WorldWideScience

Sample records for salt reactor fuels

  1. Fuel processing for molten-salt reactors

    International Nuclear Information System (INIS)

    Hightower, J.R. Jr.

    1976-01-01

    Research devoted to development of processes for the isolation of protactinium and for the removal of fission products from molten-salt breeder reactors is reported. During this report period, engineering development progressed on continuous fluorinators for uranium removal, the metal transfer process for rare-earth removal, the fuel reconstitution step, and molten salt--bismuth contactors to be used in reductive extraction processes. The metal transfer experiment MTE-3B was started. In this experiment all parts of the metal transfer process for rare-earth removal are demonstrated using salt flow rates which are about 1 percent of those required to process the fuel salt in a 1000-MW(e) MSBR. During this report period the salt and bismuth phases were transferred to the experimental vessels, and two runs with agitator speeds of 5 rps were made to measure the rate of transfer of neodymium from the fluoride salt to the Bi--Li stripper solution. The uranium removed from the fuel salt by fluorination must be returned to the processed salt in the fuel reconstitution step before the fuel salt is returned to the reactor. An engineering experiment to demonstrate the fuel reconstitution step is being installed. In this experiment gold-lined equipment will be used to avoid introducing products of corrosion by UF 6 and UF 5 . Alternative methods for providing the gold lining include electroplating and mechanical fabrication

  2. Parametric studies on the fuel salt composition in thermal molten salt breeder reactors

    International Nuclear Information System (INIS)

    Nagy, K.; Kloosterman, J.L.; Lathouwers, D.; Van der Hagen, T.H.J.J.

    2008-01-01

    In this paper the salt composition and the fuel cycle of a graphite moderated molten salt self-breeder reactor operating on the thorium cycle is investigated. A breeder molten salt reactor is always coupled to a fuel processing plant which removes the fission products and actinides from the core. The efficiency of the removal process(es) has a large influence on the breeding capacity of the reactor. The aim is to investigate the effect on the breeding ratio of several parameters such as the composition of the molten salt, moderation ratio, power density and chemical processing. Several fuel processing strategies are studied. (authors)

  3. Characterization of the molten salt reactor experiment fuel and flush salts

    International Nuclear Information System (INIS)

    Williams, D.F.; Peretz, F.J.

    1996-01-01

    Wise decisions about the handling and disposition of spent fuel from the Molten Salt Reactor Experiment (MSRE) must be based upon an understanding of the physical, chemical, and radiological properties of the frozen fuel and flush salts. These open-quotes staticclose quotes properties can be inferred from the extensive documentation of process history maintained during reactor operation and the knowledge gained in laboratory development studies. Just as important as the description of the salt itself is an understanding of the dynamic processes which continue to transform the salt composition and govern its present and potential physicochemical behavior. A complete characterization must include a phenomenological characterization in addition to the typical summary of properties. This paper reports on the current state of characterization of the fuel and flush salts needed to support waste management decisions

  4. Thermodynamic characterization of the molten salt reactor fuel - 5233

    International Nuclear Information System (INIS)

    Capelli, E.; Konings, R.J.M.; Benes, O.

    2015-01-01

    The Molten Salt Reactor (MSR) has been selected as one of the Generation IV nuclear systems. The very unique feature of this reactor concept is the liquid nature of the fuel which offers numerous advantages concerning the reactor safety. Nowadays, the research in Europe is focused on an innovative concept, the MSFR (Molten Salt Fast Reactor), that combines the generic assets of molten salt as liquid fuel with those related to fast neutron reactors and the thorium fuel cycle. For the design and safety assessment of the MSFR concept, it is extremely important to have a thorough knowledge of the physico-chemical properties of fluorides salts, which is the class of materials that is the best suited for nuclear applications. Potential chemical systems have been critically reviewed and an extensive thermodynamic database describing the most relevant systems has been created at the Institute for Transuranium Elements of the Joint Research Centre (JRC). Thermochemical equilibrium calculations are a very important tool that allows the evaluation of the performance of several salt mixtures predicting their properties and thus the optimization of the fuel composition. The work combines the experimental determination of different salt properties with the modelling of the thermodynamic functions, using the Calphad method. An overview of the experimental work and the thermodynamic assessments will be given in this paper and different fuel options for the MSFR will be discussed. (authors)

  5. Static fuel molten salt reactors - simpler, cheaper and safer

    International Nuclear Information System (INIS)

    Scott, Ian

    2015-01-01

    The many conceptual designs for Molten Salt Reactors (MSR's) today are all evolutions from the prototype MSR that went critical at Oak Ridge 50 years ago. Critically, they are based on pumping the molten fuel salt from a reaction chamber where the fuel achieves critical mass through a heat exchanger where the resulting heat is transferred to another working fluid. This basic concept was not the first idea that the Oak Ridge scientists considered. Their initial preference was to put the molten salt fuel into tubes, just like solid fuel pellets in their cladding, and circulate a coolant past the tubes. They concluded however that the low thermal conductivity of the salt meant that the tubes could be no wider than 2mm which would be entirely impractical. In this analysis they ignored the contribution of convection to heat transfer in fluids, probably because they were designing an aircraft engine where varying g forces would make convection unreliable. Moltex Energy has re-examined this decision using the modern tools of computational fluid dynamics to simulate convective flow in the molten salt and discovered that in fact tubes of similar diameter to those used for solid fuels are entirely practical. Power densities of 250kW/litre of fuel salt are readily attainable providing a higher overall power density than a PWR reactor. This discovery permits MSR's to be built without any of the complex pumping, passively safe drain systems, on line degassing, filtration and chemical processing needed in pumped MSR's. Their design is very simple and they have many intrinsic safety factors including low pressure operation, chemically unreactive fluids and strongly negative fuel thermal and coolant voiding reactivity coefficients. Most importantly, the highly radioactive fission products are retained in non-volatile form within the fuel tubes in the reactor core. Radioactive fuel salt never leaves the reactor vessel except in an immobile frozen form during

  6. Development of fuel cycle technology for molten-salt reactor systems

    International Nuclear Information System (INIS)

    Uhlir, J.

    2006-01-01

    Full text: Full text: The Molten-Salt Reactor (MSR) represents one of promising advanced reactor type assigned to the GEN IV reactor systems. It can be operated either as thorium breeder within the Th -133U fuel cycle or as actinide transmuter incinerating transuranium fuel. Essentially the main advantage of MSR comes out from the prerequisite, that this reactor type should be directly connected with the 'on-line' reprocessing of circulating liquid (molten-salt) fuel. This principle should allow very effective extraction of freshly constituted fissile material (233U). Besides, the on-line fuel salt clean up is necessary within a long run to keep the reactor in operation. As a matter of principle, it permits to clear away typical reactor poisons like xenon, krypton, lanthanides etc. and possibly also other products of burned plutonium and transmuted minor actinides. The fuel salt clean up technology should be linked with the fresh MSR fuel processing to continuously refill the new fuel (thorium or transuranics) into the reactor system. On the other hand, the technologies of fresh transuranium molten-salt fuel processing from the current LWR spent fuel and of the on-line reprocessing of MSR fuel represent two killing points of the whole MSR technology, which have to be successfully solved before MSR deployment in the future. There are three main pyrochemical partitioning techniques proposed for processing and/or reprocessing of MSR fuel: Fluoride volatilization processes, Molten salt / liquid metal extraction processes and Electrochemical separation processes. Two of them - Fluoride Volatility Method and Electrochemical separation process from fluoride media are under development in the Nuclear Research Institute Rez pic. R and D in the field of Fluoride Volatility Method is concentrated to the development and verification of experimental semi-pilot technology for LWR spent fuel reprocessing, which may result in a product the form and composition of which might be

  7. Disposition of the fluoride fuel and flush salts from the Molten Salt Reactor experiment at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Peretz, F.J.

    1996-01-01

    The Molten Salt Reactor Experiment (MSRE) is an 8 MW reactor that was operated at Oak Ridge National Laboratory (ORNL) from 1965 through 1969. The reactor used a unique liquid salt fuel, composed of a mixture of LIF, BeF 2 , ZrF 4 , and UF 4 , and operated at temperatures above 600 degrees C. The primary fuel salt circulation system consisted of the reactor vessel, a single fuel salt pump, and a single primary heat exchanger. Heat was transferred from the fuel salt to a coolant salt circuit in the primary heat exchanger. The coolant salt was similar to the fuel salt, except that it contains only LiF (66%) and BeF, (34%). The coolant salt passed from the primary heat exchanger to an air-cooled radiator and a coolant salt pump, and then returned to the primary heat exchanger. Each of the salt loops was provided with drain tanks, located such that the salt could be drained out of either circuit by gravity. A single drain tank was provided for the non-radioactive coolant salt. Two drain tanks were provided for the fuel salt. Since the fuel salt contained radioactive fuel, fission products, and activation products, and since the reactor was designed such that the fuel salt could be drained immediately into the drain tanks in the event of a problem in the fuel salt loop, the fuel salt drain tanks were provided with a system to remove the heat generated by radioactive decay. A third drain tank connected to the fuel salt loop was provided for a batch of flush salt. This batch of salt, similar in composition to the coolant salt, was used to condition the fuel salt loop after it had been exposed to air and to flush the fuel salt loop of residual fuel salt prior to accessing the reactor circuit for maintenance or experimental activities. This report discusses the disposition of the fluoride fuel and flush salt

  8. Fast molten salt reactor-transmuter for closing nuclear fuel cycle on minor actinides

    International Nuclear Information System (INIS)

    Dudnikov, A. A.; Alekseev, P. N.; Subbotin, S. A.

    2007-01-01

    Creation fast critical molten salt reactor for burning-out minor actinides and separate long-living fission products in the closed nuclear fuel cycle is the most perspective and actual direction. The reactor on melts salts - molten salt homogeneous reactor with the circulating fuel, working as burner and transmuter long-living radioactive nuclides in closed nuclear fuel cycle, can serve as an effective ecological cordon from contamination of the nature long-living radiotoxic nuclides. High-flux fast critical molten-salt nuclear reactors in structure of the closed nuclear fuel cycle of the future nuclear power can effectively burning-out / transmute dangerous long-living radioactive nuclides, make radioisotopes, partially utilize plutonium and produce thermal and electric energy. Such reactor allows solving the problems constraining development of large-scale nuclear power, including fueling, minimization of radioactive waste and non-proliferation. Burning minor actinides in molten salt reactor is capable to facilitate work solid fuel power reactors in system NP with the closed nuclear fuel cycle and to reduce transient losses at processing and fabrications fuel pins. At substantiation MSR-transmuter/burner as solvents fuel nuclides for molten-salt reactors various salts were examined, for example: LiF - BeF2; NaF - LiF - BeF2; NaF-LiF ; NaF-ZrF4 ; LiF-NaF -KF; NaCl. RRC 'Kurchatov institute' together with other employees have developed the basic design reactor installations with molten salt reactor - burner long-living nuclides for fluoride fuel composition with the limited solubility minor actinides (MAF3 10 mol %) allows to develop in some times more effective molten salt reactor with fast neutron spectrum - burner/ transmuter of the long-living radioactive waste. In high-flux fast reactors on melts salts within a year it is possible to burn ∼300 kg minor actinides per 1 GW thermal power of reactor. The technical and economic estimation given power

  9. Molten-salt reactor strategies viewed from fuel conservation effect, (1)

    International Nuclear Information System (INIS)

    Furuhashi, Akira

    1976-01-01

    Saving of material requirements in the long-term fuel cycle is studied by introducing molten-salt reactors with good neutron economy into a projection of nuclear generating capacity in Japan. In this first report an examination is made on the effects brought by the introduction of molten-salt converter reactors starting with Pu which are followed by 233 U breeders of the same type. It is shown that the sharing of some Pu in the light water- and fast breeder-reactor system with molten-salt reactors provides a more rapid transition to the self-supporting, breeding cycle than the simple fast breeding system, thus leading to an appreciable fuel conservation. Considerations are presented on the strategic repartition of generating capacity among reactor types and it is shown that all of the converted 233 U should be promptly invested to molten-salt breeders to quickly establish the dual breeding system, instead of recycling to converters themselves. (auth.)

  10. Complete Sensitivity/Uncertainty Analysis of LR-0 Reactor Experiments with MSRE FLiBe Salt and Perform Comparison with Molten Salt Cooled and Molten Salt Fueled Reactor Models

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mueller, Don [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-12-01

    In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in 7Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.

  11. Molten salt breeder reactor

    International Nuclear Information System (INIS)

    1977-01-01

    MSBR Study Group formed in October 1974 has studied molten salt breeder reactor and its various aspects. Usage of a molten salt fuel, extremely interesting as reactor chemistry, is a great feature to MSBR; there is no need for separate fuel making, reprocessing, waste storage facilities. The group studied the following, and these results are presented: molten salt technology, molten salt fuel chemistry and reprocessing, reactor characteristics, economy, reactor structural materials, etc. (Mori, K.)

  12. Molten salt reactors

    International Nuclear Information System (INIS)

    Bouchter, J.C.; Dufour, P.; Guidez, J.; Simon, N.; Renault, C.

    2014-01-01

    Molten salt reactors are one of the 6 concepts retained for the 4. generation of nuclear reactors. The principle of this reactor is very innovative: the nuclear fuel is dissolved in the coolant which allows the online reprocessing of the fuel and the online recovery of the fission products. A small prototype: the Molten Salt Reactor Experiment (MSRE - 8 MWt) was operating a few years in the sixties in the USA. The passage towards a fast reactor by the suppression of the graphite moderator leads to the concept of Molten Salt Fast Reactor (MSFR) which is presently studied through different European projects such as MOST, ALISIA and EVOL. Worldwide the main topics of research are: the adequate materials resisting to the high level of corrosiveness of the molten salts, fuel salt reprocessing, the 3-side coupling between neutron transport, thermohydraulics and thermo-chemistry, the management of the changing chemical composition of the salt, the enrichment of lithium with Li 7 in the case of the use of lithium fluoride salt and the use of MSFR using U 233 fuel (thorium cycle). The last part of the article presents a preliminary safety analysis of the MSFR. (A.C.)

  13. Prospects of subcritical molten salt reactor for minor actinides incineration in closed fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, Pavel N.; Balanin, Andrey L.; Dudnikov, Anatoly A.; Fomichenko, Petr A.; Nevinitsa, Vladimir A.; Frolov, Aleksey A.; Lubina, Anna S.; Sedov, Aleksey A.; Subbotin, Aleksey S.; Blandinsky, Viktor Yu. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2015-09-15

    A subcritical molten salt reactor is proposed for minor actinides (separated from spent fuel VVER-1000 light water reactor) incineration and for {sup 233}U conversion from {sup 232}Th. Here the subcritical molten salt reactor with fuel composition of heavy nuclide fluorides in molten LiF - NaF - KF salt and with external neutron source, based on 1 GeV proton accelerator and molten salt cooled tungsten target is considered. The paper presents the results of parametrical analysis of equilibrium nuclide composition of molten salt reactor with minor actinides feed in dependence of core dimensions, average neutron flux and external neutron source intensity. Reactor design is defined; requirements to external neutron source are posed; heavy nuclides equilibrium and fuel cycle main parameters are calculated.

  14. Fuel cycle cost analysis on molten-salt reactors

    International Nuclear Information System (INIS)

    Shimazu, Yoichiro

    1976-01-01

    An evaluation is made of the fuel cycle costs for molten-salt reactors (MSR's), developed at Oak Ridge National Laboratory. Eight combinations of conditions affecting fuel cycle costs are compared, covering 233 U-Th, 235 U-Th and 239 Pu-Th fuels, with and without on-site continuous fuel reprocessing. The resulting fuel cycle costs range from 0.61 to 1.18 mill/kWh. A discussion is also given on the practicability of these fuel cycles. The calculations indicate that somewhat lower fuel cycle costs can be expected from reactor operation in converter mode on 235 U make-up with fuel reprocessed in batches every 10 years to avoid fission product precipitation, than from operation as 233 U-Th breeder with continuous reprocessing. (auth.)

  15. Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt

    Institute of Scientific and Technical Information of China (English)

    ZHANG Da-Lin; QIU Sui-Zheng; LIU Chang-Liang; SU Guang-Hui

    2008-01-01

    The Molten Salt Reactor (MSR),one of the‘Generation Ⅳ'concepts,is a liquid-fuel reactor,which is different from the conventional reactors using solid fissile materials due to the flow effect of fuel salt.The study on its neutronice considering the fuel salt flow,which is the base of the thermal-hydraulic calculation and safety analysis,must be done.In this paper,the theoretical model on neutronics under steady condition for a single-liquid-fueled MSR is conducted and calculated by numerical method.The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes,and balance equations for six-group delayed neutron precursors considering the flow effect of fuel salt. The spatial discretization of the above models is based on the finite volume method,and the discretization equations are computed by the source iteration method.The distributions of neutron fluxes and the distributions of the delayed neutron precursors in the core are obtained.The numerical calculated results show that,the fuel salt flow has little effect on the distribution of fast and thermal neutron fluxes and the effective multiplication factor;however,it affects the distribution of the delayed neutron precursors significantly,especially the long-lived one.In addition,it could be found that the delayed neutron precursors influence the nentronics slightly under the steady condition.

  16. Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt

    International Nuclear Information System (INIS)

    Zhang Dalin; Qiu Suizheng; Su Guanghui; Liu Changliang

    2008-01-01

    The Molten Salt Reactor (MSR), one of the 'Generation IV' concepts, is a liquid-fuel reactor, which is different from the conventional reactors using solid fissile materials due to the flow effect of fuel salt. The study on its neutronics considering the fuel salt flow, which is the base of the thermal-hydraulic calculation and safety analysis, must be done. In this paper, the theoretical model on neutronics under steady condition for a single-liquid-fueled MSR is conducted and calculated by numerical method. The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six-group delayed neutron precursors considering the flow effect of fuel salt. The spatial discretization of the above models is based on the finite volume method, and the discretization equations are computed by the source iteration method. The distributions of neutron fluxes and the distributions of the delayed neutron precursors in the core are obtained. The numerical calculated results show that, the fuel salt flow has little effect on the distribution of fast and thermal neutron fluxes and the effective multiplication factor; however, it affects the distribution of the delayed neutron precursors significantly, especially the long-lived one. In addition, it could be found that the delayed neutron precursors influence the neutronics slightly under the steady condition. (authors)

  17. Parametric study of natural circulation flow in molten salt fuel in molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pauzi, Anas Muhamad, E-mail: Anas@uniten.edu.my [Centre of Nuclear Energy, Universiti Tenaga Nasional (UNITEN), Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia); Cioncolini, Andrea; Iacovides, Hector [School of Mechanical, Aerospace, and Civil Engineering (MACE), University of Manchester, Oxford Road, M13 9PL Manchester (United Kingdom)

    2015-04-29

    The Molten Salt Reactor (MSR) is one of the most promising system proposed by Generation IV Forum (GIF) for future nuclear reactor systems. Advantages of the MSR are significantly larger compared to other reactor system, and is mainly achieved from its liquid nature of fuel and coolant. Further improvement to this system, which is a natural circulating molten fuel salt inside its tube in the reactor core is proposed, to achieve advantages of reducing and simplifying the MSR design proposed by GIF. Thermal hydraulic analysis on the proposed system was completed using a commercial computation fluid dynamics (CFD) software called FLUENT by ANSYS Inc. An understanding on theory behind this unique natural circulation flow inside the tube caused by fission heat generated in molten fuel salt and tube cooling was briefly introduced. Currently, no commercial CFD software could perfectly simulate natural circulation flow, hence, modeling this flow problem in FLUENT is introduced and analyzed to obtain best simulation results. Results obtained demonstrate the existence of periodical transient nature of flow problem, hence improvements in tube design is proposed based on the analysis on temperature and velocity profile. Results show that the proposed system could operate at up to 750MW core power, given that turbulence are enhanced throughout flow region, and precise molten fuel salt physical properties could be defined. At the request of the authors and the Proceedings Editor the name of the co-author Andrea Cioncolini was corrected from Andrea Coincolini. The same name correction was made in the Acknowledgement section on page 030004-10 and in reference number 4. The updated article was published on 11 May 2015.

  18. Fuel cycle costs for molten-salt reactors

    International Nuclear Information System (INIS)

    Nagashima, Kikusaburo

    1983-01-01

    This report describes FCC (fuel cycle cost) estimates for MSCR (molten-salt converter reactor) and MSBR (molten-salt breeder reactor) compared with those for LWRs (PWR and BWR). The calculation is based on the present worth technique with a given discount rate for each cost item, which enables us to make comparison between FCC's for MSCR, MSBR and LWRs. As far as the computational results obtained here are concerned, shown that the FCC's for MSCR and MSBR are 70 -- 60 % lower than the values for LWRs. And it could be said that the FCC for MSCR (Pu-converter) is about 10 % lower than that for MSBR, because of the smaller amount of fissile inventory of MSCR than the inventory of MSBR. (author)

  19. Molten salt reactor concept

    International Nuclear Information System (INIS)

    Sood, D.D.

    1980-01-01

    Molten salt reactor is an advanced breeder concept which is suited for the utilization of thorium for nuclear power production. This reactor is based on the use of solutions of uranium or plutonium fluorides in LiF-BeF 2 -ThF 4 as fuel. Unlike the conventional reactors, no external coolant is used in the reactor core and the fuel salt itself is circulated through heat exchangers to transfer the fission produced heat to a secondary salt (NaF-NaBF 4 ) for steam generation. A part of the fuel stream is continuously processed to isolate 233 Pa, so that it can decay to fissile 233 U without getting converted to 234 Pa, and for the removal of neutron absorbing fission products. This on-line processing scheme makes this reactor concept to achieve a breeding ratio of 1.07 which is the highest for any thermal breeder reactor. Experimental studies at the Bhabha Atomic Research Centre, Bombay, have established the use of plutonium as fuel for this reactor. This molten salt reactor concept is described and the work conducted at the Bhabha Atomic Research Centre is summarised. (auth.)

  20. Molten-salt converter reactors

    International Nuclear Information System (INIS)

    Perry, A.M.

    1975-01-01

    Molten-salt reactors appear to have substantial promise as advanced converters. Conversion ratios of 0.85 to 0.9 should be attainable with favourable fuel cycle costs, with 235 U valued at $12/g. An increase in 235 U value by a factor of two or three ($10 to $30/lb. U 3 O 8 , $75/SWU) would be expected to increase the optimum conversion ratio, but this has not been analyzed in detail. The processing necessary to recover uranium from the fuel salt has been partially demonstrated in the MSRE. The equipment for doing this would be located at the reactor, and there would be no reliance on an established recycle industry. Processing costs are expected to be quite low, and fuel cycle optimization depends primarily on inventory and burnup or replacement costs for the fuel and for the carrier salt. Significant development problems remain to be resolved for molten-salt reactors, notably the control of tritium and the elimination of intergranular cracking of Hastelloy-N in contact with tellurium. However, these problems appear to be amenable to solution. It is appropriate to consider separating the development schedule for molten-salt reactors from that for the processing technology required for breeding. The Molten-Salt Converter Reactor should be a useful reactor in its own right and would be an advance towards the achievement of true breeding in thermal reactors. (author)

  1. Open problems in reprocessing of a molten salt reactor fuel

    International Nuclear Information System (INIS)

    Lelek, Vladimir; Vocka, Radim

    2000-01-01

    The study of fuel cycle in a molten salt reactor (MSR) needs deeper understanding of chemical methods used for reprocessing of spent nuclear fuel and preparation of MSR fuel, as well as of the methods employed for reprocessing of MSR fuel itself. Assuming that all the reprocessing is done on the basis of electrorefining, we formulate some open questions that should be answered before a flow sheet diagram of the reactor is designed. Most of the questions concern phenomena taking place in the vicinity of an electrode, which influence the efficiency of the reprocessing and sensibility of element separation. Answer to these questions would be an important step forward in reactor set out. (Authors)

  2. Thermodynamic characterization of salt components for the Molten Salt Reactor Fuel - 15573

    International Nuclear Information System (INIS)

    Capelli, E.; Konings, R.J.M.; Benes, A.

    2015-01-01

    Molten fluoride salts are considered as primary candidates for nuclear fuel in the Molten Salt Reactor (MSR), one of the 6 generation IV nuclear reactor designs. In order to determine the safety limits and to access the properties of the potential fuel mixtures, thermodynamic studies are very important. This study is a combination of experimental work and thermodynamic modelling and focusses on the fluoride systems with alkaline and alkaline earth fluorides as matrix and ThF 4 , UF 4 and PuF 3 as fertile and fissile materials. The purification of the single components was considered as essential first step for the study of more complex systems and ternary phase diagrams were described using Differential Scanning Calorimetry (DSC) and drop calorimetry, which are used to measure phase transitions, enthalpy of mixing and heat capacity. In addition to the calorimetric techniques, Knudsen Effusion Mass Spectrometry (KEMS) and X-ray Diffraction (XRD) were used to collect data on vapour pressure and crystal structure of fluorides. The results are then coupled with thermodynamic modelling using the Calphad method for the assessment of the phase diagrams. A thermodynamic database describing the most important systems for MSR application has been developed and it has been used to optimize the fuel composition in view of the relevant properties such as melting temperature. A reliable database of thermodynamic properties of fluoride salts has been generated. It includes the key systems for the MSR fuel and it is very useful to predict the properties of the fuel

  3. Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

    International Nuclear Information System (INIS)

    Aji, Indarta Kuncoro; Waris, Abdul; Permana, Sidik

    2015-01-01

    Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF 2 -ThF 4 - 233 UF 4 respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155

  4. Achieving salt-cooled reactor goals: economics, variable electricity, no major fuel failures - 15118

    International Nuclear Information System (INIS)

    Forsberg, C.

    2015-01-01

    The Fluoride-salt-cooled High-temperature Reactor (FHR) with a Nuclear air-Brayton Combined Cycle (NACC) and Firebrick Resistance-Heated Energy Storage (FIRES) is a new reactor concept. The FHR uses High-Temperature Gas-cooled Reactor (HTGR) coated-particle fuel and liquid-salt coolants originally developed for molten salt reactors (MSRs) where the fuel was dissolved in the coolant. The FIRES system consists of high-temperature firebrick heated to high temperatures with electricity at times of low electric prices. For a modular FHR operating with a base-load 100 MWe output, the station output can vary from -242 MWe to +242 MWe. The FHR can be built in different sizes. The reactor concept was developed using a top-down approach: markets, requirements, reactor design. The goals are: (1) increase plant revenue by 50 to 100% relative to base-load nuclear plants with capital costs similar to light-water reactors, (2) enable a zero-carbon nuclear renewable electricity grid, and (3) no potential for major fuel failure and thus no potential for major radionuclide offsite releases in a beyond-design-basis accident (BDBA). The basis for the goals and how they may be achieved is described

  5. The concept of fuel cycle integrated molten salt reactor for transmuting Pu+MA from spent LWR fuels

    International Nuclear Information System (INIS)

    Hirose, Y.; Takashima, Y.

    2001-01-01

    Japan should need a new fuel cycle, not to save spent fuels indefinitely as the reusable resources but to consume plutonium and miner actinides orderly without conventional reprocessing. The key component is a molten salt reactor fueled with the Pu+MA (PMA) separated from LWR spent fuels using fluoride volatility method. A double-tiered once-through reactor system can burn PMA down to 5% remnant ratio, and can make PMA virtually free from the HAW to be disposed geometrically. A key issue to be demonstrated is the first of all solubility behavior of trifluoride species in the molten fuel salt of 7 LiF-BeF 2 mixture. (author)

  6. Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

    Energy Technology Data Exchange (ETDEWEB)

    Aji, Indarta Kuncoro, E-mail: indartaaji@s.itb.ac.id [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia); Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Permana, Sidik [Nuclear Physics & Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia)

    2015-09-30

    Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF{sub 2}-ThF{sub 4}-{sup 233}UF{sub 4} respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.

  7. Moltex Energy's stable salt reactors

    International Nuclear Information System (INIS)

    O'Sullivan, R.; Laurie, J.

    2016-01-01

    A stable salt reactor is a molten salt reactor in which the molten fuel salt is contained in fuel rods. This concept was invented in 1951 and re-discovered and improved recently by Moltex Energy Company. The main advantage of using molten salt fuel is that the 2 problematic fission products cesium and iodine do not exist in gaseous form but rather in a form of a salt that present no danger in case of accident. Another advantage is the strongly negative temperature coefficient for reactivity which means the reactor self-regulates. The feasibility studies have been performed on a molten salt fuel composed of sodium chloride and plutonium/uranium/lanthanide/actinide trichloride. The coolant fluid is a mix of sodium and zirconium fluoride salts that will need low flow rates. The addition of 1 mol% of metal zirconium to the coolant fluid reduces the risk of corrosion with standard steels and the addition of 2% of hafnium reduces the neutron dose. The temperature of the coolant is expected to reach 650 Celsius degrees at the exit of the core. This reactor is designed to be modular and it will be able to burn actinides. (A.C.)

  8. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering; Greenspan, Ehud [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering

    2015-02-09

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X

  9. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

    International Nuclear Information System (INIS)

    Peterson, Per; Greenspan, Ehud

    2015-01-01

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m 3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m 3 . This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses novel

  10. Molten salt fueled reactors with a fast salt draining

    International Nuclear Information System (INIS)

    Ventre, Edmond; Blum, J.M.

    1976-01-01

    This invention relates to a molten salt nuclear reactor which comprises a new arrangement for shutting it down in complete safety. This nuclear reactor has a molten salt primary circuit comprising, in particular, the core of this reactor. It includes a leak tight vessel the capacity of which is appreciably greater than that of the molten salt volume of the circuit and placed so that the level of the molten salt, when all the molten salt of the circuit is contained in this vessel, is less than that of the base of the core. There are facilities for establishing and maintaining an inert gas pressure in the vessel above the molten salt, for releasing the compressed gas and for connecting the vessel to the primary circuit entering this vessel at a lower level than that of the molten salt and enabling molten salt to enter or leave the vessel according to the pressure of the inert gas. The particular advantage of this reactor is that it can be shut down safely since the draining of the primary circuit no longer results from a 'positive action' but from the suppression of an arrangement essential for the operation of the reactor consisting of the build-up of the said inert gas pressure in the said vessel [fr

  11. Molten salt reactors. The AMSTER concept

    International Nuclear Information System (INIS)

    Vergnes, J.; Garzenne, C.; Lecarpentier, D.; Mouney, H.

    2001-01-01

    This article presents the concept of actinide molten salt transmuter (AMSTER). This reactor is graphite-moderated and is dedicated to the burning of actinides. The main difference with a molten salt reactor is that its liquid fuel undergoes an on-line partial reprocessing in which fission products are extracted and heavy nuclei are reintroduced into the fuel. In order to maintain the reactivity regular injections of 235 U-salt are made. In classical reactors, fuel burn-up is limited by the swelling of the cladding and the radiation fuel pellets resistance, in AMSTER there is no limitation to the irradiation time of the fuel, so all the actinides can be burnt or transmuted. (A.C.)

  12. Investigation of molten salt fast reactor

    International Nuclear Information System (INIS)

    Kubota, Kenichi; Konomura, Mamoru

    2002-01-01

    On survey research for practicability strategy of fast reactor (FR) (phase 1), to extract future practicability image candidates of FR from wide options, in addition to their survey and investigation objects of not only solid fuel reactors of conventional research object but also molten salt reactor as a flowing fuel reactor, investigation on concept of molten salt FR plant was carried out. As a part of the first step of the survey research for practicability strategy, a basic concept on plant centered at nuclear reactor facility using chloride molten salt reactor capable of carrying out U-Pu cycle was examined, to perform a base construction to evaluate economical potential for a practical FBR. As a result, a result could be obtained that because of inferior fuel inventory and heat transmission to those in Na cooling reactor in present knowledge, mass of reactor vessel and intermediate heat exchanger were to widely increased to expect reduction of power generation unit price even on considering cheapness of its fuel cycle cost. Therefore, at present step further investigation on concept design of the chloride molten salt reactor plant system is too early in time, and it is at a condition where basic and elementary researches aiming at upgrading of economical efficiency such as wide reduction of fuel inventory, a measure expectable for remarkable rationalization effect of reprocessing system integrating a reactor to a processing facility, and so on. (G.K.)

  13. Chemistry and technology of Molten Salt Reactors - history and perspectives

    International Nuclear Information System (INIS)

    Uhlir, Jan

    2007-01-01

    Molten Salt Reactors represent one of promising future nuclear reactor concept included also in the Generation IV reactors family. This reactor type is distinguished by an extraordinarily close connection between the reactor physics and chemical technology, which is given by the specific features of the chemical form of fuel, representing by molten fluoride salt and circulating through the reactor core and also by the requirements of continuous 'on-line' reprocessing of the spent fuel. The history of Molten Salt Reactors reaches the period of fifties and sixties, when the first experimental Molten Salt Reactors were constructed and tested in ORNL (US). Several molten salt techniques dedicated to fresh molten salt fuel processing and spent fuel reprocessing were studied and developed in those days. Today, after nearly thirty years of discontinuance, a renewed interest in the Molten Salt Reactor technology is observed. Current experimental R and D activities in the area of Molten Salt Reactor technology are realized by a relatively small number of research institutions mainly in the EU, Russia and USA. The main effort is directed primarily to the development of separation processes suitable for the molten salt fuel processing and reprocessing technology. The techniques under development are molten salt/liquid metal extraction processes, electrochemical separation processes from the molten salt media, fused salt volatilization techniques and gas extraction from the molten salt medium

  14. The molten salt reactors (MSR) pyro chemistry and fuel cycle for innovative nuclear systems

    International Nuclear Information System (INIS)

    Brossard, Ph.; Garzenne, C.; Mouney, H.

    2002-01-01

    In the frame of the studies on next generation nuclear systems, and especially for the molten salt reactors and for the integrated fuel cycle (as IFR), the fuel cycle constraints must be taken into account in the preliminary studies of the system to improve the cycle and reactor optimisation. Among the purposes for next generation nuclear systems, sustainability and waste (radio-toxicity and mass) management are important goals. These goals imply reprocessing and recycling strategies. The objectives of this workshop are to present and to share the different strategies and scenarios, the needs based on these scenarios, the experimental facilities available today or in the future and their capabilities, the needs for demonstration. It aims at: identifying the needs for fuel cycle based on solid fuel or liquid fuel, and especially, the on-line reprocessing or clean up for the molten salt reactors; assessing the state-of-the-art on the pyro-chemistry applied to solid fuel and to present the research activities; assessing the state-of-the-art on liquid fuels (or others), and to present the research activities; expressing the R and D programs for pyro-chemistry, molten salt, and also to propose innovative processes; and proposing some joint activities in the frame of GEDEON and PRACTIS programs. This document brings together the transparencies of 18 contributions dealing with: scenario studies with AMSTER concept (Scenarios, MSR, breeders (Th) and burners); fuel cycle for innovative systems; current reprocessing of spent nuclear fuel (SNF) in molten salts (review of pyro-chemistry processes (non nuclear and nuclear)); high temperature NMR spectroscopies in molten salts; reductive extraction of An from molten fluorides (salt - liquid metal extraction); electrochemistry characterisation; characterisation with physical methods - extraction coefficient and kinetics; electrolytic extraction; dissolution-precipitation of plutonium in the eutectic LiCl-KCl (dissolution and

  15. Investigation of an Alternative Fuel Form for the Liquid Salt Cooled Very High Temperature Reactor (LS-VHTR)

    International Nuclear Information System (INIS)

    Casino, William A. Jr.

    2006-01-01

    Much of the recent studies investigating the use of liquid salts as reactor coolants have utilized a core configuration of graphite prismatic fuel block assemblies with TRISO particles embedded into cylindrical fuel compacts arranged in a triangular pitch lattice. Although many calculations have been performed for this fuel form in gas cooled reactors, it would be instructive to investigate whether an alternative fuel form may yield improved performance for the liquid salt-cooled Very High Temperature Reactor (LS-VHTR). This study investigates how variations in the fuel form will impact the performance of the LS-VHTR during normal and accident conditions and compares the results with a similar analysis that was recently completed for a LS-VHTR core made up of prismatic block fuel. (author)

  16. The molten salt reactor adventure

    International Nuclear Information System (INIS)

    MacPherson, H.G.

    1985-01-01

    A personal history of the development of molten salt reactors in the United States is presented. The initial goal was an aircraft propulsion reactor, and a molten fluoride-fueled Aircraft Reactor Experiment was operated at Oak Ridge National Laboratory in 1954. In 1956, the objective shifted to civilian nuclear power, and reactor concepts were developed using a circulating UF 4 -ThF 4 fuel, graphite moderator, and Hastelloy N pressure boundary. The program culminated in the successful operation of the Molten Salt Reactor Experiment in 1965 to 1969. By then the Atomic Energy Commission's goals had shifted to breeder development; the molten salt program supported on-site reprocessing development and study of various reactor arrangements that had potential to breed. Some commercial and foreign interest contributed to the program which, however, was terminated by the government in 1976. The current status of the technology and prospects for revived interest are summarized

  17. Fluoride-Salt-Cooled High-Temperature Reactor (FHR) with Silicon-Carbide-Matrix Coated-Particle Fuel

    International Nuclear Information System (INIS)

    Forsberg, C. W.; Snead, Lance Lewis; Katoh, Yutai

    2012-01-01

    The FHR is a new reactor concept that uses coated-particle fuel and a low-pressure liquid-salt coolant. Its neutronics are similar to a high-temperature gas-cooled reactor (HTGR). The power density is 5 to 10 times higher because of the superior cooling properties of liquids versus gases. The leading candidate coolant salt is a mixture of 7 LiF and BeF 2 (FLiBe) possessing a boiling point above 1300 C and the figure of merit ρC p (volumetric heat capacity) for the salt slightly superior to water. Studies are underway to define a near-term base-line concept while understanding longer-term options. Near-term options use graphite-matrix coated-particle fuel where the graphite is both a structural component and the primary neutron moderator. It is the same basic fuel used in HTGRs. The fuel can take several geometric forms with a pebble bed being the leading contender. Recent work on silicon-carbide-matrix (SiCm) coated-particle fuel may create a second longer-term fuel option. SiCm coated-particle fuels are currently being investigated for use in light-water reactors. The replacement of the graphite matrix with a SiCm creates a new family of fuels. The first motivation behind the effort is to take advantage of the superior radiation resistance of SiC compared to graphite in order to provide a stable matrix for hosting coated fuel particles. The second motivation is a much more rugged fuel under accident, repository, and other conditions.

  18. Physical and chemical feasibility of fueling molten salt reactors with TRU's trifluorides

    International Nuclear Information System (INIS)

    Ignatiev, V.; Feinberg, O.; Konakov, S.; Subbotine, S.; Surenkov, A.; Zakirov, R.

    2001-01-01

    The molten salt reactor (MSR) concept is very important for consideration as an element of future nuclear energy systems. These reactor systems are unique in many ways. Particularly, the MSRs appear to have substantial promise not only as advanced TRU free system operating in U-Th cycle, but also as transmuter of TRU. Physical and chemical feasibility of fueling MSR with TRU trifluorides is examined. Solvent compositions with and without U-Th as fissile / fertile addition are considered. The principle reactor and fuel cycle variables available for optimizing the performance of MSR as TRU transmuting system are discussed. These efforts led to the definition in minimal TRU mass flow rate, reduced total losses to waste and maximum possible burn up rate for the molten salt transmuter. The current status of technology and prospects for revisited interest are summarized. Significant chemical problems are remain to be resolved at the end of prior MSRs programs, notably, graphite life durability, tritium control, fate of noble metal fission products. Questions arising from plutonium and minor actinide fueling include: corrosion and container chemistry, new redox buffer for systems without uranium, analytical chemistry instrumentation, adequate constituent solubilities, suitable fuel processing and waste form development. However these problems appear to be soluble. (author)

  19. Mesocarbon microbead based graphite for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, Yajuan, E-mail: yajuan.zhong@gmail.com [Center for Thorium Molten Salt Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Key Laboratory of Carbon Materials, Institute of Coal Chemistry, Chinese Academy of Sciences, Taiyuan 030001 (China); Zhang, Junpeng [CAS Key Laboratory of Carbon Materials, Institute of Coal Chemistry, Chinese Academy of Sciences, Taiyuan 030001 (China); Lin, Jun, E-mail: linjun@sinap.ac.cn [Center for Thorium Molten Salt Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Xu, Liujun [Center for Thorium Molten Salt Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Zhang, Feng; Xu, Hongxia; Chen, Yu; Jiang, Haitao; Li, Ziwei; Zhu, Zhiyong [Center for Thorium Molten Salt Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Guo, Quangui [CAS Key Laboratory of Carbon Materials, Institute of Coal Chemistry, Chinese Academy of Sciences, Taiyuan 030001 (China)

    2017-07-15

    Mesocarbon microbeads (MCMB) and quasi-isostatic pressing method were used to prepare MCMB based graphite (MG) for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor (MSR). Characteristics of mercury infiltration and molten salt infiltration in MG were investigated and compared with A3-3 (graphite for spherical fuel element in high temperature gas cooled reactor) to identify the infiltration behaviors. The results indicated that MG had a low porosity about 14%, and an average pore diameter of 96 nm. Fluoride salt occupation of A3-3 (average pore diameter was 760 nm) was 10 wt% under 6.5 atm, whereas salt gain did not infiltrate in MG even up to 6.5 atm. It demonstrated that MG could inhibit the infiltration of liquid fluoride salt effectively. Coefficient of thermal expansion (CTE) of MG lies in 6.01 × 10{sup −6} K{sup −1} (α{sub ∥}) and 6.15 × 10{sup −6} K{sup −1} (α{sub ⊥}) at the temperature range of 25–700 °C. The anisotropy factor of MG calculated by CTE maintained below 1.02, which could meet the requirement of the spherical fuel element (below 1.30). The constant isotropic property of MG is beneficial for the integrity and safety of the graphite used in the spherical fuel element for a MSR.

