevier Ltd. All rights reserved.

Alternative methods for surface texturing

Surface texturing has a great potential to improve tribological performance. First, possible texturing methods were identified and classified according to their physical principles. In sequence, some alternative texturing methods are presented. Some of them are already currently used either in industry or in laboratory, and innovations or simplifications are described for them. Others are innovative techniques. Some were explored only tentatively, where basic ideas and simple experimental investigations were developed to check their validity. Others were explored in more detail, so that their practical applicability could be identified. The first texturing method was photochemical texturing using a simple and cheap apparatus. Masking with inkjet printing before chemical etching was also successful to texture metallic samples. A new method involving electrochemical texturing, without the need to previously mask the samples to be textured have been studied in terms of voltage, current, mechanical configuration of the apparatus and electrolyte flushing. Another method aims to generate randomly distributed circular pockets on steel surfaces and involves dispersion of small acid droplets in oil. The final method involves the selective formation of hard areas on a steel surface by locallised diffusion, which should then develop into a texture during wear.

Analysis of high pressure premixed flames using Equivalent Reactor Networks for predicting NOx emissions

This paper describes a computational study of lean premixed high pressure methane-air flames, using Computational Fluid Dynamics (CFD) together with a reactor network approach. A detailed chemical reaction mechanism is employed to predict pollutant concentrations, placing emphasis on nitrogen oxide emissions. The reacting flow field is divided into separate zones in which homogeneity of the physical and chemical conditions prevails. The defined zones are interconnected forming an Equivalent Reactor Network (ERN). Three flames are examined for which experimental data is available. Flame A is characterised by an equivalence ratio of 0.43 while Flames B and C are richer with equivalence ratios of 0.5 and 0.56 respectively. Computations are performed for a range of operating conditions, quantifying the effect in the emitted NOx levels. Model predictions are compared against the available experimental data. Sensitivity analysis is performed to investigate the effect of the network size, in order to define the optimum number of reactors for accurate predictions of the species mass fractions. © 2012 Elsevier Ltd. All rights reserved.

Modeling of SFR cores with Serpent-DYN3D codes sequence

DYN3D reactor dynamics nodal diffusion code was originally developed for the analysis of Light Water Reactors. In this paper, we demonstrate the feasibility of using DYN3D for modeling of fast spectrum reactors. A homogenized cross sections data library was generated using continuous energy Monte-Carlo code Serpent which provides significant modeling flexibility compared with traditional deterministic lattice transport codes and tolerable execution time. A representative sodium cooled fast reactor core was modeled with the Serpent-DYN3D code sequence and the results were compared with those produced by ERANOS code and with a 3D full core Monte-Carlo solution. Very good agreement between the codes was observed for the core integral parameters and power distribution suggesting that the DYN3D code with cross section library generated using Serpent can be reliably used for the analysis of fast reactors. © 2012 Elsevier Ltd. All rights reserved.

Comparison among MCNP-based depletion codes applied to burnup calculations of pebble-bed HTR lattices

The double-heterogeneity characterising pebble-bed high temperature reactors (HTRs) makes Monte Carlo based calculation tools the most suitable for detailed core analyses. These codes can be successfully used to predict the isotopic evolution during irradiation of the fuel of this kind of cores. At the moment, there are many computational systems based on MCNP that are available for performing depletion calculation. All these systems use MCNP to supply problem dependent fluxes and/or microscopic cross sections to the depletion module. This latter then calculates the isotopic evolution of the fuel resolving Bateman's equations. In this paper, a comparative analysis of three different MCNP-based depletion codes is performed: Montburns2.0, MCNPX2.6.0 and BGCore. Monteburns code can be considered as the reference code for HTR calculations, since it has been already verified during HTR-N and HTR-N1 EU project. All calculations have been performed on a reference model representing an infinite lattice of thorium-plutonium fuelled pebbles. The evolution of k-inf as a function of burnup has been compared, as well as the inventory of the important actinides. The k-inf comparison among the codes shows a good agreement during the entire burnup history with the maximum difference lower than 1%. The actinide inventory prediction agrees well. However significant discrepancy in Am and Cm concentrations calculated by MCNPX as compared to those of Monteburns and BGCore has been observed. This is mainly due to different Am-241 (n,γ) branching ratio utilized by the codes. The important advantage of BGCore is its significantly lower execution time required to perform considered depletion calculations. While providing reasonably accurate results BGCore runs depletion problem about two times faster than Monteburns and two to five times faster than MCNPX. © 2009 Elsevier B.V. All rights reserved.

