Academic literature on the topic '120106 Interior Design'

The ACR-1000™ developed by Atomic Energy of Canada Limited (AECL) is a 1200 MWe - pressure tube type, light-water-cooled and heavy-water-moderated reactor, which has evolved from the well-established CANDU™ line of reactors. It retains the basic proven CANDU design features while incorporating innovations and state-of-the-art technologies to ensure fully competitive safety, operation, performance and economics. The major innovation in the ACR-1000 is the use of slightly enriched uranium fuel and light water coolant. ACR-1000 is a four-quadrant design (for easier maintenance and improved reliability). There are five safety systems in the ACR-1000; (i) two independent, diverse and fast acting shutdown systems (SDS1 and SDS2), which are physically and functionally independent from each other and from the reactor regulating system; (ii) Emergency core cooling system; (iii) Emergency Feedwater system; and (iv) Containment system, which includes a strong steel-lined containment structure. In addition the Reserve water system provides feedwater to the heat transport system, steam generators, moderator and shield cooling system for beyond design basis accidents. The Level 1 Probabilistic Safety Assessment (PSA) is conducted in support of the design phase of the ACR-1000. The purpose of Level 1 PSA is to identify whether the ACR-1000 design targets and the regulatory safety goal for severe core damage frequency (SCDF) are met with adequate margin and provide design feedback. An interim Level 1 PSA was conducted for internal at-power events. Interim assessments were conducted for shutdown state, internal fire and flood at-power events. An interim seismic margin assessment was conducted for the seismic events. The Level 1 PSA results show that the ACR™ design targets and safety goal for SCDF are met with significant safety margin. Based on the ACR-1000 Level 1 PSA, the accident behaviours of the ACR-1000 are well understood and their consequences can be predicted with a high-level of confidence. It also provides sufficient assurance that the release based regulatory safety goals are achievable for ACR-1000. The Level 1 PSA results also signify a robust design that provides a strong foundation for the ACR-1000 design. The paper summarizes the Level 1 PSA program, methodology followed, the results obtained, and insights gained during the development of the ACR-1000 Level 1 PSA.

Advanced technologies in Ceramics Gas Turbine (CGT) are expected to make a great progress in energy conservation, anti-pollution, and fuel-diversification. In Japan, R&D’s in industrial usage 300 kW class CGT have been advanced under a national project entitled “New Sunshine Program”, under the subsidy of Agency of Industrial Science and Technology (AIST), Ministry of International Trade and Industry (MITI) through the period of FY1988–1996. In this project, three different type prototypes of the CGT are under development through New Energy and Industrial Technology Development Organization (NEDO). Over the last six years, the basic designs have been completed and the ceramic elements such as turbine rotors, scrolls, and combustors were successfully fabricated. To check up the whole progress of the project, an interim evaluation is scheduled by the end of FY1993. Toward this evaluation, each prototype has been programmed to demonstrate 1200°C of Turbine Inlet Temperature (TIT) and prove more than 30% of thermal efficiency. (The ultimate target in the project is 42% of thermal efficiency at 1350°C TIT.) They would also show enough environmental adaptability. In this paper, overall status of the development in the 300kW CGT project is reviewed and the items in the interim evaluation are explained.

Abstract Millions of dollars are spent annually on hydraulic fracturing for oil and gas well stimulation. Various researchers have proposed models to optimize the stimulation treatment, with limited success due to a lack of understanding of the behavior of critical parameters in the models, especially the fracturing fluids. Thus, the Gas Research Institute and the U.S. Department of Energy are sponsoring a project at the University of Oklahoma to investigate fracturing fluids in a large physical fracture model, the Fracturing Fluid Characterization Facility (FFCF). This simulator was initially expected to consist of a parallel plate simulator as large as 16 ft high by 100 ft long, with an internal pressure as high as 1200 psi, to determine critical fluid parameters under full-scale operating conditions. A smaller and more economical version of the simulator (7 ft high by 9 1/3 ft long) has since been built and moved to its own facility. As the reaction forces of the originally intended structure were tremendous (over 275,000,000 lbf for a full-scale simulator), structural analyses were essential. Thus, static strength and stability (buckling) analyses were performed on various full-scale models and on the smaller final version of the pressure retaining structure for the FFCF simulator. An interior segment of each structural model consisting of the pressure chamber and the reaction beam/ plate/tie rod assemblies was modeled in detail and analyzed by the finite element method using the industry standard code MSC/NASTRAN together with the MSC/XL pre- and postprocessor. In addition to the linear analyses for all models, a material nonlinear analysis was run for Model 1. General buckling analyses by the finite element method were performed on square plates and on retaining structure plates with various boundary conditions to determine the influence of these conditions on the critical buckling loads. Finally, the full final model was analyzed for buckling proving the stability of the design.

Within the framework of the cleaning operation of the Marcoule reprocessing plant UP1 (France), the CEA (French Atomic Energy Commission) developed a measurement system for 225-liter drums filled with bituminized radioactive sludge originating from the effluent treatment. This work was carried out for the CODEM, which is an economic interest group made up of CEA, EDF (the French public utility) and COGEMA (the operator of UP1). CODEM is in charge of UP1 dismantling operations, especially waste retrieval. The bituminized waste drums mainly contain plutonium, americium, uranium, curium and various beta emitters among which some are responsible for significant gamma irradiation, such as 137Cs. The aim of this system is to sort the packages according to their radioactive level, so as to direct them towards the French Aube Center, which is a surface repository. This means they must meet the acceptance criteria related to their activities. Otherwise, they will remain in interim storage in Marcoule, pending the choice of a final mode of management (e.g. underground disposal). The assay system, called UTM (the French acronym for “Measurement and Sorting Unit”), consists of three stations devoted to active gamma imaging, gamma-ray spectroscopy and combined passive / active neutron measurements. After nearly 3 years of optimization and design studies [1], the CEA has built a scale one mock-up of UTM, called SYMETRIC. The purpose was to validate the performances formerly assessed by numerical simulation, mainly with the computer code MCNP [2]. We present here the experimental results obtained with SYMETRIC for five real bituminized waste drums. These confirm the expected performances in the measurement time assigned for each assay, which is limited to 1200 seconds. With the help of gamma imaging, we are able to determine the density of the bituminous mix with an uncertainty of ± 10% for a confidence level of 95%. We can also measure the filling height with an accuracy of ± 2 cm. These data allow us to correct matrix effects in gamma and neutron measurements. For these assays, the main results concern the detection limits and measurement uncertainties on 241Am, 239Pu and 240Pu. These radioisotopes represent the major part of the total alpha activity, which is a very sensitive parameter for surface disposal limited to a maximum level of about 10 GBq per drum. The alpha activity must be calculated after a radioactive decay of 300 years, which is the survey period of the French Aube Center. If we can detect the former isotopes, the uncertainties on their measured activities are roughly 50%. If not, the detection limits are around a few GBq. These performances are sufficient to allow the sorting of the drums to either surface repository or interim storage. However, in order to increase the margin between the detection limits and the acceptance criterion on the total alpha activity, additional studies on the optimization of the measurement performances will be carried out. In this context, the experience gained with the SYMETRIC mock-up will be very helpful.