  20. Mesocarbon microbead based graphite for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor

    International Nuclear Information System (INIS)

    Zhong, Yajuan; Zhang, Junpeng; Lin, Jun; Xu, Liujun; Zhang, Feng; Xu, Hongxia; Chen, Yu; Jiang, Haitao; Li, Ziwei; Zhu, Zhiyong; Guo, Quangui

    2017-01-01

    Mesocarbon microbeads (MCMB) and quasi-isostatic pressing method were used to prepare MCMB based graphite (MG) for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor (MSR). Characteristics of mercury infiltration and molten salt infiltration in MG were investigated and compared with A3-3 (graphite for spherical fuel element in high temperature gas cooled reactor) to identify the infiltration behaviors. The results indicated that MG had a low porosity about 14%, and an average pore diameter of 96 nm. Fluoride salt occupation of A3-3 (average pore diameter was 760 nm) was 10 wt% under 6.5 atm, whereas salt gain did not infiltrate in MG even up to 6.5 atm. It demonstrated that MG could inhibit the infiltration of liquid fluoride salt effectively. Coefficient of thermal expansion (CTE) of MG lies in 6.01 × 10 −6 K −1 (α ∥ ) and 6.15 × 10 −6 K −1 (α ⊥ ) at the temperature range of 25–700 °C. The anisotropy factor of MG calculated by CTE maintained below 1.02, which could meet the requirement of the spherical fuel element (below 1.30). The constant isotropic property of MG is beneficial for the integrity and safety of the graphite used in the spherical fuel element for a MSR.

  1. Small molten-salt reactors with a rational thorium fuel-cycle

    International Nuclear Information System (INIS)

    Furukawa, Kazuo; Mitachi, Kohshi; Kato, Yoshio

    1992-01-01

    In the fission-energy utilization for solving global social and environmental problems including the 'Greenhouse Effect' in the next century, a new strategy should be introduced considering high safety and economy, simplicity, size-flexibility, anti-nuclear proliferation and terrorism, high temperature heat supply, etc., aiming to establish a rational breeding fuelcycle. Thorium Molten-Salt Nuclear Energy Synergetics based on [I] Th utilization, [II] fluid-fuel concept and [III] separation of fissile breeding and power generation functions would be one of the most promising approach. A design study of a standard Molten-Salt Reactor: FUJI-II (350 MWth, 155-161 MWe) ensuring fuel self-sustaining nature (conversion-ratio ∝ 1.0) in spite of small-size, and pilot-plant miniFUJI-II has been proceeded. (orig.)

  2. Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Shi, Chengbin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Cheng, Maosong, E-mail: mscheng@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Liu, Guimin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China)

    2016-08-15

    Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

  3. Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

    International Nuclear Information System (INIS)

    Shi, Chengbin; Cheng, Maosong; Liu, Guimin

    2016-01-01

    Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

  4. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  5. Molten salt reactor type

    International Nuclear Information System (INIS)

    1977-01-01

    This document is one of the three parts of a first volume devoted to the compilations of American data on the molten salt reactor concept. This part 'CIRCUITS' regroups under a condensed form - in French and using international units - the essential information contained in both basic documents of the American project for a molten-salt breeder power plant. This part is only dealing with things relating to the CEA-EDF workshop 'CIRCUITS'. It is not concerned with information on: the reactor and the moderator replacement, the primary and secondary salts, and the fuel salt reprocessing, that are dealt with in parts 'CORE' and 'CHEMISTRY' respectively. The possible evolutions in the data - and solutions - taken by the American designers for their successive projects (1970 to 1972) are shown. The MSBR power plant comprises three successive heat transfer circuits. The primary circuit (Hastelloy N), radioactive and polluted, containing the fuel salt, includes the reactor, pumps and exchangers. The secondary circuit (pipings made of modified Hastelloy N) contaminated in the exchanger, ensures the separation between the fuel and the fluid operating the turbo-alternator. The water-steam circuit feeds the turbine with steam. This steam is produced in the steam generator flowed by the secondary fluid. Some subsidiary circuits (discharge and storage of the primary and secondary salts, ventilation of the primary circuit ...) complete the three principal circuits which are briefly described. All circuits are enclosed inside the controlled-atmosphere building of the nuclear boiler. This building also ensures the biological protection and the mechanical protection against outer aggressions [fr

  6. Fuel processing for molten-salt reactors

    International Nuclear Information System (INIS)

    Hightower, J.R. Jr.

    1975-01-01

    Progress is reported on the development of processes for the isolation of protactinium and for the removal of fission products from molten-salt breeder reactors. The metal transfer experiment MTE-3 (for removing rare earths from MSRE fuel salt) was completed and the equipment used in that experiment was examined. The examination showed that no serious corrosion had occurred on the internal surfaces of the vessels, but that serious air oxidation occurred on the external surfaces of the vessels. Analyses of the bismuth phases indicated that the surfaces in contact with the salts were enriched in thorium and iron. Mass transfer coefficients in the mechanically agitated nondispersing contactors were measured in the Salt/Bismuth Flow-through Facility. The measured mass transfer coefficients are about 30 to 40 percent of those predicted by the preferred literature correlation, but were not as low as those seen in some of the runs in MTE-3. Additional studies using water--mercury systems to simulate molten salt-bismuth systems indicated that the model used to interpret results from previous measurements in the water--mercury system has significant deficiencies. Autoresistance heating studies were continued to develop a means of internal heat generation for frozen-wall fluorinators. Equipment was built to test a design of a side arm for the heating electrode. Results of experiments with this equipment indicate that for proper operation the wall temperature must be held much lower than that for which the equipment was designed. Studies with an electrical analog of the equipment indicate that no regions of abnormally high current density exist in the side arm. (JGB)

  7. Indian programme on molten salt cooled nuclear reactors

    International Nuclear Information System (INIS)

    DuIera, I.V.; Vijayan, P.K.; Sinha, R.K.

    2013-01-01

    Bhabha Atomic Research Centre (BARC) is developing a 600 MWth pebble bed high temperature reactor, cooled by natural circulation of molten fluoride salts and is capable of supplying process heat at 1000 ℃ to facilitate hydrogen production by splitting water. BARC has also initiated studies for a reactor concept in which salts of molten fluoride fuel and coolant in fluid form, flows through the reactor core of graphite moderator, resulting in nuclear fission within the molten salt. For thorium fuel cycle, this concept is very attractive, since the fuel can be re-processed on-line, enabling it to be an efficient neutron breeder. (author)

  8. Proposals on the organization of a fuel cycle of the cascade sub-critical molten salt reactor (CSMSR)

    International Nuclear Information System (INIS)

    Bychkov, A.V.; Kormilitsyn, M.V.; Melnik, M.I.; Babikov, L.G.; Ponomarev, L.I.

    2002-01-01

    At present the approach of burning out long-lived radioactive waste (RW) in the reactor core neutron flux is the most feasible one. Currently the way of closing nuclear fuel cycle (NFC) on the basis of the nuclear chemical concept of the cascade sub-critical molten salt reactor (CSMSR) is considered as the most promising one. It is characterised by a number of advantages. CSMSR controlled by a beam of protons or electrons is the optimal reactor for closing the NFC using non-aqueous fluoride methods of fuel reprocessing. They, in comparison with aqueous methods, are characterised by a small waste quantity and are less laborious because of the absence of severe requirements to the product purity. A high productivity of high-temperature electrochemical processes allows the implementation of the fuel recycling process as part of the CSMSR total technological cycle. It can be conducted in the 'on-line' mode in the bypass molten salt circuit that brings the transportation volume of high-activity materials to a minimum. In order to reprocess the CSMSR irradiated molten salt fuel on the basis of salt composition LiF-NaF-(BeF 2 ) an option, based on the following three main operations of the melt treatment, was proposed at SSC RF RIAR: (i) On-line argon treatment of molten salt fuel for removal of gaseous fission products (FP) and also FP that form volatile fluorides and aerosols; (ii) Organisation of the fuel-active metal (probably with a fine-dispersed plutonium alloy) interaction in the on-line mode for removal of 'noble' and 'semi-noble' FP and corrosion products such as Ni, Fe, Cr (when using Pu alloy it allows to regenerate at the same time of the burned-out plutonium component); (iii) Portion-by-portion (fuel composition partially being removed from the CSMSR molten salt circuit) pyroelectrochemical reprocessing of the molten salt composition aimed at the removal of lanthanides - FP followed by a return of actinides to the CSMSR fuel cycle. This technology will allow

  9. Parametric analyses of single-zone thorium-fueled molten salt reactor fuel cycle options

    International Nuclear Information System (INIS)

    Powers, J.J.; Worrall, A.; Gehin, J.C.; Harrison, T.J.; Sunny, E.E.

    2013-01-01

    Analyses of fuel cycle options based on thorium-fueled Molten Salt Reactors (MSRs) have been performed in support of fuel cycle screening and evaluation activities for the United States Department of Energy. The MSR options considered are based on thermal spectrum MSRs with 3 different separations levels: full recycling, limited recycling, and 'once-through' operation without active separations. A single-fluid, single-zone 2250 MWth (1000 MWe) MSR concept consisting of a fuel-bearing molten salt with graphite moderator and reflectors was used as the basis for this study. Radiation transport and isotopic depletion calculations were performed using SCALE 6.1 with ENDF/B-VII nuclear data. New methodology developed at Oak Ridge National Laboratory (ORNL) enables MSR analysis using SCALE, modeling material feed and removal by taking user-specified parameters and performing multiple SCALE/TRITON simulations to determine the resulting equilibrium operating conditions. Parametric analyses examined the sensitivity of the performance of a thorium MSR to variations in the separations efficiency for protactinium and fission products. Results indicate that self-sustained operation is possible with full or limited recycling but once-through operation would require an external neutron source. (authors)

  10. Subcritical enhanced safety molten-salt reactor concept

    International Nuclear Information System (INIS)

    Alekseev, P.N.; Ignatiev, V.V.; Men'shikov, L.I.; Prusakov, V.N.; Ponomarev-Stepnoy, N.N.; Subbotin, S.A.; Krasnykh, A.K.; Rudenko, V.T.; Somov, L.N.

    1995-01-01

    The nuclear power and its fuel cycle safety requirements can be met in the main by providing nuclear power with subcritical molten salt reactors (SMSR) - 'burner' with an external neutron source. The utilized molten salt fuel is the decisive advantage of the SMSR over other burners. Fissile and fertile nuclides in the burner are solved in a liquid salt in the form of fluorides. This composition acts simultaneously as: a) fuel, b) coolant, c) medium for chemical partitioning and reprocessing. The effective way of reducing the external source power consists in the cascade neutron multiplication in the system of coupled reactors with suppressed feedback between them. (author)

  11. Fast Spectrum Molten Salt Reactor Options

    Energy Technology Data Exchange (ETDEWEB)

    Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Patton, Bruce W [ORNL; Howard, Rob L [ORNL; Harrison, Thomas J [ORNL

    2011-07-01

    During 2010, fast-spectrum molten-salt reactors (FS-MSRs) were selected as a transformational reactor concept for light-water reactor (LWR)-derived heavy actinide disposition by the Department of Energy-Nuclear Energy Advanced Reactor Concepts (ARC) program and were the subject of a preliminary scoping investigation. Much of the reactor description information presented in this report derives from the preliminary studies performed for the ARC project. This report, however, has a somewhat broader scope-providing a conceptual overview of the characteristics and design options for FS-MSRs. It does not present in-depth evaluation of any FS-MSR particular characteristic, but instead provides an overview of all of the major reactor system technologies and characteristics, including the technology developments since the end of major molten salt reactor (MSR) development efforts in the 1970s. This report first presents a historical overview of the FS-MSR technology and describes the innovative characteristics of an FS-MSR. Next, it provides an overview of possible reactor configurations. The following design features/options and performance considerations are described including: (1) reactor salt options-both chloride and fluoride salts; (2) the impact of changing the carrier salt and actinide concentration on conversion ratio; (3) the conversion ratio; (4) an overview of the fuel salt chemical processing; (5) potential power cycles and hydrogen production options; and (6) overview of the performance characteristics of FS-MSRs, including general comparative metrics with LWRs. The conceptual-level evaluation includes resource sustainability, proliferation resistance, economics, and safety. The report concludes with a description of the work necessary to begin more detailed evaluation of FS-MSRs as a realistic reactor and fuel cycle option.

  12. Studies on components for a molten salt reactor

    International Nuclear Information System (INIS)

    Nejedly, M.; Matal, O.

    2003-01-01

    The aim is contribute to a design of selected components of molten salt reactors with fuel in the molten fluoride salt matrix. Molten salt reactors (MSRs) permit the utilization of plutonium and minor actinides as new nuclear fuel from a traditional nuclear power station with production of electric energy. Results of preliminary feasibility studies of an intermediate heat exchanger, a small power molten salt pump and a modular conception of a steam generator for a demonstration unit of the MSR (30 MW) are summarized. (author)

  13. Fast Thorium Molten Salt Reactors Started with Plutonium

    International Nuclear Information System (INIS)

    Merle-Lucotte, E.; Heuer, D.; Le Brun, C.; Brissot, R.; Liatard, E.; Meplan, O.; Nuttin, A.; Mathieu, L.

    2006-01-01

    One of the pending questions concerning Molten Salt Reactors based on the 232 Th/ 233 U fuel cycle is the supply of the fissile matter, and as a consequence the deployment possibilities of a fleet of Molten Salt Reactors, since 233 U does not exist on earth and is not yet produced in the current operating reactors. A solution may consist in producing 233 U in special devices containing Thorium, in Pressurized Water or Fast Neutrons Reactors. Two alternatives to produce 233 U are examined here: directly in standard Molten Salt Reactors started with Plutonium as fissile matter and then operated in the Th/ 233 U cycle; or in dedicated Molten Salt Reactors started and fed with Plutonium as fissile matter and Thorium as fertile matter. The idea is to design a critical reactor able to burn the Plutonium and the minor actinides presently produced in PWRs, and consequently to convert this Plutonium into 233 U. A particular reactor configuration is used, called 'unique channel' configuration in which there is no moderator in the core, leading to a quasi fast neutron spectrum, allowing Plutonium to be used as fissile matter. The conversion capacities of such Molten Salt Reactors are excellent. For Molten Salt Reactors only started with Plutonium, the assets of the Thorium fuel cycle turn out to be quickly recovered and the reactor's characteristics turn out to be equivalent to Molten Salt Reactors operated with 233 U only. Using a combination of Molten Salt Reactors started or operated with Plutonium and of Molten Salt Reactors started with 233 U, the deployment capabilities of these reactors fully satisfy the condition of sustainability. (authors)

  14. Assessment of the Capability of Molten Salt Reactors as a Next Generation High Temperature Reactors

    International Nuclear Information System (INIS)

    Elsheikh, B.M.

    2017-01-01

    Molten Salt Reactor according to Aircraft Reactor Experiment (ARE) and the Molten Salt Reactor Experiment (MSRE) programs, was designed to be the first full-scale, commercial nuclear power plant utilizing molten salt liquid fuels that can be used for producing electricity, and producing fissile fuels (breeding)burning actinides. The high temperature in the primary cycle enables the realization of efficient thermal conversion cycles with net thermal efficiencies reach in some of the designs of nuclear reactors greater than 45%. Molten salts and liquid salt because of their low vapor pressure are excellent candidates for meeting most of the requirements of these high temperature reactors. There is renewed interest in MSRs because of changing goals and new technologies in the use of high-temperature reactors. Molten Salt Reactors for high temperature create substantial technical challenges to have high effectiveness intermediate heat transfer loop components. This paper will discuss and investigate the capability and compatibility of molten salt reactors, toward next generation high temperature energy system and its technical challenges

  15. Preliminary design studies of the draining tanks for the Molten Salt Fast Reactor

    International Nuclear Information System (INIS)

    Merle-Lucotte, E.; Allibert, M.; Heuer, D.; Brovchenko, M.; Laureau, A.; Ghetta, V.; Rubiolo, P.

    2014-01-01

    reactor called the Molten Salt Fast Reactor (MSFR). The reference MSFR design is a 3000 MWth reactor with a total fuel salt volume of 18 m3, operated at a mean fuel temperature of 750 deg. C. The first confinement barrier of the reactor includes a salt draining system. In case of a planned reactor shut down or in case of accidents leading to an excessive increase of the temperature in the fuel circuit, the fuel configuration may be changed passively by gravitational draining of the fuel salt in dedicated draining tank located under the reactor and designed to provide adequate reactivity margins while insuring a passive cooling of the fuel salt to extract the residual heat from the short to the long term. The present preliminary assessment of this sub-critical draining system has been performed to identify the physical constraints and to give some orders of magnitude of characteristic time periods (authors)

  16. Pyrochemical reprocessing of molten salt fast reactor fuel: focus on the reductive extraction step

    Directory of Open Access Journals (Sweden)

    Rodrigues Davide

    2015-12-01

    Full Text Available The nuclear fuel reprocessing is a prerequisite for nuclear energy to be a clean and sustainable energy. In the case of the molten salt reactor containing a liquid fuel, pyrometallurgical way is an obvious way. The method for treatment of the liquid fuel is divided into two parts. In-situ injection of helium gas into the fuel leads to extract the gaseous fission products and a part of the noble metals. The second part of the reprocessing is performed by ‘batch’. It aims to recover the fissile material and to separate the minor actinides from fission products. The reprocessing involves several chemical steps based on redox and acido-basic properties of the various elements contained in the fuel salt. One challenge is to perform a selective extraction of actinides and lanthanides in spent liquid fuel. Extraction of actinides and lanthanides are successively performed by a reductive extraction in liquid bismuth pool containing metallic lithium as a reductive reagent. The objective of this paper is to give a description of the several steps of the reprocessing retained for the molten salt fast reactor (MSFR concept and to present the initial results obtained for the reductive extraction experiments realized in static conditions by contacting LiF-ThF4-UF4-NdF3 with a lab-made Bi-Li pool and for which extraction efficiencies of 0.7% for neodymium and 14.0% for uranium were measured. It was concluded that in static conditions, the extraction is governed by a kinetic limitation and not by the thermodynamic equilibrium.

  17. Conceptual design of Indian molten salt breeder reactor

    International Nuclear Information System (INIS)

    Vijayan, P.K.; Basak, A.; Dulera, I.V.; Vaze, K.K.; Basu, S.; Sinha, R.K.

    2014-01-01

    The fuel in a molten salt breeder reactor is in the form of a continuously circulating molten salt. Fluoride based salts have been almost universally proposed. A crucial part for achieving reasonable breeding in such reactors is the need to reprocess the salt continuously, either online or in batch mode. This constitutes a major technological challenge for this type of reactors. India has recently started carrying out fundamental studies so as to arrive at a conceptual design of Indian Molten Salt Breeder Reactor (IMSBR). Presently various design options and possibilities are being studied from the point of view of reactor physics and thermal hydraulic design. In parallel fundamental studies as regards various molten salts have also been initiated. This paper would discuss conceptual design of these reactors, as well as associated issues and technologies

  18. An optimized symbiotic fusion and molten-salt fission reactor system

    International Nuclear Information System (INIS)

    Blinkin, V.L.; Novikov, V.M.

    A symbiotic fusion-fission reactor system which breeds nuclear fuel is discussed. In the blanket of the controlled thermonuclear reactor (CTR) uranium-233 is generated from thorium, which circulates in the form of ThF 4 mixed with molten sodium and beryllium fluorides. The molten-salt fission reactor (MSR) burns up the uranium-233 and generates tritium for the fusion reactor from lithium, which circulates in the form of LiF mixed with BeF 2 and 233 UF 4 through the MSR core. With a CTR-MSR thermal power ratio of 1:11 the system can produce electrical energy and breed fuel with a doubling time of 4-5 years. The system has the following special features: (1) Fuel reprocessing is much simpler and cheaper than for contemporary fission reactors; reprocessing consists simply in continuous removal of 233 U from the salt circulating in the CTR blanket by the fluorination method and removal of xenon from the MSR fuel salt by gas scavenging; the MSR fuel salt is periodically exchanged for fresh salt and the 233 U is then removed from it; (2) Tritium is produced in the fission reactor, which is a much simpler system than the fusion reactor; (3) The CTR blanket is almost ''clean''; no tritium is produced in it and fission fragment activity does not exceed the activity induced in the structural materials; (4) Almost all the thorium introduced into the CTR blanket can be used for producing 233 U

  19. Molten salt small modular reactors (MSSMRs): from DMSR to SmAHTR

    International Nuclear Information System (INIS)

    LeBlanc, D.

    2013-01-01

    Molten salt reactors were developed extensively from the 1950s to 1970s as a thermal breeder alternative on the Thorium-U233 cycle. Simplified designs running as fluid fuel convertors without salt processing as well as TRISO fueled, salt cooled reactors both hold much promise as potential small modular reactors. A background will be presented along with the most likely routes forward for a Canadian development program. (author)

  20. Thermodynamic characterization of salt components for Molten Salt Reactor fuel

    NARCIS (Netherlands)

    Capelli, E.

    2016-01-01

    The Molten Salt Reactor (MSR) is a promising future nuclear fission reactor technology with excellent performance in terms of safety and reliability, sustainability, proliferation resistance and economics. For the design and safety assessment of this concept, it is extremely important to have a

  1. The risk-rewards structure of using spent nuclear fuel in molten salt reactor - 5513

    International Nuclear Information System (INIS)

    He, X.; Du, Z.; Macian-Juan, R.; Seidl, M.

    2015-01-01

    The molten salt reactor concept naturally lends itself to a re-use of fuel either by online reprocessing or by using spent nuclear fuel as part of the driver fuel. Moreover some well-known safety advantages over traditional LWR designs are promised: the primary loop can be operated at atmospheric pressure, refueling can be done online, only a minimum amount of excess reactivity needs to be stored inside the core and the continuous circulation and inter-mixing of the fuel results in a more homogenous redistribution of fission products. In this paper the feasibility of running a molten salt reactor on spent LWR fuel is discussed in a number of scenarios in order to make the various trade-offs transparent: using SNF in a classic graphite moderated MSR and doing the same for a lead-cooled dual-fluid MSR. From a commercial company's point of view the MSR concept faces already substantial risks even without the use of SNF: licensing concerns due to an enrichment of fissile nuclides typically above 5% of heavy metal mass, limited practical experience with the reliability of proposed MSR materials and almost no experience with online reprocessing. For one thing one could therefore aim for the most conservative design which would rely on the design of ORNL's graphite moderated MSR operated in the sixties. While appearing realistic from a technical perspective, the potential for SNF re-use in the sense of actinide destruction appears limited. On the other hand one can maximize the risk and the potential payoff by concentrating on the most speculative design, i.e. a dual fluid reactor with an ultra-hard neutron spectrum in order to most efficiently burn higher actinides. In this paper the neutronic design calculations for the above described MSR concepts are presented in order to maximize SNF's contribution for the reactors' energy generation and their potential for actinide destruction. Among the optimization parameters are the lattice constants, the type

  2. Nuclear power technology system with molten salt reactor for transuranium nuclides burning in closed fuel cycle

    International Nuclear Information System (INIS)

    Alekseev, P.N.; Dudnikov, A.A.; Ignatiev, V.V.; Prusakov, V.N.; Ponomarev-Stepnoy, N.N.; Subbotin, S.A.

    2000-01-01

    A concept of nuclear power technology system with homogeneous molten salt reactors for burning and transmutation of long-lived radioactive toxic nuclides is considered in the paper. Disposition of such reactors in enterprises of fuel cycle allows to provide them with power and facilitate solution of problems with rad waste with minimal losses. (Authors)

  3. Molten salt reactors - safety options galore

    International Nuclear Information System (INIS)

    Gat, U.; Dodds, H.L.

    1997-01-01

    Safety features and attributes of molten salt reactors (MSR) are described. The unique features of fluid fuel reactors of on-line continuous processing and the ability for so-called external cooling result in simple and safe designs with low excess reactivity, low fission product inventory, and small source term. These, in turn, make a criticality accident unlikely and reduce the severity of a loss of coolant to where they are no longer severe accidents. A melt down is not an accident for a reactor that uses molten fuel. The molten salts are stable, non-reactive and efficient heat transfer media that operate at high temperatures at low pressures and are highly compatible with selected structural materials. All these features reduce the accident plethora. Freeze valves can be used for added safety. An ultimate safe reactor (U.S.R) is described with safety features that are passive, inherent and non-tamperable (PINT)

  4. Studies on the molten salt reactor. Code development and neutronics analysis of MSRE-type design

    International Nuclear Information System (INIS)

    Zhuang Kun; Cao Liangzhi; Zheng Youqi; Wu Hongchun

    2015-01-01

    The molten salt reactor is characterized by its use of the fluid-fuel, which serves both as a fuel and as a coolant simultaneously. The position of delayed neutron precursors continuously changes both in the core and in the external loop due to the fuel circulation, and the fission products are extracted by an online fuel reprocessing unit, which all lead to the modeling methods for the conventional reactors using solid fuel not applicable. This study establishes suitable calculation models for the neutronics analysis of the molten salt reactor and develops a new code named MOREL based on the three-dimensional diffusion steady and transient calculations. Some numerical tests are chosen to verify the code and the numerical results indicate that MOREL can be used for the analysis of the molten salt reactor. After verification, it is applied to analyze the characteristics of a typical molten salt reactor, including the steady characteristics, the influence of fuel circulation on the kinetic behaviors. Besides, the influence of online fuel reprocessing simulation is also examined. The results show that inherent safety is the character of the molten salt reactor from the aspect of reactivity feedback and the fuel circulation has great influence on the kinetic characteristics of molten salt reactor. (author)

  5. Definition of breeding gain for molten salt reactors - 147

    International Nuclear Information System (INIS)

    Nagy, K.; Kloosterman, J.L.; Lathouwers, D.; Van der Hagen, T.H.J.J.

    2010-01-01

    The graphite-moderated Molten Salt Reactor (MSR) is a potential breeder reactor using the thorium fuel cycle. The MSR has unique properties due to the possibility of making changes to the salt composition during operation. Most important is the extraction of protactinium, which separates the fissile uranium production into two volumes: the reactor core and the external stockpile. The paper focuses on the definition of breeding gain in such a system. The prospects of using breeding gain expressions defined for solid fuel reactors are investigated and new definitions are given which incorporate the processes occurring in the reactor core and the external stockpile. The difference of the growth rate of the mass of fissile material and breeding gain is pointed out. The new definitions are applied to an optimization study of the graphite-salt lattice of a breeder MSR. (authors)

  6. Modified ADS molten salt processes for back-end fuel cycle of PWR spent fuel

    International Nuclear Information System (INIS)

    Choi, In-Kyu; Yeon, Jei-Won; Kim, Won-Ho

    2002-01-01

    The back-end fuel cycle concept for PWR spent fuel is explained. This concept is adequate for Korea, which has operated both PWR and CANDU reactors. Molten salt processes for accelerator driven system (ADS) were modified both for the transmutation of long-lived radioisotopes and for the utilisation of the remained fissile uranium in PWR spent fuels. Prior to applying molten salt processes to PWR fuel, hydrofluorination and fluorination processes are applied to obtain uranium hexafluoride from the spent fuel pellet. It is converted to uranium dioxide and fabricated into CANDU fuel. From the remained fluoride compounds, transuranium elements can be separated by the molten salt technology such as electrowinning and reductive extraction processes for transmutation purpose without weakening the proliferation resistance of molten salt technology. The proposed fuel cycle concept using fluorination processes is thought to be adequate for our nuclear program and can replace DUPIC (Direct Use of spent PWR fuel in CANDU reactor) fuel cycle. Each process for the proposed fuel cycle concept was evaluated in detail

  7. Demand driven salt clean-up in a molten salt fast reactor - Defining a priority list.

    Science.gov (United States)

    Merk, B; Litskevich, D; Gregg, R; Mount, A R

    2018-01-01

    The PUREX technology based on aqueous processes is currently the leading reprocessing technology in nuclear energy systems. It seems to be the most developed and established process for light water reactor fuel and the use of solid fuel. However, demand driven development of the nuclear system opens the way to liquid fuelled reactors, and disruptive technology development through the application of an integrated fuel cycle with a direct link to reactor operation. The possibilities of this new concept for innovative reprocessing technology development are analysed, the boundary conditions are discussed, and the economic as well as the neutron physical optimization parameters of the process are elucidated. Reactor physical knowledge of the influence of different elements on the neutron economy of the reactor is required. Using an innovative study approach, an element priority list for the salt clean-up is developed, which indicates that separation of Neodymium and Caesium is desirable, as they contribute almost 50% to the loss of criticality. Separating Zirconium and Samarium in addition from the fuel salt would remove nearly 80% of the loss of criticality due to fission products. The theoretical study is followed by a qualitative discussion of the different, demand driven optimization strategies which could satisfy the conflicting interests of sustainable reactor operation, efficient chemical processing for the salt clean-up, and the related economic as well as chemical engineering consequences. A new, innovative approach of balancing the throughput through salt processing based on a low number of separation process steps is developed. Next steps for the development of an economically viable salt clean-up process are identified.

  8. Molten salt reactors: reactor cores

    International Nuclear Information System (INIS)

    1983-01-01

    In this critical analysis of the MSBR I project are examined the problems concerning the reactor core. Advantages of breeding depend essentially upon solutions to technological problems like continuous reprocessing or graphite behavior under neutron irradiation. Graphite deformation, moderator unloading, control rods and core instrumentation require more studies. Neutronics of the core, influence of core geometry and salt composition, fuel evolution, and thermohydraulics are reviewed [fr

  9. Advanced High-Temperature Reactor for Production of Electricity and Hydrogen: Molten-Salt-Coolant, Graphite-Coated-Particle-Fuel

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    2002-01-01

    The objective of the Advanced High-Temperature Reactor (AHTR) is to provide the very high temperatures necessary to enable low-cost (1) efficient thermochemical production of hydrogen and (2) efficient production of electricity. The proposed AHTR uses coated-particle graphite fuel similar to the fuel used in modular high-temperature gas-cooled reactors (MHTGRs), such as the General Atomics gas turbine-modular helium reactor (GT-MHR). However, unlike the MHTGRs, the AHTR uses a molten salt coolant with a pool configuration, similar to that of the PRISM liquid metal reactor. A multi-reheat helium Brayton (gas-turbine) cycle, with efficiencies >50%, is used to produce electricity. This approach (1) minimizes requirements for new technology development and (2) results in an advanced reactor concept that operates at essentially ambient pressures and at very high temperatures. The low-pressure molten-salt coolant, with its high heat capacity and natural circulation heat transfer capability, creates the potential for (1) exceptionally robust safety (including passive decay-heat removal) and (2) allows scaling to large reactor sizes [∼1000 Mw(e)] with passive safety systems to provide the potential for improved economics

  10. Fuel reprocessing of the fast molten salt reactor: actinides et lanthanides extraction

    International Nuclear Information System (INIS)

    Jaskierowicz, S.

    2012-01-01

    The fuel reprocessing of the molten salt reactor (Gen IV concept) is a multi-steps process in which actinides and lanthanides extraction is performed by a reductive extraction technique. The development of an analytic model has showed that the contact between the liquid fuel LiF-ThF 4 and a metallic phase constituted of Bi-Li provide firstly a selective and quantitative extraction of actinides and secondly a quantitative extraction of lanthanides. The control of this process implies the knowledge of saline phase properties. Studies of the physico-chemical properties of fluoride salts lead to develop a technique based on potentiometric measurements to evaluate the fluoro-acidity of the salts. An acidity scale was established in order to classify the different fluoride salts considered. Another electrochemical method was also developed in order to determine the solvation properties of solutes in fluoride F- environment (and particularly ThF 4 by F-) in reductive extraction technique, a metallic phase is also involved. A method to prepare this phase was developed by electro-reduction of lithium on a bismuth liquid cathode in LiCl-LiF melt. This technique allows to accurately control the molar fraction of lithium introduced into the liquid bismuth, which is a main parameter to obtain an efficient extraction. (author)

  11. A general overview of generation IV molten salt reactor (MSR) and the use of thorium as fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yamaguchi, Carlos H.; Stefani, Giovanni L.; Santos, Thiago A., E-mail: carlos.yamaguchi@usp.br, E-mail: giovanni.stefani@ipen.br, E-mail: thiago.santos@ufabc.edu.br [Universidade de Sao Paulo (USP), SP (Brazil). Instituto de Fisica; Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais Aplicadas

    2017-07-01

    The molten salt reactors (MSRs) make use of fluoride salt as primary cooler, at low pressure. Although considered a generation IV reactor, your concept isn't new, since in the 1960 years the Oak Ridge National Laboratory created a little prototype of 8MWt. Over the 20{sup th} century, other countries, like UK, Japan, Russia, China and France also did research in the area, especially with the use of thorium as fuel. This goes with the fact that Brazil possess the biggest reserve of thorium in the world. In the center of nuclear engineering at IPEN is being created a study group connected to thorium reactors, which purpose is to investigate reactors using thorium to produce {sup 233}U and tailing burn, thus making the MSR using thorium as fuel, an object of study. This present work searches to do a general summary about the researches of MSR's, having as focus the utilization of thorium with the goal being to show it's efficiency and utilization is doable. (author)

  12. Molten salt reactor type

    International Nuclear Information System (INIS)

    1977-01-01

    This document is one of the three parts of a first volume devoted to the compilations of American data on the molten salt reactor concept. Emphasize is put essentially on the fuel salt of the primary circuit inside which fission reactions occur. The reasons why the (LiF-BeF 2 -ThF 4 -UF 4 ) salt was chosen for the M.S.B.R. concept are examined; the physical, physicochemical and chemical properties of this salt are discussed with its interactions with the structural materials and its evolution in time. An important part of this volume is devoted to the continuous reprocessing of the active salt, the project designers having deemed advisable to take advantage at best from the availability of a continuous purification, in a thermal breeding. The problem of tritium formation and distribution inside the reactor is also envisaged and the fundamentals of the chemistry of the secondary coolant salt are given. The solutions proposed are: the hydrogen scavenging of the primary circuit, a reduction in metal permeability by an oxyde layer deposition on the side in contact with the vapor, and tritium absorption through an isotope exchange with the hydroxifluoroborate [fr

  13. Combined system of accelerator molten-salt breeder (AMSB) apd molten-salt converter reactor (MSCR)

    International Nuclear Information System (INIS)

    Furukawa, K.; Kato, Y.; Ohmichi, T.; Ohno, H.

    1983-01-01

    A design and research program is discUssed of the development of accelerator molten-salt breeder (AMSB) consisting of a proton accelerator and a molten fluoride target. The target simultaneously serves as a blanket for fissionable material prodUction. An addition of some amoUnt of fissile nuclides to a melt expands the AMSB potentialities as the fissionable material production increases and the energy generation also grows up to the level of self-provision. Besides the blanket salts may be used as nuclear fuel for molten-salt converter reactor (MSCR). The combined AM SB+MSCR system has better parameters as compared to other breeder reactors, molten-salt breeder reactors (MSBR) included

  14. The introduction of the safety of molten salt reactor

    International Nuclear Information System (INIS)

    Zuo Jiaxu; Zhang Chunming

    2011-01-01

    This paper introduces the generation TV Nuclear Energy Systems and molten salt reactor which is the only fluid fuel reactor in the Gen-TV. Safety features and attributes of MSR are described. The supply of fuel and the minimum of waste are described. The clean molten salt in the secondary heat transport system transfers the heat from the primary heat exchanger to a high-temperature Brayton cycle that converts the heat to electricity. With the Brayton cycle, the thermal efficiency of the system will be improved. Base on the MSR, the thorium-uranium fuel cycle is also introduced. (authors)

  15. Molten fluoride fuel salt chemistry

    International Nuclear Information System (INIS)

    Toth, L.M.; Del Cul, G.D.; Dai, S.; Metcalf, D.H.

    1995-01-01

    The chemistry of molten fluorides is traced from their development as fuels in the Molten Salt Reactor Experiment with important factors in their selection being discussed. Key chemical characteristics such as solubility, redox behavior, and chemical activity are explained as they relate to the behavior of molten fluoride fuel systems. Development requirements for fitting the current state of the chemistry to modern nuclear fuel system are described. It is concluded that while much is known about molten fluoride behavior which can be used effectively to reduce the amount of development required for future systems, some significant molten salt chemical questions must still be addressed. copyright American Institute of Physics 1995

  16. Development of a three dimension multi-physics code for molten salt fast reactor

    International Nuclear Information System (INIS)

    Cheng Maosong; Dai Zhimin

    2014-01-01

    Molten Salt Reactor (MSR) was selected as one of the six innovative nuclear reactors by the Generation IV International Forum (GIF). The circulating-fuel in the can-type molten salt fast reactor makes the neutronics and thermo-hydraulics of the reactor strongly coupled and different from that of traditional solid-fuel reactors. In the present paper: a new coupling model is presented that physically describes the inherent relations between the neutron flux, the delayed neutron precursor, the heat transfer and the turbulent flow. Based on the model, integrating nuclear data processing, CAD modeling, structured and unstructured mesh technology, data analysis and visualization application, a three dimension steady state simulation code system (MSR3DS) for the can-type molten salt fast reactor is developed and validated. In order to demonstrate the ability of the code, the three dimension distributions of the velocity, the neutron flux, the delayed neutron precursor and the temperature were obtained for the simplified MOlten Salt Advanced Reactor Transmuter (MOSART) using this code. The results indicate that the MSR3DS code can provide a feasible description of multi-physical coupling phenomena in can-type molten salt fast reactor. Furthermore, the code can well predict the flow effect of fuel salt and the transport effect of the turbulent diffusion. (authors)

  17. Reactor physical experimental program EROS in the frame of the molten salt applying reactor concepts development

    International Nuclear Information System (INIS)

    Hron, Miloslav; Kyncl, Jan; Mikisek, Miroslav

    2009-01-01

    After the relatively broad program of experimental activities, which have been involved in the complex R and D program for the Molten Salt Reactor (MSR) - SPHINX (SPent Hot fuel Incinerator by Neutron fluX) concept development in the Czech Republic, there has been a next stage (namely large-scale experimental verification of design inputs by use of MSR-type inserted zones into the existing light water moderated experimental reactor LR-0 called EROS project) started, which will be focused to the experimental verification of the rector physical or neutronic properties of other types of reactor concepts applying molten salts in the role of liquid fuel and/or coolant. This tendency is based on the recently accepted decision of the MSR SSC of GIF to consider for further period of its activity two baseline concepts- fast neutron molten salt reactor non-moderated (FMSR-NM) as a long-term alternative to solid fuelled fast neutron reactors and simultaneously, advanced high temperature reactor (AHTR) with pebble bed type solid fuel cooled by liquid salts. There will be a brief description of the prepared and performed experimental programs in these directions (as well as the preliminary results obtained so far) introduced in the paper. (author)

  18. Calculation of β-effective of a molten salt reactor

    International Nuclear Information System (INIS)

    Hirakawa, N.; Sakaba, H.

    1987-01-01

    A method to calculate the β eff of a molten salt reactor was developed taking the effect of the flow of the molten salt into account. The method was applied to the 1000MW MSR design made by ORNL. The change in β eff due to the change in the residence time outside of the core of the fuel salt and to the change in the flow velocity when the total amount of the fuel salt is kept constant were investigated. It was found that β eff was reduced to 47.9% of the value when the fuel salt is at rest for the present design. (author)

  19. Neutronic design of a Liquid Salt-cooled Pebble Bed Reactor (LSPBR)

    International Nuclear Information System (INIS)

    De Zwaan, S. J.; Boer, B.; Lathouwers, D.; Kloosterman, J. L.

