[en] During the past several years, a comprehensive high fidelity reactor core modeling capability has been developed called the Numerical Nuclear Reactor (NNR) (Weber,2003) for detailed analysis of Light Water Reactors. The NNR achieves high fidelity with a whole-core neutron transport solution and ultra-fine-mesh computational fluid dynamics/heat transfer solution. Previous applications of the NNR have been to the steady-state analysis of both pressurized and boiling water reactors. Recently there has been interest in taking advantage of the NNR to improve the fidelity for PWR transient analysis. The work described in this paper is a preliminary demonstration of the ability of the whole core neutron transport code, DeCART, to provide a detailed intra-pin-power distribution during a control rod ejection accident. The current state of the art in analysis of this event relies upon the assembly averaged power from a whole core nodal neutronics simulator and some type of pin power reconstruction within the fuel assembly. Both methodologies are briefly presented and applied to model a super-prompt reactivity insertion accident. The difference in the results of both approaches are discussed and the benefit of the DeCART methodology is described. (authors)

[en] ANL effort is divided into five areas: (1) cooperation with Argentina to demonstrate the use of LEU-foil targets in alkaline-based processes, (2) cooperation with Indonesia in converting their HEU-based Cintichem process to LEU-foil targets, (3) technical assistance to two potential U.S. domestic suppliers (MURR and BWTX), (4) responding to the National Academies Study, and (5) participation in the IAEA CRP for Indigenous Mo-99 production. This paper presents highlights of these activities. A short description of how the dose emitted by spent HEU target material compared to spent fuel is also included. (author)

[en] Thermal stratification is an important phenomenon to be considered when removing decay heat using natural circulation in Sodium Cooled Fast Reactors (SFR) during protected loss of flow accidents (PLOF). The computational fluid dynamics (CFD) model of the hot pool of a sodium fast reactor, the Advanced Burner Test Reactor (ABTR), has been developed using STAR-CCM+. ABTR is a 250 MWt SFR pre-conceptual design for the transmutation of transuranic waste recycled from Light Water Reactor (LWR) spent fuel. The purpose of this study is to evaluate the importance of 3-D thermal effects in CFD simulations when modeling transients that involve passive safety systems. The model features a volume of fluid approach to include the compression of the cover gas as the sodium level in the tank changes during the postulated transient. The reference model treats turbulence effects in the sodium with the SST k-ω model. Steady state simulations showed that the temperature field in the hot pool is fairly uniform, except in the reflector region where temperature is higher. The calculation was initialized by performing a relatively long unsteady simulation with boundary conditions fixed at the nominal operating conditions. Transient calculations were then performed with time-dependent boundary conditions from a safety system code analysis. During transient calculations, a thermal stratification phenomenon was observed as expected by formation of temperature layers where the top layer had the highest temperature and the bottom layer had the lowest, which impedes the start of natural circulation. This has a significant impact in performance and safety and it can determine the reactor's ability to remove decay heat during transient operations. The results from CFD stand-alone simulations confirmed that 3-D thermal effects must be accounted for when performing passive safety system analyses. (authors)

[en] Highlights: ► Overview of current DEMO maintenance concepts. ► Comparison of current dextrous remote handling technologies and their rebalance to DEMO. ► Presentation of some ideas to improve the productivity and reliability of the DEMO ex-vessel transport system. ► A description of the size and type facilities that might be required in the DEMO hot cell. ► Identification of some areas that need to be developed further to meet the requirements of DEMO. -- Abstract: In Europe the work on the specification and design of a demonstration power plant (DEMO) is being carried out by EFDA in the power plant physics and technology (PPP and T) programme. DEMO will take fusion from experimental research into showing the potential for commercial power generation. During the fusion reaction, components in the tokamak become highly activated. The estimated dose rate levels after shutdown (zero decay time) due to 60 dpa accumulation in steel (blanket) and 30 dpa (divertor) are 13.1–17.4 kGy/h (blanket); 8.8–11.6 kGy/h (divertor) [1], much higher than those to be encountered at ITER. Upon removal from the tokamak, components would be transported to the hot cell facility with attention to minimizing the spread of activated dust and tritium contamination. It is proposed to use a sealed cask of ∼20 tonnes, running on air castors with 50% lifting capability redundancy. Due to the number and complexity of the routes taken by this transporter it would have to be an un-tethered semi-autonomous system. This poses some technical challenges, including providing sufficient battery capacity, reliable guidance and a fail safe un-tethered control system. The mass of the components being moved is assumed here to range from a few tonnes to in excess of one hundred tonnes. Before the removed in-vessel components can be processed in the hot cell, they would require a period of cooling, approximately 2.5 years, to allow dose rate and decay heating to reduce. This reduces the decay heating level to ∼1–1.5 kW/m³ and a contact dose rate to ∼250 Sv/h, which is more suited to dexterous man-in-the-loop remote handling (RH). During maintenance, many components would require replacement or refurbishment. The hot cell facility would have to provide all the associated RH functions and operate fully remotely. With no human intervention, an accordingly robust RH recovery system would be required. This paper describes the first steps being taken towards the design of the DEMO hot cell. It will show a comparison of the current DEMO in-vessel maintenance concepts from a hot cell perspective, describe a proposed ex-vessel transport system, and summarize the facilities that have been identified as required within the hot cell, examine current RH technology and discuss the identified critical development issues

