101 research outputs found
Physiological and growth responses of sour cherry (Prunus cerasus L.) plants subjected to short-term salinity stress
The gradual response of CAB-6P sour cherry (Prunus cerasus L.) plants to NaCl-induced salinity stress (60 mM NaCl) was investigated in a short-term hydroponic experiment, based on parameters relating to the growth, water relations, chlorophyll and mineral nutrition. The results showed that CAB-6P plants are very sensitive to salinity stress because their growth and leaf chlorophyll concentration were both affected negatively from the 3rd and 5th day, respectively, after incurring salinity stress. Since root growth was suppressed more severely than shoot growth, the shoot to root ratio was significantly increased under saline conditions. The concentrations of Na in leaves and stem of NaCl-treated plants were much lower than those measured in roots, suggesting Na exclusion mechanism from the shoot. The opposite trend was observed for Cl, indicating Cl inclusion mechanism to leaves. As regards the concentrations of Ca, Mg, P, K, Na, Fe, Zn and Mn, they were not changed in higher salinity conditions, apart from K, concentrations of which in leaves and roots were significantly increased and decreased, respectively (K translocation to leaves). Salinity further reduced K/Na ratio in root and stem as well as leaf water and osmotic potentials, whereas leaves of control and NaCl-treated plants presented similar turgor potential and K/Na ratio. These data add very important information to our knowledge about the physiological events occurring in sour cherry plants after even short-term exposure to salinity
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Generation IV Nuclear Energy System Initiative. Pin Core Subassembly Design for the Gas-Cooled Fast Reactor.
The Gas-Cooled Fast Reactor (GFR) is one of six systems selected for viability assessment in the Generation IV program. It features a closed nuclear fuel cycle, consisting of a high-temperature helium-cooled fast spectrum reactor, coupled to a direct-cycle helium turbine for electricity production. The GFR combines the advances of fast spectrum systems with those of high-temperature systems. It was clear from the very beginning that GFR design should be driven by the objective to offer a complementary approach to liquid metal cooling. On this basis, CEA and the US DOE decided to collaborate on the pre-conceptual design of a GFR. This reactor design will provide a high level of safety and full recycling of the actinides, and will also be highly proliferation resistant and economically attractive. The status of this collaborative project is that two unit sizes, 600 MWt and 2400 MWt were selected as the focus of the design and safety studies. Researchers studied fuel forms, fuel assembly/element designs, core configurations, primary and balance-of-plant layouts, and safety approaches for both of these unit sizes. Results regarding the feasibility of this GFR design are encouraging. For example, sustainability and non-proliferation goals can be met and the proposed concept has attractive safety features. These features take advantage of the helium in terms of its neutronic quasi-transparency as well as the enhanced Doppler effect in connection with candidate fuel and structural materials. The current design trend is to consider high unit power for the GFR (2400 MWt), an attractive level for the power density (100 MW/m{sup 3}), and the implementation of an innovative plate type fuel or pin type sub-assembly with carbide-based actinide compounds and SiC-based structural materials. Work is still needed to refine the safety approach, to select the main system options, and to more definitively establish economic parameters
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Generation IV nuclear energy system initiative. Large GFR core subassemblydesign for the Gas-Cooled Fast Reactor.
Gas-cooled fast reactor (GFR) designs are being developed to meet Gen IV goals of sustainability, economics, safety and reliability, and proliferation resistance and physical protection as part of an International Generation IV Nuclear Energy System Research Initiative effort. Different organizations are involved in the development of a variety of GFR design concepts. The current analysis has focused on the evaluation of low-pressure drop, pin-core designs with favorable passive cooling properties. Initial evaluation of the passive cooling safety case for the GFR during depressurized decay heat removal accidents with concurrent loss of electric power have resulted in requirements for a reduction of core power density to the 100 w/cc level and a low core pressure drop of 0.5 bars. Additional design constraints and the implementation of their constraints are evaluated in this study to enhance and passive cooling properties of the reactor. Passive cooling is made easier by a flat radial distribution of the decay heat. One goal of this study was to evaluate the radial power distribution and determine to what extent it can be flattened, since the decay heat is nearly proportional to the fission power at shutdown. In line with this investigation of the radial power profile, an assessment was also made of the control rod configuration. The layout provided a large number of control rod locations with a fixed area provided for control rods. The number of control rods was consistent with other fast reactor designs. The adequacy of the available control rod locations was evaluated. Future studies will be needed to optimize the control rod designs and evaluate the shutdown system. The case for low pressure drop core can be improved by the minimization of pressure drop sources such as the number of required fuel spacers in the subassembly design and by the details of the fuel pin design. The fuel pin design is determined by a number of neutronic, thermal-hydraulic (gas dynamics) and fuel performance considerations. For the purposes of this study, the starting point is the fuel pin design established by the CEA-ANL/US I-NERI collaboration project for the selected 2400 MWt large rector option. Structural mechanics factors are now included in the design assessment. In particular, thermal bowing establishes a bound on the minimum of fuel pin spacers required in each fuel subassembly to prevent the local flow channel restrictions and pin-to-pin mechanical interaction. There are also fabrication limitations on the maximum length of SiC fuel pin cladding which can be manufactured. This geometric limitation effects the minimum ceramic clad thickness which can be produced. This ties into the fuel pin heat transfer and temperature thresholds. All these additional design factors were included in the current iteration on the subassembly design to produce a lower core pressure drop. A more detailed definition of the fuel pin/subassembly design