    2006-01-01

    A renewed interest has been raised for liquid salt cooled nuclear reactors. The excellent heat transfer properties of liquid salt coolants provide several benefits, like lower fuel temperatures, higher coolant outlet temperatures, increased core power density and better decay heat removal. In order to benefit from the online refueling capability of a pebble bed reactor, the Liquid Salt Pebble Bed Reactor (LSPBR) is proposed. This is a high temperature pebble-bed reactor with a fuel design similar to existing HTRs, but using a liquid salt as a coolant. In this paper, the selection criteria for the liquid salt coolant are described. Based on its neutronic properties, LiF-BeF 2 (FLIBE) was selected for the LSPBR. Two designs of the LSPBR were considered: a cylindrical core and an annular core with a graphite inner reflector. Coupled neutronic-thermal hydraulic calculations were performed to obtain the steady state power distribution and the corresponding fuel temperatures. Finally, calculations were performed to investigate the decay heat removal capability in a protected loss-of-forced cooling accident. The maximum allowable power that can be produced with the LSPBR is hereby determined. (authors)

  20. Transient response of small molten salt reactor at duct blockage accident

    International Nuclear Information System (INIS)

    Yamamoto, Takahisa; Mitachi, Koshi; Ikeuchi, Koji; Suzuki, Takashi

    2005-01-01

    This paper performed transient core analysis of a small Molten Salt Reactor (MSR) at the time of a duct blockage accident. The numerical model employed in this study consists of continuity and momentum conservation equations for fuel salt flow, two group diffusion equations for fast and thermal neutron fluxes, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The analysis shows that (1) the effective multiplication factor and reactor power after the blockage accident hardly change because of the self-control performance of the MSR, (2) fuel salt and graphite moderator temperatures rise at the blockage point and its vicinity, drastically but locally, (3) the highest temperature after the blockage accident is 1 363 K, very lower than the boiling point of fuel salt and melt point of reactor vessel, (4) fast and thermal neutron fluxes distributions after the blockage accident hardly change, and (5) delayed neutron precursors accumulate at the blockage point, especially 1st delayed neutron precursor due to is large decay constant. These results lead that the safety of MSR is assured in the blockage accident. (author)

  1. Assessment of the Neutronic and Fuel Cycle Performance of the Transatomic Power Molten Salt Reactor Design

    Energy Technology Data Exchange (ETDEWEB)

    Robertson, Sean [Transatomic Power Corp., Cambridge, MA (United States); Dewan, Leslie [Transatomic Power Corp., Cambridge, MA (United States); Massie, Mark [Transatomic Power Corp., Cambridge, MA (United States); Davidson, Eva E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    This report presents results from a collaboration between Transatomic Power Corporation (TAP) and Oak Ridge National Laboratory (ORNL) to provide neutronic and fuel cycle analysis of the TAP core design through the Department of Energy Gateway for Accelerated Innovation in Nuclear (GAIN) Nuclear Energy Voucher program. The TAP concept is a molten salt reactor using configurable zirconium hydride moderator rod assemblies to shift the neutron spectrum in the core from mostly epithermal at beginning of life to thermal at end of life. Additional developments in the ChemTriton modeling and simulation tool provide the critical moderator-to-fuel ratio searches and time-dependent parameters necessary to simulate the continuously changing physics in this complex system. The implementation of continuous-energy Monte Carlo transport and depletion tools in ChemTriton provide for full-core three-dimensional modeling and simulation. Results from simulations with these tools show agreement with TAP-calculated performance metrics for core lifetime, discharge burnup, and salt volume fraction, verifying the viability of reducing actinide waste production with this concept. Additional analyses of mass feed rates and enrichments, isotopic removals, tritium generation, core power distribution, core vessel helium generation, moderator rod heat deposition, and reactivity coeffcients provide additional information to make informed design decisions. This work demonstrates capabilities of ORNL modeling and simulation tools for neutronic and fuel cycle analysis of molten salt reactor concepts.

  2. Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled molten salt reactors

    International Nuclear Information System (INIS)

    Qiu, Suizheng; Zhang, Dalin; Liu, Minghao; Liu, Limin; Xu, Rongshuan; Gong, Cheng; Su, Guanghui

    2016-01-01

    Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.

  3. Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled molten salt reactors

    Energy Technology Data Exchange (ETDEWEB)

    Qiu, Suizheng; Zhang, Dalin; Liu, Minghao; Liu, Limin; Xu, Rongshuan; Gong, Cheng; Su, Guanghui [Xi' an Jiaotong Univ. (China). State Key Laboratory of Multiphase Flow in Power Engineering

    2016-05-15

    Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.

  4. R and D of On-line Reprocessing Technology for Molten-Salt Reactor Systems

    International Nuclear Information System (INIS)

    Uhlir, Jan; Tulackova, Radka; Chuchvalcova Bimova, Karolina

    2006-01-01

    The Molten Salt Reactor (MSR) represents one of promising future nuclear reactor concept included in the Generation IV reactors family. The reactor can be operated as the thorium breeder or as the actinide transmuter. However, the future deployment of Molten-Salt Reactors will be significantly dependent on the successful mastering of advanced reprocessing technologies dedicated to their fuel cycle. Here the on-line reprocessing technology connected with the fuel circuit of MSR is of special importance because the reactor cannot be operated for a long run without the fuel salt clean-up. Generally, main MSR reprocessing technologies are pyrochemical, majority of them are fluoride technologies. The proposed flow-sheets of MSR on-line reprocessing are based on a combination of molten-salt / liquid metal extraction and electro-separation processes, which can be added to the gas extraction process already verified during the MSRE project in ORNL. The crucial separation method proposed for partitioning of actinides from fission products is based on successive Anodic dissolution and Cathodic deposition processes in molten fluoride media. (authors)

  5. The molten salt reactor: R and D status and perspectives in Europe

    International Nuclear Information System (INIS)

    Renault, Claude; Delpech, Sylvie; Merle-Lucotte, Elsa; Konings, Rudy; Hron, Miloslav; Ignatiev, Victor

    2010-01-01

    The paper concentrates on molten salt fast reactor (MSFR) concepts which are receiving most attention in the EU context. It shows the main R and D achievements and some remaining issues to be addressed in such essential areas as (a) reactor conceptual design, (b) molten salt properties, (c) fuel salt clean-up scheme and (d) high temperature materials. The status and perspectives of molten salt reactor R and D efforts in Europe are then discussed

  6. Thorium cycle and molten salt reactors: field parameters and field constraints investigations toward 'thorium molten salt reactor' definition

    International Nuclear Information System (INIS)

    Mathieu, L.

    2005-09-01

    Producing nuclear energy in order to reduce the anthropic CO 2 emission requires major technological advances. Nuclear plants of 4. generation have to respond to several constraints, as safety improvements, fuel breeding and radioactive waste minimization. For this purpose, it seems promising to use Thorium Cycle in Molten Salt Reactors. Studies on this domain have already been carried out. However, the final concept suffered from serious issues and was discontinued. A new reflection on this topic is being led in order to find acceptable solutions, and to design the Thorium Molten Salt Reactor concept. A nuclear reactor is simulated by the coupling of a neutron transport code with a materials evolution code. This allows us to reproduce the reactor behavior and its evolution all along its operation. Thanks to this method, we have studied a large number of reactor configurations. We have evaluated their efficiency through a group of constraints they have to satisfy. This work leads us to a better understanding of many physical phenomena controlling the reactor behavior. As a consequence, several efficient configurations have been discovered, allowing the emergence of new points of view in the research of Molten Salt Reactors. (author)

  7. Recommendations for a restart of Molten Salt Reactor development

    International Nuclear Information System (INIS)

    Moir, R. W.

    2007-01-01

    The concept of the molten salt reactor (MSR) refuses to go away. The Generation-IV process lists the MSR as one of the six concepts to be considered for extending fuel resources. Good fuel utilization and good economics are required to meet the often cited goal of 10 TWe globally and 1 TWe for the US by non-carbon energy sources in this century by nuclear fission. A strong incentive for the molten salt reactor design is its good fuel utilization, good economics, amazing flexibility and promised large benefits. It can: - use thorium or uranium; o be designed with lots of graphite to have a fairly thermal neutron spectrum or without graphite moderator to have a fast neutron spectrum reactor; - fission uranium isotopes and plutonium isotopes; - operate with non-weapon grade fissile fuel, or in suitable sites it can operate with enrichment between reactor-grade and weapon-grade fissile fuel; - be a breeder or near breeder; - operate at temperature >1100 degree C if carbon composites are successfully employed. Enhancing 2 32U content in the uranium to over 500 pm makes the fuel undesirable for weapons, but it should not detract from its economic use in liquid fuel reactors: a big advantage in nonproliferation. Economics of the MSR is enhanced by operating at low pressure and high temperature and may even lead to the preferred route to hydrogen production. The cost of the electricity produced from low enriched fuel averaged over the life of the entire process, has been predicted to be about 10% lower than that from LWRs, and 20% lower for high enriched fuel, with uncertainties of about 10%. The development cost has been estimated at about 1 B$ (e.g., a 100 M$/y base program for ten years) not including construction of a series of reactors leading up to the deployment of multiple commercial units at an assumed cost of 9 B$ (450 M$/y over 20 years). A benefit of liquid fuel is that smaller power reactors can faithfully test features of larger reactors, thereby reducing the

  8. Thorium and Molten Salt Reactors: Essential Questions for Classroom Discussions

    Science.gov (United States)

    DiLisi, Gregory A.; Hirsch, Allison; Murray, Meredith; Rarick, Richard

    2018-04-01

    A little-known type of nuclear reactor called the "molten salt reactor" (MSR), in which nuclear fuel is dissolved in a liquid carrier salt, was proposed in the 1940s and developed at the Oak Ridge National Laboratory in the 1960s. Recently, the MSR has generated renewed interest as a remedy for the drawbacks associated with conventional uranium-fueled light-water reactors (LWRs) in use today. Particular attention has been given to the "thorium molten salt reactor" (TMSR), an MSR engineered specifically to use thorium as its fuel. The purpose of this article is to encourage the TPT community to incorporate discussions of MSRs and the thorium fuel cycle into courses such as "Physics and Society" or "Frontiers of Physics." With this in mind, we piloted a pedagogical approach with 27 teachers in which we described the underlying physics of the TMSR and posed five essential questions for classroom discussions. We assumed teachers had some preexisting knowledge of nuclear reactions, but such prior knowledge was not necessary for inclusion in the classroom discussions. Overall, our material was perceived as a real-world example of physics, fit into a standards-based curriculum, and filled a need in the teaching community for providing unbiased references of alternative energy technologies.

  9. Thorium and Molten Salt Reactors: "Essential Questions for Classroom Discussions"

    Science.gov (United States)

    DiLisi, Gregory A.; Hirsch, Allison; Murray, Meredith; Rarick, Richard

    2018-01-01

    A little-known type of nuclear reactor called the "molten salt reactor" (MSR), in which nuclear fuel is dissolved in a liquid carrier salt, was proposed in the 1940s and developed at the Oak Ridge National Laboratory in the 1960s. Recently, the MSR has generated renewed interest as a remedy for the drawbacks associated with conventional…

  10. Metallic materials corrosion problems in molten salt reactors

    International Nuclear Information System (INIS)

    Chauvin, G.; Dixmier, J.; Jarny, P.

    1977-01-01

    The USA forecastings concerning the molten salt reactors are reviewed (mixtures of fluorides containing the fuel, operating between 560 and 700 0 C). Corrosion problems are important in these reactors. The effects of certain characteristic factors on corrosion are analyzed: humidity and metallic impurities in the salts, temperature gradients, speed of circulation of salts, tellurium from fission products, coupling. In the molten fluorides and experimental conditions, the materials with high Ni content are particularly corrosion resistant alloys (hastelloy N). The corrosion of this material is about 2.6 mg.cm -2 at 700 0 C [fr

  11. Recommendations for a restart of molten salt reactor development

    International Nuclear Information System (INIS)

    Moir, R.W.

    2008-01-01

    The concept of the molten salt reactor (MSR) refuses to go away. The Generation-IV process lists the MSR as one of the six concepts to be considered for extending fuel resources. Good fuel utilization and good economics are required to meet the often-cited goal of 10 TWe globally and 1 TWe for the US by non-carbon energy sources in this century by nuclear fission. Strong incentives for the molten salt reactor design are its good fuel utilization, good economics, amazing fuel flexibility and promised large benefits. It can: - use thorium or uranium; - be designed with lots of graphite to have a fairly thermal neutron spectrum or without graphite moderator to have an epithermal neutron spectrum; - fission uranium isotopes and plutonium isotopes; - produces less long-lived wastes than today's reactors by a factor of 10-100; - operate with non-weapon grade fissile fuel, or in suitable sites it can operate with enrichment between reactor-grade and weapon grade fissile fuel; - be a breeder or near breeder; - operate at temperature >1100 deg. C if carbon composites are successfully developed. Enhancing 232 U content in the uranium to over 500 ppm makes the fuel undesirable for weapons, but it should not detract from its economic use in liquid fuel reactors: a big advantage in nonproliferation. Economics of the MSR are enhanced by operating at low pressure and high temperature and may even lead to the preferred route to hydrogen production. The cost of the electricity produced from low enriched fuel averaged over the life of the entire process, has been predicted to be about 10% lower than that from LWRs, and 20% lower for high-enriched fuel, with uncertainties of about 10%. The development cost has been estimated at about 1 B$ (e.g., a 100 M$/year base program for 10 years) not including construction of a series of reactors leading up to the deployment of multiple commercial units at an assumed cost of 9 B$ (450 M$/year over 20 years). A benefit of liquid fuel is that

  12. Molten salt reactors and possible scenarios for future nuclear power deployment

    International Nuclear Information System (INIS)

    Merle-Lucotte, E.; Mathieu, L.; Heuer, D.; Loiseaux, J. M.; Billebaud, A.; Brissot, R.; David, S.; Garzenne, C.; Laulan, O.; Le Brun, C.; Lecarpentier, D.; Liatard, E.; Meplan, O.; Michel-Sendis, F.; Nuttin, A.; Perdu, F.

    2004-01-01

    An important fraction of the nature energy demand may be satisfied by nuclear power. In this context, the possibilities of worldwide nuclear deployment are studied. We are convinced that the Molten Salt Reactors may play a central role in this deployment. The Molten Salt Reactor needs to be coupled to a reprocessing unit in order to extract the Fission Products which poison the core. The efficiency of this reprocessing has a crucial influence on reactor behavior especially for the breeding ratio. The Molten Salt Breeder Reactor project was based on an intensive reprocessing for high breeding purposes. A new concept of Thorium Molten Salt Reactor is presented here. Including this new concept in the worldwide nuclear deployment, to satisfy these power needs, we consider three typical scenarios, based on three reactor types: Pressurized Water Reactor, Fast Neutron Reactor and Thorium Molten Salt Reactor. The aim of this paper is to demonstrate, in a first hand that a Thorium Molten Salt Reactor can be realistic, with correct temperature coefficients and at least iso-breeder with slow reprocessing and new geometry; on the other hand that such Molten Salt Reactors enable a successful nuclear deployment, while minimizing fuel and waste management problems. (authors)

  13. Coupled study of the Molten Salt Fast Reactor core physics and its associated reprocessing unit

    International Nuclear Information System (INIS)

    Doligez, X.; Heuer, D.; Merle-Lucotte, E.; Allibert, M.; Ghetta, V.

    2014-01-01

    Highlights: • The limit on the reprocessing is due to the redox potential control. • Alkali and Earth-alkaline elements do not have to be extracted. • Criticality risks have to be studied in the reprocessing unit. • The neutronics properties are not sensitive to chemical data. • The reprocessing chemistry, from a pure numerical point of view, is an issue. - Abstract: Molten Salt Reactors (MSRs) are liquid-fuel reactors, in which the fuel is also the coolant and flows through the core. A particular configuration presented in this paper called the Molten Salt Fast Reactor consists in a Molten Salt Reactor with no moderator inside the core and a salt composition that leads to a fast neutron spectrum. Previous studies showed that this concept (previously called Thorium Molten Salt Reactor – Nonmoderated) has very promising characteristics. The liquid fuel implies a special reprocessing. Each day a small amount of the fuel salt is extracted from the core for on-site reprocessing. To study such a reactor, the materials evolution within the core has to be coupled to the reprocessing unit, since the latter cleans the salt quasi continuously and feeds the reactor. This paper details the issues associated to the numerical coupling of the core and the reprocessing. It presents how the chemistry is introduced inside the classical Bateman equation (evolution of nuclei within a neutron flux) in order to carry a numerical coupled study. To achieve this goal, the chemistry has to be modeled numerically and integrated to the equations of evolution. This paper presents how is it possible to describe the whole concept (reactor + reprocessing unit) by a system of equations that can be numerically solved. Our program is a connection between MCNP and a homemade evolution code called REM. Thanks to this tool; constraints on the fuel reprocessing were identified. Limits are specified to preserve the good neutronics properties of the MSFR. In this paper, we show that the limit

  14. Molten salt burner fuel behaviour and treatment

    International Nuclear Information System (INIS)

    Ignatiev, V.V.; Zakirov, R.Y.; Grebenkine, K.F.

    2001-01-01

    The objective of this paper is to discuss the feasibility of molten salt reactor technology for treatment of Pu, minor actinides and fission products, when the reactor and fission product clean-up unit are planned as an integral system. This contribution summarises the available R and D which led to selection of the fuel compositions for the molten salt reactor of the TRU burner type (MSB). Special characteristics of behaviour of TRUs and fission products during power operation of MSB concepts are presented. The present paper briefly reviews the processing developments underlying the prior molten salt reactor programmes and relates them to the separation requirements of the MSB concept, including the permissible range of processing cycle times and removal times. Status and development needs in the thermodynamic properties of fluorides, fission product clean-up methods and container materials compatibility with the working fluids for the fission product clean-up unit are discussed. (authors)

  15. Study of the pyrochemical treatment-recycling process of the Molten Salt Reactor fuel

    International Nuclear Information System (INIS)

    Boussier, H.; Heuer, D.

    2010-01-01

    The Separation Processes Studies Laboratory (Commissariat a l'energie Atomique) has made a preliminary assessment of the reprocessing system associated with Molten Salt Fast Reactor (MSFR). The scheme studied in this paper is based on the principle of reductive extraction and metal transfer that constituted the core process designed for the Molten Salt Breeder Reactor (MSBR), although the flow diagram has been adapted to the current needs of the Molten Salt Reactor Fast (MSFR).

  16. Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-02-01

    The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include TRISO particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Several preconceptual and conceptual design efforts that have been conducted on FHR concepts bear a significant influence on the FHR DR design. Specific designs include the Oak Ridge National Laboratory (ORNL) advanced high-temperature reactor (AHTR) with 3400/1500 MWt/megawatts of electric output (MWe), as well as a 125 MWt small modular AHTR (SmAHTR) from ORNL. Other important examples are the Mk1 pebble bed FHR (PB-FHR) concept from the University of California, Berkeley (UCB), and an FHR test reactor design developed at the Massachusetts Institute of Technology (MIT). The MIT FHR test reactor is based on a prismatic fuel platform and is directly relevant to the present FHR DR design effort. These FHR concepts are based on reasonable assumptions for credible commercial prototypes. The FHR DR concept also directly benefits from the operating experience of the Molten Salt Reactor Experiment (MSRE), as well as the detailed design efforts for a large molten salt reactor concept and its breeder variant, the Molten Salt Breeder Reactor. The FHR DR technology is most representative of the 3400 MWt AHTR

  17. Development of a safety analysis code for molten salt reactors

    International Nuclear Information System (INIS)

    Zhang Dalin; Qiu Suizheng; Su Guanghui

    2009-01-01

    The molten salt reactor (MSR) well suited to fulfill the criteria defined by the Generation IV International Forum (GIF) is presently revisited all around the world because of different attractive features of current renewed relevance. The MSRs are characterized by using the fluid-fuel, so that their technologies are fundamentally different from those used in the conventional solid-fuel reactors. In this work, in particular, the attention is focused on the safety characteristic analysis of the MSRs, in which a point kinetic model considering the flow effects of the fuel salt is established for the MSRs and calculated by developing a microcomputer code coupling with a simplified heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the molten salt actinide recycler and transmuter system (MOSART) by simulating three types of basic transient conditions including the unprotected loss of flow, unprotected overcooling accident and unprotected transient overpower. Some reasonable results are obtained for the MOSART, which show that the MOSART conceptual design is an inherently stable reactor design. The present study provides some valuable information for the research and design of the new generation MSRs.

  18. Numerical study on heat transfer characteristics of liquid-fueled molten salt using OpenFOAM

    International Nuclear Information System (INIS)

    Jeong, Yeong Shin; Bang, In Cheol

    2017-01-01

    To pursue sustainability and safety enhancement of nuclear energy, molten salt reactor is regarded as a promising candidate among various types of gen-IV reactors. Besides, pyroprocessing, which treats molten salt containing fission products, should consider safety related to decay heat from fuel material. For design of molten salt-related nuclear system, it is required to consider both thermal-hydraulic characteristics and neutronic behaviors for demonstration. However, fundamental heat transfer study of molten salt in operation condition is not easy to be experimentally studied due to its large scale, high temperature condition as well as difficulties of treating fuel material. >From that reason, numerical study can have benefit to investigate behaviors of liquid-fueled molten salt in real condition. In this study, open source CFD package OpenFOAM was used to analyze liquid-fueled molten salt loop having internal heat source as a first step of research. Among various molten salts considered as a candidate of liquid fueled molten salt reactors, in this study, FLiBe was chosen as liquid salt. For simulating heat generation from fuel material within fluid flow, volumetric heat source was set for fluid domain and OpenFOAM solver was modified as fvOptions as customized. To investigate thermal-hydraulic behavior of molten salt, CFD model was developed and validated by comparing experimental results in terms of heat transfer and pressure drop. As preliminary stage, 2D cavity simulations were performed to validate the modeling capacity of modified solver of OpenFOAM by comparison with those of ANSYS-CFX. In addition, cases of external heat flux and internal heat source were compared to configure the effect of heat source setting in various operation condition. As a result, modified solver of OpenFOAM considering internal heat source have sufficient modeling capacity to simulate liquid-fueled molten salt systems including heat generation cases. (author)

  19. Advancing Molten Salts and Fuels at Sandia National Laboratories

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, Salvador B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-09-26

    SNL has a combination of experimental facilities, nuclear engineering, nuclear security, severe nuclear accidents, and nuclear safeguards expertise that can enable significant progress towards molten salts and fuels for Molten Salt Reactors (MSRs). The following areas and opportunities are discussed in more detail in this white paper.

  20. Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Mueller, Donald E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

    2016-08-31

    Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

  1. Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

    International Nuclear Information System (INIS)

    Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.; Powers, Jeffrey J.

    2016-01-01

    Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 "7LiF-BeF_2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

  2. Fusion reactor fuel processing

    International Nuclear Information System (INIS)

    Johnson, E.F.

    1972-06-01

    For thermonuclear power reactors based on the continuous fusion of deuterium and tritium the principal fuel processing problems occur in maintaining desired compositions in the primary fuel cycled through the reactor, in the recovery of tritium bred in the blanket surrounding the reactor, and in the prevention of tritium loss to the environment. Since all fuel recycled through the reactor must be cooled to cryogenic conditions for reinjection into the reactor, cryogenic fractional distillation is a likely process for controlling the primary fuel stream composition. Another practical possibility is the permeation of the hydrogen isotopes through thin metal membranes. The removal of tritium from the ash discharged from the power system would be accomplished by chemical procedures to assure physiologically safe concentration levels. The recovery process for tritium from the breeder blanket depends on the nature of the blanket fluids. For molten lithium the only practicable possibility appears to be permeation from the liquid phase. For molten salts the process would involve stripping with inert gas followed by chemical recovery. In either case extremely low concentrations of tritium in the melts would be desirable to maintain low tritium inventories, and to minimize escape of tritium through unwanted permeation, and to avoid embrittlement of metal walls. 21 refs

  3. Issues of high-burnup fuel for advanced nuclear reactors

    International Nuclear Information System (INIS)

    Belac, J.; Milisdoerfer, L.

    2004-12-01

    A brief description is given of nuclear fuels for Generation III+ and IV reactors, and the major steps needed for a successful implementation of new fuels in prospective types of newly designed power reactors are outlined. The following reactor types are discussed: gas cooled fast reactors, heavy metal (lead) cooled fast reactors, molten salt cooled reactors, sodium cooled fast reactors, supercritical water cooled reactors, and very high temperature reactors. The following are regarded as priority areas for future investigations: (i) spent fuel radiotoxicity; (ii) proliferation volatility; (iii) neutron physics characteristics and inherent safety element assessment; technical and economic analysis of the manufacture of advanced fuels; technical and economic analysis of the fuel cycle back end, possibilities of spent nuclear fuel reprocessing, storage and disposal. In parallel, work should be done on the validation and verification of analytical tools using existing and/or newly acquired experimental data. (P.A.)

  4. Assessment of Silicon Carbide Composites for Advanced Salt-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Katoh, Yutai [ORNL; Wilson, Dane F [ORNL; Forsberg, Charles W [ORNL

    2007-09-01

    The Advanced High-Temperature Reactor (AHTR) is a new reactor concept that uses a liquid fluoride salt coolant and a solid high-temperature fuel. Several alternative fuel types are being considered for this reactor. One set of fuel options is the use of pin-type fuel assemblies with silicon carbide (SiC) cladding. This report provides (1) an initial viability assessment of using SiC as fuel cladding and other in-core components of the AHTR, (2) the current status of SiC technology, and (3) recommendations on the path forward. Based on the analysis of requirements, continuous SiC fiber-reinforced, chemically vapor-infiltrated SiC matrix (CVI SiC/SiC) composites are recommended as the primary option for further study on AHTR fuel cladding among various industrially available forms of SiC. Critical feasibility issues for the SiC-based AHTR fuel cladding are identified to be (1) corrosion of SiC in the candidate liquid salts, (2) high dose neutron radiation effects, (3) static fatigue failure of SiC/SiC, (4) long-term radiation effects including irradiation creep and radiation-enhanced static fatigue, and (5) fabrication technology of hermetic wall and sealing end caps. Considering the results of the issues analysis and the prospects of ongoing SiC research and development in other nuclear programs, recommendations on the path forward is provided in the order or priority as: (1) thermodynamic analysis and experimental examination of SiC corrosion in the candidate liquid salts, (2) assessment of long-term mechanical integrity issues using prototypical component sections, and (3) assessment of high dose radiation effects relevant to the anticipated operating condition.

  5. A simplified burnup calculation strategy with refueling in static molten salt reactor

    International Nuclear Information System (INIS)

    Srivastava, A.K.; Gupta, Anurag; Krishnani, P.D.

    2015-01-01

    Molten Salt Reactors, by nature can be refuelled and reprocessed online. Thus, a simulation methodology has to be developed which can consider online refueling and reprocessing aspect of the reactor. To cater such needs a simplified burnup calculation strategy to account for refueling and removal of molten salt fuel at any desired burnup has been identified in static molten salt reactor in batch mode as a first step of way forward. The features of in-house code ITRAN has been explored for such calculations. The code also enables us to estimate the reactivity introduced in the system due to removal of any number of considered nuclides at any burnup. The effect of refueling fresh fuel and removal of burned fuel has been studied in batch mode with in-house code ITRAN. The effect of refueling and burnup on change in reactivity per day has been analyzed. The analysis of removal of 233 Pa at a particular burnup has been carried out. The similar analysis has been performed for some other nuclides also. (author)

  6. Molten salt related extensions of the SIMMER-III code and its application for a burner reactor

    International Nuclear Information System (INIS)

    Wang Shisheng; Rineiski, Andrei; Maschek, Werner

    2006-01-01

    Molten salt reactors (MSRs) can be used as effective burners of plutonium (Pu) and minor actinides (MAs) from light water reactor (LWR) spent fuel. In this paper a study was made to examine the thermal hydraulic behaviour of the conceptual design of the molten salt advanced reactor transmuter (MOSART) [Ignatiev, V., Feynberg, O., Myasnikov, A., Zakirov, R., 2003a. Neutronic properties and possible fuel cycle of a molten salt transmuter. Proceedings of the 2003 ANS/ENS International Winter Meeting (GLOBAL 2003), Hyatt Regency, New Orleans, LA, USA 16-20 November 2003]. The molten salt fuel is a ternary NaF-LiF-BeF 2 system fuelled with ca. 1 mol% typical compositions of transuranium-trifluorides (PuF 3 , etc.) from light water reactor spent fuel. The MOSART reactor core does not contain graphite structure elements to guide the flow, so the neutron spectrum is rather hard in order to improve the burning performance. Without those structure elements in the core, the molten salt in core flows freely and the flow pattern could be potentially complicated and may affect significantly the fuel temperature distribution in the core. Therefore, some optimizations of the salt flow pattern may be needed. Here, the main attention has been paid to the fluid dynamic simulations of the MOSART core with the code SIMMER-III [Kondo, Sa., Morita, K., Tobita, Y., Shirakawa, K., 1992. SIMMER-III: an advanced computer program for LMFBR severe accident analysis. Proceedings of the ANP' 92, Tokyo, Japan; Kondo, Sa., Tobita, Y., Morita, K., Brear, D.J., Kamiyama, K., Yamano, H., Fujita, S., Maschek, W., Fischer, E.A., Kiefhaber, E., Buckel, G., Hesselschwerdt, E., Flad, M., Costa, P., Pigny, S., 1999. Current status and validation of the SIMMER-III LMFR safety analysis code. Proceedings of the ICONE-7, Tokyo, Japan], which was originally developed for the safety assessment of sodium-cooled fast reactors and recently extended by the authors for the thermo-hydraulic and neutronic models so as

  7. The integral fast reactor fuel cycle

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1990-01-01

    The liquid-metal reactor (LMR) has the potential to extend the uranium resource by a factor of 50 to 100 over current commercial light water reactors (LWRs). In the integral fast reactor (IFR) development program, the entire reactor system - reactor, fuel cycle, and waste process - is being developed and optimized at the same time as a single integral entity. A key feature of the IFR concept is the metallic fuel. The lead irradiation tests on the new U-Pu-Zr metallic fuel in the Experimental Breeder Reactor II have surpassed 185000 MWd/t burnup, and its high burnup capability has now been fully demonstrated. The metallic fuel also allows a radically improved fuel cycle technology. Pyroprocessing, which utilizes high temperatures and molten salt and molten metal solvents, can be advantageously utilized for processing metal fuels because the product is metal suitable for fabrication into new fuel elements. Direct production of a metal product avoids expensive and cumbersome chemical conversion steps that would result from use of the conventional Purex solvent extraction process. The key step in the IFR process is electrorefining, which provides for recovery of the valuable fuel constituents, uranium and plutonium, and for removal of fission products. A notable feature of the IFR process is that the actinide elements accompany plutonium through the process. This results in a major advantage in the high-level waste management

  8. Neutronic study of a nuclear reactor of fused salts

    International Nuclear Information System (INIS)

    Garcia B, F. B.; Francois L, J. L.

    2012-10-01

    The reactors of fused salts called Molten Salt Reactor have presented a resurgence of interest in the last decade, due to they have a versatility in particular to operate, either with a thermal or fast neutrons spectrum. The most active development was by the middle of 1950 and principles of 1970 in the Oak Ridge National Laboratory. In this work some developed models are presented particularly and studied with the help of the MCNPX code, for the development of the neutronic study of this reactor, starting of proposed models and from a simple and homogeneous geometry until other more complex models and approximate to more real cases. In particular the geometry conditions and criticality of each model were analyzed, the isotopic balance, as well as the concentrations of the salts and different assigned fuel types. (Author)

  9. Safety studies dedicated to molten salt reactors with a fast neutron spectrum and operated in the Thorium fuel cycle - Innovative concept of Molten Salt Fast Reactor

    International Nuclear Information System (INIS)

    Brovchenko, Mariya

    2013-01-01

    The nuclear reactors of the 4. generation must allow an optimized use of natural resources, while performing at a high safety level. The framework of this thesis is the deployment study of one of such a system, an innovative and still little studied Molten Salt Fast Reactor. An excellent safety is an ultimate requirement of the nuclear energy deployment, so it is important to raise this question at the current early stage of the MSFR concept development. This concept was the subject of a neutronic tool benchmark within a European project EVOL. Definition, calculations and results analyses were performed during this thesis. Comparisons of static neutronic and burn-up calculations, performed by the project participants, concluded to a good agreement between the different codes and methods used and pointed out the sensibility of the nuclear database choice on the results. With the aim of safety analysis of the MSFR, the decay heat was studied in detail. The tool used for the decay heat calculation was developed and validated, to finally evaluate the decay heat in the reactor. The decay heat source presented in different zones was quantified, concluding to a high importance of the cooling of the fuel salt and the bubbling system enclosing a part of the fission products. The safety analysis methodology was also studied in this thesis. Even if the safety principles are directly transposable to the MSFR, the precise recommendations are not. This is due to the specificity of the design that relies on the liquid state of the fuel, on the reprocessing systems located in the reactor and the embryonic stage of the design. First, a preliminary transposition work of some criteria to the MSFR design was realized, resulting amongst other things in a list of accidental scenarios particular for MSFR. Finally, a preliminary physical study of some types of accidental scenarios was performed, that can be used as a basis for further analyses with more sophisticated tools. (author) [fr

  10. The United States fluoride-salt-cooled high-temperature reactor program

    International Nuclear Information System (INIS)

    Holcomb, David E.

    2013-01-01

    The United States is pursuing the development of fluoride-salt-cooled high-temperature reactors (FHRs) through the Department of Energy's Office of Nuclear Energy (DOE-NE). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. FHRs, in principle, have the potential to economically generate large amounts of electricity while maintaining full passive safety. FHRs, however, remain a longer-term power production option. A principal development focus is, thus, on shortening, to the extent possible, the overall development time by focusing initial efforts on the longest lead-time issues. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid-metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High-temperature gas-cooled reactors provide experience with coated-particle fuel and graphite components. Light-water reactors show the potential of transparent, high-heat-capacity coolants with low chemical reactivity. The FHR development efforts include both reactor concept and technology developments and are being broadly pursued. Oak Ridge National Laboratory (ORNL) provides technical leadership to the effort and is performing concept development on both a large base-load-type FHR as well as a small modular reactor (SMR) in addition to performing a broad scope of technology developments. Idaho National Laboratory (INL) is providing coated-particle fuel irradiation testing as well as developing high-temperature steam generator technology. The Massachusetts Institute of Technology (MIT

  11. An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aufiero, M.; Cammi, A.; Fiorina, C. [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy); Leppänen, J. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland); Luzzi, L., E-mail: lelio.luzzi@polimi.it [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy); Ricotti, M.E. [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy)

    2013-10-15

    In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

  12. Accelerator molten-salt breeding and thorium fuel cycle

    International Nuclear Information System (INIS)

    Furukawa, Kazuo; Nakahara, Yasuaki; Kato, Yoshio; Ohno, Hideo; Mitachi, Kohshi.

    1990-01-01

    The recent efforts at the development of fission energy utilization have not been successful in establishing fully rational technology. A new philosophy should be established on the basis of the following three principles: (1) thorium utilization, (2) molten-salt fuel concept, and (3) separation of fissile-breeding and power-generating functions. Such philosophy is called 'Thorium Molten-Salt Nuclear Energy Synergetics [THORIMS-NES]'. The present report first addresses the establishment of 233 U breeding fuel cycle, focusing on major features of the Breeding and Chemical Processing Centers and a small molten-salt power station (called FUJI-II). The development of fissile producing breeders is discussed in relation to accelerator molten-salt breeder (AMSB), impact fusion molten-salt breeder, and inertial-confined fusion hybrid molten-salt breeder. Features of the accelerator molten-salt breeder are described, focusing on technical problems with accelerator breeders (or spallators), design principle of the accelerator molten-salt breeder, selection of molten salt compositions, and nuclear- and reactor-chemical aspects of AMSB. Discussion is also made of further research and development efforts required in the future for AMSB. (N.K.)

  13. Tritium control and capture in salt-cooled fission and fusion reactors: Status, challenges, and path forward

    International Nuclear Information System (INIS)

    Forsberg, Charles W.; Lam, Stephen; Carpenter, David M.; Whyte, Dennis G.; Scarlat, Raluca

    2017-01-01

    Three advanced nuclear power systems use liquid salt coolants that generate tritium and thus face the common challenges of containing and capturing tritium to prevent its release to the environment. The Fluoride-salt-cooled High-temperature Reactor (FHR) uses clean fluoride salt coolants and the same graphite-matrix coated-particle fuel as high-temperature gas-cooled reactors. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt with release of fission product tritium into the salt. In most FHR and MSR systems, the base-line salts contain lithium where isotopically separated "7Li is proposed to minimize tritium production from neutron interactions with the salt. The Chinese Academy of Science plans to start operation of a 2-MWt molten salt test reactor by 2020. For high-magnetic-field fusion machines, the use of lithium enriched in "6Li is proposed to maximize tritium generation the fuel for a fusion machine. Advances in superconductors that enable higher power densities may require the use of molten lithium salts for fusion blankets and as coolants. Recent technical advances in these three reactor classes have resulted in increased government and private interest and the beginning of a coordinated effort to address the tritium control challenges in 700 °C liquid salt systems. We describe characteristics of salt-cooled fission and fusion machines, the basis for growing interest in these technologies, tritium generation in molten salts, the environment for tritium capture, models for high-temperature tritium transport in salt systems, alternative strategies for tritium control, and ongoing experimental work. Several methods to control tritium appear viable. Finally, limited experimental data is the primary constraint for designing efficient cost-effective methods of tritium control.

  14. Validation of the TRACE code for the system dynamic simulations of the molten salt reactor experiment and the preliminary study on the dual fluid molten salt reactor

    International Nuclear Information System (INIS)

    He, Xun

    2016-01-01

    Molten Salt Reactor (MSR), which was confirmed as one of the six Generation IV reactor types by the GIF (Generation IV International Forum in 2008), recently draws a lot of attention all around the world. Due to the application of liquid fuels the MSR can be regarded as the most special one among those six GEN-IV reactor types in a sense. A unique advantage of using liquid nuclear fuel lies in that the core melting accident can be thoroughly eliminated. Besides, a molten salt reactor can have several fuel options, for instance, the fuel can be based on "2"3"5U, "2"3"2Th-"2"3"3U, "2"3"8U-"2"3"9Pu cycle or even the spent nuclear fuel (SNF), so the reactor can be operated as a breeder or as an actinides burner both with fast, thermal or epi-thermal neutron spectrum and hence, it has excellent features of the fuel sustainability and for the non-proliferation. Furthermore, the lower operating pressure not only means a lower risk of the explosion as well as the radioactive leakage but also implies that the reactor vessel and its components can be lightweight, thus lowering the cost of equipments. So far there is no commercial MSR being operated. However, the MSR concept and its technical validation dates back to the 1960s to 1970s, when the scientists and engineers from ORNL (Oak Ridge National Laboratory) in the United States managed to build and run the world's first civilian molten salt reactor called MSRE (Molten Salt Reactor Experiment). The MSRE was an experimental liquid-fueled reactor with 10 MW thermal output using "4LiF-BeF_2-ZrF_4-UF_4 as the fuel also as the coolant itself. The MSRE is usually taken as a very important reference case for many current researches to validate their codes and simulations. Without exception it works also as a benchmark for this thesis. The current thesis actually consists of two main parts. The first part is about the validation of the current code for the old MSRE concept, while the second one is about the demonstration of a new

  15. Validation of the TRACE code for the system dynamic simulations of the molten salt reactor experiment and the preliminary study on the dual fluid molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    He, Xun

    2016-06-14

    Molten Salt Reactor (MSR), which was confirmed as one of the six Generation IV reactor types by the GIF (Generation IV International Forum in 2008), recently draws a lot of attention all around the world. Due to the application of liquid fuels the MSR can be regarded as the most special one among those six GEN-IV reactor types in a sense. A unique advantage of using liquid nuclear fuel lies in that the core melting accident can be thoroughly eliminated. Besides, a molten salt reactor can have several fuel options, for instance, the fuel can be based on {sup 235}U, {sup 232}Th-{sup 233}U, {sup 238}U-{sup 239}Pu cycle or even the spent nuclear fuel (SNF), so the reactor can be operated as a breeder or as an actinides burner both with fast, thermal or epi-thermal neutron spectrum and hence, it has excellent features of the fuel sustainability and for the non-proliferation. Furthermore, the lower operating pressure not only means a lower risk of the explosion as well as the radioactive leakage but also implies that the reactor vessel and its components can be lightweight, thus lowering the cost of equipments. So far there is no commercial MSR being operated. However, the MSR concept and its technical validation dates back to the 1960s to 1970s, when the scientists and engineers from ORNL (Oak Ridge National Laboratory) in the United States managed to build and run the world's first civilian molten salt reactor called MSRE (Molten Salt Reactor Experiment). The MSRE was an experimental liquid-fueled reactor with 10 MW thermal output using {sup 4}LiF-BeF{sub 2}-ZrF{sub 4}-UF{sub 4} as the fuel also as the coolant itself. The MSRE is usually taken as a very important reference case for many current researches to validate their codes and simulations. Without exception it works also as a benchmark for this thesis. The current thesis actually consists of two main parts. The first part is about the validation of the current code for the old MSRE concept, while the second

  16. Preliminary safety analysis of molten salt breeder reactor

    International Nuclear Information System (INIS)

    Cheng Maosong; Dai Zhimin

    2013-01-01

    Background: The molten salt reactor is one of the six advanced reactor concepts identified by the Generation IV International Forum as a candidate for cooperative development, which is characterized by remarkable advantages in inherent safety, fuel cycle, miniaturization, effective utilization of nuclear resources and proliferation resistance. ORNL finished the conceptual design of Molten Salt Breeder Reactor (MSBR) based on the design, building and operation of Molten Salt Reactor Experiment (MSRE). Purpose: We attempt to implement the preliminary safety analysis of MSBR in order to provide a reference for the design and optimization of MSBR in the future. Methods: According to the conceptual design of MSBR, a model of safety analysis using point kinetics coupled with the simplified heat transfer mechanism is presented. The model is applied to simulate the transient phenomena of MSBR initiated by an abnormal step reactivity addition and an abnormal ramp reactivity addition at full-power equilibrium condition. Results: The thermal power in the core increases rapidly at the beginning and is accompanied by a rise of the fuel and graphite temperatures after 100, 300, 500 and 600 pcm reactivity addition. The maximum outlet temperature of the fuel in the core is at 1250℃ in 500 pcm reactivity addition, but up to 1350℃ in 600 pcm reactivity addition. The maximum of the power and the temperature are delayed and lower in the ramp reactivity addition rather than in the step reactivity addition. Conclusions: Based on the results, when the reactivity inserted is less than 500 pcm in maximum at full power equilibrium condition, the structural material in Hastelloy-N is not melted and can keep integrity without external control action. And it is necessary to try to avoid inserting a reactivity at short time. (authors)

  17. Simulation of Molten Salt Reactor dynamics

    International Nuclear Information System (INIS)

    Krepel, J.; Rohde, U.; Grundmann, U.

