An investigation into the residual stresses in an aluminium 2024 test weld

This paper uses finite element (FE) analysis to examine the residual stresses generated during the TIG welding of aluminium aerospace alloys. It also looks at whether such an approach could be useful for evaluating the effectiveness of various residual stress control techniques. However, such simulations cannot be founded in a vacuum. They require accurate measurements to refine and validate them. The unique aspect of this work is that two powerful engineering techniques are combined: FE modelling and neutron diffraction. Weld trials were performed and the direct measurement of residual strain made using the ENGIN neutron diffraction strain scanning facility. The predicted results show an excellent agreement with experimental values. Finally this model is used to simulate a weld made using a "Low Stress No Distortion" (LSND) technique. Although the stress reduction predicted is only moderate, the study suggests the approach to be a quick and efficient means of optimising such techniques.

Spoilage bacteria of Penaeus indicus

Bacteria isolated from raw (untreated and unprocessed) prawn (Penaeus indicus) stored at 28±2°C, 4°C and-18°C were tested for spoilage potential, namely, production of protease, lipase, amylase, reduction of trimethylamineoxide (TMAO) to trimethylamine (TMA), production of off odours from flesh broth and halo zone around the colony grown on flesh agar. About 63 % of the total isolates tested were potential spoilers. Members of Vibrio, Pseudomonas and Acinetobacter were found to be dominant potential spoilers at all temperatures.

Properties of tetrahedral amorphous carbon prepared by vacuum arc deposition

The structural, optical, electrical and physical properties of amorphous carbon deposited from the filtered plasma stream of a vacuum arc were investigated. The structure was determined by electron diffraction, neutron diffraction and energy loss spectroscopy and the tetrahedral coordination of the material was confirmed. The measurements gave a nearest neighbour distance of 1.53 Å, a bond angle of 110 and a coordination number of four. A model is proposed in which the compressive stress generated in the film by energetic ion impact produces pressure and temperature conditions lying well inside the region of the carbon phase diagram within which diamond is stable. The model is confirmed by measurements of stress and plasmon energy as a function of ion energy. The model also predicts the formation of sp2-rich materials on the surface owing to stress relaxation and this is confirmed by a study of the surface plasmon energy. Some nuclear magnetic resonance, infrared and optical properties are reported and the behaviour of diodes using tetrahedral amorphous carbon is discussed. © 1991.

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.

The effect of beam interruptions on the integrity of ADSR fuel pin cladding: A thermo-mechanical analysis

During its lifetime in the core, the cladding of an Accelerator Driven Subcritical Reactor (ADSR) fuel pin is expected to experience variable stresses due to frequent interruptions in the accelerator proton beam. This paper investigates the thermal fatigue damage in the cladding due to repetitive and unplanned beam interruptions under certain operational conditions. Beam trip data was obtained for four operating high power proton accelerators, among which the Spallation Neutron Source (SNS) superconducting accelerator was selected for further analysis. 9Cr-1Mo-Nb-V (T91) steel was selected as the cladding material because of its proven compatibility with proposed ADSR design concepts. The neutronic, thermal and stress analyses were performed using the PTS-ADS, a code that has been specifically developed for studying the dynamic response to beam-induced transients in accelerator driven subcritical systems. The lifetime of the fuel cladding in the core was estimated for three levels of allowed pin power and specific operating conditions. © 2012 Elsevier Ltd. All rights reserved.

Interfacial recognition of human prostate-specific antigen by immobilized monoclonal antibody: effects of solution conditions and surface chemistry.

The specific recognition between monoclonal antibody (anti-human prostate-specific antigen, anti-hPSA) and its antigen (human prostate-specific antigen, hPSA) has promising applications in prostate cancer diagnostics and other biosensor applications. However, because of steric constraints associated with interfacial packing and molecular orientations, the binding efficiency is often very low. In this study, spectroscopic ellipsometry and neutron reflection have been used to investigate how solution pH, salt concentration and surface chemistry affect antibody adsorption and subsequent antigen binding. The adsorbed amount of antibody was found to vary with pH and the maximum adsorption occurred between pH 5 and 6, close to the isoelectric point of the antibody. By contrast, the highest antigen binding efficiency occurred close to the neutral pH. Increasing the ionic strength reduced antibody adsorbed amount at the silica-water interface but had little effect on antigen binding. Further studies of antibody adsorption on hydrophobic C8 (octyltrimethoxysilane) surface and chemical attachment of antibody on (3-mercaptopropyl)trimethoxysilane/4-maleimidobutyric acid N-hydroxysuccinimide ester-modified surface have also been undertaken. It was found that on all surfaces studied, the antibody predominantly adopted the 'flat on' orientation, and antigen-binding capabilities were comparable. The results indicate that antibody immobilization via appropriate physical adsorption can replace elaborate interfacial molecular engineering involving complex covalent attachments.