is proposed here to meet these limitations. This subassembly design was then evaluated under low pressure natural convection conditions to assess its acceptability for the decay heat removal accidents. A number of integrated decay heat removal (DHR) loop plus core calculations were performed to scope the thermal-hydraulic response of the subassembly design to the accidents of interest. It is evident that there is a large sensitivity to the guard containment back pressure for these designs. The implication of this conclusion and possible design modifications to reduce this sensitivity will be explored under the auspices of the International GENIV GFR collaborative R&D plan. Chapter 2 describes the core reference design for the 2,400 MWt GFR being evaluated. The methodology, modeling, and codes used in the analysis of the fuel pin structural behavior are described in Chapter 3. Chapter 4 provides the result of the thermal-hydraulic study of the assembly design for the accidents of interest. An evaluation of the performance and control rod reactivity control is also presented in Chapter 2
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Proliferation Resistance and Physical Protection Evaluation Methodology Development and Applications
An overview of the technical progress and accomplishments on the evaluation methodology for proliferation resistance and physical protection of Generation IV nuclear energy Systems
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Advanced Safeguards Approaches for New TRU Fuel Fabrication Facilities
This second report in a series of three reviews possible safeguards approaches for the new transuranic (TRU) fuel fabrication processes to be deployed at AFCF – specifically, the ceramic TRU (MOX) fuel fabrication line and the metallic (pyroprocessing) line. The most common TRU fuel has been fuel composed of mixed plutonium and uranium dioxide, referred to as “MOX”. However, under the Advanced Fuel Cycle projects custom-made fuels with higher contents of neptunium, americium, and curium may also be produced to evaluate if these “minor actinides” can be effectively burned and transmuted through irradiation in the ABR. A third and final report in this series will evaluate and review the advanced safeguards approach options for the ABR. In reviewing and developing the advanced safeguards approach for the new TRU fuel fabrication processes envisioned for AFCF, the existing international (IAEA) safeguards approach at the Plutonium Fuel Production Facility (PFPF) and the conceptual approach planned for the new J-MOX facility in Japan have been considered as a starting point of reference. The pyro-metallurgical reprocessing and fuel fabrication process at EBR-II near Idaho Falls also provided insight for safeguarding the additional metallic pyroprocessing fuel fabrication line planned for AFCF
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Advanced Safeguards Approaches for New Fast Reactors
This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors
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Evaluation Methodology For Proliferation Resistance And Physical Protection Of Generation IV Nuclear Energy Systems: An Overview
This paper provides an overview of the methodology approach developed by the Generation IV International Forum Expert Group on Proliferation Resistance & Physical Protection for evaluation of Proliferation Resistance and Physical Protection robustness of Generation IV nuclear energy systems options. The methodology considers a set of alternative systems and evaluates their resistance or robustness to a collection of potential threats. For the challenges considered, the response of the system to these challenges is assessed and expressed in terms of outcomes. The challenges to the system are given by the threats posed by potential proliferant States and sub-national adversaries on the nuclear systems. The characteristics of the Generation IV systems, both technical and institutional, are used to evaluate their response to the threats and determine their resistance against the proliferation threats and robustness against sabotage and theft threats. System response encompasses three main elements: 1.System Element Identification. The nuclear energy system is decomposed into smaller elements (subsystems) at a level amenable to further analysis. 2.Target Identification and Categorization. A systematic process is used to identify and select representative targets for different categories of pathways, within each system element, that actors (proliferant States or adversaries) might choose to use or attack. 3.Pathway Identification and Refinement. Pathways are defined as potential sequences of events and actions followed by the proliferant State or adversary to achieve its objectives (proliferation, theft or sabotage). For each target, individual pathway segments are developed through a systematic process, analyzed at a high level, and screened where possible. Segments are connected into full pathways and analyzed in detail. The outcomes of the system response are expressed in terms of PR&PP measures. Measures are high-level characteristics of a pathway that include information important to the evaluation methodology users and to the decisions of a proliferant State or adversary. They are first evaluated for segments and then aggregated for complete pathways. Results are aggregated as appropriate to permit pathway comparisons and system assessment. The paper highlights the current achievements in the development of the Proliferation Resistance and Physical Protection Evaluation Methodology. The way forward is also briefly presented together with some conclusions
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Gas-Cooled Fast Reactor (GFR) FY05 Annual Report
The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection. Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom and Switzerland), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report outlines the current design status of the GFR, and includes work done in the areas mentioned above for this fiscal year. In addition, this report fulfills the Level 2 milestones, ''Complete annual status report on GFR reactor design'', and ''Complete annual status report on pre-conceptual GFR reactor designs'' in work package GI0401K01. GFR funding for FY05 included FY04 carryover funds, and was comprised of multiple tasks. These tasks involved a consortium of national laboratories and universities, including the Idaho National Laboratory (INL), Argonne National Laboratory (ANL), Brookhaven National Laboratory (BNL), Oak Ridge National Laboratory (ORNL), Auburn University (AU), Idaho State University (ISU), and the University of Wisconsin-Madison (UW-M). The total funding for FY05 was 174K. The cost breakdown can be seen in Table 1
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