    2005-01-01

    Dynamics of the Molten Salt Reactor - one of the 'Generation IV' concepts - was studied in this paper. The graphite-moderated channel type MSR was selected for the numerical simulation of the reactor with liquid fuel. The MSR dynamics is very specific because of two physical peculiarities of the liquid fueled reactor: the delayed neutrons precursors are drifted by the fuel flow and the fission energy is immediately released directly into the coolant. Presently, there are not many accessible numerical codes appropriate for the MSR simulation, therefore the DYN3D-MSR code was developed based on the FZR in-house code DYN3D. It allows calculating of full 3D transient neutronics in combination with parallel channel type thermal-hydraulics. By means of DYN3D-MSR, several transients typical for the liquid fuel system were analyzed. Those transients were initiated by reactivity insertion, by overcooling of fuel at the core inlet, by the fuel pump start-up or coast-down, or by the blockage of selected fuel channels. In these considered transients, the response of the MSR is characterized by the immediate change of the fuel temperature with changing power and fast negative temperature feedback to the power. The response through the graphite temperature is slower. Furthermore, for big MSR cores fueled with U233 the graphite feedback coefficient can be positive. In this case the addition of erbium to the graphite can ensure the inherent safety features. The DYN3D-MSR code has been shown to be an effective tool for MSR dynamics studies. (author)

  18. Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors

    International Nuclear Information System (INIS)

    Petrovic, Bojan; Maldonado, Ivan

    2016-01-01

    The research performed in this project addressed the issue of low heavy metal loading and the resulting reduced cycle length with increased refueling frequency, inherent to all FHR designs with solid, non-movable fuel based on TRISO particles. Studies performed here focused on AHTR type of reactor design with plate ('plank') fuel. Proposal to FY12 NEUP entitled 'Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors' was selected for award, and the 3-year project started in August 2012. A 4-month NCE was granted and the project completed on December 31, 2015. The project was performed by Georgia Tech (Prof. Bojan Petrovic, PI) and University of Tennessee (Prof. Ivan Maldonado, Co-PI), with a total funding of $758,000 over 3 years. In addition to two Co-PIs, the project directly engaged 6 graduate students (at doctoral or MS level) and 2 postdoctoral researchers. Additionally, through senior design projects and graduate advanced design projects, another 23 undergraduate and 12 graduate students were exposed to and trained in the salt reactor technology. We see this as one of the important indicators of the project's success and effectiveness. In the process, 1 journal article was published (with 3 journal articles in preparation), together with 8 peer-reviewed full conference papers, 8 peer-reviewed extended abstracts, as well as 1 doctoral dissertation and 2 master theses. The work included both development of models and methodologies needed to adequately analyze this type of reactor, fuel, and its fuel cycle, as well as extensive analyses and optimization of the fuel and core design.

  19. Preliminary analysis on in-core fuel management optimization of molten salt pebble-bed reactor

    International Nuclear Information System (INIS)

    Xia Bing; Jing Xingqing; Xu Xiaolin; Lv Yingzhong

    2013-01-01

    The Nuclear Hot Spring (NHS) is a molten salt pebble-bed reactor featured by full power natural circulation. The unique horizontal coolant flow of the NHS demands the fuel recycling schemes based on radial zoning refueling and the corresponding method of fuel management optimization. The local searching algorithm (LSA) and the simulated annealing algorithm (SAA), the stochastic optimization methods widely used in the refueling optimization problems in LWRs, were applied to the analysis of refueling optimization of the NHS. The analysis results indicate that, compared with the LSA, the SAA can survive the traps of local optimized solutions and reach the global optimized solution, and the quality of optimization of the SAA is independent of the choice of the initial solution. The optimization result gives excellent effects on the in-core power flattening and the suppression of fuel center temperature. For the one-dimensional zoning refueling schemes of the NHS, the SAA is an appropriate optimization method. (authors)

  20. Thorium converter (ThorCon) - a doable molten salt reactor

    International Nuclear Information System (INIS)

    Myneni, Ganapati

    2015-01-01

    ThorCon mass-producible nuclear power plants are being built to generate electricity cheaper than coal, at a scale to make a real improvement in world poverty and environment, now. ThorCon irradiated materials and fuel salt are designed to be replaced in four-year cycles with no impact on electricity generation. This flex-fuel plant and its replaceable reactor cans can operate with mixtures of thorium and uranium at multiple enrichments. Fuel salt can be NaF/BeF 2 or LiF/BeF 2 if available. ThorCon's design exceeds current nuclear power safety practice. The team calls for regulatory participation in rigorous testing of a full-scale prototype to develop licensing guidance

  1. Pyrochemical recovery of actinide elements from spent light water reactor fuel

    International Nuclear Information System (INIS)

    Johnson, G.K.; Pierce, R.D.; Poa, D.S.; McPheeters, C.C.

    1994-01-01

    Argonne National Laboratory is investigating salt transport and lithium pyrochemical processes for recovery of transuranic (TRU) elements from spent light water reactor fuel. The two processes are designed to recover the TRU elements in a form compatible with the Integral Fast Reactor (IFR) fuel cycle. The IFR is uniquely effective in consuming these long-lived TRU elements. The salt transport process uses calcium dissolved in Cu-35 wt % Mg in the presence of a CaCl 2 salt to reduce the oxide fuel. The reduced TRU elements are separated from uranium and most of the fission products by using a MgCl 2 transport salt. The lithium process, which does not employ a solvent metal, uses lithium in the presence of a LiCl salt as the reductant. After separation from the salt, the reduced metal is introduced into an electrorefiner, which separates the TRU elements from the uranium and fission products. In both processes, reductant and reduction salt are recovered by electrochemical decomposition of the oxide reaction product

  2. Transient thermal characteristics of a core channel in a molten salt reactor

    International Nuclear Information System (INIS)

    Sakashita, H.; Ishiguro, R.; Sugiyama, K.

    1987-01-01

    The present paper deals with the thermal characteristics of Molten Salt Reactor (MSR). Analyses of the fundamental behavior of internal heat generating fluid and graphite contiguous to the fluid are performed. As a result, it is known that the transient thermal characteristics of MSR differ fundamentally from those of a solid-fuel reactor, and the simplified method of thermal analysis which is commonly used for solid-fuel reactors gives optimistic predictions than the actual phenomena. (author)

  3. Three-dimensional numerical investigation of a Molten Salt reactor concept with the code CFX-5.5

    International Nuclear Information System (INIS)

    Yamaji, B.; Csom, G.; Aszodi, A.

    2002-01-01

    Partitioning and transmutation of actinides and long-lived fission products is a promising option to extend the possibilities and enhance the environmentally acceptable capabilities of nuclear energy. Also the possible implementation of the thorium cycle is considered as a way to reduce the problem of energy resources in the future. For both objectives different molten salt reactor concepts were proposed mainly based on the Molten Salt Reactor Experiment of the Oak Ridge National Laboratory. Not only critical reactors but also accelerator-driven subcritical systems (ADSs) have advantages worth considering for those aims, especially those ones with liquid fuel, such as molten salts. By using liquid fuel which is the coolant medium, too, a basically different thermalhydraulic behavior is expected than in the case of solid fuel and water coolant. In this work our purpose is to present the possible use of Computational Fluid Dynamics (CFD) technology in molten salt thermal hydraulics. The simulations were performed with the three-dimensional code CFX-5.5.(author)

  4. Simulation tools and new developments of the molten salt fast reactor

    International Nuclear Information System (INIS)

    Heuer, D.; Merle-Lucotte, E.; Allibert, M.; Doligez, X.; Ghetta, V.

    2010-01-01

    In the MSFR (Molten Salt Fast Reactor), the liquid fuel processing is part of the reactor where a small side stream of the molten salt is processed for fission product removal and then returned to the reactor. Because of this design characteristic, the MSFR can thus operate with a widely varying fuel composition. Our reactor's studies of the MSFR concept rely on numerical simulations making use of the MCNP neutron transport code coupled with a code for Bateman's equations computing the population of any nucleus inside any part of the reactor at any moment. The classical Bateman's equations have been modified by adding 2 terms representing the reprocessing capacities and an online addition. We have thus coupled neutronic and reprocessing simulation codes in a numerical tool used to calculate the extraction efficiencies of fission products, their location in the whole system and radioprotection issues. The very preliminary results show the potential of the neutronic-reprocessing coupling we have developed. We also show that these studies are limited by the uncertainties on the design and the knowledge of the chemical reprocessing processes. (A.C.)

  5. Molten salt reactors. Synthesis of studies realized between 1973 and 1983. Carbon-materials file

    International Nuclear Information System (INIS)

    1983-03-01

    The study of a molten salt fueled reactor requires a thorough examination of carbon containing materials for moderator, reflectors and structural materials. Are examined: texture, structure, physical and mechanical properties, chemical purity, neutron irradiation, salt-graphite and salt-lead interactions for different types of graphite. [fr

  6. Method for calculating the steady-state distribution of tritium in a molten-salt breeder reactor plant

    International Nuclear Information System (INIS)

    Briggs, R.B.; Nestor, C.W.

    1975-04-01

    Tritium is produced in molten salt reactors primarily by fissioning of uranium and absorption of neutrons by the constituents of the fuel carrier salt. At the operating temperature of a large power reactor, tritium is expected to diffuse from the primary system through pipe and vessel walls to the surroundings and through heat exchanger tubes into the secondary system which contains a coolant salt. Some tritium will pass from the secondary system into the steam power system. This report describes a method for calculating the steady state distribution of tritium in a molten salt reactor plant and a computer program for making the calculations. The method takes into account the effects of various processes for removing tritium, the addition of hydrogen or hydrogenous compounds to the primary and secondary systems, and the chemistry of uranium in the fuel salt. Sample calculations indicate that 30 percent or more of the tritium might reach the steam system in a large power reactor unless special measures are taken to confine the tritium. (U.S.)

  7. ANALISIS TRANSIEN PADA PASSIVE COMPACT MOLTEN SALT REACTOR (PCMSR

    Directory of Open Access Journals (Sweden)

    M. Makrus Imron

    2015-04-01

    Full Text Available Penggunaan bahan bakar cair berupa garam LiF-BeF2-ThF4-UF4 pada Passive Compact Molten Salt Reactor (PCMSR meyebabkan pengendalian daya pada PCMSR dapat dilakukan dengan mengendalikan laju aliran bahan bakar dan pendingin. Sedangkan dari sistem keselamatan, penggunaan bahan bakar cair menjadikan PCMSR memiliki karakter keselamatan melekat (inherent safety yang baik. Pada penelitian ini telah dilakukan analisis transien PCMSR pada tiga kondisi, yaitu: ketika terjadi perubahan laju aliran bahan bakar, ketika terjadi perubahan laju aliran pendingin dan ketika terdapat kegagalan pada sistem pelepasan panas (loss of heat sink. Penelitian dilakukan dengan memodelkan reaktor pada kondisi tunak menggunakan paket program. Standart Reactor Analysis Code (SRAC. Selanjutnya dari keluaran paket program SRAC diperoleh data data yang meliputi fluks netron,konstanta grup, kontanta peluran prekusor netron, fraksi netron kasip untuk perhitungan transien. Penelitian ini menunjukkan bahwa penurunan laju aliran bahan bakar sebesar 50 % dari laju bahan bakar sebelumnya, menyebabkan daya pada PCMSR turun menjadi 78 % dari daya sebelumnya. Dan penurunan laju aliran pendingin sebesar 50 % dari laju pendingin sebelumnya, menyebabkan daya pada PCMSR turun menjadi 63 % dari daya sebelumnya. Sedangkan pada saat terjadi loss of heat sink daya PCMSR menunjukkan penurunan. Kata kunci: PCMSR, transien, daya, laju aliran.   The use of liquid fuels in the form of molten salts LiF-BeF2-ThF4-UF4 in Passive Compact Molten Salt Reactor (PCMSR makes power control at PCMSR can be done by controlling the flow rate of fuel and coolant. In addition, from safety systems aspect, the use of liquid fuels makes PCMSR has good inherent safety characteristics. In this study transient analysis has been carried out on three conditions of PCMSR, namely when the fuel flow rate is changing, when the coolant flow rate is changing and when there is loss of heat sink condition. This research is

  8. Subcritical molten salt reactor with fast/intermediate spectrum for minor actinides transmutation

    International Nuclear Information System (INIS)

    Degtyarev, Alexey M.; Feinberg, Olga S.; Kolyaskin, Oleg E.; Myasnikov, Andrey A.; Karmanov, Fedor I.; Kuznetsov, Andrey Yu.; Ponomarev, Leonid I.; Seregin, Mikhail B.; Sidorkin, Stanislav F.

    2011-01-01

    The subcritical molten-salt reactor for transmutation of Am and Cm with the fast-intermediate neutron spectrum is suggested. It is shown that ∼10 such reactor-burners is enough to support the future nuclear power based on the fast reactors as well as for the transmutation of Am and Cm accumulated in the spent fuel storages. (author)

  9. Molten salts as possible fuel fluids for TRU fuelled systems: ISTC no. 1606 approach

    International Nuclear Information System (INIS)

    Ignatiev, V.; Zakirov, R.; Grebenkine, K.

    2001-01-01

    The principle attraction of the molten salt reactor (MSR) technology is the use of fuel/fertile material flexibility (easy of fuel preparation and processing) for gaining additional profits as compared with solid materials. This approach presents important departures from traditional philosophy, applied in current nuclear power plants, and to some extent contradicts the straightforward interpretation of the defence-in-depth principal. Nevertheless we understand there may be potential to use MSR technology to support back end fuel cycle technologies in future commercial environment. The paper aims at reviewing results of the work performed in Russia, relevant to the problems of MSR technology development. Also this contribution aims at evaluation of remaining uncertainties for molten salt burner concept implementation. Fuel properties and behaviour, container materials, and clean-up of fuels with emphasis on experiments will be of priority. Recommendations are made regarding the types of experimental studies needed on a way to implement molten salt technology to the back-end of the fuel cycle. To better understand the potential and limitations of the molten salts as a fuel for reactor of incinerator type, Russian Institutes have submitted to the ISTC the Task no. 1606 Experimental Study of Molten Salt Technology for Safe and Low Waste Treatment of Plutonium and Minor Actinides in Accelerator Driven and Critical Systems. The project goals, technical approach and expected specific results are discussed. (author)

  10. The Molten Salt Reactor option for beneficial use of fissile material from dismantled weapons

    International Nuclear Information System (INIS)

    Gat, U.; Engel, J.R.; Dodds, H.L.

    1991-01-01

    The Molten Salt Reactor (MSR) option for burning fissile fuel from dismantled weapons is examined. It is concluded that MSRs are very suitable for beneficial utilization of the dismantled fuel. The MSRs can utilize any fissile fuel in continuous operation with no special modifications, as demonstrated in the Molten Salt Reactor Experiment. Thus MSRs are flexible while maintaining their economy. MSRs further require a minimum of special fuel preparation and can tolerate denaturing and dilution of the fuel. Fuel shipments can be arbitrarily small, all of which supports nonproliferation and averts diversion. MSRs have inherent safety features which make them acceptable and attractive. They can burn a fuel type completely and convert it to other fuels. MSRs also have the potential for burning the actinides and delivering the waste in an optimal form, thus contributing to the solution of one of the major remaining problems for deployment of nuclear power. 19 refs

  11. Neutronics of a liquid salt cooled - very high temperature reactor

    International Nuclear Information System (INIS)

    Zakova, J.

    2007-01-01

    During last few years, the interest in the innovative, Liquid Salt cooled - Very High Temperature Reactor (LS-VHTR), has been growing. The preconceptual design of the LS-VHTR was suggested in Oak Ridge National Laboratory (ORNL) [1] and nowadays, several research institutions contribute to the development of this concept. The LS-VHTR design utilises a prismatic, High Temperature Reactor (HTR) fuel [2] in combination with liquid salt as a coolant. This connection of high-performance fuel and a coolant with enhanced heat transfer abilities enables efficient and economical operation. Main objective of the LS-VHTR operation may be either an efficient electricity production or a heat supply for a production of hydrogen or, combination of both. The LS-VHTR is moderated by graphite. The graphite matrix of the fuel blocks, as well as the inner and outer core reflectors serve as a thermal buffer in case of an accident, and they provide a strong thermal feedback during normal reactor operation. The high inherent safety of the LS-VHTR meets the strict requirements on future reactor systems, as defined by the Gen IV project. This work, purpose, scope, contribution to the state-of-art: The design, used in the present work is based on the first ORNL suggestion [1]. Recent study is focused on comparison of the neutronic performance of two types of fuel in the LS-VHTR core, whereas, in all previous works, only uranium fuel has been investigated. The first type of fuel, which has been employed in the present analysis, is based on the spent Light Water Reactor (LWR) fuel, whereas the second one consists of enriched uranium oxide. The results of such a comparison bring a valuable knowledge about limits and possibilities of the LS-VHTR concept, when employed as a spent fuel burner. Method:It is used a 3-D drawing of the LS-VHTR core, which contains 324x10 hexagonal fuel blocks. Each fuel block contains 216x10 fuel pins, which consists of TRISO particles incorporated into a graphite

  12. Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Maldonado, Ivan [Univ. of Tennessee, Knoxville, TN (United States)

    2016-04-14

    The research performed in this project addressed the issue of low heavy metal loading and the resulting reduced cycle length with increased refueling frequency, inherent to all FHR designs with solid, non-movable fuel based on TRISO particles. Studies performed here focused on AHTR type of reactor design with plate (“plank”) fuel. Proposal to FY12 NEUP entitled “Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors” was selected for award, and the 3-year project started in August 2012. A 4-month NCE was granted and the project completed on December 31, 2015. The project was performed by Georgia Tech (Prof. Bojan Petrovic, PI) and University of Tennessee (Prof. Ivan Maldonado, Co-PI), with a total funding of $758,000 over 3 years. In addition to two Co-PIs, the project directly engaged 6 graduate students (at doctoral or MS level) and 2 postdoctoral researchers. Additionally, through senior design projects and graduate advanced design projects, another 23 undergraduate and 12 graduate students were exposed to and trained in the salt reactor technology. We see this as one of the important indicators of the project’s success and effectiveness. In the process, 1 journal article was published (with 3 journal articles in preparation), together with 8 peer-reviewed full conference papers, 8 peer-reviewed extended abstracts, as well as 1 doctoral dissertation and 2 master theses. The work included both development of models and methodologies needed to adequately analyze this type of reactor, fuel, and its fuel cycle, as well as extensive analyses and optimization of the fuel and core design.

  13. The molten salt reactors (MSR) pyro chemistry and fuel cycle for innovative nuclear systems; Congres sur les reacteurs a sels fondus (RSF) pyrochimie et cycles des combustibles nucleaires du futur

    Energy Technology Data Exchange (ETDEWEB)

    Brossard, Ph. [GEDEON, Groupement de Recherche CEA CNRS EDF FRAMATOME (France); Garzenne, C.; Mouney, H. [and others

    2002-07-01

    In the frame of the studies on next generation nuclear systems, and especially for the molten salt reactors and for the integrated fuel cycle (as IFR), the fuel cycle constraints must be taken into account in the preliminary studies of the system to improve the cycle and reactor optimisation. Among the purposes for next generation nuclear systems, sustainability and waste (radio-toxicity and mass) management are important goals. These goals imply reprocessing and recycling strategies. The objectives of this workshop are to present and to share the different strategies and scenarios, the needs based on these scenarios, the experimental facilities available today or in the future and their capabilities, the needs for demonstration. It aims at: identifying the needs for fuel cycle based on solid fuel or liquid fuel, and especially, the on-line reprocessing or clean up for the molten salt reactors; assessing the state-of-the-art on the pyro-chemistry applied to solid fuel and to present the research activities; assessing the state-of-the-art on liquid fuels (or others), and to present the research activities; expressing the R and D programs for pyro-chemistry, molten salt, and also to propose innovative processes; and proposing some joint activities in the frame of GEDEON and PRACTIS programs. This document brings together the transparencies of 18 contributions dealing with: scenario studies with AMSTER concept (Scenarios, MSR, breeders (Th) and burners); fuel cycle for innovative systems; current reprocessing of spent nuclear fuel (SNF) in molten salts (review of pyro-chemistry processes (non nuclear and nuclear)); high temperature NMR spectroscopies in molten salts; reductive extraction of An from molten fluorides (salt - liquid metal extraction); electrochemistry characterisation; characterisation with physical methods - extraction coefficient and kinetics; electrolytic extraction; dissolution-precipitation of plutonium in the eutectic LiCl-KCl (dissolution and

  14. Process technology for the molten-salt reactor 233U--Th cycle

    International Nuclear Information System (INIS)

    Hightower, J.R. Jr.

    1975-01-01

    After a brief description of the design features of the molten-salt breeder reactor, fuel processing for removal of 233 Pa and fission products is examined. Some recent developments in processing technology are discussed

  15. Thorium cycle and molten salt reactors: field parameters and field constraints investigations toward 'thorium molten salt reactor' definition; Cycle thorium et reacteurs a sel fondu: exploration du champ des parametres et des contraintes definissant le 'Thorium Molten Salt Reactor'

    Energy Technology Data Exchange (ETDEWEB)

    Mathieu, L

    2005-09-15

    Producing nuclear energy in order to reduce the anthropic CO{sub 2} emission requires major technological advances. Nuclear plants of 4. generation have to respond to several constraints, as safety improvements, fuel breeding and radioactive waste minimization. For this purpose, it seems promising to use Thorium Cycle in Molten Salt Reactors. Studies on this domain have already been carried out. However, the final concept suffered from serious issues and was discontinued. A new reflection on this topic is being led in order to find acceptable solutions, and to design the Thorium Molten Salt Reactor concept. A nuclear reactor is simulated by the coupling of a neutron transport code with a materials evolution code. This allows us to reproduce the reactor behavior and its evolution all along its operation. Thanks to this method, we have studied a large number of reactor configurations. We have evaluated their efficiency through a group of constraints they have to satisfy. This work leads us to a better understanding of many physical phenomena controlling the reactor behavior. As a consequence, several efficient configurations have been discovered, allowing the emergence of new points of view in the research of Molten Salt Reactors. (author)

  16. Waste management in IFR [Integral Fast Reactor] fuel cycle

    International Nuclear Information System (INIS)

    Johnson, T.R.; Battles, J.E.

    1991-01-01

    The fuel cycle of the Integral Fast Reactor (IFR) has important potential advantage for the management of high-level wastes. This sodium-cooled, fast reactor will use metal fuels that are reprocessed by pyrochemical methods to recover uranium, plutonium, and the minor actinides from spent core and blanket fuel. More than 99% of all transuranic (TRU) elements will be recovered and returned to the reactor, where they are efficiently burned. The pyrochemical processes being developed to treat the high-level process wastes are capable of producing waste forms with low TRU contents, which should be easier to dispose of. However, the IFR waste forms present new licensing issues because they will contain chloride salts and metal alloys rather than glass or ceramic. These fuel processing and waste treatment methods can also handle TRU-rich materials recovered from light-water reactors and offer the possibility of efficiently and productively consuming these fuel materials in future power reactors

  17. Fission product behavior in the Molten Salt Reactor Experiment

    International Nuclear Information System (INIS)

    Compere, E.L.; Kirslis, S.S.; Bohlmann, E.G.; Blankenship, F.F.; Grimes, W.R.

    1975-10-01

    Essentially all the fission product data for numerous and varied samples taken during operation of the Molten Salt Reactor Experiment or as part of the examination of specimens removed after particular phases of operation are reported, together with the appropriate inventory or other basis of comparison, and relevant reactor parameters and conditions. Fission product behavior fell into distinct chemical groups. Evidence for fission product behavior during operation over a period of 26 months with 235 U fuel (more than 9000 effective full-power hours) was consistent with behavior during operation using 233 U fuel over a period of about 15 months (more than 5100 effective full-power hours)

  18. The molten salt reactor option for beneficial use of fissile material from dismantled weapons

    International Nuclear Information System (INIS)

    Gat, U.; Engel, J.R.

    1991-01-01

    The Molten Salt Reactor (MSR) option for burning fissile fuel from dismantled weapons is examined and is found very suitable for the beneficial use of this fuel. MSRs can utilize any fissile fuel in continuous operation with no special modifications, as demonstrated in the Molten Salt Reactor Experiment. Thus, MSRs are flexible while maintaining their economy. Furthermore, MSRs require only a minimum of special fuel preparation. They can tolerate denaturing and dilution of their fuel. The size of fuel shipments can be determined to optimize safety and security-all of which supports nonproliferation and resists diversion. In addition, MSRs have inherent safety features that make them acceptable and attractive. They can burn fissile material completely or can convert it to other fuels. MSRs also have the potential for burning the actinides and delivering the waste in an optimal form, thus contributing to the solution of one of the major remaining problems in the deployment of nuclear power

  19. Thermal-Hydraulics Phenomena Important in Modeling and Simulation of Liquid-Fuel Molten Salt Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bajorek, Stephen; Diamond, David J.

    2018-11-11

    This paper discusses liquid-fuel molten salt reactors, how they will operate under normal, transient, and accident conditions, and the results of an expert elicitation to determine the corresponding thermalhydraulic phenomena important to understanding their behavior. Identifying these phenomena will enable the U.S. Nuclear Regulatory Commission (NRC) to develop or identify modeling functionalities and tools required to carry out confirmatory analyses that examine the validity and accuracy of an applicant’s calculations and help determine the margin of safety in plant design. NRC frequently does an expert elicitation using a Phenomena Identification and Ranking Table (PIRT) to identify and evaluate the state of knowledge of important modeling phenomena. However, few details about the design of these reactors and the sequence of events during accidents are known, so the process used was considered a preliminary PIRT. A panel met to define phenomena that would need to be modeled and considered the impact/importance of each phenomenon with respect to specific figures-of-merit (FoMs) (e.g., salt temperature, velocity, and composition). Each FoM reflected a potential impact on radionuclide release or loss of a barrier to release. The panel considered what the path forward might be with respect to being able to model the phenomenon in a simulation code. Results are explained for both thermal and fast spectrum designs.

  20. Impact on breeding rate of different Molten Salt reactor core structures

    International Nuclear Information System (INIS)

    Wang Haiwei; Mei Longwei; Cai Xiangzhou; Chen Jingen; Guo Wei; Jiang Dazhen

    2013-01-01

    Background: Molten Salt Reactor (MSR) has several advantages over the other Generation IV reactor. Referred to the French CNRS research and compared to the fast reactor, super epithermal neutron spectrum reactor type is slightly lower and beading rate reaches 1.002. Purpose: The aim is to explore the best conversion zone layout scheme in the super epithermal neutron spectrum reactor. This study can make nuclear fuel as one way to solve the energy problems of mankind in future. Methods: Firstly, SCALE program is used for molten salt reactor graphite channel, molten salt core structure, control rods, graphite reflector and layer cladding structure. And the SMART modules are used to record the important actinides isotopes and their related reaction values of each reaction channel. Secondly, the thorium-uranium conversion rate is calculated. Finally, the better molten salt reactor core optimum layout scheme is studied comparing with various beading rates. Results: Breading zone layout scheme has an important influence on the breading rate of MSR. Central graphite channels in the core can get higher neutron flux irradiation. And more 233 Th can convert to 233 Pa, which then undergoes beta decay to become 233 U. The graphite in the breading zone gets much lower neutron flux irradiation, so the life span of this graphite can be much longer than that of others. Because neutron flux irradiation in the uranium molten salt graphite has nearly 10 times higher than the graphite in the breading zone, it has great impact on the thorium-uranium conversion rates. For the super epithermal neutron spectrum molten salt reactors, double salt design cannot get higher thorium-uranium conversion rates. The single molten salt can get the same thorium-uranium conversion rate, meanwhile it can greatly extend the life of graphite in the core. Conclusions: From the analysis of calculation results, Blanket breeding area in different locations in the core can change the breeding rates of thorium

  1. A Novel Molten Salt Reactor Concept to Implement the Multi-Step Time-Scheduled Transmutation Strategy

    International Nuclear Information System (INIS)

    Csom, Gyula; Feher, Sandor; Szieberthj, Mate

    2002-01-01

    Nowadays the molten salt reactor (MSR) concept seems to revive as one of the most promising systems for the realization of transmutation. In the molten salt reactors and subcritical systems the fuel and material to be transmuted circulate dissolved in some molten salt. The main advantage of this reactor type is the possibility of the continuous feed and reprocessing of the fuel. In the present paper a novel molten salt reactor concept is introduced and its transmutation capabilities are studied. The goal is the development of a transmutation technique along with a device implementing it, which yield higher transmutation efficiencies than that of the known procedures and thus results in radioactive waste whose load on the environment is reduced both in magnitude and time length. The procedure is the multi-step time-scheduled transmutation, in which transformation is done in several consecutive steps of different neutron flux and spectrum. In the new MSR concept, named 'multi-region' MSR (MRMSR), the primary circuit is made up of a few separate loops, in which salt-fuel mixtures of different compositions are circulated. The loop sections constituting the core region are only neutronically and thermally coupled. This new concept makes possible the utilization of the spatial dependence of spectrum as well as the advantageous features of liquid fuel such as the possibility of continuous chemical processing etc. In order to compare a 'conventional' MSR and a proposed MRMSR in terms of efficiency, preliminary calculational results are shown. Further calculations in order to find the optimal implementation of this new concept and to emphasize its other advantageous features are going on. (authors)

  2. Safeguards operations in the integral fast reactor fuel cycle

    International Nuclear Information System (INIS)

    Goff, K.M.; Benedict, R.W.; Brumbach, S.B.; Dickerman, C.E.; Tompot, R.W.

    1994-01-01

    Argonne National Laboratory is currently demonstrating the fuel cycle for the Integral Fast Reactor (IFR), an advanced reactor concept that takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel-cycle economics, environmental protection, and safeguards. The IFR fuel cycle employs a pyrometallurgical process using molten salts and liquid metals to recover actinides from spent fuel. The safeguards aspects of the fuel cycle demonstration must be approved by the United States Department of Energy, but a further goal of the program is to develop a safeguards system that could gain acceptance from the Nuclear Regulatory Commission and International Atomic Energy Agency. This fuel cycle is described with emphasis on aspects that differ from aqueous reprocessing and on its improved safeguardability due to decreased attractiveness and diversion potential of all process streams, including the fuel product

  3. A scaled experimental study of control blade insertion dynamics in Pebble-Bed Fluoride-Salt-Cooled High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Buster, Grant C., E-mail: grant.buster@gmail.com; Laufer, Michael R.; Peterson, Per F.

    2016-07-15

    Highlights: • A granular dynamics scaling methodology is discussed. • Control blade insertion in a representative pebble-bed core is experimentally studied. • Control blade insertion forces and pebble displacements are experimentally measured. • X-ray tomography techniques are used to observe pebble displacement distributions. - Abstract: Direct control element insertion into a pebble-bed reactor core is proposed as a viable control system in molten-salt-cooled pebble-bed reactors. Unlike helium-cooled pebble-bed reactors, this reactor type uses spherical fuel elements with near-neutral buoyancy in the molten-salt coolant, thus reducing contact forces on the fuel elements. This study uses the X-ray Pebble Bed Recirculation Experiment facility to measure the force required to insert a control element directly into a scaled pebble-bed. The required control element insertion force, and therefore the contact force on fuel elements, is measured to be well below recommended limits. Additionally, X-ray tomography is used to observe how the direct insertion of a control element physically displaces spherical fuel elements. The tomography results further support the viability of direct control element insertion into molten-salt-cooled pebble-bed reactor cores.

  4. Pre-Conceptual Design of a Fluoride-Salt-Cooled Small Modular Advanced High Temperature Reactor (SmAHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Greene, Sherrell R [ORNL; Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Carbajo, Juan J [ORNL; Ilas, Dan [ORNL; Cisneros, Anselmo T [ORNL; Varma, Venugopal Koikal [ORNL; Corwin, William R [ORNL; Wilson, Dane F [ORNL; Yoder Jr, Graydon L [ORNL; Qualls, A L [ORNL; Peretz, Fred J [ORNL; Flanagan, George F [ORNL; Clayton, Dwight A [ORNL; Bradley, Eric Craig [ORNL; Bell, Gary L [ORNL; Hunn, John D [ORNL; Pappano, Peter J [ORNL; Cetiner, Sacit M [ORNL

    2011-02-01

    This document presents the results of a study conducted at Oak Ridge National Laboratory during 2010 to explore the feasibility of small modular fluoride salt-cooled high temperature reactors (FHRs). A preliminary reactor system concept, SmATHR (for Small modular Advanced High Temperature Reactor) is described, along with an integrated high-temperature thermal energy storage or salt vault system. The SmAHTR is a 125 MWt, integral primary, liquid salt cooled, coated particle-graphite fueled, low-pressure system operating at 700 C. The system employs passive decay heat removal and two-out-of-three , 50% capacity, subsystem redundancy for critical functions. The reactor vessel is sufficiently small to be transportable on standard commercial tractor-trailer transport vehicles. Initial transient analyses indicated the transition from normal reactor operations to passive decay heat removal is accomplished in a manner that preserves robust safety margins at all times during the transient. Numerous trade studies and trade-space considerations are discussed, along with the resultant initial system concept. The current concept is not optimized. Work remains to more completely define the overall system with particular emphasis on refining the final fuel/core configuration, salt vault configuration, and integrated system dynamics and safety behavior.

  5. Sustainability of thorium-uranium in pebble-bed fluoride salt-cooled high temperature reactor

    International Nuclear Information System (INIS)

    Zhu, G.; Zou, Y.; Xu, H.

    2016-01-01

    Sustainability of thorium fuel in a Pebble-Bed Fluoride salt-cooled High temperature Reactor (PBFHR) is investigated to find the feasible region of high discharge burnup and negative Flibe (2LiF-BeF_2) salt Temperature Reactivity Coefficient (TRC). Dispersion fuel or pellet fuel with SiC cladding and SiC matrix is used to replace the tri-structural-isotropic (TRISO) coated particle system for increasing fuel loading and decreasing excessive moderation. To analyze the neutronic characteristics, an equilibrium calculation method of thorium fuel self-sustainability is developed. We have compared two refueling schemes (mixing flow pattern and directional flow pattern) and two kinds of reflector materials (SiC and graphite). This method found that the feasible region of breeding and negative Flibe TRC is between 20 vol% and 62 vol% fuel loading in the fuel. A discharge burnup could be achieved up to about 200 MWd/kgHM. The case with directional flow pattern and SiC reflector showed superior burnup characteristics but the worst radial power peak factor, while the case with mixing flow pattern and SiC reflector, which was the best tradeoff between discharge burnup and radial power peak factor, could provide burnup of 140 MWd/kgHM and about 1.4 radial power peak factor with 50 vol% dispersion fuel. In addition, Flibe salt displays good neutron properties as a coolant of quasi-fast reactors due to the strong "9Be(n,2n) reaction and low neutron absorption of "6Li (even at 1000 ppm) in fast spectrum. Preliminary thermal hydraulic calculation shows a good safety margin. The greatest challenge of this reactor may be the decades irradiation time of the pebble fuel. (A.C)

  6. The source term and waste optimization of molten salt reactors with processing

    International Nuclear Information System (INIS)

    Gat, U.; Dodds, H.L.

    1993-01-01

    The source term of a molten salt reactor (MSR) with fuel processing is reduced by the ratio of processing time to refueling time as compared to solid fuel reactors. The reduction, which can be one to two orders of magnitude, is due to removal of the long-lived fission products. The waste from MSRs can be optimized with respect to its chemical composition, concentration, mixture, shape, and size. The actinides and long-lived isotopes can be separated out and returned to the reactor for transmutation. These features make MSRs more acceptable and simpler in operation and handling

  7. Thermodynamics of soluble fission products cesium and iodine in the Molten Salt Reactor

    NARCIS (Netherlands)

    Capelli, E.; Beneš, O.; Konings, R.J.M.

    2018-01-01

    The present study describes the full thermodynamic assessment of the Li,Cs,Th//F,I system. The existing database for the relevant fluoride salts considered as fuel for the Molten Salt Reactor (MSR) has been extended with two key fission products, cesium and iodine. A complete evaluation of all

  8. Identification and evaluation of alternatives for the disposition of fluoride fuel and flush salts from the molten salt reactor experiment at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1996-01-01

    This document presents an initial identification and evaluation of the alternatives for disposition of the fluoride fuel and flush salts stored in the drain tanks at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). It will serve as a resource for the U.S. Department of Energy contractor preparing the feasibility study for this activity under the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA). This document will also facilitate further discussion on the range of credible alternatives, and the relative merits of alternatives, throughout the time that a final alternative is selected under the CERCLA process

  9. Reactor chemical considerations of the accelerator molten-salt breeders

    International Nuclear Information System (INIS)

    Furukawa, Kazuo; Kato, Yoshio; Ohno, Hideo; Ohmichi, Toshihiko

    1982-01-01

    A single phase of the molten fluoride mixture is simultaneously functionable as a nuclear reaction medium, a heat medium and a chemical processing medium. Applying this characteristics of molten salts, the single-fluid type accelerator molten-salt breeder (AMSB) concept was proposed, in which 7 LiF-BeF 2 -ThF 4 was served as a target-and-blanket salt (Fig. 1 and Table 1), and the detailed discussion on the chemical aspects of AMSB are presented (Tables 2 -- 4 and Fig.2). Owing to the small total amount of radiowaste and the low concentrations of each element in target salt, AMSB would be chemically managable. The performance of the standard-type AMSB is improved by adding 0.3 -- 0.8 m/o 233 UF 4 as follows(Tables 1 and 4, and Figs. 2 and 3): (a) this ''high-gain'' type AMSB is feasible to design chemically, in which still only small amount of radiowaste is included ; (b) the fissile material production rate will be increased significantly; (c) this target salt is straightly fed as an 233 U additive to the fuel of molten-salt converter reactor (MSCR) ; (d) the dirty fuel salt suctioned from MSCR is batch-reprocessed in the safeguarded regional center, in which many AMSB are facilitated ; (e) the isolated 233 UF 4 is blended in the target salt sent to many MSCRs, and the cleaned residual fertile salt is used as a diluent of AMSB salt ; (f) this simple and rational thorium fuel breeding cycle system is also suitable for the nuclear nonproliferation and for the fabrication of smaller size power-stations. (author)

  10. Neutronic study of a nuclear reactor of fused salts; Estudio neutronico de un reactor nuclear de sales fundidas

    Energy Technology Data Exchange (ETDEWEB)

    Garcia B, F. B.; Francois L, J. L., E-mail: faviolabelen@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    The reactors of fused salts called Molten Salt Reactor have presented a resurgence of interest in the last decade, due to they have a versatility in particular to operate, either with a thermal or fast neutrons spectrum. The most active development was by the middle of 1950 and principles of 1970 in the Oak Ridge National Laboratory. In this work some developed models are presented particularly and studied with the help of the MCNPX code, for the development of the neutronic study of this reactor, starting of proposed models and from a simple and homogeneous geometry until other more complex models and approximate to more real cases. In particular the geometry conditions and criticality of each model were analyzed, the isotopic balance, as well as the concentrations of the salts and different assigned fuel types. (Author)

  11. Graphite and carbonaceous materials in a molten salt nuclear reactor

    International Nuclear Information System (INIS)

    Rousseau, Ginette; Lecocq, Alfred; Hery, Michel.