Growth of large sized y Ba 2Cu 3O 7 single crystals using the top seeded melt growth process

Neutron scattering experiments are fundamental to the study of magnetic order and related phenomena in a range of superconducting and magnetic materials. Traditional methods of crystal growth, however, do not yield single crystals of sufficient size for practical neutron scattering measurements. In this paper, we demonstrate the growth of relatively pure, large Y Ba 2Cu 3O 7 single crystals up to 30mm in diameter using a top seeded melt growth process. The characterization of the microstructural and magnetic properties of these crystals indicates that they contain <2% of impurity phases and, hence, exhibit only weak flux pinning behaviour. © 2012 IOP Publishing Ltd.

Structural and thermodynamic properties of different phases of supercooled liquid water.

Computer simulation results are reported for a realistic polarizable potential model of water in the supercooled region. Three states, corresponding to the low density amorphous ice, high density amorphous ice, and very high density amorphous ice phases are chosen for the analyses. These states are located close to the liquid-liquid coexistence lines already shown to exist for the considered model. Thermodynamic and structural quantities are calculated, in order to characterize the properties of the three phases. The results point out the increasing relevance of the interstitial neighbors, which clearly appear in going from the low to the very high density amorphous phases. The interstitial neighbors are found to be, at the same time, also distant neighbors along the hydrogen bonded network of the molecules. The role of these interstitial neighbors has been discussed in connection with the interpretation of recent neutron scattering measurements. The structural properties of the systems are characterized by looking at the angular distribution of neighboring molecules, volume and face area distribution of the Voronoi polyhedra, and order parameters. The cumulative analysis of all the corresponding results confirms the assumption that a close similarity between the structural arrangement of molecules in the three explored amorphous phases and that of the ice polymorphs I(h), III, and VI exists.

Structure of coexisting liquid phases of supercooled water: analogy with ice polymorphs.

The structural changes occurring in supercooled liquid water upon moving from one coexisting liquid phase to the other have been investigated by computer simulation using a polarizable interaction potential model. The obtained results favorably compare with recent neutron scattering data of high and low density water. In order to assess the physical origin of the observed structural changes, computer simulation of several ice polymorphs has also been carried out. Our results show that there is a strict analogy between the structure of various disordered (supercooled) and ordered (ice) phases of water, suggesting that the occurrence of several different phases of supercooled water is rooted in the same physical origin that is responsible for ice polymorphism.

A preliminary study of target multiplicity for ADSRs

The Accelerator Driven Subcritical Reactor (ADSR) concept is based on the coupling of a particle accelerator to a subcritical reactor core by means of a neutron spallation target interface. This paper investigates the benefits of multiple spallation targets in ADSRs. The motivation behind this is, firstly, to improve the overall reliability of the accelerator-reactor system, and, secondly, to evaluate other potential advantages such as lower beam power requirements. The results show that a system containing two or three spallation targets, coupled to independent accelerators, offers better neutronic performance. This is demonstrated through the increased effective multiplication factor (keff) in the two- and three-target configurations and a more uniform neutron flux distribution. A multiple-target ADSR also proves effective in mitigating the impact of frequent beam interruptions, a pressing issue that needs to be addressed for the ADSR concept to advance. Assuming no simultaneous beam shutdowns, the two- and three-target configurations reduce the risk of fuel cladding failure due to thermal cyclic fatigue. © 2013 Elsevier B.V. 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.

Flexible conversion ratio fast reactors: Overview

Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. The performance achievable by the unity conversion ratio cores of these reactors was compared to an existing supercritical carbon dioxide-cooled (S-CO2) fast reactor design and an uprated version of an existing sodium-cooled fast reactor. All concepts have cores rated at 2400 MWt. The cores of the liquid-cooled reactors are placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchangers (IHXs) coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. The S-CO2 reactor is directly coupled to the S-CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced reactor vessel auxiliary cooling system (RVACS) and a passive secondary auxiliary cooling system (PSACS). The selection of the water-cooled versus air-cooled heat sink for the PSACS as well as the analysis of the probability that the PSACS may fail to complete its mission was performed using risk-informed methodology. In addition to these features, all reactors were designed to be self-controllable. Further, the liquid-cooled reactors utilized common passive decay heat removal systems whereas the S-CO2 uses reliable battery powered blowers for post-LOCA decay heat removal to provide flow in well defined regimes and to accommodate inadvertent bypass flows. The multiple design limits and challenges which constrained the execution of the four fast reactor concepts are elaborated. These include principally neutronics and materials challenges. The neutronic challenges are the large positive coolant reactivity feedback, small fuel temperature coefficient, small effective delayed neutron fraction, large reactivity swing and the transition between different conversion ratio cores. The burnup, temperature and fluence constraints on fuels, cladding and vessel materials are elaborated for three categories of material - materials currently available, available on a relatively short time scale and available only with significant development effort. The selected fuels are the metallic U-TRU-Zr (10% Zr) for unity conversion ratio and TRU-Zr (75% Zr) for zero conversion ratio. The principal selected cladding and vessel materials are HT-9 and A533 or A508, respectively, for current availability, T-91 and 9Cr-1Mo steel for relatively short-term availability and oxide dispersion strengthened ferritic steel (ODS) available only with significant development. © 2009 Elsevier B.V. All rights reserved.

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.

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 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.