Advanced Molten-Salt Reactor Using High-Temperature Tech [pres. slides]

Mini Reactors Are Going Places and Pack a Lot of Power
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The October issue of Annals of Nuclear Energy publishes a paper "Development and verification of the neutron diffusion solver for the GeN-Foam multi-physics platform" by C. Fiorina, N. Kerkar, K.

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No results for Advanced Molten Salt Reactor Using High Temperature Tech Pres Slides in Books. Try checking your spelling or use more general terms. The Holts book Advanced Molten Salt Reactor Using High Temperature Tech [ pres. the interface of their deviations prior and look all their such AF, heading.

Mikityuk, P. Rubiolo and A. We contributed to the part devoted to the European Sodium Fast Reactor calculations. Brankov, G. Khvostov, K. Pautz, R. Restani, S.

Very-high-temperature reactor

Abolhassani, G. Ledergerber, W.

Wiesenack, to which we contributed in frame of the cooperation with the Fuel Modeling Group of our lab. This year we host one internship and three MS students. Their topics span from sodium boiling simulation to re-analysis of the molten-salt reactor experiment. Two lecturers and two instructors are from our group. Pautz, W.

Job opportunity at the Fast Reactors Group: we are looking for a Postdoctoral Fellow to work on static and transient analysis of the Generation IV sodium-cooled fast reactor. The March issue of Progress in Nuclear Energy publishes the second part of the paper on "The effectiveness of full actinide recycle as a nuclear waste management strategy when implemented over a limited timeframe -- Part II: Thorium fuel cycle" by B.

Lindley University of Cambridge , C. Fiorina PSI , R. Gregg NNL , F. Franceschini Westinghouse , G. Switzerland will be represented by Paul Scherrer Institut and, in particular, by our group. We contributed to two papers presented at the conference: one on fission gas trapping in the BWR fuel A and second on calculational benchmark of the sodium-cooled fast reactor fuel behaviour A The January issue of Annals of Nuclear Energy publishes a paper "Static and transient analysis of a medium-sized sodium cooled fast reactor loaded with oxide, nitride, carbide and metallic fuels" by Y.

Zhang and K. We performed a preliminary safety analysis of the BN type sodium cooled fast reactor loaded with different types of fuels with the TRACE code, by using the safety related parameters obtained from the Serpent calculations. Jeremie Nebes successfully passed the MS exam. The main goal was to evaluate the feasibility and main issues of Computational Fluid Dynamics simulations of sodium boiling in the fuel rod bundle. Euratom Work Program is oficially published.

Why the molten salt fast reactor (MSFR) is the “best” Gen IV reactor

Home Contact. Some advanced reactors would use new or non-conventional fuel forms, such as metallic fuels or dissolved molten fuels. Another concept proposed by Canada is generically called CANDU-SCWR [ 19 ], it is a pressured tube type reactor with fuel channels separating the light water coolant from the heavy water moderator. Built by scientists, for scientists. Nuclear plants of any sort may not be competitive in the emerging electricity market, as renewables get ever cheaper and their market share expands, but some nuclear options might be able to compete in the heat and synfuel markets. Lindsay Krall and Allison Macfarlane have written an important article in the Bulletin of the Atomic Scientists debunking claims that certain Generation IV reactor concepts promise major advantages with respect to nuclear waste management. New safety provisions or design improvements can be identified, developed, and implemented relatively early.

Fiorina, I. Clifford, M. Aufiero, K. Mikityuk, which summarizes our 2-year efforts on launching a new open-source computational engine for nuclear reactor analysis. Nikitin, E. Fridman, and K. Hyemin Kim successfully passed the MS exam. The experiment, completed 50 years ago, was partly re-analyzed with the modern codes to improve our understanding of the liquid-fuel reactor physics and to validate the computational tools to be used in design and safety demonstration of the future MSRs.

Jongsoo Choe successfully passed the MS exam. Molten Salt Reactors were simulated using Serpent and EQL0D to correlate decay heat power not only with reactor operation time and time after reactor shut-down, but also with efficiency of the on-line fission gas removal system. Benoit Soubelet successfully passed the MS exam. The topic of his MS study was "Time efficient fluid dynamics analysis of sodium fast reactor wire wrapped rod bundles".

The November issue of Progress in Nuclear Energy publishes the first part of the paper on "The effectiveness of full actinide recycle as a nuclear waste management strategy when implemented over a limited timeframe -- Part I: Uranium fuel cycle" by B. She was in Team 4 which won the prize for the best team report on "Future experimentation". Aufiero, A. Bidaud, M. Hursin, J. Palmiotti, S. Pelloni, P. Perko, S.

Pelloni, K. Mikityuk, J. Krepel, M. Szieberth, G. Girardin, B. Vrban, J. Cerba, M. Halasz, S. Feher, T. Reiss, J. Kloosterman, R. Stainsby, C. We contributed to 9 papers presented at the conference, 6 of them related to Sodium-cooled Fast Reactor and 3 -- to Molten Salt Reactor. The proposal was highly evaluated and granted by Euratom.

Non-LWR Vision and Strategy and Implementation Action Plans

The FAST group contributes with the neutronics and transient analysis. Matteo Zanetti, PhD student from Politecnico di Milano joins us for three-month internship to complete the work on development and implementation of the delayed neutron precursors transport in a molten salt reactor. This is the second attachment of Matteo to our group, the first one was in The study is supported by the PSI grant for the summer student.