    1982-09-01

    A project for a molten salt 1000 MWe reactor is studied by EDF-CEA teams. The design provides for a chromesco 3 vessel housing graphite structures in which the salt circulates. The salt (Th, U, Be and Li fluorides) is cooled by direct contact with lead. The graphites and carbonated materials, inert with respect to lead and the fuel salt, are being considered not only as moderators, but as reflectors and in the construction of the sections where the heat exchange takes place. On the basis of the problems raised in the operation of the reactor, a study programme on French experimental materials (Le Carbone Lorraine, SERS, SEP) has been defined. Hence, depending on the function or functions that the material is to ensure in the structure, the criteria of choice which follow will have to be examined: behaviour under irradiation, insertion of a fluid in the material, thermal properties required, mechanical properties required, utilization [fr

  12. Analysis of the transmutational characteristics of a novel molten salt reactor concept

    International Nuclear Information System (INIS)

    Csom, Gy.; Feher, S.; Szieberth, M.

    2001-01-01

    One of the arguments most frequently brought up by the opponents of the utilization of nuclear energy is the requirement that the radioactive waste and the long-lived radioisotopes accumulated in the spent fuel should be isolated for a very long time from the biosphere. The solution is the elimination of long-lived actinides (plutonium isotopes and minor actinides) and long-lived fission products by transforming (transmuting) them into short-lived or stable nuclei. The high neutron flux required for transmutation can be realized in nuclear installations. these may be conventional therma; and fast reactors, furthermore dedicated devices, namely thermal and fast reactors and accelerator driven subcritical systems (ADSs), which are specifically designed for this purpose. Some of the most promising systems are the molten salt reactors and subcritical systems, in which the fuel and material to be transmuted circulate dissolved in some molten salt. In the present paper this transmutational device, as well as recommendations for the improvement are discussed in detail (Authors)

  13. Simulation tools and new developments of the molten salt fast reactor

    International Nuclear Information System (INIS)

    Merle-Lucotte, E.; Doligez, X.; Heuer, D.; Allibert, M.; Ghetta, V.

    2010-01-01

    Starting from the Molten Salt Breeder Reactor project of Oak-Ridge, we have performed parametric studies in terms of safety coefficients, reprocessing requirements and breeding capabilities. In the frame of this major re-evaluation of the molten salt reactor (MSR), we have developed a new concept called Molten Salt Fast Reactor or MSFR, based on the Thorium fuel cycle and a fast neutron spectrum. This concept has been selected for further studies by the MSR steering committee of the Generation IV International Forum in 2009. Our reactor's studies of the MSFR concept rely on numerical simulations making use of the MCNP neutron transport code coupled with a code for materials evolution which resolves the Bateman's equations giving the population of each nucleus inside each part of the reactor at each moment. Because of MSR's fundamental characteristics compared to classical solid-fuelled reactors, the classical Bateman equations have to be modified by adding two terms representing the reprocessing capacities and the fertile or fissile alimentation. We have thus coupled neutronic and reprocessing simulation codes in a numerical tool used to calculate the extraction efficiencies of fission products, their location in the whole system (reactor and reprocessing unit) and radioprotection issues. (authors)

  14. Molten Salt Fuel Cycle Requirements for ADTT Applications

    International Nuclear Information System (INIS)

    Del Cul, G.D.; Toth, L.M.; Williams, D.F.

    1999-01-01

    The operation of an ADT system with the associated nuclear reactions has a profound effect upon the chemistry of the fuel - especially with regards to container compatibility and the chemical separations that may be required. The container can be protected by maintaining the redox chemistry within a relatively narrow, non-corrosive window. Neutron economy as well as other factors require a sophisticated regime of fission product separations. Neither of these control requirements has been demonstrated on the scale or degree of sophistication necessary to support an ADT device. We review the present situation with respect to fluoride salts, and focus on the critical issues in these areas which must be addressed. One requirement for advancement in this area - a supply of suitable materials - will soon be fulfilled by the remediation of ORNLs Molten Salt Reactor Experiment, and the removal of a total of 11,000 kg of enriched (Li-7 > 99.9%) coolant, flush, and fuel salts

  15. Transient analyses for a molten salt fast reactor with optimized core geometry

    Energy Technology Data Exchange (ETDEWEB)

    Li, R., E-mail: rui.li@kit.edu [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Wang, S.; Rineiski, A.; Zhang, D. [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Merle-Lucotte, E. [Laboratoire de Physique Subatomique et de Cosmologie – IN2P3 – CNRS/Grenoble INP/UJF, 53, rue des Martyrs, 38026 Grenoble (France)

    2015-10-15

    Highlights: • MSFR core is analyzed by fully coupling neutronics and thermal-hydraulics codes. • We investigated four types of transients intensively with the optimized core geometry. • It demonstrates MSFR has a high safety potential. - Abstract: Molten salt reactors (MSRs) have encountered a marked resurgence of interest over the past decades, highlighted by their inclusion as one of the six candidate reactors of the Generation IV advanced nuclear power systems. The present work is carried out in the framework of the European FP-7 project EVOL (Evaluation and Viability Of Liquid fuel fast reactor system). One of the project tasks is to report on safety analyses: calculations of reactor transients using various numerical codes for the molten salt fast reactor (MSFR) under different boundary conditions, assumptions, and for different selected scenarios. Based on the original reference core geometry, an optimized geometry was proposed by Rouch et al. (2014. Ann. Nucl. Energy 64, 449) on thermal-hydraulic design aspects to avoid a recirculation zone near the blanket which accumulates heat and very high temperature exceeding the salt boiling point. Using both fully neutronics thermal-hydraulic coupled codes (SIMMER and COUPLE), we also re-confirm the efforts step by step toward a core geometry without the recirculation zone in particular as concerns the modifications of the core geometrical shape. Different transients namely Unprotected Loss of Heat Sink (ULOHS), Unprotected Loss of Flow (ULOF), Unprotected Transient Over Power (UTOP), Fuel Salt Over Cooling (FSOC) are intensively investigated and discussed with the optimized core geometry. It is demonstrated that due to inherent negative feedbacks, an MSFR plant has a high safety potential.

  16. Proposed Guidance for Preparing and Reviewing Molten Salt Nonpower Reactor Licence Applications (NUREG-1537)

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [ORNL; Flanagan, George F. [ORNL; Voth, Marcus [Boston Government Services, LLC

    2018-05-01

    Development of non-power molten salt reactor (MSR) test facilities is under consideration to support the analyses needed for development of a full-scale MSR. These non-power MSR test facilities will require review by the US Nuclear Regulatory Commission (NRC) staff. This report proposes chapter adaptations for NUREG-1537 in the form of interim staff guidance to address preparation and review of molten salt non-power reactor license applications. The proposed adaptations are based on a previous regulatory gap analysis of select chapters from NUREG-1537 for their applicability to non-power MSRs operating with a homogeneous fuel salt mixture.

  17. Evaluation of Fluorine-Trapping Agents for Use During Storage of the MSRE Fuel Salt

    Energy Technology Data Exchange (ETDEWEB)

    Brynestad, J.; Williams, D.F.

    1999-05-01

    A fundamental characteristic of the room temperature Molten Salt Reactor Experiment (MSRE) fuel is that the radiation from the retained fission products and actinides interacts with this fluoride salt to produce fluorine gas. The purpose of this investigation was to identify fluorine-trapping materials for the MSRE fuel salt that can meet both the requirement of interim storage in a sealed (gastight) container and the vented condition required for disposal at the Waste Isolation Pilot Plant (WIPP). Sealed containers will be needed for interim storage because of the large radon source that remains even in fuel salt stripped of its uranium content. An experimental program was undertaken to identify the most promising candidates for efficient trapping of the radiolytic fluorine generated by the MSRE fuel salt. Because of the desire to avoid pressurizing the closed storage containers, an agent that traps fluorine without the generation of gaseous products was sought.

  18. Evaluation of Fluorine-Trapping Agents for Use During Storage of the MSRE Fuel Salt

    International Nuclear Information System (INIS)

    Brynestad, J.; Williams, D.F.

    1999-01-01

    A fundamental characteristic of the room temperature Molten Salt Reactor Experiment (MSRE) fuel is that the radiation from the retained fission products and actinides interacts with this fluoride salt to produce fluorine gas. The purpose of this investigation was to identify fluorine-trapping materials for the MSRE fuel salt that can meet both the requirement of interim storage in a sealed (gastight) container and the vented condition required for disposal at the Waste Isolation Pilot Plant (WIPP). Sealed containers will be needed for interim storage because of the large radon source that remains even in fuel salt stripped of its uranium content. An experimental program was undertaken to identify the most promising candidates for efficient trapping of the radiolytic fluorine generated by the MSRE fuel salt. Because of the desire to avoid pressurizing the closed storage containers, an agent that traps fluorine without the generation of gaseous products was sought

  19. Radionuclide release from research reactor spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Curtius, H., E-mail: h.curtius@fz-juelich.de [Forschungszentrum Juelich, Institut fuer Energieforschung, IEF-6 Sicherheitsforschung und Reaktortechnik, Geb. 05.3, D-52425 Juelich (Germany); Kaiser, G.; Mueller, E.; Bosbach, D. [Forschungszentrum Juelich, Institut fuer Energieforschung, IEF-6 Sicherheitsforschung und Reaktortechnik, Geb. 05.3, D-52425 Juelich (Germany)

    2011-09-01

    Numerous investigations with respect to LWR fuel under non oxidizing repository relevant conditions were performed. The results obtained indicate slow corrosion rates for the UO{sub 2} fuel matrix. Special fuel-types (mostly dispersed fuels, high enriched in {sup 235}U, cladded with aluminium) are used in German research reactors, whereas in German nuclear power plants, UO{sub 2}-fuel (LWR fuel, enrichment in {sup 235}U up to 5%, zircaloy as cladding) is used. Irradiated research reactor fuels contribute less than 1% to the total waste volume. In Germany, the state is responsible for fuel operation and for fuel back-end options. The institute for energy research (IEF-6) at the Research Center Juelich performs investigation with irradiated research reactor spent fuels under repository relevant conditions. In the study, the corrosion of research reactor spent fuel has been investigated in MgCl{sub 2}-rich salt brine and the radionuclide release fractions have been determined. Leaching experiments in brine with two different research reactor fuel-types were performed in a hot cell facility in order to determine the corrosion behaviour and the radionuclide release fractions. The corrosion of two dispersed research reactor fuel-types (UAl{sub x}-Al and U{sub 3}Si{sub 2}-Al) was studied in 400 mL MgCl{sub 2}-rich salt brine in the presence of Fe{sup 2+} under static and initially anoxic conditions. Within these experimental parameters, both fuel types corroded in the experimental time period of 3.5 years completely, and secondary alteration phases were formed. After complete corrosion of the used research reactor fuel samples, the inventories of Cs and Sr were quantitatively detected in solution. Solution concentrations of Am and Eu were lower than the solubility of Am(OH){sub 3}(s) and Eu(OH){sub 3}(s) solid phases respectively, and may be controlled by sorption processes. Pu concentrations may be controlled by Pu(IV) polymer species, but the presence of Pu(V) and Pu

  20. Molten Salt Reactor in the Overview and Perspective of Technological Assessment

    International Nuclear Information System (INIS)

    Julia Abdul Karim; Khaironie Md Takip; Muhammad Khairul Arif Mustafa; Mohd Hairie Rabir; Lanyau, T.; Tom, P.P.

    2016-01-01

    Full text: A Molten Salt Reactor (MSR) is unique in its characteristics that offer safer operation, deliver efficient power output that can assure in the sustainable energy production without CO_2 emissions. Several concepts of this kind of reactor have been proposed by stake holder with different design and configuration and up to date they are exasperating to obtain an optimum workable solution to the fuel salt composition in the foresee of neutronic properties, operating temperature, actinide and fission products solubility, chemical control and processing, materials compatibility and handling of waste. Hence, these key issues are wide open as the potential Research and Development in the specific areas of studies. In addition to that, concern arise in the viewpoint of socioeconomic, politics, public acceptance, safety and security, proven technology, proliferation resistance and physical protection that also need to give special attention in problem solving. The worldwide collaboration through Gen IV International Forum has discussed the potential of MSR and addresses on the issues globally. Recently, Malaysia has taken an initiative aiming to participate in MSR studies due to its potential as an energy source using thorium. Therefore, this paper is focusing on the technology assessment for Thorium-breeding Molten Salt Reactor (TMSR) especially on the ability of utilizing thorium as fuel. This assessment also will help to enhance the understanding of thorium beneficiation to cater for the energy demand. (author)

  1. Study of trans-uranian incineration in molten salt reactor

    International Nuclear Information System (INIS)

    Valade, M.

    2000-01-01

    With the emergence of new options for nuclear power, molten salt reactors are envisaged for waste management. The aim of this thesis is to show how molten salt reactors can help to solve the transuranics issue. Their high versatility regarding to isotopic vector allows to accommodate large fractions of minor actinides as compared to solid fuel system. In this thesis, a neutronics study of molten salt reactors, MSR, has been conducted. For this purpose, two reference systems were considered, TIER1 and AMSTER. In the case of TIER1, an optimisation was made to reach an equilibrium. The analysis of both systems showed the main characteristics of MSR: their link to chemistry and on line reprocessing. In this work, several methods to drive the system to a state of equilibrium have been implemented and compared. During this process the isotopic composition and neutron spectrum, thus the nuclear reaction cross sections, vary tremendously. It is essential to take these evolutions into account in order to accurately estimate the equilibrium state. This has been accomplished inside the multi-recycling procedure we set with ERANOS. A dedicated calculation schema has been realized to simulate superthermal systems with this computation code. These results were checked through a benchmark against other computer codes. Then, with multi-recycling method, several molten salt systems have been compared in order to define the optimal reactor for transuranics incineration. Nevertheless, a final choice can not only be done using only neutronics characteristics since chemistry and thermal-hydraulics constraints are really important for MSR. Moreover, a complete safety study would be required. (author)

  2. Conception of electron beam-driven subcritical molten salt ultimate safety reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abalin, S.S.; Alekseev, P.N.; Ignat`ev, V.V. [Kurchatov Institute, Moscow (Russian Federation)] [and others

    1995-10-01

    This paper is a preliminary sketch of a conception to develop the {open_quotes}ultimate safety reactor{close_quotes} using modern reactor and accelerator technologies. This approach would not require a long-range R&D program. The ultimate safety reactor could produce heat and electric energy, expand the production of fuel, or be used for the transmutation of long-lived wastes. The use of the combined double molten salt reactor system allows adequate neutron multiplication to permit using an electron accelerator for the initial neutron flux. The general parameters of such a system are discussed in this paper.

  3. Molten Salt Breeder Reactor Analysis Based on Unit Cell Model

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yongjin; Choi, Sooyoung; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-05-15

    Contemporary computer codes like the MCNP6 or SCALE are only good for solving a fixed solid fuel reactor. However, due to the molten-salt fuel, MSR analysis needs some functions such as online reprocessing and refueling, and circulating fuel. J. J. Power of Oak Ridge National Laboratory (ORNL) suggested in 2013 a method for simulating the Molten Salt Breeder Reactor (MSBR) with SCALE, which does not support continuous material processing. In order to simulate MSR characteristics, the method proposes dividing a depletion time into short time intervals and batchwise reprocessing and refueling at each step. We are applying this method by using the MCNP6 and PYTHON and NEWT-TRITON-PYTHON and PYTHON code systems to MSBR. This paper contains various parameters to analyze the MSBR unit cell model such as the multiplication factor, breeding ratio, change of amount of fuel, amount of fuel feeding, and neutron flux distribution. The result of MCNP6 and NEWT module in SCALE show some difference in depletion analysis, but it still seems that they can be used to analyze MSBR. Using these two computer code system, it is possible to analyze various parameters for the MSBR unit cells such as the multiplication factor, breeding ratio, amount of material, total feeding, and neutron flux distribution. Furthermore, the two code systems will be able to be used for analyzing other MSR model or whole core models of MSR.

  4. Molten Salt Breeder Reactor Analysis Based on Unit Cell Model

    International Nuclear Information System (INIS)

    Jeong, Yongjin; Choi, Sooyoung; Lee, Deokjung

    2014-01-01

    Contemporary computer codes like the MCNP6 or SCALE are only good for solving a fixed solid fuel reactor. However, due to the molten-salt fuel, MSR analysis needs some functions such as online reprocessing and refueling, and circulating fuel. J. J. Power of Oak Ridge National Laboratory (ORNL) suggested in 2013 a method for simulating the Molten Salt Breeder Reactor (MSBR) with SCALE, which does not support continuous material processing. In order to simulate MSR characteristics, the method proposes dividing a depletion time into short time intervals and batchwise reprocessing and refueling at each step. We are applying this method by using the MCNP6 and PYTHON and NEWT-TRITON-PYTHON and PYTHON code systems to MSBR. This paper contains various parameters to analyze the MSBR unit cell model such as the multiplication factor, breeding ratio, change of amount of fuel, amount of fuel feeding, and neutron flux distribution. The result of MCNP6 and NEWT module in SCALE show some difference in depletion analysis, but it still seems that they can be used to analyze MSBR. Using these two computer code system, it is possible to analyze various parameters for the MSBR unit cells such as the multiplication factor, breeding ratio, amount of material, total feeding, and neutron flux distribution. Furthermore, the two code systems will be able to be used for analyzing other MSR model or whole core models of MSR

  5. Sustainability of thorium-uranium in pebble-bed fluoride salt-cooled High Temperature Reactor - 15171

    International Nuclear Information System (INIS)

    Zhu, G.; Zou, Y.; Xu, Hongjie

    2015-01-01

    Sustainability of thorium fuel in a pebble-bed fluoride salt-cooled high temperature reactor (PB-FHR) is investigated to find the feasible region of high discharge burnup and negative FLiBe (2LiF-BeF 2 ) salt temperature reactivity coefficient (TRC). Dispersion fuel or pellet fuel with SiC cladding and SiC matrix is used to replace the tri-structural-isotropic (TRISO) coated particle system for increasing heavy metal loading and decreasing excessive moderation. In order to analyze the neutronic characteristics, an equilibrium calculation method of thorium fuel self-sustainability is developed. We have compared 2 refueling schemes (mixing flow pattern and directional flow pattern) and 2 kinds of reflector materials (SiC and graphite). This method has found that the feasible regions of breeding and negative FLiBe TRC is between 20 vol% and 62 vol% heavy metal loading in the fuel. A discharge burnup could be achieved up to about 200 MWd/kgHM. The case with directional flow pattern and SiC reflector showed superior burnup characteristics but the worst radial power peak factor, while the case with mixing flow pattern and SiC reflector, which was the best tradeoff between discharge burnup and radial power peak factor, could provide burnup of 140 MWd/kgHM and about 1.4 radial power peak factor with 50 vol% dispersion fuel. In addition, FLiBe salt displays good neutron properties as a coolant of quasi-fast reactors due to the strong 9 Be(n,2n) reaction and low neutron absorption of 6 Li (even at 1000 ppm) in fast spectrum. Preliminary thermal hydraulic calculation shows good safety margins. The greatest challenge of this reactor may be the very long irradiation time of the pebble fuel. (authors)

  6. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    Science.gov (United States)

    Cisneros, Anselmo Tomas, Jr.

    The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP

  7. Molten salt reactor related research in Switzerland

    International Nuclear Information System (INIS)

    Krepel, Jiri; Hombourger, Boris; Fiorina, Carlo

    2015-01-01

    Switzerland represented by the Paul Scherrer Institute (PSI) is a member of the Generation IV International Forum (GIF). In the past, the research at PSI focused mainly on HTR, SFR, and GFR. Currently, a research program was established also for Molten Salt Reactors (MSR). Safety is the key point and main interest of the MSR research at the Nuclear Energy and Safety (NES) department of PSI. However, it cannot be evaluated without knowing the system design, fuel chemistry, salt thermal-hydraulics features, safety and fuel cycle approach, and the relevant material and chemical limits. Accordingly, sufficient knowledge should be acquired in the other individual fields before the safety can be evaluated. The MSR research at NES may be divided into four working packages (WP): WP1: MSR core design and fuel cycle, WP2: MSR fuel behavior at nominal and accidental conditions, WP3: MSR thermal-hydraulics and decay heat removal system, WP4: MSR safety, fuel stream, and relevant limits. The WPs are proposed so that there are research topics which can be independently studied within each of them. The work plan of the four WPs is based on several ongoing or past national and international projects relevant to MSR, where NES/PSI participates. At the current stage, the program focuses on several specific and design independent studies. The safety is the key point and main long-term interest of the MSR research at NES. (author)

  8. Molten salt fueled nuclear facility with steam-and gas turbine cycles of heat transformation

    International Nuclear Information System (INIS)

    Ananich, P.I.; Bunin, E.N.; Kazazyan, V.T.; Nemtsev, V.A.; Sikorin, S.N.

    2001-01-01

    The molten salt fueled nuclear facilities with fuel circulating in the primary circuit have a series of the potential advantages in comparison with the traditional thermal and fast reactors with solid fuel elements. These advantages are ensured by the possibility to receive effective neutron balance in the core, minimum margin reactivity, more deep fuel burnup, unbroken correctness of the fuel physical and chemical properties and by low prices of the fuel cycle. The neutron and thermal-physical calculations of the various variants of the MSFNF with steam-water and gas turbine power circuits and their technical and economical comparison are carried out in this article. Calculations of molten salt nuclear power plant with gas turbine power circuit have been carried out using chemically reacting working body ''nitrin'' (N304 + 1%NO). The molten salt fueled reactors with the thermal power near of 2300 MW with two fuel compositions have been considered. The base variant has been taken the design of NPP with VVER NP-1000 when comparing the results of the calculations. Its economical performances are presented in prices of 1990. The results of the calculations show that it is difficult to determine the advantages of any one of the variants considered in a unique fashion. But NPP with MSR possesses large reserves in the process of optimization of cycle and energy equipment parameters that can improve its technical and economical performances sufficiently. (author)

  9. Simulation tool of the on-line reprocessing unit of a molten salt reactor

    International Nuclear Information System (INIS)

    Simon, Nicole; Conocar, Olivier; Boussier, Hubert; Gastaldi, Olivier; Penit, Thomas; Walle, Eric; Huguet, Anne

    2006-01-01

    Molten salt reactor (MSR) is an interesting technology selected in the frame of the Generation IV forum. In the MSR, actinides are diluted in a molten salt which is both the fuel and the coolant. The ability of such a reactor is the reducing of the long-lived wastes due to high burn-up and the on-site simplified reprocessing directly connected with the core which preserve the salt properties necessary for its correct operation. Here is defined a flexible computer reprocessing code which can use data from neutronic calculations and can be coupled to a neutronic code. The code allow the description the whole behaviour of MSR, including, a coupled manner, both the design of the core and the optimised reprocessing scheme effects. (authors)

  10. Development status and potential program for development of proliferation-resistant molten-salt reactors

    International Nuclear Information System (INIS)

    Engel, J.R.; Bauman, H.F.; Dearing, J.F.; Grimes, W.R.; McCoy, H.E. Jr.

    1979-03-01

    Preliminary studies of existing and conceptual molten-salt reactor (MSR) designs have led to the identification of conceptual systems that are technologically attractive when operated with denatured uranium as the principal fissile fuel. These denatured MSRs would also have favorable resource-utilization characteristics and substantial resistance to proliferation of weapons-usable nuclear materials. The report presents a summary of the current status of technology and a discussion of the major technical areas of a possible base program to develop commercial denatured MSRs. The general areas treated are (1) reactor design and development, (2) safety and safety related technology, (3) fuel-coolant behavior and fuel processing, and (4) reactor materials. A substantial development effort could lead to authorization for construction of a molten-salt test reactor about 5 years after the start of the program and operation of the unit about 10 years later. A prototype commercial denatured MSR could be expected to begin operating 25 years from the start of the program. The postulated base program would extend over 32 years and would cost about $700 million (1978 dollars, unescalated). Additional costs to construct the MSTR, $600 million, and the prototype commercial plant, $1470 million, would bring the total program cost to about $2.8 billion. Additional allowances probably should be made to cover contingencies and incidental technology areas not explicitly treated in this preliminary review

  11. Development status and potential program for development of proliferation-resistant molten-salt reactors

    Energy Technology Data Exchange (ETDEWEB)

    Engel, J.R.; Bauman, H.F.; Dearing, J.F.; Grimes, W.R.; McCoy, H.E. Jr.

    1979-03-01

    Preliminary studies of existing and conceptual molten-salt reactor (MSR) designs have led to the identification of conceptual systems that are technologically attractive when operated with denatured uranium as the principal fissile fuel. These denatured MSRs would also have favorable resource-utilization characteristics and substantial resistance to proliferation of weapons-usable nuclear materials. The report presents a summary of the current status of technology and a discussion of the major technical areas of a possible base program to develop commercial denatured MSRs. The general areas treated are (1) reactor design and development, (2) safety and safety related technology, (3) fuel-coolant behavior and fuel processing, and (4) reactor materials. A substantial development effort could lead to authorization for construction of a molten-salt test reactor about 5 years after the start of the program and operation of the unit about 10 years later. A prototype commercial denatured MSR could be expected to begin operating 25 years from the start of the program. The postulated base program would extend over 32 years and would cost about $700 million (1978 dollars, unescalated). Additional costs to construct the MSTR, $600 million, and the prototype commercial plant, $1470 million, would bring the total program cost to about $2.8 billion. Additional allowances probably should be made to cover contingencies and incidental technology areas not explicitly treated in this preliminary review.

  12. Assessment of the dry process fuel sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok

    2004-04-01

    The feasibility of using dry-processed oxide fuel in a Sodium-cooled Fast Reactor (SFR) was analyzed for the equilibrium fuel cycle of two reference cores: Hybrid BN-600 benchmark core with a enlarged lattice pitch and modified BN-600 core. The dry process technology assumed in this study based on the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was {approx}50% and most of the fission products were removed.

  13. Assessment of the dry process fuel sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Choi, Hang Bok

    2004-04-01

    The feasibility of using dry-processed oxide fuel in a Sodium-cooled Fast Reactor (SFR) was analyzed for the equilibrium fuel cycle of two reference cores: Hybrid BN-600 benchmark core with a enlarged lattice pitch and modified BN-600 core. The dry process technology assumed in this study based on the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was ∼50% and most of the fission products were removed

  14. Neutronic analysis of two-fluid thorium molten salt reactor

    International Nuclear Information System (INIS)

    Frybort, Jan; Vocka, Radim

    2009-01-01

    The aim of this paper is to evaluate features of the two-fluid MSBR through a parametric study and compare its properties to one-fluid MSBR concepts. The starting point of the analysis is the original ORNL 1000 MWe reactor design, although simplified to some extent. We studied the influence of dimensions of distinct reactor parts - fuel and fertile channels radius, plenum height, design etc. - on fundamental reactor properties: breeding ratio and doubling time, reactor inventory, graphite lifetime, and temperature feedback coefficients. The calculations were carried out using MCNP5 code. Based on obtained results we proposed an improved reactor design. Our results show clear advantages of the concept with two separate fluoride salts if compared to the one fluid concept in breading, doubling time, and temperature feedback coefficients. Limitations of the two-fluid concept - particularly the graphite lifetime - are also pointed out. The reactor design can be a subject of further optimizations, namely from the viewpoint of reactor safety. (author)

  15. Thorium-based Molten Salt Reactor (TMSR) project in China

    International Nuclear Information System (INIS)

    Dai, Zhimin; Liu, Wei

    2013-01-01

    Making great efforts in development of nuclear energy is one of the long-term-plan in China's energy strategies. The advantages of Thorium-based nuclear energy are: rich resource in nature, less nuclear waste, low toxicity, nuclear non-proliferation and so on. Furthermore, China is a country with abundant thorium, thus it is necessary to develop the Thorium-based Molten Salt Reactor (TMSR) in China. Shanghai Institute of Applied Physics, Chinese Academy of Sciences (SINAP) had designed and constructed the first China's light-water reactor and developed a zero-power thorium-based molten salt reactor successfully in the early 1970s. The applied research project 'thorium molten salt reactor nuclear power system' by SINAP together with several other institutes had been accepted and granted by China government in 2011. The whole project has been divided into three stages: Firstly, built a 2 MW-zero-power high temperature solid molten salt reactor in 2015 and a 2 MW-zero-power high temperature liquid molten salt reactor in 2017. Secondly, in 2020 built a 10 MW high temperature liquid molten salt reactor. Thirdly, on the base of previous work, a 100 MW high temperature molten salt reactor should be achieving in 2030. After more than one years of efforts, a high quality scientific research team has been formed, which is able to design the molten salt reactor, the molten salt loop and related key equipment, the systems of molten salt preparation, purification and the radioactive gas removal. In the past one year, the initial physical design of high temperature molten salt reactor has been completed; the nuclear chemistry and radiation chemical laboratory has been built, a high temperature salt (HTS) loop and radioactive gas removal experiment device system have been successfully developed and constructed. Further, the preliminary study on reactor used carbon-carbon composite material has been investigated. (author)

  16. Improvements and validation of the transient analysis code MOREL for molten salt reactors

    International Nuclear Information System (INIS)

    Zhuang Kun; Zheng Youqi; Cao Liangzhi; Hu Tianliang; Wu Hongchun

    2017-01-01

    The liquid fuel salt used in the molten salt reactors (MSRs) serves as the fuel and coolant simultaneously. On the one hand, the delayed neutron precursors circulate in the whole primary loop and part of them decay outside the core. On the other hand, the fission heat is carried off directly by the fuel flow. These two features require new analysis method with the coupling of fluid flow, heat transfer and neutronics. In this paper, the recent update of MOREL code is presented. The update includes: (1) the improved quasi-static method for the kinetics equation with convection term is developed. (2) The multi-channel thermal hydraulic model is developed based on the geometric feature of MSR. (3) The Variational Nodal Method is used to solve the neutron diffusion equation instead of the original analytic basis functions expansion nodal method. The update brings significant improvement on the efficiency of MOREL code. And, the capability of MOREL code is extended for the real core simulation with feedback. The numerical results and experiment data gained from molten salt reactor experiment (MSRE) are used to verify and validate the updated MOREL code. The results agree well with the experimental data, which prove the new development of MOREL code is correct and effective. (author)

  17. Inertia-confining thermonuclear molten salt reactors

    International Nuclear Information System (INIS)

    Furukawa, Kazuo; Yamanaka, Chiyoe; Nakai, Sadao; Imon, Shunji; Nakajima, Hidenori; Nakamura, Norio; Kato, Yoshio.

    1984-01-01

    Purpose: To increase the heat generating efficiency while improving the reactor safety and thereby maintaining the energy balance throughout the reactor. Constitution: In an inertia-confining type D-T thermonuclear reactor, the blanket is made of lithium-containing fluoride molten salts (LiF.BeF 2 , LiF.NaF.KF, LiF.KF, etc) which are cascaded downwardly in a large thickness (50 - 100 cm) along the inner wall of the thermonuclear reaction vessel, and neutrons generated by explosive compression are absorbed to lithium in the molten salts to produce tritium, Heat transportation is carried out by the molten salts. (Ikeda, J.)

  18. Transient Analyses for a Molten Salt Transmutation Reactor Using the Extended SIMMER-III Code

    International Nuclear Information System (INIS)

    Wang, Shisheng; Rineiski, Andrei; Maschek, Werner; Ignatiev, Victor

    2006-01-01

    Recent developments extending the capabilities of the SIMMER-III code for the dealing with transient and accidents in Molten Salt Reactors (MSRs) are presented. These extensions refer to the movable precursor modeling within the space-time dependent neutronics framework of SIMMER-III, to the molten salt flow modeling, and to new equations of state for various salts. An important new SIMMER-III feature is that the space-time distribution of the various precursor families with different decay constants can be computed and took into account in neutron/reactivity balance calculations and, if necessary, visualized. The system is coded and tested for a molten salt transmuter. This new feature is also of interest in core disruptive accidents of fast reactors when the core melts and the molten fuel is redistributed. (authors)

  19. Improved inherent safety in liquid fuel reactors

    International Nuclear Information System (INIS)

    Taube, M.

    1982-01-01

    The molten salt reactor system divided into core (thermal and fast) and breeding zone (fission breeder reactor, fusion hybrid system, accelerator-spallation system) has some unique inherent safety properties: a) reduced inventory of fission products during normal operation due to on-line chemical reprocessing and in-core gas purging; b) fast removal of freshly bred fissile nuclides and fission products from the breeding zone (the so called suppressed fission system); c) pressureless fuel and primary coolant system; d) elimination of the possibility of a violent exoenergetic chemical reaction with air, water or metals; e) elimination of the possibility of gaseous hydrogen production during an accident; f) provides a non-engineered feature of dumping of fuel from the core and heat exchanger to a safe drain tank; g) presence of a large heat sink in the form of an inactive diluent salt; h) possibility of natural convection heat removal during an accident and even normal operation (by means of gas lifting); i) dissipation of the remaining decayheat by spraying water on the containment from outside, which allows to manage the worst accident; i) Even in the case of the destruction in the war by conventional or nuclear weapon the contaminated land is significantly reduced. The world-wide present activity in the field of molten salt technology is reviewed. (orig.)

  20. Actinide recycle potential in the Integral Fast Reactor (IFR) fuel cycle

    International Nuclear Information System (INIS)

    Chang, Y.I.; Till, C.E.

    1990-01-01

    In the Integral Fast Reactor (IFR) development program, the entire reactor system -- reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. The use of metallic fuel in the IFR allows a radically improved fuel cycle technology. Pyroprocessing, which utilizes high temperatures and molten salt and molten metal solvents, can be advantageously utilized for processing metal fuels because the product is metal suitable for fabrication into new fuel elements. The key step in the IFR process is electrorefining, which provides for recovery of the valuable fuel constituents, uranium and plutonium, and for removal of fission products. In the electrorefining operation, uranium and plutonium are selectively transported from an anode to a cathode, leaving impurity elements, mainly fission products, either in the anode compartment or in a molten salt electrolyte. A notable feature of the IFR process is that the actinide elements accompany plutonium through the process. This results in a major advantage in the high-level waste management, because these actinides are automatically recycled back into the reactor for in-situ burning. Based on the recent IFR process development, a preliminary assessment has also been made to investigate the feasibility of further adapting the pyrochemical processes to directly extract actinides from LWR spent fuel. The results of this assessment indicate very promising potential and two most promising flowsheet options have been identified for further research and development. This paper also summarizes current thinking on the rationale for actinide recycle, its ramifications on the geologic repository and the current high-level waste management plans, and the necessary development programs. 5 refs., 4 figs., 4 tabs

  1. The integral fast reactor fuels reprocessing laboratory at Argonne National Laboratory, Illinois

    International Nuclear Information System (INIS)

    Wolson, R.D.; Tomczuk, Z.; Fischer, D.F.; Slawecki, M.A.; Miller, W.E.

    1986-09-01

    The processing of Integral Fast Reactor (IFR) metal fuel utilizes pyrochemical fuel reprocessing steps. These steps include separation of the fission products from uranium and plutonium by electrorefining in a fused salt, subsequent concentration of uranium and plutonium for reuse, removal, concentration, and packaging of the waste material. Approximately two years ago a facility became operational at Argonne National Laboratory-Illinois to establish the chemical feasibility of proposed reprocessing and consolidation processes. Sensitivity of the pyroprocessing melts to air oxidation necessitated operation in atmosphere-controlled enclosures. The Integral Fast Reactor Fuels Reprocessing Laboratory is described

  2. Treatment of waste salt from the advanced spent fuel conditioning process (I): characterization of Zeolite A in Molten LiCl Salt

    International Nuclear Information System (INIS)

    Kim, Jeong Guk; Lee, Jae Hee; Yoo, Jae Hyung; Kim, Joon Hyung

    2004-01-01

    The oxide fuel reduction process based on the electrochemical method (Advanced spent fuel Conditioning Process; ACP) and the long-lived radioactive nuclides partitioning process based on electro-refining process, which are being developed ay the Korea Atomic Energy Research Institute (KAERI), are to generate two types of molten salt wastes such as LiCl salt and LiCl-KCl eutectic salt, respectively. These waste salts must meet some criteria for disposal. A conditioning process for LiCl salt waste from ACP has been developed using zeolite A. This treatment process of waste salt using zeolite A was first developed by US ANL (Argonne National Laboratory) for LiCl-KCl eutectic salt waste from an electro-refining process of EBR (Experimental Breeder Reactor)-II spent fuel. This process has been developed recently, and a ceramic waste form (CWF) is produced in demonstration-scale V-mixer (50 kg/batch). However, ANL process is different from KAERI treatment process in waste salt, the former is LiCl-KCl eutectic salt and the latter is LiCl salt. Because of melting point, the immobilization of eutectic salt is carried out at about 770 K, whereas LiCl salt at around 920 K. Such difference has an effect on properties of immobilization media, zeolite A. Here, zeolite A in high-temperature (923 K) molten LiCl salt was characterized by XRD, Ion-exchange, etc., and evaluated if a promising media or not

  3. Integrating the fuel cycle at IFR [Integral Fast Reactor

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.

    1992-01-01

    During the past few years Argonne National Laboratory has been developing the Integral Fast Reactor (IFR), an advanced liquid metal reactor. Much of the IFR technology stems from Argonne National Laboratory's experience with the Experimental Breeder Reactors, EBR 1 and 2. The unique aspect of EBR 2 is its success with high-burnup metallic fuel. Irradiation tests of the new U-Pu-Zr fuel for the IFR have now reached a burnup level of 20%. The results to date have demonstrated excellent performance characteristics of the metallic fuel in both steady-state and off-normal operating conditions. EBR 2 is now fully loaded with the IFR fuel alloys and fuel performance data are being generated. In turn, metallic fuel becomes the key factor in achieving a high degree of passive safety in the IFR. These characteristics were demonstrated dramatically by two landmark tests conducted at EBR 2 in 1986: loss of flow without scram; and loss of heat sink without scram. They demonstrated that the combination of high heat conductivity of metallic fuel and thermal inertia of the large sodium pool can shut the reactor down during potentially severe accidents without depending on human intervention or the operation of active engineered components. The IFR metallic fuel is also the key factor in compact pyroprocessing. Pyroprocessing uses high temperatures, molten salt and metal solvents to process metal fuels. The result is suitable for fabrication into new fuel elements. Feasibility studies are to be conducted into the recycling of actinides from light water reactor spent fuel in the IFR using the pyroprocessing approach to extract the actinides (author)

  4. Neutronics Phenomena Important in Modeling and Simulation of Liquid-Fuel Molten Salt Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, David J.

    2018-11-11

    This paper discusses liquid-fuel molten salt reactors, how they will operate under normal, transient, and accident conditions, and the results of an expert elicitation to determine the corresponding neutronic phenomena important to understanding their behavior. Identifying these phenomena will enable the U.S. Nuclear Regulatory Commission (NRC) to develop or identify modeling functionalities and tools required to carry out confirmatory analyses that examine the validity and accuracy of applicants’ calculations and help determine the margin of safety in plant design. NRC frequently does an expert elicitation using a Phenomena Identification and Ranking Table (PIRT) to identify and evaluate the state of knowledge of important modeling phenomena. However, few details about the design of these reactors and the sequence of events during accidents are known, so the process used was considered a preliminary PIRT. A panel met to define phenomena that would need to be modeled and considered the impact/importance of each phenomenon with respect to specific figures-of-merit (FoMs) (e.g., power distribution, fluence, kinetics parameters and reactivity). Each FoM reflected a potential impact on radionuclide release or loss of a barrier to release. The panel considered what the path forward might be with respect to being able to model the phenomenon in a simulation code. Results are explained for both thermal and fast spectrum designs.

  5. Tests of prototype salt stripper system for IFR fuel cycle

    International Nuclear Information System (INIS)

    Carls, E.L.; Blaskovitz, R.J.; Johnson, T.R.; Ogata, T.

    1993-01-01

    One of the waste treatment steps for the on-site reprocessing of spent fuel from the Integral Fast Reactor fuel cycles is stripping of the electrolyte salt used in the electrorefining process. This involves the chemical reduction of the actinides and rare earth chlorides forming metals which then dissolve in a cadmium pool. To develop the equipment for this step, a prototype salt stripper system has been installed in an engineering scale argon-filled glovebox. Pumping trails were successful in transferring 90 kg of LiCl-KCl salt containing uranium and rare earth metal chlorides at 500 degree C from an electrorefiner to the stripper vessel at a pumping rate of about 5 L/min. The freeze seal solder connectors which were used to join sections of the pump and transfer line performed well. Stripping tests have commenced employing an inverted cup charging device to introduce a Cd-15 wt % Li alloy reductant to the stripper vessel

  6. The procedure and results of calculations of the equilibrium isotopic composition of a demonstration subcritical molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nevinitsa, V. A., E-mail: Neviniza-VA@nrcki.ru; Dudnikov, A. A.; Blandinskiy, V. Yu.; Balanin, A. L.; Alekseev, P. N. [National Research Centre Kurchatov Institute (Russian Federation); Titarenko, Yu. E.; Batyaev, V. F.; Pavlov, K. V.; Titarenko, A. Yu., E-mail: yuri.titarenko@itep.ru [Institute for Theoretical and Experimental Physics (Russian Federation)

    2015-12-15

    A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing {sup 233}U from {sup 232}Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.

