162,075 research outputs found
ELSY Neutronic Analysis by deterministic and Monte Carlo methods: an innovative concept for the control rod systems
No full textThis paper deals with the neutronic design of ELSY (the European Lead-cooled SYstem), a 600 MWe Fast Reactor developed within the 6th EURATOM Framework Programme. ELSY aims at being an “adiabatic” system (as far as possible) in order to fulfill both the requirements of sustainability and proliferation resistance. It represents the European solution for the Lead Fast Reactor (LFR), one of the six candidate typologies proposed by the Generation-IV International Forum (GIF).
The analysis of the ELSY reference configuration, with typical pure MOX loading, is here presented. An introductory investigation of the adiabatic and, possibly, the burner options viability is also achieved by providing a rough estimate of the Minor Actinides (MAs) equilibrium concentrations and time constants. One of the main challenge-points in the design of the core, made up of wrapper-less square Fuel Assemblies (FAs) according to the common scheme of PWRs, is the small delta-T between the coolant average outlet temperature (480 °C) and the allowable cladding one (550 °C): it requires a rather flat radial power distribution, obtained by segmenting the core in three zones with different enrichments.
Three different control sets have been introduced in order to achieve the required reliability for reactor shutdown and safety systems: eight traditional concept Control Rod (CR) assemblies together with two independent systems of sparse control “Finger Absorber” Rods (FARs), small B4C rods that can be inserted, in principle, in the center of each FA.
One of the two finger absorber systems includes a subset of rods devoted to the regulation of the criticality swing during the cycle: their number can be limited indeed since the small reactivity swing (some hundreds pcm) due to the about unitary breeding ratio. Such an innovative solution can also be positioned in order to maintain an optimal power flattening during the fuel cycle.
To verify the feasibility of this solution, a very detailed neutronic analysis, adopting both deterministic and stochastic approaches, has been carried out. It becomes crucial indeed to estimate accurately the self-shielding phenomenon of the innovative FARs in order to achieve the aimed performances (a reactivity worth of about 3000 pcm for scram)
[Report to Chief J. E. Curry, by an unknown author #1]
No full textReport to Chief J. E. Curry, by an unknown author. The report contains a list of officers who gave depositions to the United States Attorney
[Report to Chief J. E. Curry, by an unknown author #2]
No full textReport to Chief J. E. Curry, by an unknown author. The report contains a list of officers who gave depositions to the United States Attorney
Extension of the FAST code system for the modelling and simulation of MSR dynamics
No full textIn this work, the PSI FAST code system is extended for the modeling and simulation of Molten Salt Reactor (MSR) dynamics. The thermal-hydraulic code TRACE has been provided with a module for 1-D Delayed Neutron Precursor (DNP) balance and decay heat modeling in fluid fuel, and with built-in MSR materials. The reactor power is determined by means of a Point-Kinetics approach that makes use of power-weighted values of temperature and DNP distributions in the core. To validate the module, models for the comparison with experiments performed in the Molten Salt Reactor Experiment (MSRE) have been developed. The MSRE was a graphite moderated reactor built and operated in the sixties at Oak Ridge National Laboratory (ORNL). The neutronic characteristics of the reactor have been determined by means of the Monte Carlo code SERPENT. Two models of the MSRE plant with different detailing have been set up in order to determine the importance of the plant components. In particular, different descriptions of the external cooling loop have been tested and compared. Some significant transients have been considered for the module assessment, by comparison with available experimental data from ORNL, both in time and frequency domain
MSFR TRU-BURNING POTENTIAL AND COMPARISON WITH AN SFR
No full textTransmutation of the legacy TRansUranics (TRU) from Light Water Reactor operation has become in recent years a main objective for the development of Fast Reactors (FR). In fact, an effective TRU-burning requires fuel multi-recycling and a fast-neutron-spectrum reduces the endogenous generation of Cm and Cf isotopes, thus
benefitting fuel handling and in-core radiotoxicity generation. However, achievement of high TRU-burning rates requires low-Conversion-Ratio (CR) reactors with a high fraction of Minor Actinides (MA) in the core, requiring remote fuel fabrication behind thick shielding.
Problems of fuel handling are exacerbated if Th is used as fertile isotope (e.g. to enhance safety or TRU-burning rate), since Th-232 irradiation causes the build-up of U-232, whose progeny emits high energy gamma rays. Use of a liquid fuel with online reprocessing would avoid most of the issues related to reprocessing, manufacturing and transporting highly radioactive recycled fuel. The logical technology for the adoption of liquid fuel is the Molten Salt Reactor (MSR). Among MSRs, the Molten Salt Fast Reactor (MSFR) is in principle better suited for TRU burning as it combines the advantages of a liquid fuel with those of a fast-spectrum and of Th use. Objective of this work is to evaluate the MSFR potential benefits in terms of TRU burning through a comparative analysis with a sodium-cooled FR. The comparison is
based on TRU- and MA-burning rates, as well as on the in-core evolution of radiotoxicity and decay heat.