Evzen Losa, PhD student from Czech Technical University in Prague, joins us for nine months to work on 3D modeling of closed equilibrium fuel cycle in advanced fast reactors as well as in VVER, using our calculational tools. This is already the second attachment of Evzen, the first one was in Two of them will work on topics related to sodium boiling and other two -- on neutronics and transient behaviour of molten-salt reactor.

The VHTR will be controlled like current PBMR designs if it utilizes a pebble bed core, the control rods will be inserted in the surrounding graphite reflector. Control can also be attained by adding pebbles containing neutron absorbers.

Generation IV

The high-temperature, high- neutron dose, and, if using a molten salt coolant, the corrosive environment, [1] p46 of the VHTR require materials that exceed the limitations of current nuclear reactors. Some materials suggested include nickel-base superalloys , silicon carbide , specific grades of graphite, high- chromium steels, and refractory alloys.

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The design takes advantage of the inherent safety characteristics of a helium-cooled, graphite-moderated core with specific design optimizations. The graphite has large thermal inertia and the helium coolant is single phase, inert, and has no reactivity effects. The core is composed of graphite, has a high heat capacity and structural stability even at high temperatures.

Reactor is designed for 60 years of service. From Wikipedia, the free encyclopedia. A type of nuclear reactor. Nuclear technology portal. February Oak Ridge National Laboratory. Archived from the original PDF on 16 July Retrieved 20 November Name of the figure Page no. The nuclear power industry has been developing and improving reactor technology for more than five decades and is starting to build the next generation of nuclear power reactors to fill new orders.

Several generations of reactors are commonly distinguished. Generation I reactors were developed in s, and outside the UK none are still running today. Generation II reactors are typified by the present US and French fleets and most in operation elsewhere. The first are in operation in Japan and others are under construction or ready to be ordered. Generation IV designs are still on the drawing board and will not be operational before at the earliest. These and other nuclear power units now operating have been found to be safe and reliable, but they are being superseded by better designs.

Reactor suppliers in North America, Japan, Europe, Russia and elsewhere have a dozen new nuclear reactor designs at advanced stages of planning. They are cooled and moderated by high-pressure liquid water. The hot radioactive water that leaves the pressure vessel is looped through a steam generator, which in turn heats a secondary non-radioactive loop of water to steam that can run turbines. They are the majority of current reactors. United States Naval reactors are of this type.

In a PWR, the primary coolant water is pumped under high pressure to the reactor core where it is heated by the energy generated by the fission of atoms. The heated water then flows to a steam generator where it transfers its thermal energy to a secondary system where steam is generated and flows to turbines which, in turn, spin an electric generator. PWRs were originally designed to serve as nuclear propulsion for nuclear submarines and were used in the original design of the second commercial power plant at Shipping port Atomic Power Station.

A boiling water reactor is cooled and moderated by water like a PWR, but at a lower pressure, which allows the water to boil inside the pressure vessel producing the steam that runs the turbines. Unlike a PWR, there is no primary and secondary loop. The thermal efficiency of these reactors can be higher, and they can be simpler, and even potentially more stable and safe. Instead of using a single large pressure vessel as in a PWR, the fuel is contained in hundreds of pressure tubes. These reactors are fuelled with natural uranium and are thermal neutron reactor designs.

Experimental

PHWRs can be refuelled while at full power, which makes them very efficient in their use of uranium it allows for precise flux control in the core. They can have a high thermal efficiency compared with PWRs due to higher operating temperatures. There are a number of operating reactors of this design, mostly in the United Kingdom, where the concept was developed. Mango stations are either shut down or will be in the near future. However, the AGCRs have an anticipated life of a further 10 to 20 years. This is a thermal neutron reactor design. Decommissioning costs can be high due to large volume of reactor core.

There were two main types of generation I GCR 1. The Magnox reactors developed by the UnitedKingdom 2. The main difference between these two types is in the fuel cladding material. Both types used fuel cladding materials that were unsuitable for medium term storage under water, making reprocessing an essential part of the cycle.

Both types were also designed and used to produce weapons-grade plutonium, but at the 9. These reactors can function much like a PWR in terms of efficiency, and do not require much high-pressure containment, as the liquid metal does not need to be kept at high pressure, even at very high temperatures. The Monju reactor in Japan suffered a sodium leak in and was restarted in May These reactors are fast neutron, not thermal neutron designs. These reactors come in two types: 2. Using lead as the liquid metal provides excellent radiation shielding, and allows for operation at very high temperatures.

Also, lead is mostly transparent to neutrons, so fewer neutrons are lost in the coolant, and the coolant does not become radioactive. Unlike sodium, lead is mostly inert, so there is less risk of explosion or accident, but such large quantities of lead may be problematic from toxicology and disposal points of view. Often a reactor of this type would use a lead-bismuth eutectic mixture. In this case, the bismuth would present some minor radiation problems, as it is not quite as transparent to neutrons, and can be transmuted to a radioactive isotope more lead. The Russian Alfa class submarine uses a lead-bismuth-cooled fast reactor.

The sodium is relatively easy to obtain and work with, and it also manages to actually prevent corrosion on the various reactor parts immersed in it. However, sodium explodes violently when exposed to water, so care must be taken, but such explosions would not be vastly more violent than for example a leak of superheated fluid from a SCWR or PWR. EBR-I, the first reactor to have a core meltdown, was of this type.

The result is an efficient, low-maintenance, very safe reactor with inexpensive, standardized fuel. The prototype was the AVR.