  7. Fluoride Salt-Cooled High-Temperature Reactor Technology Development and Demonstration Roadmap

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Mays, Gary T [ORNL; Pointer, William David [ORNL; Robb, Kevin R [ORNL; Yoder Jr, Graydon L [ORNL

    2013-11-01

    Fluoride salt-cooled High-temperature Reactors (FHRs) are an emerging reactor class with potentially advantageous performance characteristics, and fully passive safety. This roadmap describes the principal remaining FHR technology challenges and the development path needed to address the challenges. This roadmap also provides an integrated overview of the current status of the broad set of technologies necessary to design, evaluate, license, construct, operate, and maintain FHRs. First-generation FHRs will not require any technology breakthroughs, but do require significant concept development, system integration, and technology maturation. FHRs are currently entering early phase engineering development. As such, this roadmap is not as technically detailed or specific as would be the case for a more mature reactor class. The higher cost of fuel and coolant, the lack of an approved licensing framework, the lack of qualified, salt-compatible structural materials, and the potential for tritium release into the environment are the most obvious issues that remain to be resolved.

  8. Simulated first operating campaign for the Integral Fast Reactor fuel cycle demonstration

    International Nuclear Information System (INIS)

    Goff, K.M.; Mariani, R.D.; Benedict, R.W.; Park, K.H.; Ackerman, J.P.

    1993-01-01

    This report discusses the Integral Fast Reactor (IFR) which is an innovative liquid-metal-cooled reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid-metal cooling to offer significant improvements in reactor safety, operation, fuel cycle-economics, environmental protection, and safeguards. Over the next few years, the IFR fuel cycle will be demonstrated at Argonne-West in Idaho. Spent fuel from the Experimental Breeder Reactor II (EBR-II) win be processed in its associated Fuel Cycle Facility (FCF) using a pyrochemical method that employs molten salts and liquid metals in an electrorefining operation. As part of the preparation for the fuel cycle demonstration, a computer code, PYRO, was developed at Argonne to model the electrorefining operation using thermodynamic and empirical data. This code has been used extensively to evaluate various operating strategies for the fuel cycle demonstration. The modeled results from the first operating campaign are presented. This campaign is capable of processing more than enough material to refuel completely the EBR-II core

  9. Design of a heterogeneous subcritical nuclear reactor with molten salts based on thorium

    International Nuclear Information System (INIS)

    Medina C, D.; Hernandez A, P.; Letechipia de L, C.; Vega C, H. R.; Sajo B, L.

    2015-09-01

    This paper presents the design of a heterogeneous subcritical nuclear reactor with molten salts based on thorium, with graphite moderator and a 252 Cf source, whose dose levels at the periphery allows its use in teaching and research activities. The design was realized by the Monte Carlo method, where the geometry, dimensions and the fuel was varied in order to obtain the best design. The result was a cubic reactor of 110 cm of side, with graphite moderator and reflector. In the central part having 9 ducts of 3 cm in diameter, eight of them are 110 cm long, which were placed on the Y axis; the separation between each duct is 10 cm. The central duct has 60 cm in length and this contains the 252 Cf source, also there are two irradiation channels and the other six contain a molten salt ( 7 LiF - BeF 2 - ThF 4 - UF 4 ) as fuel. For the design the k eff was calculated, neutron spectra and ambient dose equivalent. In the first instance the above was calculated for a virgin fuel, was called case 1; then a percentage of 233 U was used and the percentage of Th was decreased and was called case 2. This with the purpose of comparing two different fuels operating within the reactor. For the two irradiation ducts three positions are used: center, back and front, in each duct in order to have different flows. (Author)

  10. Feasibility study for fast reactor and related fuel cycle. Preliminary studies in 1998

    International Nuclear Information System (INIS)

    Hayafune, Hiroki; Enuma, Yasuhiro; Kubota, Kenichi; Yoshida, Masashi; Uno, Osamu; Ishikawa, Hiroyasu; Kobayashi, Jun; Umetsu, Youichiro; Ichimiya, Masakazu

    1999-10-01

    Prior to the feasibility study for fast reactors (FRs) starting from the 1999 fiscal year, planned in the medium and long-term program of JNC, preliminarily studies were performed on 'FR systems except sodium cooled MOX fueled reactors'. Small scale or module type reactors, heavy metal (Pb or Pb-Bi) cooled reactors, gas cooled reactors, light water cooled reactors, and molten salt reactors were studied on the basis of literature. They were evaluated from the viewpoint of the technical possibility (the structure integrity, earthquake resistance, safety, productivity, operability, maintenance repair, difficulty of the development), the long-term targets (market competitiveness as an energy system, utilization of uranium resources, reduction of radioactive waste, security of the non-proliferation), and developmental risk. As the result, the following concepts should be studied for future commercialized FRs. Small scale and module type reactor: Middle-sized reactor with an excellent economical efficiency. Small power reactor with a multipurpose design concept. Gas cooled reactor: CO2 gas cooled reactor, He gas cooled reactor. Heavy metal cooled reactor: Russian type lead cooled reactor. Light water cooled reactor: Light water cooled high converter reactor and super critical pressure light water cooled reactor. Molten salt reactor: Trichloride molten salt reactor which matches the U-Pu cycle. (author)

  11. Safe actinide disposition in molten salt reactors

    International Nuclear Information System (INIS)

    Gat, U.

    1997-01-01

    Safe molten salt reactors (MSR) can readily accommodate the burning of all fissile actinides. Only minor compromises associated with plutonium are required. The MSRs can dispose safely of actinides and long lived isotopes to result in safer and simpler waste. Disposing of actinides in MSRs does increase the source term of a safety optimized MSR. It is concluded that the burning and transmutation of actinides in MSRs can be done in a safe manner. Development is needed for the processing to handle and separate the actinides. Calculations are needed to establish the neutron economy and the fuel management. 9 refs

  12. Treatment of wastes in the Integral Fast Reactor (IFR) fuel cycle

    International Nuclear Information System (INIS)

    Ackerman, J.P.; Johnson, T.R.; Chow, L.S.H.; Carls, E.L.; Hannum, W.H.; Laidler, J.J.

    1997-01-01

    In both the reactor portion and the fuel-cycle portion of the Integral Fast Reactor (IFR), handling, treatment and disposal of wastes are simpler than in current fuel cycles. The vast majority (> 99.9%) of the very-long-lived radioactive TRU elements are not sent to the repository; rather, they are recycled. High-level waste volume from the IFR process (called ''the pyroprocess'') is lower than that from either the direct disposal of spent fuel or from conventional PUREX-type reprocessing. The quantity of low-level waste is very low. In the pyroprocess, the actinides are recovered and separated from the bulk of the fission products by an electrorefining step wherein the actinides are electrotransported from chopped fuel elements and deposited at cathodes. The volatile fission products xenon, krypton, and tritium are collected for long-term storage and decay. Zirconium and the ''noble metal'' fission products (those that are less easily oxidized than zirconium) remain in the anode compartment, to be removed with the fuel cladding fragments and made into a metal waste form. The remaining fission products collect in the salt as chlorides. A process has been developed to periodically remove the contaminated salt from the electrorefiner, separate most of the fission products, and return the purified salt in a form that is ready for continuing use. To clean up the electrorefiner salt, the fission products are removed by ion exchange onto a column of Zeolite A. After the purification step, the column material and the contained fission products are converted to a mineral waste form for disposal. The processes and equipment for waste isolation and conversion to suitable disposal forms are described in this paper. (author)

  13. Flow effect on {sup 135}I and {sup 135}Xe evolution behavior in a molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Jianhui; Guo, Chen [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); Cai, Xiangzhou [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Yu, Chenggang; Zou, Chunyan [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); Han, Jianlong [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Chen, Jingen, E-mail: chenjg@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China)

    2017-04-01

    Highlights: • {sup 135}Xe and {sup 135}I evolution law in a molten salt reactor is analytically deduced. • The circulation of fuel salt through the primary loop decreases the concentration of {sup 135}I and {sup 135}Xe. • {sup 135}I and {sup 135}Xe concentration reduction is independent with the mass flow rate at normal core operating condition. • Increasing the external core volume would raise {sup 135}I and {sup 135}Xe concentration reduction caused by the flow effect. - Abstract: Molten Salt Reactor (MSR) employs fissile material dissolved in the fluoride salt as fuel which continuously circulates through the primary loop with the flow cycle time being a few tens of seconds. The nuclei evolution law is quite different from that in a solid fuel reactor. In this paper, we analytically deduce the nuclei evolution law of {sup 135}Xe and {sup 135}I which are entrained in the flowing salt, evaluate its concentration changing with the burnup time, and validate the result with the SCALE6. The circulation of fuel salt could decrease the concentration of {sup 135}Xe and {sup 135}I, and the reduction can achieve to around 40% and 50% for {sup 135}Xe and {sup 135}I respectively at a small power level (e.g., 2 MW) when the core has the same fuel salt volume as that of the outer-loop. Furthermore, it can be found that the reduction is inversely proportional to the core to outer-loop volume ratio, but uncorrelated with the mass flow rate under normal operating condition of a MSR. At low core power scale, the flow effect on {sup 135}Xe concentration reduction is apparent, but it is mitigated as the core power scale increases because of the rise of {sup 135}I concentration, which raises its decay to {sup 135}Xe and compensates the loss of {sup 135}Xe due to decay at the outer-loop. The decreased {sup 135}Xe concentration results in a core reactivity increase varying from around 150 pcm to 1000 pcm depending on the core power and core to outer-loop volume ratio.

  14. An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

    2009-11-01

    This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

  15. Waste removal in pyrochemical fuel processing for the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Ackerman, J.P.; Johnson, T.R.; Laidler, J.J.

    1994-01-01

    Electrorefining in a molten salt electrolyte is used in the Integral Fast Reactor fuel cycle to recover actinides from spent fuel. Processes that are being developed for removing the waste constituents from the electrorefiner and incorporating them into the waste forms are described in this paper. During processing, halogen, chalcogen, alkali, alkaline earth, and rare earth fission products build up in the molten salt as metal halides and anions, and fuel cladding hulls and noble metal fission products remain as metals of various particle sizes. Essentially all transuranic actinides are collected as metals on cathodes, and are converted to new metal fuel. After processing, fission products and other waste are removed to a metal and a mineral waste form. The metal waste form contains the cladding hulls, noble metal fission products, and (optionally) most rare earths in a copper or stainless steel matrix. The mineral waste form contains fission products that have been removed from the salt into a zeolite or zeolite-derived matrix

  16. Generation IV reactors: reactor concepts

    International Nuclear Information System (INIS)

    Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

    2003-01-01

    Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

  17. Amster: a molten-salt reactor concept generating its own 233U and incinerating transuranium elements

    International Nuclear Information System (INIS)

    Lecarpentier, D.; Garzenne, C.; Vergnes, J.; Mouney, H.; Delpech, M.

    2002-01-01

    In the coming century, sustainable development of atomic energy will require the development of new types of reactors able to exceed the limits of the existing reactor types, be it in terms of optimum use of natural fuel resources, reduction in the production of long-lived radioactive waste, or economic competitiveness. Of the various candidates with the potential to meet these needs, molten-salt reactors are particularly attractive, in the light of the benefits they offer, arising from two fundamental features: - A liquid fuel does away with the constraints inherent in solid fuel, leading to a drastic simplification of the fuel cycle, in particular making in possible to carry out on-line pyrochemical reprocessing; - Thorium cycle and thermal spectrum breeding. The MSBR concept proposed by ORNL in the 1970's thus gave a breeding factor of 1.06, with a doubling time of about 25 years. However, given the tight neutron balance of the thorium cycle (the η of 233 U is about 2.3), MSBR performance is only possible if there are strict constraints set on the in-line reprocessing unit: all the 233 Pa must be removed from the core so that it can decay on the 233 U in no more than about ten days (or at least 15 tonnes of salt to be extracted from the core daily), and the absorbing fission products, in particular the rare earths, must be extracted in about fifty days. With the AMSTER MSR concept, which we initially developed for incinerating transuranium elements, we looked to reduce the mass of salt to be reprocessed in order to minimise the size and complexity of the reprocessing unit coupled to the reactor, and the quantity of transuranium elements sent for disposal, as this is directly proportional to the mass of salt reprocessed for extraction of the fission products. Given that breeding was not an absolute necessity, because the reactor can be started by incinerating the transuranium elements from the spent fuel assemblies of current reactors, or if necessary by loading

  18. Pyrochemical head-end treatment for fast reactor fuel elements

    International Nuclear Information System (INIS)

    Avogadro, A.

    1978-01-01

    The paper presents the R and D work performed at Ispra and Mol during the period 1965-1975 in order to find a way to overcome technical and economical difficulties arising when the conventional reprocessing is applied to fast reactor fuel elements. The work had been directed towards 3 specific topics: a) liquid-metal decladding of spent stainless steel - clad fuels (solinox process). b) oxidative pulverisation by fused salts and extraction of volatile fission products (satex process). c) Pyrochemical separation of plutonium from the bulk of the fuel

  19. Neutronics calculations for denatured molten salt reactors: Assessing resource requirements and proliferation-risk attributes

    International Nuclear Information System (INIS)

    Ahmad, Ali; McClamrock, Edward B.; Glaser, Alexander

    2015-01-01

    Highlights: • We study the proliferation-risk and resource attributes of denatured MSRs. • MSRs offer significantly better resource efficiency compared to light-water reactors. • Denatured single-fluid MSRs reactors offer promising non-proliferation attributes. - Abstract: Molten salt reactors (MSRs) are often advocated as a radical but worthwhile alternative to traditional reactor concepts based on solid fuels. This article builds upon the existing research into MSRs to model and simulate the operation of thorium-fueled single-fluid and two-fluid reactors. The analysis is based on neutronics calculations and focuses on denatured MSR systems. Resource utilization and basic proliferation-risk attributes are compared to those of standard light-water reactors. Depending on specific design choices, even fully denatured reactors could reduce uranium and enrichment requirements by a factor of 3–4. Overall, denatured single-fluid designs appear as the most promising candidate technology minimizing both design complexity and overall proliferation risks despite being somewhat less attractive from the perspective of resource utilization

  20. Preliminary Study for Inventories of Minor Actinides in Thorium Molten Salt Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Choong Wie; Kim, Hee Reyoung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2015-05-15

    It has different characteristic with the conventional reactors which use a solid fuel. It can continually supply the fuel by online refueling and reprocessing of minor actinides so that those can be separated and eliminated from the reactor. The MSR maintains steady state except initial stage and the reactor becomes stable. In this research, considering online refueling, bubbling and reprocessing, the basic concept for evaluation of the inventory of minor actinide in the molten salt reactor is driven using the Bateman equation. The simulation results, where REM and MCNP code from CNRS (Centre National de la Recherche Scientifique) applied to the concept equation are analyzed. The analysis of the basic concept was carried out for evaluation of the inventory of the minor actinides in MSR. It was thought that the inventories of the minor actinides should be evaluated by solving the modified Bateman equation due to the MSR characteristic of online refueling, chemical reprocessing and bubbling.

  1. Preliminary Study for Inventories of Minor Actinides in Thorium Molten Salt Reactor

    International Nuclear Information System (INIS)

    Lee, Choong Wie; Kim, Hee Reyoung

    2015-01-01

    It has different characteristic with the conventional reactors which use a solid fuel. It can continually supply the fuel by online refueling and reprocessing of minor actinides so that those can be separated and eliminated from the reactor. The MSR maintains steady state except initial stage and the reactor becomes stable. In this research, considering online refueling, bubbling and reprocessing, the basic concept for evaluation of the inventory of minor actinide in the molten salt reactor is driven using the Bateman equation. The simulation results, where REM and MCNP code from CNRS (Centre National de la Recherche Scientifique) applied to the concept equation are analyzed. The analysis of the basic concept was carried out for evaluation of the inventory of the minor actinides in MSR. It was thought that the inventories of the minor actinides should be evaluated by solving the modified Bateman equation due to the MSR characteristic of online refueling, chemical reprocessing and bubbling

  2. Thorium fuel-cycle development through plutonium incineration by THORIMS-NES (Thorium Molten-Salt nuclear energy synergetics)

    International Nuclear Information System (INIS)

    Furukawa, K.; Furuhashi, A.; Chigrinov, S.E.

    1996-01-01

    Thorium fuel-cycle has benefit on not-only trans-U element reduction but also their incineration. The disadvantage of high gamma activity of fuel, which is useful for improving the resistance to nuclear proliferation and terrorism, can overcome by molten fluorides fuel, and practically by THORIMS-NES, symbiotically coupled with fission Molten-Salt Reactor (FUJI) and fissile-producing Accelerator Molten-Salt Breeder (AMSB). This will have wide excellent advantages in global application, and will be deployed by incinerating Pu and Producing 233 U. Some details of this strategy including time schedule are presented. 14 refs, 2 figs, 4 tabs

  3. DESAIN KONSEP TANGKI PENAMPUNG BAHAN BAKAR PASSIVE COMPACT MOLTEN SALT REACTOR

    Directory of Open Access Journals (Sweden)

    A. Hadiwinata

    2015-04-01

    Full Text Available Passive Compact Molten Salt Reactor (PCMSR merupakan pengembangan dari reaktor MSR. Desain reaktor PCMSR membutuhkan tempat khusus penampung sementara bahan bakar pada saat terjadi insiden, misalnya kecelakaan yang menyebabkan peningkatan suhu bahan bakar. Tangki penampung bahan bakar tersusun dari 3 bagian yang saling terhubung yaitu bagian penampung cairan bahan bakar, cerobong (chimney, dan penukar kalor. Dalam penelitian ini, tangki dimodelkan secara lump dan dilakukan variasi daya awal reaktor dan ketinggian cerobong. Syarat batas model ditetapkan suhu bahan bakar maksimum 1400 °C, yang didasarkan pada titik didih larutan garam LiF-BeF2-ThF4-UF4. Analisis dilakukan dengan cara menghitung rugi tekanan total dan transfer kalor untuk variasi daya awal antara 1800-3000 MWth dan ketinggian cerobong antara 1-10 m. Hasil penelitian menunjukan semakin besar daya reaktor, maka tinggi tangki penampung bahan bakar dan tinggi alat penukar kalor yang dibutuhkan akan semakin besar, tejadi kenaikan suhu fluida pendingin dan suhu udara pendingin, dan menyebabkan kenaikan laju aliran masa fluida pendingin, sedangkan laju aliran masa udara menurun. Peningkatan ketinggian cerobong menyebabkan ketinggian tangki penampung bahan bakar dan ketinggian alat penukar kalor semakin menurun, penurunan suhu fluida pendingin, tetapi suhu udara meningkat, dan menyebabkan peningkatan laju aliran masa fluida pendingin, tetapi laju aliran masa udara akan semakin menurun. Kata kunci: PCMSR, cerobong, alat penukar kalor, variasi daya.   The Passsive Compact Molten Salat Reactor (PCMSR reactor is developed from MSR reactor. The PCMSR reactor design requires special place to temporarily storage for reactor fuel when incident occurs, such as when there is an accident which caused the temperature of the fuel increases. The tank consist of three interconnected parts, the reservoir liquid fuel, chimney, and the heat exchanger. In this research, the tank system is modeled based on

  4. An overview of radiolysis studies for the molten salt reactor remediation project

    International Nuclear Information System (INIS)

    Icenhour, A.S.; Williams, D.F.; Trowbridge, L.D.; Toth, L.M.; Del Cul, G.D.

    2001-01-01

    A number of radiolysis experiments have been performed in support of the remediation of the Molten Salt Reactor Experiment (MSRE)at the Oak Ridge National Laboratory.Materials studied included simulated MSRE fuel salt,fluorinated charcoal, NH 4 F,2NaFUF 6 ,UO 2 F 2 uranium oxides with a known residual fluoride content,and uranium oxides with a known moisture content.The results from these studies were used as part of the basis for the interim or long-term storage of materials removed from the MSRE. (author)

  5. Quantity and management of spent fuel from prototype and research reactors in Germany

    International Nuclear Information System (INIS)

    Dorr, Sabine; Bollingerfehr, Wilhelm; Filbert, Wolfgang; Tholen, Marion

    2013-01-01

    Within the scope of an R and D project (project identification number FKZ 02 S 8679) sponsored by BMBF (Federal Ministry of Education and Research), the current state of storage and management of fuel elements from prototype and research reactors was established, and an approach for their future storage/management was developed. The spent fuels from prototype and research reactors in Germany that require disposal were specified and were described in regard to their repository-relevant characteristics. As there are currently no casks licensed for disposal in Germany, descriptions of casks that were considered to be suitable were provided. Based on the information provided on the spent fuel from prototype and research reactors and the potential casks, a technical disposal concept was developed. In this context, concepts to integrate the spent fuel from prototype and research reactors into existing disposal concepts for spent fuel from German nuclear power plants and for waste from reprocessing were developed for salt and clay formations. (authors)

  6. Molten salt reactors: A new beginning for an old idea

    International Nuclear Information System (INIS)

    LeBlanc, David

    2010-01-01

    Molten salt reactors have seen a marked resurgence of interest over the past decade, highlighted by their inclusion as one of six Generation IV reactor types. The most active development period however was between the mid 1950s and early 1970s at Oak Ridge National Laboratories (ORNL) and any new re-examination of this concept must bear in mind the far different priorities then in place. High breeding ratios and short doubling times were paramount and this guided the evolution of the Molten Salt Breeder Reactor (MSBR) program. As the inherent advantages of the molten salt concept have become apparent to an increasing number of researchers worldwide it is important to not simply look to continue where ORNL left off but to return to basics in order to offer the best design using updated goals and abilities. A major potential change to the traditional Single Fluid, MSBR design and a subject of this presentation is a return to the mode of operation that ORNL proposed for the majority of its MSR program. That being the Two Fluid design in which separate salts are used for fissile 233 UF 4 and fertile ThF 4 . Oak Ridge abandoned this promising route due to what was known as the 'plumbing problem'. It will be shown that a simple yet crucial modification to core geometry can solve this problem and enable the many advantages of the Two Fluid design. In addition, another very promising route laid out by ORNL was simplified Single Fluid converter reactors that could obtain far superior lifetime uranium utilization than LWR or CANDU without the need for any fuel processing beyond simple chemistry control. Updates and potential improvements to this very attractive concept will also be explored.

  7. Dynamics and control of molten-salt breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sing, Vikram; Lish, Matthew R.; Chvala, Ondrej; Upadhyaya, Belle R. [Dept. of Nuclear Engineering, University of Tennessee, Knoxville (United States)

    2017-08-15

    Preliminary results of the dynamic analysis of a two-fluid molten-salt breeder reactor (MSBR) system are presented. Based on an earlier work on the preliminary dynamic model of the concept, the model presented here is nonlinear and has been revised to accurately reflect the design exemplified in ORNL-4528. A brief overview of the model followed by results from simulations performed to validate the model is presented. Simulations illustrate stable behavior of the reactor dynamics and temperature feedback effects to reactivity excursions. Stable and smooth changes at various nodal temperatures are also observed. Control strategies for molten-salt reactor operation are discussed, followed by an illustration of the open-loop load-following capability of the molten-salt breeder reactor system. It is observed that the molten-salt breeder reactor system exhibits “self-regulating” behavior, minimizing the need for external controller action for load-following maneuvers.

  8. Dynamics and control of molten-salt breeder reactor

    Directory of Open Access Journals (Sweden)

    Vikram Singh

    2017-08-01

    Full Text Available Preliminary results of the dynamic analysis of a two-fluid molten-salt breeder reactor (MSBR system are presented. Based on an earlier work on the preliminary dynamic model of the concept, the model presented here is nonlinear and has been revised to accurately reflect the design exemplified in ORNL-4528. A brief overview of the model followed by results from simulations performed to validate the model is presented. Simulations illustrate stable behavior of the reactor dynamics and temperature feedback effects to reactivity excursions. Stable and smooth changes at various nodal temperatures are also observed. Control strategies for molten-salt reactor operation are discussed, followed by an illustration of the open-loop load-following capability of the molten-salt breeder reactor system. It is observed that the molten-salt breeder reactor system exhibits “self-regulating” behavior, minimizing the need for external controller action for load-following maneuvers.

  9. Status of the Integral Fast Reactor fuel cycle demonstration and waste management practices

    International Nuclear Information System (INIS)

    Benedict, R.W.; Goff, K.M.; McFarlane, H.F.

    1994-01-01

    Over the past few years, Argonne National Laboratory has been preparing for the demonstration of the fuel cycle for the Integral Fast Reactor (IFR), an advanced reactor concept that takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety and operations, fuel-cycle economics, environmental protection, and safeguards. The IFR fuel cycle, which will be demonstrated at Argonne-West in Idaho, employs a pyrometallurgical process using molten salts and liquid metals to recover actinides from spent fuel. The required facility modifications and process equipment for the demonstration are nearing completion. Their status and the results from initial fuel fabrication work, including the waste management aspects, are presented. Additionally, estimated compositions of the various process waste streams have been made, and characterization and treatment methods are being developed. The status of advanced waste processing equipment being designed and fabricated is described

  10. Use of thorium in the generation IV Molten Salt reactors and perspectives for Brazil

    International Nuclear Information System (INIS)

    Seneda, Jose A.; Lainetti, Paulo E.O.

    2013-01-01

    Interest in thorium stems mainly from the fact that it is expected a substantial increase in uranium prices over the next fifty years. The reactors currently in operation consume 65,500 tons of uranium per year. Each electrical gigawatt (GWe) additional need about 200 tU mined per year. So advanced fuel cycles, which increase the reserves of nuclear materials are interesting, particularly the use of thorium to produce the fissile isotope 233 U. It is important to mention some thorium advantages. Thorium is three to five times more abundant than uranium in the earth's crust. Thorium has only one oxidation state. Additionally, thoria produces less radiotoxicity than the UO 2 because it produces fewer amounts of actinides, reducing the radiotoxicity of long life nuclear waste. ThO 2 has higher corrosion resistance than UO 2 , besides being chemically stable due to their low water solubility. The burning of Pu in a reactor based in thorium also decreases the inventories of Pu from the current fuel cycles, resulting in lower risks of material diversion for use in nuclear weapons. There are some ongoing projects in the world, taking into consideration the proposed goals for Generation IV reactors, namely: sustainability, economics, safety and reliability, proliferation resistance and physical protection. Some developments on the use of thorium in reactors are underway, with the support of the IAEA and some governs. Can be highlighted some reactor concepts using thorium as fuel: CANDU; ADTR -Accelerator Driven Thorium Reactor; AHWR -Advanced Heavy Water Reactor proposed by India (light water cooled and moderated by heavy water) and the MSR -Molten Salt Reactor. The latter is based on a reactor concept that has already been successfully tested in the U.S. in the 50s, for use in aircrafts. In this paper, we discuss the future importance of thorium, particularly for Brazil, which has large mineral reserves of this strategic element, the characteristics of the molten salt

  11. Use of thorium in the generation IV Molten Salt reactors and perspectives for Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Seneda, Jose A.; Lainetti, Paulo E.O., E-mail: jaseneda@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    Interest in thorium stems mainly from the fact that it is expected a substantial increase in uranium prices over the next fifty years. The reactors currently in operation consume 65,500 tons of uranium per year. Each electrical gigawatt (GWe) additional need about 200 tU mined per year. So advanced fuel cycles, which increase the reserves of nuclear materials are interesting, particularly the use of thorium to produce the fissile isotope {sup 233}U. It is important to mention some thorium advantages. Thorium is three to five times more abundant than uranium in the earth's crust. Thorium has only one oxidation state. Additionally, thoria produces less radiotoxicity than the UO{sub 2} because it produces fewer amounts of actinides, reducing the radiotoxicity of long life nuclear waste. ThO{sub 2} has higher corrosion resistance than UO{sub 2}, besides being chemically stable due to their low water solubility. The burning of Pu in a reactor based in thorium also decreases the inventories of Pu from the current fuel cycles, resulting in lower risks of material diversion for use in nuclear weapons. There are some ongoing projects in the world, taking into consideration the proposed goals for Generation IV reactors, namely: sustainability, economics, safety and reliability, proliferation resistance and physical protection. Some developments on the use of thorium in reactors are underway, with the support of the IAEA and some governs. Can be highlighted some reactor concepts using thorium as fuel: CANDU; ADTR -Accelerator Driven Thorium Reactor; AHWR -Advanced Heavy Water Reactor proposed by India (light water cooled and moderated by heavy water) and the MSR -Molten Salt Reactor. The latter is based on a reactor concept that has already been successfully tested in the U.S. in the 50s, for use in aircrafts. In this paper, we discuss the future importance of thorium, particularly for Brazil, which has large mineral reserves of this strategic element, the

  12. Engineering Evaluation of Proposed Alternative Salt Transfer Method for the Molten Salt Reactor Experiment for the Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    Carlberg, Jon A.; Roberts, Kenneth T.; Kollie, Thomas G.; Little, Leslie E.; Brady, Sherman D.

    2009-01-01

    This evaluation was performed by Pro2Serve in accordance with the Technical Specification for an Engineering Evaluation of the Proposed Alternative Salt Transfer Method for the Molten Salt Reactor Experiment at the Oak Ridge National Laboratory (BJC 2009b). The evaluators reviewed the Engineering Evaluation Work Plan for Molten Salt Reactor Experiment Residual Salt Removal, Oak Ridge National Laboratory, Oak Ridge, Tennessee (DOE 2008). The Work Plan (DOE 2008) involves installing a salt transfer probe and new drain line into the Fuel Drain Tanks and Fuel Flush Tank and connecting them to the new salt transfer line at the drain tank cell shield. The probe is to be inserted through the tank ball valve and the molten salt to the bottom of the tank. The tank would then be pressurized through the Reactive Gas Removal System to force the salt into the salt canisters. The Evaluation Team reviewed the work plan, interviewed site personnel, reviewed numerous documents on the Molten Salt Reactor (Sects. 7 and 8), and inspected the probes planned to be used for the transfer. Based on several concerns identified during this review, the team recommends not proceeding with the salt transfer via the proposed alternate salt transfer method. The major concerns identified during this evaluation are: (1) Structural integrity of the tanks - The main concern is with the corrosion that occurred during the fluorination phase of the uranium removal process. This may also apply to the salt transfer line for the Fuel Flush Tank. Corrosion Associated with Fluorination in the Oak Ridge National Laboratory Fluoride Volatility Process (Litman 1961) shows that this problem is significant. (2) Continued generation of Fluorine - Although the generation of Fluorine will be at a lower rate than experienced before the uranium removal, it will continue to be generated. This needs to be taken into consideration regardless of what actions are taken with the salt. (3) More than one phase of material

  13. On-line reprocessing of a molten salt reactor: a simulation tool

    International Nuclear Information System (INIS)

    Simon, Nicole; Gastaldi, Olivier; Penit, Thomas; Cohin, Olivier; Campion, Pierre-Yves

    2008-01-01

    The molten salt reactor (MSR) is one of the concepts studied in the frame of GEN IV road-map. Due to the specific features of its liquid fuel, the reprocessing unit may be directly connected to the reactor. A modelling of this unit is presented. The final objective is to create a flexible computer reprocessing code which can use data from neutron calculations and can be coupled to a neutron code. Such a code allows the description of the whole behaviour of MSR, including, in a coupled manner, both the design of the core and the optimised reprocessing scheme effects. (authors)

  14. Molten Salt Breeder Reactor Analysis Methods

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jinsu; Jeong, Yongjin; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2015-05-15

    Utilizing the uranium-thorium fuel cycle shows considerable potential for the possibility of MSR. The concept of MSBR should be revised because of molten salt reactor's advantage such as outstanding neutron economy, possibility of continuous online reprocessing and refueling, a high level of inherent safety, and economic benefit by keeping off the fuel fabrication process. For the development of MSR research, this paper provides the MSBR single-cell, two-cell and whole core model for computer code input, and several calculation results including depletion calculation of each models. The calculations are carried out by using MCNP6, a Monte Carlo computer code, which has CINDER90 for depletion calculation using ENDF-VII nuclear data. From the calculation results of various reactor design parameters, the temperature coefficients are all negative at the initial state and MTC becomes positive at the equilibrium state. From the results of core rod worth, the graphite control rod alone cannot makes the core subcritical at initial state. But the equilibrium state, the core can be made subcritical state only by graphite control rods. Through the comparison of the results of each models, the two-cell method can represent the MSBR core model more accurately with a little more computational resources than the single-cell method. Many of the thermal spectrum MSR have adopted a multi-region single-fluid strategy.

  15. Method for converting UF5 to UF4 in a molten fluoride salt

    International Nuclear Information System (INIS)

    Bennett, M.R.; Bamberge, C.E.; Kelmers, A.D.

    1980-01-01

    The subject relates to fuel preparation for molten salt breeder reactors, and more particularly to the reconstitution of spent molten fuel salt after fission product removal. During the course of reactor operation, fission products including rare earths and bred-in protactinium build up in the fuel salt and adversely affect the nuclear properties of the fuel. In order to more efficiently operate the reactor, the level of neutron poison fission products must be kept at a minimum. This is accomplished by continuously removing spent fuel from the primary circuit, processing it to remove fission products, and returning the reprocessed molten salt to the primary circuit. It is desirable for safety and economy that the fuel processing plant be a component of the reactor itself and that the salt be kept in the molten state throughout the processing system. (auth)

  16. MSR - SPHINX concept program Eros (Experimental zero power Salt reactor SR-0) - The proposed experimental program as a basis for validation of reactor physics methods

    Energy Technology Data Exchange (ETDEWEB)

    Hron, M.; Juricek, V.; Kyncl, J.; Mikisek, M.; Rypar, V. [Nuclear Research Institute Rez plc, Rez (Czech Republic)

    2007-07-01

    The Molten Salt Reactor (MSR) - SPHINX (SPent Hot fuel Incinerator by Neutron fluX) concept solves this principal problem of spent fuel treatment by means of so-called nuclear incineration. It means the burning of fissionable part of its inventory and transmutation of other problematic radionuclides by use of nuclear reactions with neutrons in a MSR-SPHINX system. This reactor system is an actinide burner (most in resonance neutron spectrum) and a radionuclide transmuter in a well-thermalized neutron spectrum. In the frame of the physical part, there are computational analyses and experimental activities. The experimental program has been focused, in its first stage, on a short-term irradiation of small size samples of molten-salt systems as well as structural materials proposed for the MSR blanket in the field of high neutron flux of research reactors. The proposed next stage of the program will focus on a large-scale experimental verification of design inputs by use of MSR-type inserting zones into the existing light water moderated experimental reactor LR-0, which may allow us to modify it into the experimental zero power salt reactor SR-0. There will be a detail description of the proposed program given in the paper together with the so far performed experiments and their first results. These realized experiments help us also to verify computational codes used, and to recognize some anomalies related to molten fluorides utilization. (authors)

  17. Interim Report on Fluid-Fuel Thermal Breeder Reactors (Revised)

    International Nuclear Information System (INIS)

    MacPherson, H. G.; Alexander, L. G.; Carter, W. L.; Chapman, R. H.; Kinyon, B. W.; Miller, J. W.

    1960-01-01

    The merits of aqueous-homogeneous ), graphite-moderated molten salt (MSBR) , and graphite-moderated liquid-bismuth (LBBR) breeder reactors operated at nearly comparable fuel-cycle costs (~1.5 mills/kwhr) were evaluated. The net electrical plant capability was assumed to be 1000 MwE, and the fuel and fertile streams were processed continuously on-site. The specific powers based on fuel were 1.2, 1.2, and 0.5 MwE/kg respectively, and 5.9, 3.7, and 5.3 MwE/tonne based on thorium. Net breeding ratios were 1.10, 1.07, and 1.07, giving doubling times of 5-1/2, 11, and 25 full power years . The fuel-cycle costs at the design points selected were 1.4, 1.3, and 1.6 mills/kwhr . The AHBR has an advantage in breeding ratio and doubling time because D 2 O is superior to graphite as a moderator in breeder reactors. MSBR has an advantage in fuel-cycle costs and in inventory of uranium in the fertile stream as a result of using a solution blanket.

  18. Electrometallurgical treatment of aluminum-matrix fuels

    International Nuclear Information System (INIS)

    Willit, J.L.; Gay, E.C.; Miller, W.E.; McPheeters, C.C.; Laidler, J.J.

    1996-01-01

    The electrometallurgical treatment process described in this paper builds on our experience in treating spent fuel from the Experimental Breeder Reactor (EBR-II). The work is also to some degree, a spin-off from applying electrometallurgical treatment to spent fuel from the Hanford single pass reactors (SPRs) and fuel and flush salt from the Molten Salt Reactor Experiment (MSRE) in treating EBR-II fuel, we recover the actinides from a uranium-zirconium fuel by electrorefining the uranium out of the chopped fuel. With SPR fuel, uranium is electrorefined out of the aluminum cladding. Both of these processes are conducted in a LiCl-KCl molten-salt electrolyte. In the case of the MSRE, which used a fluoride salt-based fuel, uranium in this salt is recovered through a series of electrochemical reductions. Recovering high-purity uranium from an aluminum-matrix fuel is more challenging than treating SPR or EBR-II fuel because the aluminum- matrix fuel is typically -90% (volume basis) aluminum

  19. Molten salt reactors: chemistry

    International Nuclear Information System (INIS)

    1983-01-01

    This work is a critical analysis of the 1000 MW MSBR project. Behavior of rare gases in the primary coolant circuit, their extraction from helium. Coating of graphite by molybdenum, chemistry of protactinium and niobium produced in the molten salt, continuous reprocessing of the fuel salt and use of stainless steel instead of hastelloy are reviewed [fr

  20. Assessment of Candidate Molten Salt Coolants for the Advanced High Temperature Reactor (AHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Williams, D.F.

    2006-03-24

    The Advanced High-Temperature Reactor (AHTR) is a novel reactor design that utilizes the graphite-matrix high-temperature fuel of helium-cooled reactors, but provides cooling with a high-temperature fluoride salt. For applications at temperatures greater than 900 C the AHTR is also referred to as a Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR). This report provides an assessment of candidate salts proposed as the primary coolant for the AHTR based upon a review of physical properties, nuclear properties, and chemical factors. The physical properties most relevant for coolant service were reviewed. Key chemical factors that influence material compatibility were also analyzed for the purpose of screening salt candidates. Some simple screening factors related to the nuclear properties of salts were also developed. The moderating ratio and neutron-absorption cross-section were compiled for each salt. The short-lived activation products, long-lived transmutation activity, and reactivity coefficients associated with various salt candidates were estimated using a computational model. Table A presents a summary of the properties of the candidate coolant salts. Certain factors in this table, such as melting point, vapor pressure, and nuclear properties, can be viewed as stand-alone parameters for screening candidates. Heat-transfer properties are considered as a group in Sect. 3 in order to evaluate the combined effects of various factors. In the course of this review, it became apparent that the state of the properties database was strong in some areas and weak in others. A qualitative map of the state of the database and predictive capabilities is given in Table B. It is apparent that the property of thermal conductivity has the greatest uncertainty and is the most difficult to measure. The database, with respect to heat capacity, can be improved with modern instruments and modest effort. In general, ''lighter'' (low-Z) salts tend to

  1. Molten salts and nuclear energy production

    International Nuclear Information System (INIS)

    Le Brun, Christian

    2007-01-01

    Molten salts (fluorides or chlorides) were considered near the beginning of research into nuclear energy production. This was initially due to their advantageous physical and chemical properties: good heat transfer capacity, radiation insensitivity, high boiling point, wide range solubility for actinides. In addition it was realised that molten salts could be used in numerous situations: high temperature heat transfer, core coolants with solid fuels, liquid fuel in a molten salt reactor, solvents for spent nuclear solid fuel in the case of pyro-reprocessing and coolant and tritium production in the case of fusion. Molten salt reactors, one of the six innovative concepts chosen by the Generation IV international forum, are particularly interesting for use as either waste incinerators or thorium cycle systems. As the neutron balance in the thorium cycle is very tight, the possibility to perform online extraction of some fission product poisons from the salt is very attractive. In this article the most important questions that must be addressed to demonstrate the feasibility of molten salt reactor will be reviewed

  2. Accelerator molten-salt breeder reactor

    International Nuclear Information System (INIS)

    Furukawa, Kazuo; Kuroi, Hideo; Kato, Yoshio; Oomichi, Toshihiko.