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.

Investigation of fuel assembly design options for high conversion thorium fuel cycle in PWRs

This study explores the basic possibility of achieving a self-sustainable Th-U233 fuel cycle that can be adopted in the current generation of Pressurized Water Reactors. This study outlines some fuel design strategies to achieve (or to approach as closely as possible) a sustainable fuel cycle. Major design tradeoffs in the core design are discussed. Preliminary neutronic analysis performed on the fuel assembly level with BOXER computer code suggests that net breeding of U233 is feasible in principle within a typical PWR operating envelope. However, some reduction in the core power density and/or shorter than typical fuel cycle length would most likely be required in order to achieve such performance.

Effect of95 241Am(n, γ) reaction branching ratio on fuel cycle and reactor design characteristics

The growing interest in innovative reactors and advanced fuel cycle designs requires more accurate prediction of various transuranic actinide concentrations during irradiation or following discharge because of their effect on reactivity or spent-fuel emissions, such as gamma and neutron activity and decay heat. In this respect, many of the important actinides originate from the 241Am(n,γ) reaction, which leads to either the ground or the metastable state of 242Am. The branching ratio for this reaction depends on the incident neutron energy and has very large uncertainty in the current evaluated nuclear data files. This study examines the effect of accounting for the energy dependence of the 241Am(n,γ) reaction branching ratio calculated from different evaluated data files for different reactor and fuel types on the reactivity and concentrations of some important actinides. The results of the study confirm that the uncertainty in knowing the 241Am(n,γ) reaction branching ratio has a negligible effect on the characteristics of conventional light water reactor fuel. However, in advanced reactors with large loadings of actinides in general, and 241Am in particular, the branching ratio data calculated from the different data files may lead to significant differences in the prediction of the fuel criticality and isotopic composition. Moreover, it was found that neutron energy spectrum weighting of the branching ratio in each analyzed case is particularly important and may result in up to a factor of 2 difference in the branching ratio value. Currently, most of the neutronic codes have a single branching ratio value in their data libraries, which is sometimes difficult or impossible to update in accordance with the neutron spectrum shape for the analyzed system.

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.

Fertile-free annular fuel for plutonium recycling

Up to 50% increase in the power density of the existing pressurized water reactor (PWR)-type reactors can be achieved by the use of internally and externally cooled annular fuel geometry. As a result, the accumulated stock-piles of Pu, especially if incorporated infertile-free inert matrix, can be burnt at a substantially higher rate as compared with the conventional mixed oxide-fueled reactors operating at standard power density. In this work, we explore the basic feasibility of a PWR core fully loaded with Pu incorporated infertile-free fuel of annular internally and externally cooled geometry and operating at 150% of nominal power density. We evaluate basic burnable poison designs, fuel management strategies, and reactivity feedback coefficients. The three-dimensional full core neutronic analysis performed with Studsvik Core Management System showed that the design of such a Pu-loaded annular fuel core is feasible but significantly more challenging than the Pu fertile-free core with solid fuel pins operating at nominal power density. The main difficulty arises from the fact that the annular fuel core requires at least 50% higher initial Pu loading in order to maintain the standard fuel cycle length of 18 months. Such a high Pu loading results in hardening of the neutron spectrum and consequent reduction in reactivity worth of all reactivity control mechanisms and, in some cases, positive moderator temperature coefficient (MTC). The use of isotopically enriched Gd and Er burnable poisons was found to be beneficial with respect to maximizing Pu burnup and reducing power peaking factors. Overall, the annular fertile-free Pu-loaded high-power-density core appears to be feasible, although it still has relatively high power peaking and potential for slightly positive MTC at beginning of cycle. However, we estimate that limiting the power density to 140% of the nominal case would assure acceptable core power peaking and negative MTC at all times during the cycle.