Solubility issues limit the TRU-burning rate to 1/3 that achievable in traditional low-CR FRs. The softer spectrum also determines notable radiotoxicity and decay heat of the equilibrium actinide inventory. On the other hand, the liquid fuel suggests the possibility of using a Pu- free feed composed only of Th and MA, thus maximizing the MA burning rate. This is generally not possible in traditional low-CR FRs due to safety deterioration and decay heat of reprocessed fuel. In addition, the high specific power and the lack of out-of-core cooling times foster a quick transition toward equilibrium, which improves the MSFR capability to burn an initial fissile loading, and makes the MSFR a promising system for a
quick (i.e., in a reactor lifetime) transition from the current U-based fuel cycle to a novel closed Th cycle
THE MSFR AS A FLEXIBLE CR REACTOR: THE VIEWPOINT OF SAFETY
No full textIn recent years efforts have been spent in the development of innovative reactors capable of operating with flexible Conversion Ratio (CR). Fast Reactors (FR) are natural candidates since they allow to achieve high CR, as well as an efficient TRU burning through a low CR and the closure of the fuel cycle. Among the fast-spectrum systems, a peculiar role is played by the Molten Salt Fast Reactor. This reactor lacks the sound technological basis available for the solid-fuelled liquid-metal-cooled FRs, but it shows fuel cycle potential benefits: it uses Th, which features vast natural resources and mitigates waste management issues due to a low generation of TRUs; it can naturally operate with flexible CR without design modifications thanks to the online reprocessing system; it can achieve high CR, with doubling times of the order of
40 years or lower; it can achieve good TRU-burning rates and very high burning rates of minor actinides. However, such fuel cycle flexibility implies a wide variety of fuel salt compositions. Along with the variation of the fuel salt properties, concerns arise for the varying safety features of the core, especially when using the MSFR as TRU- burner. This work first summarizes results regarding the fuel cycle performances of the MSFR when used as breeder, iso-breeder or burner reactor. Subsequently, safety parameters are computed for each fuel cycle option and a simple approach based on reactivity and energy balances is employed to predict the reactor steady-state after major accidental transient initiators, thus giving indications of its inherent safety features for different fuel cycle strategies
Murder on the mountain: author talk with Peter J. Wosh
No full textAuthor talk by Peter J. Wosh on May 5th, 2022, on his book, "Murder on the Mountain: crime, passion, and punishment in gilded age New Jersey.
Mr. Melvin J. Collier, RWWL AUC, June 2011
No full textThis video is a conversation with Mr. Melvin J. Collier. Mr. Collier talks about his book, "From Mississippi to Africa: A Journey of Discovery". Daniel Le, AUC Woodruff Library, is the interviewer
Hybrib Spectrum Molten Salt Reactor
No full textThe Molten Salt Reactor (MSR) concept has a unique feature, if compared to the majority of other reactor designs, that its fuel is liquid. This property creates, on one hand, several technical challenges; on the other hand, it offers flexibility in shaping and designing of the active core. Accordingly, single fuel fluid can, for instance, circulate through several core zones with different moderation ratios. This possibility was already considered in the past; however, in relation to the thermal MSR. In the presented study, an extreme case of hybrid spectrum MSR is proposed and preliminarily analyzed. It is concluded that hybrid spectrum may provide several advantages and could be applicable especially during the initial phase of the fuel cycle or by the transition to equilibrium cycle
Coupled computational fluid dynamics and computational thermodynamics simulations for fission product retention and release: A molten salt fast reactor application
No full textThis study presents a computational capability for fission product retention and release in two-phase, multi-species systems representing Molten Salt Reactors (MSR) with coupled thermal-hydraulics and fuel coolant chemical behaviours. This is demonstrated through four simulated cases centred on the proposed Molten Salt Fast Reactor (MSFR). This is achieved by two-way coupling the Computational Fluid Dynamics (CFD) code OpenFOAM and the Computational Thermodynamics (CT) code Thermochimica, using the Joint Research Centre Molten Salt Database (JRCMSD). Local chemical equilibrium is assumed, implying that chemical kinetics are predominantly governed by mass transport. Four simulations address normal operating conditions, exploring: (i) dilution of fission products injected within the molten salt coolant, (ii) molten salt coolant evaporation rate, (iii) release of radioactive gaseous species, (iv) shifts in the UF4/UF3 ratio, and (v) comparison of vapour pressures of gaseous species. The influence of temperature-dependent viscosity on retaining fission products, compared to consistent values, is also discussed. The feasibility of integrating CFD with Thermochimica showed promising results, broadening insights into multiphysics systems and setting the stage for its application in more intricate scenarios
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