    1979-01-01

    Purpose: To obtain fission products and to transmute transuranium elements and other radioactive wastes by the use of Accelerator Molten-Salt Breeder Reactor. Constitution: Beams from an accelerator pipe at one end of a target vessel is injected through a window into target molten salts filled inside of the target vessel. The target molten salts are subjected to pump recycling or spontaneous convection while forcively cooled by blanket molten salts in an outer vessel. Then, energy is recovered from the blanket molten salts or the target molten salts at high temperatures through electric power generation or the like. Those salts containing such as thorium 232 and uranium 238 are used as the blanket molten salts so that fission products may be produced by neutrons generated in the target molten salts. PbCl 2 -PbF 2 and LiF-BeF 2 -ThF 4 can be used as the target molten salts and as the blanket molten salts respectively. (Seki, T.)

  3. Study of the moderating effect of salts on the sodium-water reaction on the cleaning of irradiated fuel assemblies from fast neutron reactors, using fluid sodium heat transfer

    International Nuclear Information System (INIS)

    Lacroix, Marie

    2014-01-01

    Within the framework of the development of generation IV reactors one of the research tracks is related to the development of fast neutron reactors using fluid sodium heat transfer. The CEA (French Alternative Energies and Atomic Energy Commission) plans to build a prototype of reactor of this type called 'ASTRID'. To address development requirements for this prototype, research is in progress on the reactor's availability and in particular on the reduction of the washing duration for residual sodium fuel assemblies during their discharge. In fact, because sodium is very reactive with water (presently the only available process), the washing is done, for example, by very gradual addition. A solution currently being studied at the CEA and which is the subject of this thesis report consists of the addition of an aqueous salts solutions to the washing water in order to slow down the kinetic reaction. This doctoral dissertation describes the various salts, which have been evaluated and aims to explain their action mode. (author) [fr

  4. Molten Salt Fuel Version of Laser Inertial Fusion Fission Energy (LIFE)

    International Nuclear Information System (INIS)

    Moir, R.W.; Shaw, H.F.; Caro, A.; Kaufman, L.; Latkowski, J.F.; Powers, J.; Turchi, P.A.

    2008-01-01

    Molten salt with dissolved uranium is being considered for the Laser Inertial Confinement Fusion Fission Energy (LIFE) fission blanket as a backup in case a solid-fuel version cannot meet the performance objectives, for example because of radiation damage of the solid materials. Molten salt is not damaged by radiation and therefore could likely achieve the desired high burnup (>99%) of heavy atoms of 238 U. A perceived disadvantage is the possibility that the circulating molten salt could lend itself to misuse (proliferation) by making separation of fissile material easier than for the solid-fuel case. The molten salt composition being considered is the eutectic mixture of 73 mol% LiF and 27 mol% UF 4 , whose melting point is 490 C. The use of 232 Th as a fuel is also being studied. ( 232 Th does not produce Pu under neutron irradiation.) The temperature of the molten salt would be ∼550 C at the inlet (60 C above the solidus temperature) and ∼650 C at the outlet. Mixtures of U and Th are being considered. To minimize corrosion of structural materials, the molten salt would also contain a small amount (∼1 mol%) of UF 3 . The same beryllium neutron multiplier could be used as in the solid fuel case; alternatively, a liquid lithium or liquid lead multiplier could be used. Insuring that the solubility of Pu 3+ in the melt is not exceeded is a design criterion. To mitigate corrosion of the steel, a refractory coating such as tungsten similar to the first wall facing the fusion source is suggested in the high-neutron-flux regions; and in low-neutron-flux regions, including the piping and heat exchangers, a nickel alloy, Hastelloy, would be used. These material choices parallel those made for the Molten Salt Reactor Experiment (MSRE) at ORNL. The nuclear performance is better than the solid fuel case. At the beginning of life, the tritium breeding ratio is unity and the plutonium plus 233 U production rate is ∼0.6 atoms per 14.1 MeV neutron

  5. Assessment of the Dry Processed Oxide Fuel in Liquid Metal Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok

    2005-09-15

    The neutronic feasibility of the dry process oxide fuel was assessed for the sodium-cooled and lead-cooled fast reactors (SFR and LFR, respectively), which were recommended as Generation-IV (Gen-IV) reactor systems by the Gen-IV international forum. The reactor analysis was performed for the equilibrium fuel cycle of two core configurations: Hybrid BN-600 benchmark core with an enlarged lattice pitch and a modified BN-600 core. The dry process technology assumed in this study is the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic (TRU) enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a fissile self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was {approx}50% and most of the fission products were removed. If the design criteria used in this study is proved to be acceptable through a detailed physics design and thermal hydraulic analysis in the future, it is practically possible to construct an equilibrium fuel cycle of the SFR and LFR systems based on the oxide fuel by utilizing the dry process technology.

  6. Assessment of the Dry Processed Oxide Fuel in Liquid Metal Fast Reactors

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Choi, Hang Bok

    2005-09-01

    The neutronic feasibility of the dry process oxide fuel was assessed for the sodium-cooled and lead-cooled fast reactors (SFR and LFR, respectively), which were recommended as Generation-IV (Gen-IV) reactor systems by the Gen-IV international forum. The reactor analysis was performed for the equilibrium fuel cycle of two core configurations: Hybrid BN-600 benchmark core with an enlarged lattice pitch and a modified BN-600 core. The dry process technology assumed in this study is the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic (TRU) enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a fissile self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was ∼50% and most of the fission products were removed. If the design criteria used in this study is proved to be acceptable through a detailed physics design and thermal hydraulic analysis in the future, it is practically possible to construct an equilibrium fuel cycle of the SFR and LFR systems based on the oxide fuel by utilizing the dry process technology

  7. Transport properties of molten-salt reactor fuel mixtures: the case of Na, Li, Be/F and Li, Be, Th/F salts

    International Nuclear Information System (INIS)

    Ignatiev, V.; Merzlyakov, A.; Afonichkin, V.; Khokhlov, V.; Salyulev, A.

    2003-01-01

    In this paper we have compiled transport properties information, available, on two types of FLiBe based salt mixtures (Na,Li,Be/F and Li,Be,Th/F) that are presently of importance in the design of innovative molten-salt burner reactors. Estimated and/or experimental values measured (particularly, from prior US and Russian studies, as well our recent studies) are given for the following properties: viscosity, thermal conductivity, phase transition behaviour, heat capacity, density and thermal expansion. (author)

  8. Transport properties of molten-salt reactor fuel mixtures: the case of Na, Li, Be/F and Li, Be, Th/F salts

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V; Merzlyakov, A [Kurchatov Institute - KI (Russian Federation); Afonichkin, V; Khokhlov, V; Salyulev, A [Institute of High Temperature Electrochemisty (IHTE), RF Yuri Golovatov, Konstantin Grebenkine, Vladimir Subbotin Institute of Technical Physics (VNIITF) (Russian Federation)

    2003-07-01

    In this paper we have compiled transport properties information, available, on two types of FLiBe based salt mixtures (Na,Li,Be/F and Li,Be,Th/F) that are presently of importance in the design of innovative molten-salt burner reactors. Estimated and/or experimental values measured (particularly, from prior US and Russian studies, as well our recent studies) are given for the following properties: viscosity, thermal conductivity, phase transition behaviour, heat capacity, density and thermal expansion. (author)

  9. Corrosion behavior of spent MTR fuel elements in a drowned salt mine repository

    International Nuclear Information System (INIS)

    Brodda, B.G.; Fachinger, J.

    1995-01-01

    Spent MTR fuel from German Material Test Reactors will not be reprocessed, but stored in a final salt repository in the deep geologic underground. Fuel elements will be placed in POLLUX containers, which are assumed to resist the corrosive attack of an accidentally formed concentrated salt brine for about 500 years. After a container failure the brine would contact the fuel element, corrode the aluminum plating and possibly leach radionuclides from the fuel. A source term for the calculation of radionuclide mobilization results from the investigation of the behavior of MTR fuel in this scenario, which has to be considered for the long-term safety analysis of a deep mined rock salt repository. Experiments with the different plating materials show that the considered aluminum alloys will not resist the corrosive attack of a brine solution, especially in the presence of iron, under the conditions in a drowned salt mine repository. Although differences in the corrosion rates of about two orders of magnitude were observed when applying different parameter sets, the deterioration must be considered to be almost instantaneous in geological terms. Radionuclides are mobilized from irradiated MTR fuel, when the meat of the fuel element becomes accessible to the brine solution. It seems, however, that the radionuclides are effectively trapped by the aluminum hydroxide formed, as the activity concentrations in the brine solution soon reach a constant level with the progressing corrosion of the cladding aluminum. In the presence of iron a more significant initial release was observed, but also in this case an equilibrium activity seems to be reached as a consequence of radionuclide trapping

  10. Evaluation of potential for MSRE spent fuel and flush salt storage and treatment at the INEL

    International Nuclear Information System (INIS)

    Ougouag, A.M.; Ostby, P.A.; Nebeker, R.L.

    1996-09-01

    The potential for interim storage as well as for treatment of the Molten Salt Reactor Experiment spent fuel at INEL has been evaluated. Provided that some minimal packaging and chemical stabilization prerequisites are satisfied, safe interim storage of the spent fuel at the INEL can be achieved in a number of existing or planned facilities. Treatment by calcination in the New Waste Calcining Facility at the INEL can also be a safe, effective, and economical alternative to treatment that would require the construction of a dedicated facility. If storage at the INEL is chosen for the Molten Salt Reactor Experiment (MSRE) spent fuel salts, their transformation to the more stable calcine solid would still be desirable as it would result in a lowering of risks. Treatment in the proposed INEL Remote-Handled Immobilization Facility (RHIF) would result in a waste form that would probably be acceptable for disposal at one of the proposed national repositories. The cost increment imputable to the treatment of the MSRE salts would be a small fraction of the overall capital and operating costs of the facility or the cost of building and operating a dedicated facility. Institutional and legal issues regarding shipments of fuel and waste to the INEL are summarized. The transfer of MSRE spent fuel for interim storage or treatment at the INEL is allowed under existing agreements between the State of idaho and the Department of energy and other agencies of the Federal Government. In contrast, current agreements preclude the transfer into Idaho of any radioactive wastes for storage or disposal within the State of Idaho. This implies that wastes and residues produced from treating the MSRE spent fuel at locations outside Idaho would not be acceptable for storage in Idaho. Present agreements require that all fuel and high-level wastes stored at the INEL, including MSRE spent fuel if received at the INEL, must be moved to a location outside Idaho by the year 2035

  11. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Greenwood, Michael Scott [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  12. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    International Nuclear Information System (INIS)

    Qualls, A. L.; Betzler, Benjamin R.; Brown, Nicholas R.; Carbajo, Juan; Greenwood, Michael Scott; Hale, Richard Edward; Harrison, Thomas J.; Powers, Jeffrey J.; Robb, Kevin R.; Terrell, Jerry W.

    2015-01-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  13. Cermet fuel reactors

    International Nuclear Information System (INIS)

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

    1987-09-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs

  14. A method for detecting the rupture of a fuel element in a fast neutron breeder reactor

    International Nuclear Information System (INIS)

    Cohen, Paul.

    1974-01-01

    The method according to the invention is characterized by the steps of balancing a cooling sodium sample driven through a nozzle into a molten salt constituted by a baryum-iodide strontium-iodide mixture, so that a substantial portion of radioactive iodine contingently present in the liquid sodium accumulates in the molten salt through isotopic exchange, separating the molten salt from sodium, balancing (if required) the molten salt with nonradioactive sodium and separating the molten salt from the sodium, and controlling the molten salt in order to determine the presence of iodine, such presence being-indicative of the rupture (or burst) of a fuel element sheath. Such a method is suitable in particular for detecting the rupture of a fuel element in a sodium-cooled fast breeder-reactor [fr

  15. Development of fluoride reprocessing technology for molten salt transmutation reactor systems in the Czech Republic

    International Nuclear Information System (INIS)

    Uhlir, J.; Hosnedl, P.; Matal, O.

    2000-01-01

    At present, the transmutation of spent nuclear fuel is considered a prospective alternative conception with respect to the current conception based on the non-reprocessed spent fuel disposal into a deep geological repository. The Czech research and development programme in the area of partitioning is directed primarily on the development of the fuel cycle technology for the accelerator - driven subcritical reactor with a liquid fuel based on fluoride melts. The final objective of the research programme is the development of pyrochemical technologies suitable for a continuous or semi-continuous separation process which would allow practically perfect utilization of the transmutation potentialities of the reactor system. The present research is directed particularly on the development of suitable fluoride separation methods the target of which is the removal of the uranium component from spent nuclear fuel and on the research of the electro-separation procedures and further on the development of appropriate construction materials and equipment for the technology of fluoride salt melts. (authors)

  16. Molten salts in nuclear reactors

    International Nuclear Information System (INIS)

    Dirian, J.; Saint-James

    1959-01-01

    Collection of references dealing with the physicochemical studies of fused salts, in particular the alkali and alkali earth halides. Numerous binary, ternary and quaternary systems of these halides with those of uranium and thorium are examined, and the physical properties, density, viscosity, vapour pressure etc... going from the halides to the mixtures are also considered. References relating to the corrosion of materials by these salts are included and the treatment of the salts with a view to recuperation after irradiation in a nuclear reactor is discussed. (author) [fr

  17. Molten salt/metal extractions for recovery of transuranic elements

    International Nuclear Information System (INIS)

    Chow, L.S.; Basco, J.K.; Ackerman, J.P.; Johnson, T.R.

    1992-01-01

    The integral fast reactor (EFR) is an advanced reactor concept that incorporates metallic driver and blanket fuels, an inherently safe, liquid-sodium-cooled, pool-type, reactor design, and on-site pyrochemical reprocessing (including electrorefining) of spent fuels and wastes. This paper describes a pyrochemical method that is being developed at Argonne National Laboratory to recover transuranic elements from the EFR electrorefiner process salt. The method uses multistage extractions between molten chloride salts and cadmium metal at high temperatures. The chemical basis of the salt extraction method, the test equipment, and a test plan are discussed

  18. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Sakurai, Shungo; Ogiya, Shunsuke.

    1990-01-01

    In a fuel assembly, if the entire fuels comprise mixed oxide fuels, reactivity change in cold temperature-power operation is increased to worsen the reactor shutdown margin. The reactor shutdown margin has been improved by increasing the burnable poison concentration thereby reducing the reactivity of the fuel assembly. However, since unburnt poisons are present at the completion of the reactor operation, the reactivity can not be utilized effectively to bring about economical disadvantage. In view of the above, the reactivity change between lower temperature-power operations is reduced by providing a non-boiling range with more than 9.1% of cross sectional area at the inside of a channel at the central portion of the fuel assembly. As a result, the amount of the unburnt burnable poisons is decreased, the economy of fuel assembly is improved and the reactor shutdown margin can be increase. (N.H.)

  19. Pyrometallurgical processing of Integral Fast Reactor metal fuels

    International Nuclear Information System (INIS)

    Battles, J.E.; Miller, W.E.; Gay, E.C.

    1991-01-01

    The pyrometallurgical process for recycling spent metal fuels from the Integral Fast Reactor is now in an advanced state of development. This process involves electrorefining spent fuel with a cadmium anode, solid and liquid cathodes, and a molten salt electrolyte (LiCl-KCl) at 500 degrees C. The initial process feasibility and flowsheet verification studies have been conducted in a laboratory-scale electrorefiner. Based on these studies, a dual cathode approach has been adopted, where uranium is recovered on a solid cathode mandrel and uranium-plutonium is recovered in a liquid cadmium cathode. Consolidation and purification (salt and cadmium removal) of uranium and uranium-plutonium products from the electrorefiner have been successful. The process is being developed with the aid of an engineering-scale electrorefiner, which has been successfully operated for more than three years. In this electrorefiner, uranium has been electrotransported from the cadmium anode to a solid cathode in 10 kg quantities. Also, anodic dissolution of 10 kg batches of chopped, simulated fuel (U--10% Zr) has been demonstrated. Development of the liquid cadmium cathode for recovering uranium-plutonium is under way

  20. The thermodynamics of pyrochemical processes for liquid metal reactor fuel cycles

    International Nuclear Information System (INIS)

    Johnson, I.

    1987-01-01

    The thermodynamic basis for pyrochemical processes for the recovery and purification of fuel for the liquid metal reactor fuel cycle is described. These processes involve the transport of the uranium and plutonium from one liquid alloy to another through a molten salt. The processes discussed use liquid alloys of cadmium, zinc, and magnesium and molten chloride salts. The oxidation-reduction steps are done either chemically by the use of an auxiliary redox couple or electrochemically by the use of an external electrical supply. The same basic thermodynamics apply to both the salt transport and the electrotransport processes. Large deviations from ideal solution behavior of the actinides and lanthanides in the liquid alloys have a major influence on the solubilities and the performance of both the salt transport and electrotransport processes. Separation of plutonium and uranium from each other and decontamination from the more noble fission product elements can be achieved using both transport processes. The thermodynamic analysis is used to make process design computations for different process conditions

  1. Fuel cycle options for light water reactors in Germany

    International Nuclear Information System (INIS)

    Broecking, D.; Mester, W.

    1999-01-01

    In Germany 19 nuclear power plants with an electrical output of 22 GWe are in operation. Annually about 450 t of spent fuel are unloaded from the reactors. Currently most of the spent fuel elements are shipped to France and the United Kingdom for reprocessing according to contracts which have been signed since the late 70es. By the amendment of the Atomic Energy Act in 1994 the previous priority for reprocessing of spent nuclear fuel was substituted by a legal equivalency of the reprocessing and direct disposal option. As a consequence some utilities take into consideration the direct disposal of their spent fuel for economical reasons. The separated plutonium will be recycled as MOX fuel in light water reactors. About 30 tons of fissile plutonium will be available to German utilities for recycling by the year 2000. Twelve German reactors are already licensed for the use of MOX fuel, five others have applied for MOX use. Eight reactors are currently using MOX fuel or used it in the past. The spent fuel elements which shall be disposed of without reprocessing will be stored in two interim dry storage facilities at Gorleben and Ahaus. The storage capacities are 3800 and 4200 tHM, respectively. The Gorleben salt dome is currently investigated to prove its suitability as a repository for high level radioactive waste, either in a vitrified form or as conditioned spent fuel. The future development of the nuclear fuel cycle and radioactive waste management depends on the future role of nuclear energy in Germany. According to estimations of the German utilities no additional nuclear power plants are needed in the near future. Around the middle of the next decade it will have to be decided whether existing plants should be substituted by new ones. For the foreseeable time German utilities are interested in a highly flexible approach to the nuclear fuel cycle and waste management keeping open both spent fuel management options: the closed fuel cycle and direct disposal of

  2. Pyro-electrochemical reprocessing of irradiated MOX fast reactor fuel, testing of the reprocessing process with direct MOX fuel production

    Energy Technology Data Exchange (ETDEWEB)

    Kormilitzyn, M.V.; Vavilov, S.K.; Bychkov, A.V.; Skiba, O.V.; Chistyakov, V.M.; Tselichshev, I.V

    2000-07-01

    One of the advanced technologies for fast reactor fuel recycle is pyro-electrochemical molten salt technology. In 1998 we began to study the next phase of the irradiated oxide fuel reprocessing new process MOX {yields} MOX. This process involves the following steps: - Dissolution of irradiated fuel in molten alkaline metal chlorides, - Purification of melt from fission products that are co-deposited with uranium and plutonium oxides, - Electrochemical co-deposition of uranium and plutonium oxides under the controlled cathode potential, - Production of granulated MOX (crushing,salt separation and sizing), and - Purification of melt from fission products by phosphate precipitation. In 1998 a series of experiments were prepared and carried out in order to validate this process. It was shown that the proposed reprocessing flowsheet of irradiated MOX fuel verified the feasibility of its decontamination from most of its fission products (rare earths, cesium) and minor-actinides (americium, curium)

  3. Pyro-electrochemical reprocessing of irradiated MOX fast reactor fuel, testing of the reprocessing process with direct MOX fuel production

    International Nuclear Information System (INIS)

    Kormilitzyn, M.V.; Vavilov, S.K.; Bychkov, A.V.; Skiba, O.V.; Chistyakov, V.M.; Tselichshev, I.V.

    2000-01-01

    One of the advanced technologies for fast reactor fuel recycle is pyro-electrochemical molten salt technology. In 1998 we began to study the next phase of the irradiated oxide fuel reprocessing new process MOX → MOX. This process involves the following steps: - Dissolution of irradiated fuel in molten alkaline metal chlorides, - Purification of melt from fission products that are co-deposited with uranium and plutonium oxides, - Electrochemical co-deposition of uranium and plutonium oxides under the controlled cathode potential, - Production of granulated MOX (crushing,salt separation and sizing), and - Purification of melt from fission products by phosphate precipitation. In 1998 a series of experiments were prepared and carried out in order to validate this process. It was shown that the proposed reprocessing flowsheet of irradiated MOX fuel verified the feasibility of its decontamination from most of its fission products (rare earths, cesium) and minor-actinides (americium, curium)

  4. Advanced Research Reactor Fuel Development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, C. K.; Park, H. D.; Kim, K. H. (and others)

    2006-04-15

    RERTR program for non-proliferation has propelled to develop high-density U-Mo dispersion fuels, reprocessable and available as nuclear fuel for high performance research reactors in the world. As the centrifugal atomization technology, invented in KAERI, is optimum to fabricate high-density U-Mo fuel powders, it has a great possibility to be applied in commercialization if the atomized fuel shows an acceptable in-reactor performance in irradiation test for qualification. In addition, if rod-type U-Mo dispersion fuel is developed for qualification, it is a great possibility to export the HANARO technology and the U-Mo dispersion fuel to the research reactors supplied in foreign countries in future. In this project, reprocessable rod-type U-Mo test fuel was fabricated, and irradiated in HANARO. New U-Mo fuel to suppress the interaction between U-Mo and Al matrix was designed and evaluated for in-reactor irradiation test. The fabrication process of new U-Mo fuel developed, and the irradiation test fuel was fabricated. In-reactor irradiation data for practical use of U-Mo fuel was collected and evaluated. Application plan of atomized U-Mo powder to the commercialization of U-Mo fuel was investigated.

  5. Waste management analysis for the nuclear fuel cycle. I. Actinide recovery from aqueous salt wastes

    International Nuclear Information System (INIS)

    Martella, L.L.; Navratil, J.D.

    1979-01-01

    A preliminary feasibility study of solvent extraction methods has been completed for removing actinides from selected salt wastes likely to be produced during reactor fuel fabrication and reprocessing. The use of a two-step solvent extraction system, tributyl phosphate (TBP) followed by a bidentate organophosphorus extractant (DHDECMP), appears most efficient for removing actinides from salt waste. The TBP step would remove most of the plutonium and >99.99% of the uranium. The second step, using DHDECMP, would remove >99.91% of the americium, the remaining plutonium (>99.98%), and other actinides from the acidified salt waste

  6. Study of the pyrochemical treatment-recycling process of the Molten Salt Reactor fuel; Estudio de sistema de un proceso de tratamiento-reciclaje piroquimico del combustible de un reactor de sales fundidas

    Energy Technology Data Exchange (ETDEWEB)

    Boussier, H.; Heuer, D.

    2010-07-01

    The Separation Processes Studies Laboratory (Commissariat a l'energie Atomique) has made a preliminary assessment of the reprocessing system associated with Molten Salt Fast Reactor (MSFR). The scheme studied in this paper is based on the principle of reductive extraction and metal transfer that constituted the core process designed for the Molten Salt Breeder Reactor (MSBR), although the flow diagram has been adapted to the current needs of the Molten Salt Fast Reactor (MSFR).

  7. Research reactor de-fueling and fuel shipment

    International Nuclear Information System (INIS)

    Ice, R.D.; Jawdeh, E.; Strydom, J.

    1998-01-01

    Planning for the Georgia Institute of Technology Research Reactor operations during the 1996 Summer Olympic Games began in early 1995. Before any details could be outlined, several preliminary administrative decisions had to be agreed upon by state, city, and university officials. The two major administrative decisions involving the reactor were (1) the security level and requirements and (2) the fuel status of the reactor. The Georgia Tech Research Reactor (GTRR) was a heavy-water moderated and cooled reactor, fueled with high-enriched uranium. The reactor was first licensed in 1964 with an engineered lifetime of thirty years. The reactor was intended for use in research applications and as a teaching facility for nuclear engineering students and reactor operators. Approximately one year prior to the olympics, the Georgia Tech administration decided that the GTRR fuel would be removed. In addition, a heightened, beyond regulatory requirements, security system was to be implemented. This report describes the scheduling, operations, and procedures

  8. The effects of core zoning on optimization of design analysis of molten salt reactor

    International Nuclear Information System (INIS)

    Guo, Zhangpeng; Wang, Chenglong; Zhang, Dalin; Chaudri, Khurrum Saleem; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2013-01-01

    Highlights: • 1/8 of core is simulated by MCNP and thermal-hydraulic code simultaneously. • Effects of core zoning are studied by dividing the core into two regions. • Both the neutronics and thermal-hydraulic behavior are investigated. • The flat flux distribution is achieved in the optimization analysis. • The flat flux can lead to worse thermal-hydraulic behavior occasionally. - Abstract: The molten salt reactor (MSR) is one of six advanced reactor types in the frame of the Generation 4 International Forum. In this study, a multiple-channel analysis code (MAC) is developed to analyze thermal-hydraulics behavior and MCNP4c is used to study the neutronics behavior of Molten Salt Reactor Experiment (MSRE). The MAC calculates thermal-hydraulic parameters, namely temperature distribution, flow distribution and pressure drop. The MCNP4c performs the analysis of effective multiplication factor, neutron flux, power distribution and conversion ratio. In this work, the modification of core configuration is achieved by different core zoning and various fuel channel diameters, contributing to flat flux distribution. Specifically, the core is divided into two regions and the effects of different core zoning on the both neutronics and thermal-hydraulic behavior of moderated molten salt reactor are investigated. We conclude that the flat flux distribution cannot always guarantee better performance in thermal-hydraulic perspective and can decreases the graphite lifetime significantly

  9. Separation and Recovery of Uranium Metal from Spent Light Water Reactor Fuel via Electrolytic Reduction and Electrorefining

    International Nuclear Information System (INIS)

    Herrmann, S.D.; Li, S.X.

    2010-01-01

    A series of bench-scale experiments was performed in a hot cell at Idaho National Laboratory to demonstrate the separation and recovery of uranium metal from spent light water reactor (LWR) oxide fuel. The experiments involved crushing spent LWR fuel to particulate and separating it from its cladding. Oxide fuel particulate was then converted to metal in a series of six electrolytic reduction runs that were performed in succession with a single salt loading of molten LiCl - 1 wt% Li2O at 650 C. Analysis of salt samples following the series of electrolytic reduction runs identified the diffusion of select fission products from the spent fuel to the molten salt electrolyte. The extents of metal oxide conversion in the post-test fuel were also quantified, including a nominal 99.7% conversion of uranium oxide to metal. Uranium metal was then separated from the reduced LWR fuel in a series of six electrorefining runs that were performed in succession with a single salt loading of molten LiCl-KCl-UCl3 at 500 C. Analysis of salt samples following the series of electrorefining runs identified additional partitioning of fission products into the molten salt electrolyte. Analyses of the separated uranium metal were performed, and its decontamination factors were determined.

  10. An experimental test facility to support development of the fluoride-salt-cooled high-temperature reactor

    International Nuclear Information System (INIS)

    Yoder, Graydon L.; Aaron, Adam; Cunningham, Burns; Fugate, David; Holcomb, David; Kisner, Roger; Peretz, Fred; Robb, Kevin; Wilgen, John; Wilson, Dane

    2014-01-01

    Highlights: • • A forced convection test loop using FLiNaK salt was constructed to support development of the FHR. • The loop is built of alloy 600, and operating conditions are prototypic of expected FHR operation. • The initial test article is designed to study pebble bed heat transfer cooled by FLiNaK salt. • The test facility includes silicon carbide test components as salt boundaries. • Salt testing with silicon carbide and alloy 600 confirmed acceptable loop component lifetime. - Abstract: The need for high-temperature (greater than 600 °C) energy transport systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The fluoride-salt-cooled high-temperature reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during the development process. The experimental facility described here has been constructed to support the development of the fluoride-salt-cooled high-temperature reactor concept. The facility is capable of operating at up to 700 °C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system, a trace heating system, and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride-salt heat transfer inside a heated pebble bed

  11. Material test reactor fuel research at the BR2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

    2012-03-15

    The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

  12. LIFE Materails: Molten-Salt Fuels Volume 8

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R; Brown, N; Caro, A; Farmer, J; Halsey, W; Kaufman, L; Kramer, K; Latkowski, J; Powers, J; Shaw, H; Turchi, P

    2008-12-11

    The goals of the Laser Inertial Fusion Fission Energy (LIFE) is to use fusion neutrons to fission materials with no enrichment and minimum processing and have greatly reduced wastes that are not of interest to making weapons. Fusion yields expected to be achieved in NIF a few times per day are called for with a high reliable shot rate of about 15 per second. We have found that the version of LIFE using TRISO fuel discussed in other volumes of this series can be modified by replacing the molten-flibe-cooled TRISO fuel zone with a molten salt in which the same actinides present in the TRISO particles are dissolved in the molten salt. Molten salts have the advantage that they are not subject to radiation damage, and hence overcome the radiation damage effects that may limit the lifetime of solid fuels such as TRISO-containing pebbles. This molten salt is pumped through the LIFE blanket, out to a heat exchanger and back into the blanket. To mitigate corrosion, steel structures in contact with the molten salt would be plated with tungsten or nickel. The salt will be processed during operation to remove certain fission products (volatile and noble and semi-noble fission products), impurities and corrosion products. In this way neutron absorbers (fission products) are removed and neutronics performance of the molten salt is somewhat better than that of the TRISO fuel case owing to the reduced parasitic absorption. In addition, the production of Pu and rare-earth elements (REE) causes these elements to build up in the salt, and leads to a requirement for a process to remove the REE during operation to insure that the solubility of a mixed (Pu,REE)F3 solid solution is not exceeded anywhere in the molten salt system. Removal of the REE will further enhance the neutronics performance. With molten salt fuels, the plant would need to be safeguarded because materials of interest for weapons are produced and could potentially be removed.

  13. LIFE Materails: Molten-Salt Fuels Volume 8

    International Nuclear Information System (INIS)

    Moir, R.; Brown, N.; Caro, A.; Farmer, J.; Halsey, W.; Kaufman, L.; Kramer, K.; Latkowski, J.; Powers, J.; Shaw, H.; Turchi, P.

    2008-01-01

    The goals of the Laser Inertial Fusion Fission Energy (LIFE) is to use fusion neutrons to fission materials with no enrichment and minimum processing and have greatly reduced wastes that are not of interest to making weapons. Fusion yields expected to be achieved in NIF a few times per day are called for with a high reliable shot rate of about 15 per second. We have found that the version of LIFE using TRISO fuel discussed in other volumes of this series can be modified by replacing the molten-flibe-cooled TRISO fuel zone with a molten salt in which the same actinides present in the TRISO particles are dissolved in the molten salt. Molten salts have the advantage that they are not subject to radiation damage, and hence overcome the radiation damage effects that may limit the lifetime of solid fuels such as TRISO-containing pebbles. This molten salt is pumped through the LIFE blanket, out to a heat exchanger and back into the blanket. To mitigate corrosion, steel structures in contact with the molten salt would be plated with tungsten or nickel. The salt will be processed during operation to remove certain fission products (volatile and noble and semi-noble fission products), impurities and corrosion products. In this way neutron absorbers (fission products) are removed and neutronics performance of the molten salt is somewhat better than that of the TRISO fuel case owing to the reduced parasitic absorption. In addition, the production of Pu and rare-earth elements (REE) causes these elements to build up in the salt, and leads to a requirement for a process to remove the REE during operation to insure that the solubility of a mixed (Pu,REE)F3 solid solution is not exceeded anywhere in the molten salt system. Removal of the REE will further enhance the neutronics performance. With molten salt fuels, the plant would need to be safeguarded because materials of interest for weapons are produced and could potentially be removed.

  14. Coupled neutronics/thermal-hydraulics for analysis of molten salt reactor

    International Nuclear Information System (INIS)

    Guo, Zhangpeng; Zhou, Jianjun; Zhang, Dalin; Chaudri, Khurrum Saleem; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2013-01-01

    Highlights: ► A multiple-channel analysis code (MAC) is developed to be coupled with MCNP. ► 1/8 of core is simulated in MCNP and thermal-hydraulic code. ► The coupling calculation can achieve stable state after a few iterations. ► The coupling calculation results are in reasonable agreement with the analytic solutions of the ORNL. ► Parametric studies of MSR are performed to provide valuable information for future design MSR. -- Abstract: The Generation IV International Forum (GIF) selected molten salt reactor (MSR) among six advanced reactor types. It is characterized by a liquid circulating fuel that also serves as coolant. In this study, a multiple-channel analysis code (MAC) is developed and it is coupled with MCNP4c to analyze the neutronics/thermal-hydraulics behavior of molten salt reactor experiment (MSRE). The MAC calculates thermal-hydraulic parameters, such as temperature distribution, flow distribution and pressure drop. MCNP4c performs the analysis of effective multiplication factor, neutron flux and power distribution. A linkage code is developed to exchange data between MAC and MCNP to implement coupling iteration process until the power convergence is achieved. The coupling calculation can achieve converged solution after a few iterations. The results are in reasonable agreement with the analytic solutions from the ORNL. For further design analysis, parametric studies are performed to provide valuable information for new design of MSR. The effect of inlet temperature, graphite to molten salt volume ratio (G/Ms) from varying channel diameter and different power levels on the effective multiplication factor, neutron flux, graphite lifetime and temperature distribution are discussed in detail

  15. Molten-salt reactor information system

    International Nuclear Information System (INIS)

    Haubenreich, P.N.; Cardwell, D.W.; Engel, J.R.

    1975-06-01

    The Molten-Salt Reactor Information System (MSRIS) is a computer-based file of abstracts of documents dealing with the technology of molten-salt reactors. The file is stored in the IBM-360 system at ORNL, and may be searched through the use of established interactive computer programs from remote terminals connected to the computer via telephone lines. The system currently contains 373 entries and is subject to updating and expansion as additional information is developed. The nature and general content of the data file, a general approach for obtaining information from it, and the manner in which material is added to the file are described. Appendixes provide the list of keywords currently in use, the subject categories under which information is filed, and simplified procedures for searching the file from remote terminals. (U.S.)

  16. Cermet fuel reactors

    International Nuclear Information System (INIS)

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.

    1987-01-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. The concept evolved in the 1960's with the objective of developing a reactor design which could be used for a wide range of mobile power generation systems including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests and in-reactor irradiation tests using cermet fuel were carried out by General Electric in the 1960's as part of the 710 Development Program and by Argonne National laboratory in a subsequent activity. Cermet fuel development programs are currently underway at Argonne National laboratory and Pacific Northwest Laboratory as part of the Multi-Megawatt Space Power Program. Key features of the cermet fueled reactor design are 1) the ability to achieve very high coolant exit temperatures, and 2) thermal shock resistance during rapid power changes, and 3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, there is a potential for achieving a long operating life because of 1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and 2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core

  17. The fuel of nuclear reactors

    International Nuclear Information System (INIS)

    1995-03-01

    This booklet is a presentation of the different steps of the preparation of nuclear fuels performed by Cogema. The documents starts with a presentation of the different French reactor types: graphite moderated reactors, PWRs using MOX fuel, fast breeder reactors and research reactors. The second part describes the fuel manufacturing process: conditioning of nuclear materials and fabrication of fuel assemblies. The third part lists the different companies involved in the French nuclear fuel industry while part 4 gives a short presentation of the two Cogema's fuel fabrication plants at Cadarache and Marcoule. Part 5 and 6 concern the quality assurance, the safety and reliability aspects of fuel elements and the R and D programs. The last part presents some aspects of the environmental and personnel protection performed by Cogema. (J.S.)

  18. Analysis of minor actinides transmutation for a Molten Salt Fast Reactor

    International Nuclear Information System (INIS)

    Yu, Chenggang; Li, Xiaoxiao; Cai, Xiangzhou; Zou, Chunyan; Ma, Yuwen; Han, Jianlong; Chen, Jingen

    2015-01-01

    Highlights: • The transmutation of MA in a 500 MWth MSFR is analyzed. • A larger MA loading can enhance the MA transmutation and deepen the burnup. • The MA transmutation efficiency can reach 95%. • The FTC can satisfy the safe operating requirement during the entire operating. - Abstract: As one of the six candidate reactors chosen by the Generation IV International Forum (GIF), Molten Salt Fast Reactor (MSFR) has many outstanding advantages and features for advanced nuclear fuel utilization. Effective transmutation of minor actinides (MA) could be attained in this kind of fast reactor, which is of importance in the future closed nuclear fuel cycle scenario. In this work, we attempt to study the MA transmutation capability in a MSFR with power of 500 MWth by analyzing the neutronics characteristics for different MA loadings. The calculated results show that MA loading plays an important role in the reactivity evolution of the MSFR. A larger MA loading is favorable to improving the MA transmutation performance and simultaneously to reducing the fissile consumption. When MA = 18.17 mol%, the transmutation fraction can achieve to about 95% on iso-breeding. We also find that although the fuel temperature coefficient (FTC) decreases with the increasing MA loading, it is still negative enough to keep the safety of the MSFR during the whole operation time. The MA contribution to the effective delayed neutron fraction (EDNF) and the intensity of spontaneous fission neutron (ISFN) are also analyzed. Also MA loading can affect the EDNF during the operation and the ISFN of the MSFR is dominated by 244 Cm. Finally, we analyze the effect of the core power on MA transmutation capability. The result shows that for all the operating powers the depletion ratio of MA to HN increases with time and reaches a maximum value. And additional MA should be fed into the fuel salt before the MA depletion ratio reaches the peak value to improve its transmutation capability. The net

  19. Development of strong-sense validation benchmarks for the fluoride salt-cooled high-temperature reactor

    International Nuclear Information System (INIS)

    Blandford, E. D.

    2012-01-01

    The Fluoride salt-cooled High-temperature Reactor (FHR) is a class of reactor concepts currently under development for the U. S. Dept. of Energy. The FHR is defined as a Generation IV reactor that features low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. Recent experimental work using simulant fluids have been performed to demonstrate key 'proof of principle' FHR concepts and have helped inform the reactor design process. An important element of developing FHR technology is to sufficiently validate the predictive accuracy of the computer codes used to model system response. This paper presents a set of thermal-hydraulics experiments, defined as Strong-Sense Benchmarks (SSB's), which will help establish the FHR validation domain for simulant fluid suitability. These SSB's are more specifically designed to investigate single-phase natural circulation which is the dominant mode of FHR decay heat removal during off-normal conditions. SSB s should be viewed as engineering reference standards and differ from traditional confirmatory experiments in the sense that they are more focused on fundamental physics as opposed to reproducing high levels of physical similarity with the prototypical design. (authors)

  20. Reactor transients tests for SNR fuel elements in HFR reactor

    International Nuclear Information System (INIS)

    Plitz, H.

    1989-01-01

    In HFR reactor, fuel pins of LMFBR reactors are putted in irradiation specimen capsules cooled with sodium for reactor transients tests. These irradiation capsules are instrumented and the experiences realized until this day give results on: - Fuel pins subjected at a continual variation of power - melting fuel - axial differential elongation of fuel pins

  1. Analysis of the reactivity coefficients of the advanced high-temperature reactor for plutonium and uranium fuels

    Energy Technology Data Exchange (ETDEWEB)

    Zakova, Jitka [Department of Nuclear and Reactor Physics, Royal Institute of Technology, KTH, Roslagstullsbacken 21, S-10691, Stockholm (Sweden)], E-mail: jitka.zakova@neutron.kth.se; Talamo, Alberto [Nuclear Engineering Division, Argonne National Laboratory, ANL, 9700 South Cass Avenue, Argonne, IL 60439 (United States)], E-mail: alby@anl.gov

    2008-05-15

    The conceptual design of the advanced high-temperature reactor (AHTR) has recently been proposed by the Oak Ridge National Laboratory, with the intention to provide and alternative energy source for very high temperature applications. In the present study, we focused on the analyses of the reactivity coefficients of the AHTR core fueled with two types of fuel: enriched uranium and plutonium from the reprocessing of light water reactors irradiated fuel. More precisely, we investigated the influence of the outer graphite reflectors on the multiplication factor of the core, the fuel and moderator temperature reactivity coefficients and the void reactivity coefficient for five different molten salts: NaF, BeF{sub 2}, LiF, ZrF{sub 4} and Li{sub 2}BeF{sub 4} eutectic. In order to better illustrate the behavior of the previous parameters for different core configurations, we evaluated the moderating ratio of the molten salts and the absorption rate of the key fuel nuclides, which, of course, are driven by the neutron spectrum. The results show that the fuel and moderator temperature reactivity coefficients are always negative, whereas the void reactivity coefficient can be set negative provided that the fuel to moderator ratio is optimized (the core is undermoderated) and the moderating ratio of the coolant is large.