The smallest thermal nuclear reactor

There is a growing need for very small nuclear reactors for space applications and as portable high-intensity neutron sources. This technical note investigates the question of what is the smallest possible thermal reactor. It was found that the smallest reactor is a spherically shaped solution of 242mAm(NO3)3 in water. The weight of such a reactor is 4.95 kg with 0.7 kg of 242mAm nuclear fuel. The radius of the reactor in this case is 9.6 cm.

Sensitivity analysis and optimization of the nuclear fuel cycle

A sensitivity study has been conducted to assess the robustness of the conclusions presented in the MIT Fuel Cycle Study. The Once Through Cycle (OTC) is considered as the base-line case, while advanced technologies with fuel recycling characterize the alternative fuel cycles. The options include limited recycling in LWRs and full recycling in fast reactors and in high conversion LWRs. Fast reactor technologies studied include both oxide and metal fueled reactors. The analysis allowed optimization of the fast reactor conversion ratio with respect to desired fuel cycle performance characteristics. The following parameters were found to significantly affect the performance of recycling technologies and their penetration over time: Capacity Factors of the fuel cycle facilities, Spent Fuel Cooling Time, Thermal Reprocessing Introduction Date, and incore and Out-of-core TRU Inventory Requirements for recycling technology. An optimization scheme of the nuclear fuel cycle is proposed. Optimization criteria and metrics of interest for different stakeholders in the fuel cycle (economics, waste management, environmental impact, etc.) are utilized for two different optimization techniques (linear and stochastic). Preliminary results covering single and multi-variable and single and multi-objective optimization demonstrate the viability of the optimization scheme.

A survey of alternative once-through fast reactor core designs

Reprocessing of Light Water Reactor (LWR) spent fuel to recover plutonium or transuranics for use in Sodium cooled Fast Reactors (SFRs) is a distant prospect in the U.S.A. This has motivated our evaluation of potentially cost-effective operation of uranium startup fast reactors (USFRs) in a once-through mode. This review goes beyond findings reported earlier based on a UC fueled MgO reflected SFR to describe a broader parametric study of options. Cores were evaluated for a variety of fuel/coolant/reflector combinations: UC/UZr/UO 2/UN;Na/Pb; MgO/SS/Zr. The challenge is achieving high burnup while minimizing enrichment and respecting both cladding fluence/dpa and reactivity lifetime limits. These parametric studies show that while UC fuel is still the leading contender, UO 2 fuel and ZrH 1.7 moderated metallic fuel are also attractive if UC proves to be otherwise inadequate. Overall, these findings support the conclusion that a competitive fuel cycle cost and uranium utilization compared to LWRs is possible for SFRs operated on a once-through uranium fueled fuel cycle. In addition, eventual transition to TRU recycle mode is studied, as is a small test reactor to demonstrate key features.

The use of Th in HTR: State of the art and implementation in Th/Pu fuel cycles

Nowadays nuclear is the only greenhouse-free source that can appreciably respond to the increasing worldwide energy demand. The use of Thorium in the nuclear energy production may offer some advantages to accomplish this task. Extensive R&D on the thorium fuel cycle has been conducted in many countries around the world. Starting from the current nuclear waste policy, the EU-PUMA project focuses on the potential benefits of using the HTR core as a Pu/MA transmuter. In this paper the following aspects have been analysed: (1) the state-of-the-art of the studies on the use of Th in different reactors, (2) the use of Th in HTRs, with a particular emphasis on Th-Pu fuel cycles, (3) an original assessment of Th-Pu fuel cycles in HTR. Some aspects related to Thorium exploitation were outlined, particularly its suitability for working in pebble-bed HTR in a Th-Pu fuel cycle. The influence of the Th/Pu weight fraction at BOC in a typical HTR pebble was analysed as far as the reactivity trend versus burn-up, the energy produced per Pu mass, and the Pu isotopic composition at EOC are concerned. Although deeper investigations need to be performed in order to draw final conclusions, it is possible to state that some optimized Th percentage in the initial Pu/Th fuel could be suggested on the basis of the aim we are trying to reach. Copyright © 2009 Guido Mazzini et al.