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[en] Misuse of declared nuclear facilities is one of the important proliferation threats. The robustness of a facility against these threats is characterized by a number of proliferation resistance (PR) measures. This paper evaluates and compares PR measures for several misuse scenarios using a Markov model approach to implement the pathway analysis methodology being developed by the PR and PP (Proliferation Resistance and Physical Protection) Expert Group. Different misue strategies can be adopted by a proliferator and each strategy is expected to have different impacts on the proliferator's success. Selected as the probabilistic measure to represent proliferation resistance, the probabilities of the proliferator's success of misusing a hypothetical ESFR (Example Sodium Fast Reactor) facility system are calculated using the Markov model based on the pathways constructed for individual misuse scenarios. Insights from a comparison of strategies that are likely to be adopted by the proliferator are discussed in this paper.
[en] The lattice sites and spatial disorder of isolated Mn2+ ions in calcite were examined with x-ray standing waves, and the structure of the surrounding ions was examined with extended x-ray absorption fine-structure spectroscopy. The Mn2+ ion is found to be on-center substitutional at the Ca2+ site, with spatial disorder comparable to that of Ca2+. The first-neighbor Mn-O distance is found to be the same as that in the isostructural MnCO3. The radial distance of the closest Mn-Ca shell is reduced by ∼2% from the undistorted Ca-Ca distance. Based on these measurements, an atomic-scale structural model of the Mn2+ site suggests that the intramolecular distortion in the first-neighbor CO32- anions plays a key role in establishing the conserved first-neighbor Mn-O distance while maintaining ordering with respect to the lattice. The CaCO3:Mn2+ structure is shown to be characteristically distinct from those of analogous impurities in monatomic ionic crystals
[en] The Gas Turbine High-Temperature Gas-Cooled Reactor combines a helium-cooled reactor core of established design with a closed cycle helium turbine power conversion system. The design considerations which mitigate the consequences of failure of the rotating machinery located within the reactor vessel are discussed. The methods of analysis and a summary of results are presented for the failure modes of most concern. The spectrum of potential incidents which have been evaluated includes turbine blade, rim, and disk failures. The requirements and design methods for rotor containment are discussed. The turbomachine maintains a pressure ratio of about two between the high-pressure and low-pressure portions of the loop; postulated failures can therefore lead to rapid rates of pressure change. The preliminary evaluation of this internal pressure equilibration is presented
[en] Dehydration is widely involved in tobacco processing such as tobacco leaf curing, tobacco trip redrying and cut tobacco drying, which plays a key role due to its effect on the physical and chemical quality of tobacco. The current drying methods in tobacco processing mainly use heat conduction, heat convection or their combination to dehydrate tobacco materials. However, radiation heat transfer as one of basic heat transferways has not been investigated in the tobacco drying. In the present work, infrared radiation dryer was designed to explore the tobacco infrared radiation drying characteristics. The effect of radiation heat transfer conditions and vacuum on the drying kinetics and temperature of tobacco leaves was investigated. Diffusion coefficient of middle tobacco leaves C2F is between 0.848×10-10 1.597×10-10 m2/s. At the same time, the pore structure andpetroleum ether tobacco extracts in dried tobacco were also analyzed in order to explore the different effects of infrared radiation drying and traditional drying technology on tobacco quality. (Author)
[en] Detailed reactor physics and safety analyses are being performed for the 20 MW D2O-moderated research reactor at the National Institute of Standards and Technology (NIST). The analyses employ state-of-the-art calculational methods and will contribute to an update to the Final Safety Analysis Report (FSAR). Three-dimensional MCNP Monte Carlo neutron and photon transport calculations are performed to determine power and reactivity parameters, including feedback coefficients and control element worths. The core depletion and determination of the fuel compositions are performed with MONTEBURNS to model the reactor at the beginning, middle, and end-of-cycle. The time-dependent analysis of the primary loop is determined with a RELAP5 transient analysis model that includes the pump, heat exchanger, fuel element geometry, and flow channels. A statistical analysis used to assure protection from critical heat flux (CHF) is performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF are determined with MCNP. Evaluations have been performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. These analyses are significantly more rigorous than those performed previously. They have provided insights into reactor behavior and additional assurance that previous analyses were conservative and the reactor was being operated safely
[en] The objective of the present work is to study and model the interfacial structure development of air-water two-phase flow in a confined test section. Experiments of a total of 9 flow conditions in a cap-turbulent and churn-turbulent flow regimes are carried out in a vertical air-water upward two-phase flow experimental loop with a test section of 20-cm in width and 1-cm in gap. The miniaturized four-sensor conductivity probes are used to measure local two-phase parameters at three different elevations for each flow condition. The bubbles captured by the probes are categorized into two groups in view of the two-group interfacial area transport equation, i.e., spherical/distorted bubbles as Group 1 and cap/churn-turbulent bubbles as Group 2. The acquired parameters are time-averaged local void fraction, interfacial velocity, bubble number frequency, interfacial area concentration, and bubble Sauter mean diameter for both groups of bubbles. Also, the line-averaged and area-averaged data are presented and discussed. The comparisons of these parameters at different elevations demonstrate the development of interfacial structure along the flow direction due to bubble interactions
[en] Detailed reactor physics and safety analyses have been performed for the 20 MW D2O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional MCNP Monte Carlo neutron and photon transport calculations were performed to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim arms. The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model including the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP. Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. In both the startup and maximum reactivity insertion accidents, the core power transient is terminated by a reactor trip at 30 MW. The calculations show that both the peak reactor power and the excursion energy depend on the negative reactivity insertion from reactor trip. Two cases were considered for loss of electrical power. In the first case offsite power is lost, resulting in an immediate scram caused by loss of power to the control rod system. In the second case power is lost to only the three operating primary pumps, resulting in a slightly delayed scram when loss-of-flow is detected as the pumps coast down. In both instances, RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur, because of the momentum of the coolant flowing through the fuel channels and the negative scram reactivity insertion. For both the primary pump seizure and inadvertent throttling of a flow control valve, the RELAP5 analyses indicate that the reduction in power following the trip is sufficient to ensure that there is adequate margin to CHF and the fuel cladding does not fail. The analysis of the loss-of-flow accident in the extremely unlikely case where both shutdown pumps fail shows that the cooling provided by the D2O is sufficient to ensure the cladding does not fail. The power distributions were examined for a set of fuel misloadings in which a fresh fuel element is moved from a peripheral low-reactivity location to a central high-reactivity location. The calculations show that there is adequate margin to CHF and the cladding does not fail
[en] Microstructural features and the formation mechanisms of interlocked microstructures of acicular ferrite in a low-carbon high-strength steel weld metal were investigated by means of computer-aided three-dimensional reconstruction technique and electron backscattered diffraction analysis. Multiple nucleation on inclusions, sympathetic nucleation or repeated nucleation, hard impingement, mutual intersection, and fixed orientation relationships of acicular ferrite grains were observed. They were all responsible for the formation of interlocked microstructures in the weld metal. During the process of isothermal transformation, the pre-formed acicular ferrite laths or plates partitioned austenite grains into many small and separate regions, and the growth of later formed acicular ferrite grains was confined in these small regions. Thus, the crystallographic grain size became smaller with the increasing holding time. Highlights: ► Acicular ferrite is formed by multiple nucleation and sympathetic nucleation. ► Hard impingement and intersection of ferrite grains occur at later stages. ► The pre-formed ferrite laths partition austenite grains into smaller regions. ► The growth of later formed ferrite grains is confined in the smaller regions.