دانلود رایگان ترجمه مقاله تحلیل تولید فرآورده های Tritium برای یک راکتور تست با دمای بالا – 2013
دانلود رایگان مقاله انگلیسی تجزیه و تحلیل تولید ترتیم (Tritium) و راهبردهای مدیریت برای راکتور تست دمای بالا خنک شده با نمک فلوراید (FHTR) به همراه ترجمه فارسی
عنوان فارسی مقاله: | تجزیه و تحلیل تولید ترتیم (Tritium) و راهبردهای مدیریت برای راکتور تست دمای بالا خنک شده با نمک فلوراید (FHTR) |
عنوان انگلیسی مقاله: | (Tritium Production Analysis and Management Strategies for a Fluoride-salt-cooled High-temperature Test Reactor (FHTR |
رشته های مرتبط: | فیزیک، فیزیک هسته ای، مهندسی هسته ای، راکتور (هسته ای) |
فرمت مقالات رایگان | مقالات انگلیسی و ترجمه های فارسی رایگان با فرمت PDF میباشند |
کیفیت ترجمه | کیفیت ترجمه این مقاله خوب میباشد |
توضیحات | ترجمه این مقاله با کیفیت مناسب انجام شده است. |
کد محصول | f360 |
مقاله انگلیسی رایگان (PDF) |
دانلود رایگان مقاله انگلیسی |
ترجمه فارسی رایگان (PDF) |
دانلود رایگان ترجمه مقاله |
خرید ترجمه با فرمت ورد |
خرید ترجمه مقاله با فرمت ورد |
جستجوی ترجمه مقالات | جستجوی ترجمه مقالات |
بخشی از ترجمه فارسی مقاله: 1 – مقدمه |
بخشی از مقاله انگلیسی: 1 Introduction 1.1 Overview of the Fluoride-salt-cooled High-temperature Reactor (FHR) Research in the nuclear and materials engineering fields worldwide is currently investigating a new reactor concept called the Fluoride-salt-cooled High-temperature Reactor (FHR). The design features a low-pressure, high-temperature liquid fluoride-salt coolant, coated-particle fuel, increased efficiency for electricity generation, process heat production, and fully passive decay heat rejection. These qualities contribute to the generation of electricity at a low cost while maintaining passive safety features, overcoming current limitations of Light Water Reactors (LWRs) [1]. The concept of the FHR must undergo further design study, materials and fuel irradiation tests, and operational and safety evaluation before a full-scale power reactor can be built. The development of a Fluoride-salt-cooled High-temperature Test Reactor (FHTR) will allow for the testing of critical concepts in the FHR including the testing of materials, coolants, and the neutronics. An interest in Molten Salt Reactors (MSRs) originated first in the 1950s and 1960s as highly experimental designs. The interest evolved around the potential ability of the MSR to burn actinides, operate as a breeding reactor for Thorium, provide a simplified fuel cycle, and make way for limited transport of radioactive materials. Advances in HighTemperature Gas-cooled Reactor (HTGR) fuel technology, materials, and MSR technology led to the development of the FHR concept. The MSR and FHR designs are mainly different in the fuel aspect: the fuel in MSRs was dissolved in the molten salt coolant whereas the fuel for the FHR is the same type of solid fuel developed for HTGRs. The primary coolant candidate for the FHR is flibe (66.7%LiF-33.3%BeF 2) because it has excellent heat transfer and neutronics properties, with a high melting point of 460’C and boiling point of 1430*C that allow reactor operation at very high temperatures [2]. As a result, flibe is also the primary coolant candidate for the FHTR. Tritium is a radioactive isotope of hydrogen with an atomic mass of three that may be produced from neutron capture reactions with the components of flibe, primary 6Li. Tritium has a half-life of 12.3 years and is a pure beta emitter. At high temperatures, tritium behaves like hydrogen and may diffuse through materials and escape from the reactor. 1.2 Overview of the FHTR [3] Before the FHR becomes commercially viable, a test reactor is necessary to study important aspects relevant to neutronics, thermal hydraulics, materials, and most relevant to this case, methods for managing tritium. A preliminary 100 MW FHTR has been designed by the 2012 MIT Nuclear Systems Design Class to prove commercial feasibility of and provide accelerated testing for the FHR. The MIT FHTR design has notable features that include: a large central thermal test position with a neutron flux greater than 3E14 n/cm 2 -s for accelerated fuel and materials testing, numerous in/ex-core testing positions in contact with the primary coolant, a coated-particle-based fuel form, and high inlet (>600’C) and outlet (-700’C) temperatures to avoid salt freezing and provide the expected hightemperature environment of the commercial design. Numerous safety features have been suggested in the FHTR design for tritium control and decay heat removal. Figure 1 shows the core geometry layout of the MIT FHTR design and Table 1 shows the core’s relevant dimensions [4]. Table 2 shows the flux levels in several parts of the FHTR. |