[en] Highlights: • The proposed models are 400 times less computationally expensive than CFD simulations. • The proposed models show good duct wall temperature agreement with CFD simulations. • The paper provides an efficient tool for coupled radial core expansion calculation. - Abstract: Porous medium models have been established for predicting duct wall temperature of sodium fast reactor rod bundle assembly, which is much less computationally expensive than conventional CFD simulations that explicitly represent the wire-wrap and fuel pin geometry. Three porous medium models are proposed in this paper. Porous medium model 1 takes the whole assembly as one porous medium of uniform characteristics in the conventional approach. Porous medium model 2 distinguishes the pins along the assembly's edge from those in the interior with two distinct regions, each with a distinct porosity, resistance, and volumetric heat source. This accounts for the different fuel-to-coolant volume ratio in the two regions, which is important for predicting the temperature of the assembly's exterior duct wall. In Porous medium model 3, a precise resistance distribution was employed to define the characteristic of the porous medium. The results show that both porous medium model 2 and 3 can capture the average duct wall temperature well. Furthermore, the local duct wall variations due to different sub-channel patterns in bare rod bundles are well captured by porous medium model 3, although the wire effect on the duct wall temperature in wire wrap rod bundle has not been fully reproduced yet.

[en] The DEMO Active Maintenance Facility (AMF) would be used for the storage, handling and processing of In-Vessel Components (IVC) throughout their time on site, the only exception being the time that they are installed in the vessel. It is anticipated that all handling operations associated with used components will have to be carried out using remote handling techniques. During plasma operations the In-Vessel Components are exposed to high levels of neutron activation. This activation results in high radiation dose rates and decay heating. This presents a significant problem for Remote Handling Equipment (RHE) in the AMF. The high dose rates require the equipment to be sufficiently radiation tolerant to allow it to work reliably for long periods. The decay heating requires forced cooling of newly removed IVC's while they are in storage. The duration of the storage is dependent on the decay heating reducing to a level that has been nominally set at <50 °C without active cooling in room temperature air. This paper summarises the progress made in 2012 on the conceptual design of the AMF and its facilities. The layout and proposed function of the main areas will be described along with the principles applied. The design of the AMF has evolved from a simple representation of the required facilities in 2011 to a concept that can be developed to support maintenance of DEMO

[en] To effectively use a passive sampler for monitoring trace contaminants in the gas-phase, its sampling characteristics as a function of ambient wind conditions must be known. In this study two commonly used passive samplers were evaluated using computational fluid dynamics. Contaminant uptake by the polyurethane foam (PUF) was modeled using a species transport model. The external-internal flow interactions in the sampler were characterized, and the uptake rates of contaminant species were quantified. The simulations show that flow fields in the samplers have strong velocity gradients, and single-point velocity measurements do not capture flow interactions accurately. Sampling rates calculated for a PUF in freestream are in good agreement with sampling rates for PUFs in the passive samplers studied for the same average velocity over the PUF. The calculated sampling rates are in general agreement with those obtained experimentally by other researchers. - The effect of wind speed on sampling rates of two commonly used passive samplers was investigated using computational fluid dynamic techniques

[en] SAS4A/SASSYS-1 is a software simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced nuclear reactors. This report summarizes recent tasks to modernize the SAS4A/SASSYS-1 code system to improve internal data management and to update the code documentation to reflect recent code developments. The motivation for performing these updates stems from the relevance of SAS4A/SASSYS-1 to a number of U.S. Department of Energy programs as well as domestic and international collaborations.

[en] Micro aerial vehicles, particularly quadrotors, have been used in a wide range of applications. However, the literature on aerial manipulation and grasping is limited and the work is based on quasi-static models. In this paper, we draw inspiration from agile, fast-moving birds such as raptors, that are able to capture moving prey on the ground or in water, and develop similar capabilities for quadrotors. We address dynamic grasping, an approach to prehensile grasping in which the dynamics of the robot and its gripper are significant and must be explicitly modeled and controlled for successful execution. Dynamic grasping is relevant for fast pick-and-place operations, transportation and delivery of objects, and placing or retrieving sensors. We show how this capability can be realized (a) using a motion capture system and (b) without external sensors relying only on onboard sensors. In both cases we describe the dynamic model, and trajectory planning and control algorithms. In particular, we present a methodology for flying and grasping a cylindrical object using feedback from a monocular camera and an inertial measurement unit onboard the aerial robot. This is accomplished by mapping the dynamics of the quadrotor to a level virtual image plane, which in turn enables dynamically-feasible trajectory planning for image features in the image space, and a vision-based controller with guaranteed convergence properties. We also present experimental results obtained with a quadrotor equipped with an articulated gripper to illustrate both approaches. (papers)

[en] This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the experienced user-base and the experimental validation base was decaying away quickly.