  2. Analysis of the reactivity coefficients of the advanced high-temperature reactor for plutonium and uranium fuels

    International Nuclear Information System (INIS)

    Zakova, Jitka; Talamo, Alberto

    2008-01-01

    The conceptual design of the advanced high-temperature reactor (AHTR) has recently been proposed by the Oak Ridge National Laboratory, with the intention to provide and alternative energy source for very high temperature applications. In the present study, we focused on the analyses of the reactivity coefficients of the AHTR core fueled with two types of fuel: enriched uranium and plutonium from the reprocessing of light water reactors irradiated fuel. More precisely, we investigated the influence of the outer graphite reflectors on the multiplication factor of the core, the fuel and moderator temperature reactivity coefficients and the void reactivity coefficient for five different molten salts: NaF, BeF 2 , LiF, ZrF 4 and Li 2 BeF 4 eutectic. In order to better illustrate the behavior of the previous parameters for different core configurations, we evaluated the moderating ratio of the molten salts and the absorption rate of the key fuel nuclides, which, of course, are driven by the neutron spectrum. The results show that the fuel and moderator temperature reactivity coefficients are always negative, whereas the void reactivity coefficient can be set negative provided that the fuel to moderator ratio is optimized (the core is undermoderated) and the moderating ratio of the coolant is large

  3. Ceramics as nuclear reactor fuels

    International Nuclear Information System (INIS)

    Reeve, K.D.

    1975-01-01

    Ceramics are widely accepted as nuclear reactor fuel materials, for both metal clad ceramic and all-ceramic fuel designs. Metal clad UO 2 is used commercially in large tonnages in five different power reactor designs. UO 2 pellets are made by familiar ceramic techniques but in a reactor they undergo complex thermal and chemical changes which must be thoroughly understood. Metal clad uranium-plutonium dioxide is used in present day fast breeder reactors, but may eventually be replaced by uranium-plutonium carbide or nitride. All-ceramic fuels, which are necessary for reactors operating above about 750 0 C, must incorporate one or more fission product retentive ceramic coatings. BeO-coated BeO matrix dispersion fuels and silicate glaze coated UO 2 -SiO 2 have been studied for specialised applications, but the only commercial high temperature fuel is based on graphite in which small fuel particles, each coated with vapour deposited carbon and silicon carbide, are dispersed. Ceramists have much to contribute to many aspects of fuel science and technology. (author)

  4. The Advanced High-Temperature Reactor (AHTR) for Producing Hydrogen to Manufacture Liquid Fuels

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Peterson, P.F.; Ott, L.

    2004-01-01

    Conventional world oil production is expected to peak within a decade. Shortfalls in production of liquid fuels (gasoline, diesel, and jet fuel) from conventional oil sources are expected to be offset by increased production of fuels from heavy oils and tar sands that are primarily located in the Western Hemisphere (Canada, Venezuela, the United States, and Mexico). Simultaneously, there is a renewed interest in liquid fuels from biomass, such as alcohol; but, biomass production requires fertilizer. Massive quantities of hydrogen (H2) are required (1) to convert heavy oils and tar sands to liquid fuels and (2) to produce fertilizer for production of biomass that can be converted to liquid fuels. If these liquid fuels are to be used while simultaneously minimizing greenhouse emissions, nonfossil methods for the production of H2 are required. Nuclear energy can be used to produce H2. The most efficient methods to produce H2 from nuclear energy involve thermochemical cycles in which high-temperature heat (700 to 850 C) and water are converted to H2 and oxygen. The peak nuclear reactor fuel and coolant temperatures must be significantly higher than the chemical process temperatures to transport heat from the reactor core to an intermediate heat transfer loop and from the intermediate heat transfer loop to the chemical plant. The reactor temperatures required for H2 production are at the limits of practical engineering materials. A new high-temperature reactor concept is being developed for H2 and electricity production: the Advanced High-Temperature Reactor (AHTR). The fuel is a graphite-matrix, coated-particle fuel, the same type that is used in modular high-temperature gas-cooled reactors (MHTGRs). The coolant is a clean molten fluoride salt with a boiling point near 1400 C. The use of a liquid coolant, rather than helium, reduces peak reactor fuel and coolant temperatures 100 to 200 C relative to those of a MHTGR. Liquids are better heat transfer fluids than gases

  5. Numerical research on natural convection in molten salt reactor with non-uniformly distributed volumetric heat generation

    International Nuclear Information System (INIS)

    Qian Libo; Qiu Suizheng; Zhang Dalin; Su Guanghui; Tian Wenxi

    2010-01-01

    Molten salt reactor is one of the six Generation IV systems capable of breeding and transmutation of actinides and long-lived fission products, which uses the liquid molten salt as the fuel solvent, coolant and heat generation simultaneously. The present work presents a numerical investigation on natural convection with non-uniform heat generation through which the heat generated by the fluid fuel is removed out of the core region when the reactor is under post-accident condition or zero-power condition. The two-group neutron diffusion equation is applied to calculated neutron flux distribution, which leads to non-uniform heat generation. The SIMPLER algorithm is used to calculate natural convective heat transfer rate with isothermal or adiabatic rigid walls. These two models are coupled through the temperature field and heat sources. The peculiarities of natural convection with non-uniform heat generation are investigated in a range of Ra numbers (10 3 ∼ 10 7 ) for the laminar regime of fluid motion. In addition, the numerical results are also compared with those containing uniform heat generation.

  6. Program management plan for the Molten Salt Reactor Experiment Remediation Project at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1996-09-01

    The primary mission of the Molten Salt Reactor Experiment (MSRE) Remediation Project is to effectively implement the risk-reduction strategies and technical plans to stabilize and prevent further migration of uranium within the MSRE facility, remove the uranium and fuel salts from the system, and dispose of the fuel and flush salts by storage in appropriate depositories to bring the facility to a surveillance and maintenance condition before decontamination and decommissioning. This Project Management Plan (PMP) for the MSRE Remediation Project details project purpose; technical objectives, milestones, and cost objectives; work plan; work breakdown structure (WBS); schedule; management organization and responsibilities; project management performance measurement planning, and control; conduct of operations; configuration management; environmental, safety, and health compliance; quality assurance; operational readiness reviews; and training

  7. Program management plan for the Molten Salt Reactor Experiment Remediation Project at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    The primary mission of the Molten Salt Reactor Experiment (MSRE) Remediation Project is to effectively implement the risk-reduction strategies and technical plans to stabilize and prevent further migration of uranium within the MSRE facility, remove the uranium and fuel salts from the system, and dispose of the fuel and flush salts by storage in appropriate depositories to bring the facility to a surveillance and maintenance condition before decontamination and decommissioning. This Project Management Plan (PMP) for the MSRE Remediation Project details project purpose; technical objectives, milestones, and cost objectives; work plan; work breakdown structure (WBS); schedule; management organization and responsibilities; project management performance measurement planning, and control; conduct of operations; configuration management; environmental, safety, and health compliance; quality assurance; operational readiness reviews; and training.

  8. Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

    Energy Technology Data Exchange (ETDEWEB)

    Laureau, A., E-mail: laureau.axel@gmail.com; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

    2017-05-15

    Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

  9. Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

    International Nuclear Information System (INIS)

    Laureau, A.; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

    2017-01-01

    Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

  10. Fuel exchanger in FBR type reactor

    International Nuclear Information System (INIS)

    Shinden, Kazuhiko; Tanaka, Osamu.

    1990-01-01

    The present invention concerns a fuel exchanger for exchanging fuels in an LMFBR type reactor using liquid metals as coolants. An outer gripper cylinder rotating device for rotating an outer gripper cylinder that holds a gripper is driven, to lower the gripper driving portion and the outer gripper cylinder, fuels are caught by the finger at the top end of the outer gripper cylinder and elevated to extract the fuels from the reactor core. Then, the gripper driving portion casing and the outer gripper cylinder are rotated to rotate the fuels caught by the gripper. Subsequently, the gripper driving portion and the outer gripper cylinder are lowered to charge the fuels in the reactor core. This can directly shuffle the fuels in the reactor core without once transferring the fuels into a reactor storing pot and replacing with other fuels, thereby shortening the shuffling time. (I.N.)

  11. The multi region molten-salt reactor concept

    International Nuclear Information System (INIS)

    Gyula, Csom; Sandor, Feher; Szieberth, M.; Szabolcs, Czifrus

    2003-01-01

    The molten-salt reactor (MSR) concept is one of the most promising systems for the realisation of transmutation. The objective is the development of a transmutation technique along with a device implementing it, which yield higher transmutation efficiencies than that of the known procedures. The procedure is the multi-step transmutation, in which the transformation is carried out in several consecutive steps of different neutron flux and spectrum. In order to implement this, a multi-region transmutation device, i.e. nuclear reactor or sub-critical system is proposed, in which several separate flow-through irradiation rooms are formed with various neutron spectra and fluxes. The paper presents calculations that were performed for a special 5-region version of the multi-region molten-salt reactor. (author)

  12. Reactor fuel charging equipment

    International Nuclear Information System (INIS)

    Wade, Elman.

    1977-01-01

    In many types of reactor fuel charging equipment, tongs or a grab, attached to a trolley, housed in a guide duct, can be used for withdrawing from the core a selected spent fuel assembly or to place a new fuel assembly in the core. In these facilities, the trolley may have wheels that roll on rails in the guide duct. This ensures the correct alignment of the grab, the trolley and fuel assembly when this fuel assembly is being moved. By raising or lowering such a fuel assembly, the trolley can be immerged in the coolant bath of the reactor, whereas at other times it can be at a certain level above the upper surface of the coolant bath. The main object of the invention is to create a fuel handling apparatus for a sodium cooled reactor with bearings lubricated by the sodium coolant and in which the contamination of these bearings is prevented [fr

  13. Fluoride partitioning R and D programme for molten salt transmutation reactor systems in the Czech Republic

    International Nuclear Information System (INIS)

    Uhlir, J.; Priman, V.; Vanicek, J.

    2001-01-01

    The transmutation of spent nuclear fuel is considered a prospective alternative conception to the current conception based on the non-reprocessed spent fuel disposal into underground repository. The Czech research and development programme in the field of partitioning and transmutation is founded on the Molten Salt Transmutation Reactor system concept with fluoride salts based liquid fuel, the fuel cycle of which is grounded on pyrochemical / pyrometallurgical fluoride partitioning of spent fuel. The main research activities in the field of fluoride partitioning are oriented mainly towards technological research of Fluoride Volatility Method and laboratory research on electro-separation methods from fluoride melts media. The Czech national conception in the area of P and T research issues from the national power industry programme and from the Czech Power Company intentions of the extensive utilization of nuclear power in our country. The experimental R and D work is concentrated mainly in the Nuclear Research Institute Rez plc that plays a role of main nuclear research workplace for the Czech Power Company. (author)

  14. Immobilization of IFR salt wastes in mortar

    International Nuclear Information System (INIS)

    Fisher, D.F.; Johnson, T.R.

    1988-01-01

    Portland cement-base mortars are being considered for immobilizing chloride salt wastes from the fuel cycle of an integral fast reactor (IFR). The IFR is a sodium-cooled fast reactor with metal fuel. It has a close-coupled fuel cycle in which fission products are separated from the actinides in an electrochemical cell operating at 500 degrees C. This cell has a cadmium anode and a liquid salt electrolyte. The salt will be a low-melting mixture of alkaline and alkaline earth chlorides. This paper discusses one method being considered for immobilizing this treated salt, to disperse it in a portland cement-base motar, which would then be sealed in corrosion-resistant containers. For this application, the grout must be sufficiently fluid that it can be pumped into canisters where it will solidify into a strong, leach-resistant material

  15. Power from plutonium: fast reactor fuel

    International Nuclear Information System (INIS)

    Bishop, J.F.W.

    1981-01-01

    Points of similarity and of difference between fast reactor fuel and fuels for AGR and PWR plants are established. The flow of uranium and plutonium in fast and thermal systems is also mentioned, establishing the role of the fast reactor as a plutonium burner. A historical perspective of fast reactors is given in which the substantial experience accumulated in test and prototype is indicated and it is noted that fast reactors have now entered the commercial phase. The relevance of the data obtained in the test and prototype reactors to the behaviour of commercial fast reactor fuel is considered. The design concepts employed in fuel are reviewed, including sections on core support styles, pin support and pin detail. This is followed by a discussion of current issues under the headings of manufacture, performance and reprocessing. This section includes a consideration of gel fuel, achievable burn-up, irradiation induced distortions and material choices, fuel form, and fuel failure mechanisms. Future development possibilities are also discussed and the Paper concludes with a view on the logic of a UK fast reactor strategy. (U.K.)

  16. Potentialities of the molten salt reactor concept for a sustainable nuclear power production based on thorium cycle in epithermal spectrum

    International Nuclear Information System (INIS)

    Nuttin, Alexis

    2002-01-01

    In the case of a significant nuclear contribution to world energy needs, the problem of present nuclear waste management pose the sustainability of the PWR fuel cycle back into question. Studies on storage and incineration of these wastes should therefore go hand in hand with studies on innovative systems dedicated to a durable nuclear energy production, as reliable, clean and safe as possible. We are here interested in the concept of molten salt reactor, whose fuel is liquid. This particularity allows an online pyrochemical reprocessing which gives the possibility to overcome some neutronic limits. In the late sixties, the MSBR (Molten Salt Breeder Reactor) project of a graphite-moderated fluoride molten salt reactor proved thus that breeding is attainable with thorium in a thermal spectrum, provided that the online reprocessing is appropriate. By means of simulation tools developed around the Monte Carlo code MCNP, we first re-evaluate the performance of a reference system, which is inspired by the MSBR project. The complete study of the pre-equilibrium transient of this 2,500 MWth reactor, started with 232 Th/ 233 U fuel, allows us to validate our reference choices. The obtained equilibrium shows an important reduction of inventories and induced radio-toxicities in comparison with the other possible fuel cycles. The online reprocessing is efficient enough to make the system breed, with a doubling time of about thirty years at equilibrium. From the reference system, we then test different options in terms of neutron economy, transmutation and control of reactivity. We find that the online reprocessing brings most of its flexibility to this system, which is particularly well adapted to power generation with thorium. The study of transition scenarios to this fuel cycle quantifies the limits of a possible deployment from the present French power stock, and finally shows that a rational management of the available plutonium would be necessary in any case. (author)

  17. Removal of uranium and salt from the Molten Salt Reactor Experiment

    International Nuclear Information System (INIS)

    Peretz, F.J.; Rushton, J.E.; Faulkner, R.L.; Walker, K.L.; Del Cul, G.D.

    1998-01-01

    In 1994, migration of 233 U was discovered to have occurred at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). This paper describes the actions now underway to remove uranium from the off-gas piping and the charcoal bed, to remove and stabilize the salts, and to convert the uranium to a stable oxide for long-term storage

  18. Removal of uranium and salt from the Molten Salt Reactor Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Peretz, F.J.; Rushton, J.E.; Faulkner, R.L.; Walker, K.L.; Del Cul, G.D.

    1998-06-01

    In 1994, migration of {sup 233}U was discovered to have occurred at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). This paper describes the actions now underway to remove uranium from the off-gas piping and the charcoal bed, to remove and stabilize the salts, and to convert the uranium to a stable oxide for long-term storage.

  19. CANDU reactors with reactor grade plutonium/thorium carbide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

    2011-08-15

    Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

  20. Research reactor fuel - an update

    International Nuclear Information System (INIS)

    Finlay, M.R.; Ripley, M.I.

    2003-01-01

    In the two years since the last ANA conference there have been marked changes in the research reactor fuel scene. A new low-enriched uranium (LEU) fuel, 'monolithic' uranium molybdenum, has shown such promise in initial trials that it may be suitable to meet the objectives of the Joint Declaration signed by Presidents Bush and Putin to commit to converting all US and Russian research reactors to LEU by 2012. Development of more conventional aluminium dispersion UMo LEU fuel has continued in the meantime and is entering the final qualification stage of multiple full sized element irradiations. Despite this progress, the original 2005 timetable for UMo fuel qualification has slipped and research reactors, including the RRR, may not convert from silicide to UMo fuel before 2007. The operators of the Swedish R2 reactor have been forced to pursue the direct route of qualifying a UMo lead test assembly (LTA) in order to meet spent fuel disposal requirements of the Swedish law. The LTA has recently been fabricated and is expected to be loaded shortly into the R2 reactor. We present an update of our previous ANA paper and details of the qualification process for UMo fuel

  1. Study on Utilization of Super Grade Plutonium in Molten Salt Reactor FUJI-U3 using CITATION Code

    Science.gov (United States)

    Wulandari, Cici; Waris, Abdul; Pramuditya, Syeilendra; Asril, Pramutadi AM; Novitrian

    2017-07-01

    FUJI-U3 type of Molten Salt Reactor (MSR) has a unique design since it consists of three core regions in order to avoid the replacement of graphite as moderator. MSR uses floride as a nuclear fuel salt with the most popular chemical composition is LiF-BeF2-ThF4-233UF4. ThF4 and 233UF4 are the fertile and fissile materials, respectively. On the other hand, LiF and BeF2 working as both fuel and heat transfer medium. In this study, the super grade plutonium will be utilized as substitution of 233U since plutonium is easier to be obtained compared to 233U as main fuel. Neutronics calculation was performed by using PIJ and CITATION modules of SRAC 2002 code with JENDL 3.2 as nuclear data library.

  2. Asymptotic estimation of reactor fueling optimal strategy

    International Nuclear Information System (INIS)

    Simonov, V.D.

    1985-01-01

    The problem of improving the technical-economic factors of operating. and designed nuclear power plant blocks by developino. internal fuel cycle strategy (reactor fueling regime optimization), taking into account energy system structural peculiarities altogether, is considered. It is shown, that in search of asymptotic solutions of reactor fueling planning tasks the model of fuel energy potential (FEP) is the most ssuitable and effective. FEP represents energy which may be produced from the fuel in a reactor with real dimensions and power, but with hypothetical fresh fuel supply, regime, providing smilar burnup of all the fuel, passing through the reactor, and continuous overloading of infinitely small fuel portion under fule power, and infinitely rapid mixing of fuel in the reactor core volume. Reactor fuel run with such a standard fuel cycle may serve as FEP quantitative measure. Assessment results of optimal WWER-440 reactor fresh fuel supply periodicity are given as an example. The conclusion is drawn that with fuel enrichment x=3.3% the run which is 300 days, is economically justified, taking into account that the cost of one energy unit production is > 3 cop/KW/h

  3. Molten salt breeder reactor

    International Nuclear Information System (INIS)

    Furukawa, Kazuo; Tsukada, Kineo; Nakahara, Yasuaki; Oomichi, Toshihiko; Oono, Hideo.

    1982-01-01

    Purpose: To simplify the structure, as well as improve the technical reliability and safety by the elimination of a proton beam entering window. Constitution: The nuclear reactor container main body is made of Hastelloy N and provided at the inner surface with two layers of graphite shields except for openings. An aperture was formed in the upper surface of the container, through which protons accelerated by a linear accelerator are directly entered to the liquid surface of molten salts such as 7LiF-BeF 2 -ThF 4 , 7LiF-NaF-ThF 4 , 7LiF-Rb-UF 4 , NaF-KF-UF 4 and the like. The heated molten salts are introduced by way of a pipeway into a heat exchanger where the heat is transferred to coolant salts and electric generation is conducted by way of heated steams. (Furukawa, Y.)

  4. Improvement to molten salt reactors

    International Nuclear Information System (INIS)

    Bienvenu, Claude.

    1975-01-01

    The invention proposes a molten salt nuclear reactor whose core includes a mass of at least one fissile element salt to which can be added other salts to lower the melting temperature of the mass. This mass also contains a substance with a low neutron capture section that does not give rise to a chemical reaction or to an azeotropic mixture with these salts and having an atmospheric boiling point under that of the mass in operation. Means are provided for collecting this substance in the vapour state and returning it as a liquid to the mass. The kind of substance chosen will depend on that of the molten salts (fissile element salts and, where required, salts to lower the melting temperature). In actual practice, the substance chosen will have an atmospheric pressure boiling point of between 600 and 1300 0 C and a melting point sufficiently below 600 0 C to prevent solidification and clogging in the return line of the substance from the exchanger. Among the materials which can be considered for use, mention is made of magnesium, rubidium, cesium and potassium but metal cesium is not employed in the case of many fissile salts, such as fluorides, which it would reduced to the planned working temperatures [fr

  5. Thorium molten-salt nuclear energy synergetics

    International Nuclear Information System (INIS)

    Furukawa, Kazuo

    1989-01-01

    One of the most practical and rational approaches for establishing the idealistic Thorium resource utilization program has been presented, which might be effective to solve the principal energy problems, concerning safety, proliferation and terrorism, resource, power size and fuel cycle economy, for the next century. The first step will be the development of Small Molten-Salt Reactors as a flexible power station, which is suitable for early commercialization of Th reactors not necessarily competing with proven Large Solid-Fuel Reactors. Therefore, the more detailed design works and practical R and D planning should be performed under the international cooperations soon, soundly depending on the basic technology established by ORNL already. R and D cost would be surprisingly low. This reactor(MSR) seems to be idealistic not only in power-size, siting, safety, safeguard and economy, but also as an effective partner of Molten-Salt Fissile Breeders(MSB) in order to establish the simplest and economical Thorium molten-salt breeding fuel cycle named THORIMS-NES in all over the world including the developing countries and isolated areas. This would be one of the most practical replies to the Lilienthal's appeal of 'A NEW START' in Nuclear Energy. (author)

  6. Fluoride-Salt-Cooled High-Temperature Reactor (FHR) for Power and Process Heat

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, Charles [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Peterson, Per [Univ. of California, Berkeley, CA (United States); Sridharan, Kumar [Univ. of Wisconsin, Madison, WI (United States)

    2015-01-21

    In 2011 the U.S. Department of Energy through its Nuclear Energy University Program (NEUP) awarded a 3- year integrated research project (IRP) to the Massachusetts Institute of Technology (MIT) and its partners at the University of California at Berkeley (UCB) and the University of Wisconsin at Madison (UW). The IRP included Westinghouse Electric Company and an advisory panel chaired by Regis Matzie that provided advice as the project progressed. The first sentence of the proposal stated the goals: The objective of this Integrated Research Project (IRP) is to develop a path forward to a commercially viable salt-cooled solid-fuel high-temperature reactor with superior economic, safety, waste, nonproliferation, and physical security characteristics compared to light-water reactors. This report summarizes major results of this research.

  7. Fluoride-Salt-Cooled High-Temperature Reactor (FHR) for Power and Process Heat

    International Nuclear Information System (INIS)

    Forsberg, Charles; Hu, Lin-wen; Peterson, Per; Sridharan, Kumar

    2015-01-01

    In 2011 the U.S. Department of Energy through its Nuclear Energy University Program (NEUP) awarded a 3- year integrated research project (IRP) to the Massachusetts Institute of Technology (MIT) and its partners at the University of California at Berkeley (UCB) and the University of Wisconsin at Madison (UW). The IRP included Westinghouse Electric Company and an advisory panel chaired by Regis Matzie that provided advice as the project progressed. The first sentence of the proposal stated the goals: The objective of this Integrated Research Project (IRP) is to develop a path forward to a commercially viable salt-cooled solid-fuel high-temperature reactor with superior economic, safety, waste, nonproliferation, and physical security characteristics compared to light-water reactors. This report summarizes major results of this research.

  8. Fast reactor fuel reprocessing. An Indian perspective

    International Nuclear Information System (INIS)

    Natarajan, R.; Raj, Baldev

    2005-01-01

    The Department of Atomic Energy (DAE) envisioned the introduction of Plutonium fuelled fast reactors as the intermediate stage, between Pressurized Heavy Water Reactors and Thorium-Uranium-233 based reactors for the Indian Nuclear Power Programme. This necessitated the closing of the fast reactor fuel cycle with Plutonium rich fuel. Aiming to develop a Fast Reactor Fuel Reprocessing (FRFR) technology with low out of pile inventory, the DAE, with over four decades of operating experience in Thermal Reactor Fuel Reprocessing (TRFR), had set up at the India Gandhi Center for Atomic Research (IGCAR), Kalpakkam, R and D facilities for fast reactor fuel reprocessing. After two decades of R and D in all the facets, a Pilot Plant for demonstrating FRFR had been set up for reprocessing the FBTR (Fast Breeder Test Reactor) spent mixed carbide fuel. Recently in this plant, mixed carbide fuel with 100 GWd/t burnup fuel with short cooling period had been successfully reprocessed for the first time in the world. All the challenging problems encountered had been successfully overcome. This experience helped in fine tuning the designs of various equipments and processes for the future plants which are under construction and design, namely, the DFRP (Demonstration Fast reactor fuel Reprocessing Plant) and the FRP (Fast reactor fuel Reprocessing Plant). In this paper, a comprehensive review of the experiences in reprocessing the fast reactor fuel of different burnup is presented. Also a brief account of the various developmental activities and strategies for the DFRP and FRP are given. (author)

  9. Characters of neutron noise in full-size molten salt reactor

    International Nuclear Information System (INIS)

    Wang, Jiangmeng; Cao, Xinrong

    2015-01-01

    Highlights: • The larger system size makes full-size MSR deviate from point kinetic behavior. • The increasing velocity has non-monotonic effect on the effective delayed neutron fraction. • The amplitude of Green’s function at low frequencies is inversely proportional to the effective delayed neutron fraction. • The range of plateau region is smaller due to the more prominent point kinetic effect. - Abstract: In the present paper, the frequency-dependent and space-dependent behavior of the neutron noise in a full-size Molten Salt Reactor (MSR) is investigated. The theoretical models considering the fuel circulation are established based on one-group neutron diffusion theory. Green’s function of the neutron noise induced by a propagating perturbation is introduced with linear noise theory, due to the small perturbation. The equations are numerically calculated by developing a code, in which the eigenfunction expansion method is adopted. The static results show that the effective delayed neutron fraction changes non-monotonically with the increasing fuel velocity. In the dynamic case, the results are compared to those obtained in 1D MSR and a traditional reactor, in order to figure out the effects of both the fuel circulation and the system size. It is found that there is no difference in 1D and 3D MSR systems from the view of fuel circulation, i.e., the fuel circulation enhances the spatial neutronic coupling and leads to the stronger point kinetic effect. The more prominent space-dependent effect founded in 3D traditional reactors is also observed in the MSR, due to the looser neutronic coupling and the unique singularity of Green’s function in the location of the perturbation. Another interesting finding is that Green’s function for low frequencies changes non-monotonically with increasing velocity. The largest magnitude of Green’s function is observed at the velocity where the effective delayed neutron fraction reaches its minimum. Finally, the

  10. Fuel assembly and reactor core

    International Nuclear Information System (INIS)

    Aoyama, Motoo; Koyama, Jun-ichi; Uchikawa, Sadao; Bessho, Yasunori; Nakajima, Akiyoshi; Maruyama, Hiromi; Ozawa, Michihiro; Nakamura, Mitsuya.

    1990-01-01

    The present invention concerns fuel assemblies charged in a BWR type reactor and the reactor core. The fuel assembly comprises fuel rods containing burnable poisons and fuel rods not containing burnable poisons. Both of the highest and the lowest gadolinia concentrations of the fuel rods containing gadolinia as burnable poisons are present in the lower region of the fuel assembly. This can increase the spectral shift effect without increasing the maximum linear power density. (I.N.)

  11. High Temperature Fluoride Salt Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Aaron, Adam M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cunningham, Richard Burns [Univ. of Tennessee, Knoxville, TN (United States); Fugate, David L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holcomb, David Eugene [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kisner, Roger A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peretz, Fred J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yoder, Jr, Graydon L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    Effective high-temperature thermal energy exchange and delivery at temperatures over 600°C has the potential of significant impact by reducing both the capital and operating cost of energy conversion and transport systems. It is one of the key technologies necessary for efficient hydrogen production and could potentially enhance efficiencies of high-temperature solar systems. Today, there are no standard commercially available high-performance heat transfer fluids above 600°C. High pressures associated with water and gaseous coolants (such as helium) at elevated temperatures impose limiting design conditions for the materials in most energy systems. Liquid salts offer high-temperature capabilities at low vapor pressures, good heat transport properties, and reasonable costs and are therefore leading candidate fluids for next-generation energy production. Liquid-fluoride-salt-cooled, graphite-moderated reactors, referred to as Fluoride Salt Reactors (FHRs), are specifically designed to exploit the excellent heat transfer properties of liquid fluoride salts while maximizing their thermal efficiency and minimizing cost. The FHR s outstanding heat transfer properties, combined with its fully passive safety, make this reactor the most technologically desirable nuclear power reactor class for next-generation energy production. Multiple FHR designs are presently being considered. These range from the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) [1] design originally developed by UC-Berkeley to the Small Advanced High-Temperature Reactor (SmAHTR) and the large scale FHR both being developed at ORNL [2]. The value of high-temperature, molten-salt-cooled reactors is also recognized internationally, and Czechoslovakia, France, India, and China all have salt-cooled reactor development under way. The liquid salt experiment presently being developed uses the PB-AHTR as its focus. One core design of the PB-AHTR features multiple 20 cm diameter, 3.2 m long fuel channels

  12. Method of reactor fueling

    International Nuclear Information System (INIS)

    Saito, Toshiro.

    1983-01-01

    Purpose: To decrease the cost and shorten the working time by saving fueling neutron detectors and their components. Method: Incore drive tubes for the neutron source range monitor (SRM) and intermediate range monitor (IRM) are disposed respectively within in a reactor core and a SRM detector assembly is inserted to the IRM incore drive tube which is most nearest to the neutron source upon reactor fueling. The reactor core reactivity is monitored by the SRM detector assembly. The SRM detector asesembly inserted into the IRM drive tube is extracted at the time of charging fuels up to the frame connecting the SRM and, thereafter, IRM detection assembly is inserted into the IRM drive tube and the SRM detector assembly is inserted into the SRM drive tube respectively for monitoring the reactor core. (Sekiya, K.)

  13. Thermodynamics of soluble fission products cesium and iodine in the Molten Salt Reactor

    Science.gov (United States)

    Capelli, E.; Beneš, O.; Konings, R. J. M.

    2018-04-01

    The present study describes the full thermodynamic assessment of the Li,Cs,Th//F,I system. The existing database for the relevant fluoride salts considered as fuel for the Molten Salt Reactor (MSR) has been extended with two key fission products, cesium and iodine. A complete evaluation of all the common-ion binary and ternary sub-systems of the LiF-ThF4-CsF-LiI-ThI4-CsI system has been performed and the optimized parameters are presented in this work. New equilibrium data have been measured using Differential Scanning Calorimetry and were used to assess the reciprocal ternary systems and confirm the extrapolated phase diagrams. The developed database significantly contributes to the understanding of the behaviour of cesium and iodine in the MSR, which strongly depends on their concentration and chemical form. Cesium bonded with fluorine is well retained in the fuel mixture while in the form of CsI the solubility of these elements is very limited. Finally, the influence of CsI and CsF on the physico-chemical properties of the fuel mixture was calculated as function of composition.

  14. Gaseous fuel reactors for power systems

    Science.gov (United States)

    Kendall, J. S.; Rodgers, R. J.

    1977-01-01

    Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

  15. Status of the French research in the field of molten salt nuclear reactors

    International Nuclear Information System (INIS)

    Hery, M.; Israel, M.; Fauger, P.; Lecocq, A.

    1977-01-01

    The research program of the CEA in the field of molten salt nuclear reactors has been concerned with MSBR type reactors (Molten Salt Breeder Reactor). The papers written after having performed the theoretical analysis are entitled: core, circuits, chemistry and economy; they include some criticisms and suggestions. The experimental studies consisted in: graphite studies, chemical studies of the salt, metallic materials, the salt loop and the lead loop [fr

  16. An aqueous lithium salt blanket option for fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Steiner, D.; Varsamis, G. (Rensselaer Polytechnic Inst., Troy, NY (USA). Dept. of Nuclear Engineering and Engineering Physics); Deutsch, L.; Rathke, J. (Grumman Corp., Bethpage, NY (USA). Advanced Energy Systems); Gierszewski, P. (Canadian Fusion Fuels Technology Project (CFFTP), Mississauga, ON (Canada))

    1989-04-01

    An aqueous lithium salt blanket (ALSB) concept is proposed which could be the basis for either a power reactor blanket or a test module in an engineering test reactor. The design is based on an austenitic stainless steel structure, a beryllium multiplier, and a salt breeder concentration of about 32 g LiNO/sub 3/ per 100 cm/sup 3/ of H/sub 2/O. To limit tritium release rates, the salt breeder solution is separated from the water coolant circuit. The overall tritium system cost for a 2400 MW (fusion power) reactor is estimated to be 180 million Dollar US87 installed. (orig.).

  17. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Butterfield, C.E.; Waite, E.

    1982-01-01

    A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

  18. Fast reactor fuel design and development

    International Nuclear Information System (INIS)

    Bishop, J.F.W.; Chamberlain, A.; Holmes, J.A.G.

    1977-01-01

    Fuel design parameters for oxide and carbide fast reactor fuels are reviewed in the context of minimising the total uranium demands for a combined thermal and fast reactor system. The major physical phenomena conditioning fast reactor fuel design, with a target of high burn-up, good breeding and reliable operation, are characterised. These include neutron induced void swelling, irradiation creep, pin failure modes, sub-assembly structural behaviour, behaviour of defect fuel, behaviour of alternative fuel forms. The salient considerations in the commercial scale fabrication and reprocessing of the fuels are reviewed, leading to the delineation of possible routes for the manufacture and reprocessing of Commercial Reactor fuel. From the desiderata and restraints arising from Surveys, Performance and Manufacture, the problems posed to the Designer are considered, and a narrow range of design alternatives is proposed. The paper concludes with a consideration of the development areas and the conceptual problems for fast reactors associated with those areas

  19. Molten salt reactor as asymptotic safety nuclear system

    International Nuclear Information System (INIS)

    Novikov, V.M.; Ignatyev, V.V.

    1989-01-01

    Safety is becoming the main and priority problem of the nuclear power development. An increase of the active safety measures could hardly be considered as the proper way to achieve the asymptotically high level of nuclear safety. It seem that the more realistic way to achieve such a goal is to minimize risk factors and to maximize the use of inherent and passive safety properties. The passive inherent safety features of the liquid fuel molten salt reactor (MSR) technology are making it attractive for future energy generation. The achievement of the asymptotic safety in MSR is being connected with the minimization of such risk factors as a reactivity excess, radioactivity stored, decay heat, non nuclear energy stored in core. In this paper safety peculiarities of the different MSR concepts are discussed

  20. Fuel assembly for FBR type reactor and reactor core thereof

    International Nuclear Information System (INIS)

    Kobayashi, Kaoru.

    1998-01-01

    The present invention provides a fuel assembly to be loaded to a reactor core of a large sized FBR type reactor, in which a coolant density coefficient can be reduced without causing power peaking in the peripheral region of neutron moderators loaded in the reactor core. Namely, the fuel assembly for the FBR type reactor comprises a plurality of fission product-loaded fuel rods and a plurality of fertile material-loaded fuel rods and one or more rods loading neutron moderators. In this case, the plurality of fertile material-loaded fuel rods are disposed to the peripheral region of the neutron moderator-loaded rods. The plurality of fission product-loaded fuel rods are disposed surrounding the peripheral region of the plurality of fertile material-loaded fuel rods. The neutron moderator comprises zirconium hydride, yttrium hydride and calcium hydride. The fission products are mixed oxide fuels. The fertile material comprises depleted uranium or natural uranium. (I.S.)

  1. Advanced Nuclear Fuels for More Capable and Sustainable Exploration

    Data.gov (United States)

    National Aeronautics and Space Administration — Molten salt reactors are a subtype of reactor that uses nuclear fuel dissolved in a molten salt liquid medium (such as LiF-BeF2-UF4) as both fuel and coolant. The...

  2. Design of a heterogeneous subcritical nuclear reactor with molten salts based on thorium; Diseno de un reactor nuclear subcritico heterogeneo con sales fundidas a base de torio

    Energy Technology Data Exchange (ETDEWEB)

    Medina C, D.; Hernandez A, P.; Letechipia de L, C.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Sajo B, L., E-mail: dmedina_c@hotmail.com [Universidad Simon Bolivar, Laboratorio de Fisica Nuclear, Apdo. Postal 89000, Caracas 1080-A (Venezuela, Bolivarian Republic of)

    2015-09-15

    This paper presents the design of a heterogeneous subcritical nuclear reactor with molten salts based on thorium, with graphite moderator and a {sup 252}Cf source, whose dose levels at the periphery allows its use in teaching and research activities. The design was realized by the Monte Carlo method, where the geometry, dimensions and the fuel was varied in order to obtain the best design. The result was a cubic reactor of 110 cm of side, with graphite moderator and reflector. In the central part having 9 ducts of 3 cm in diameter, eight of them are 110 cm long, which were placed on the Y axis; the separation between each duct is 10 cm. The central duct has 60 cm in length and this contains the {sup 252}Cf source, also there are two irradiation channels and the other six contain a molten salt ({sup 7}LiF - BeF{sub 2} - ThF{sub 4} - UF{sub 4}) as fuel. For the design the k{sub eff} was calculated, neutron spectra and ambient dose equivalent. In the first instance the above was calculated for a virgin fuel, was called case 1; then a percentage of {sup 233}U was used and the percentage of Th was decreased and was called case 2. This with the purpose of comparing two different fuels operating within the reactor. For the two irradiation ducts three positions are used: center, back and front, in each duct in order to have different flows. (Author)

  3. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  4. A calculational procedure for neutronic and depletion analysis of Molten-Salt reactors based on SCALE6/TRITON

    International Nuclear Information System (INIS)

    Sheu, R.J.; Chang, J.S.; Liu, Y.-W. H.

    2011-01-01

    Molten-Salt Reactors (MSRs) represent one of the selected categories in the GEN-IV program. This type of reactor is distinguished by the use of liquid fuel circulating in and out of the core, which makes it possible for online refueling and salt processing. However, this operation characteristic also complicates the modeling and simulation of reactor core behaviour using conventional neutronic codes. The TRITON sequence in the SCALE6 code system has been designed to provide the combined capabilities of problem-dependent cross-section processing, rigorous treatment of neutron transport, and coupled with the ORIGEN-S depletion calculations. In order to accommodate the simulation of dynamic refueling and processing scheme, an in-house program REFRESH together with a run script are developed for carrying out a series of stepwise TRITON calculations, that makes the work of analyzing the neutronic properties and performance of a MSR core design easier. As a demonstration and cross check, we have applied this method to reexamine the conceptual design of Molten Salt Actinide Recycler & Transmuter (MOSART). This paper summarizes the development of the method and preliminary results of its application on MOSART. (author)

  5. Actinide transmutation in nuclear reactors

    International Nuclear Information System (INIS)

    Bultman, J.H.

    1995-01-01

    An optimization method is developed to maximize the burning capability of the ALMR while complying with all constraints imposed on the design for reliability and safety. This method leads to a maximal transuranics enrichment, which is being limited by constraints on reactivity. The enrichment can be raised by using the neutrons less efficiently by increasing leakage from the fuel. With the developed optimization method, a metallic and an oxide fueled ALMR were optimized. Both reactors perform equally well considering the burning of transuranics. However, metallic fuel has a much higher heat conductivity coefficient, which in general leads to better safety characteristics. In search of a more effective waste transmuter, a modified Molten Salt Reactor was designed. A MSR operates on a liquid fuel salt which makes continuous refueling possible, eliminating the issue of the burnup reactivity loss. Also, a prompt negative reactivity feedback is possible for an overmoderated reactor design, even when the Doppler coefficient is positive, due to the fuel expansion with fuel temperature increase. Furthermore, the molten salt fuel can be reprocessed based on a reduction process which is not sensitive to the short-lived spontaneously fissioning actinides. (orig./HP)

  6. Actinide transmutation in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bultman, J H

    1995-01-17

    An optimization method is developed to maximize the burning capability of the ALMR while complying with all constraints imposed on the design for reliability and safety. This method leads to a maximal transuranics enrichment, which is being limited by constraints on reactivity. The enrichment can be raised by using the neutrons less efficiently by increasing leakage from the fuel. With the developed optimization method, a metallic and an oxide fueled ALMR were optimized. Both reactors perform equally well considering the burning of transuranics. However, metallic fuel has a much higher heat conductivity coefficient, which in general leads to better safety characteristics. In search of a more effective waste transmuter, a modified Molten Salt Reactor was designed. A MSR operates on a liquid fuel salt which makes continuous refueling possible, eliminating the issue of the burnup reactivity loss. Also, a prompt negative reactivity feedback is possible for an overmoderated reactor design, even when the Doppler coefficient is positive, due to the fuel expansion with fuel temperature increase. Furthermore, the molten salt fuel can be reprocessed based on a reduction process which is not sensitive to the short-lived spontaneously fissioning actinides. (orig./HP).

  7. Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle descriptions

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-01

    The Nonproliferation Alternative Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates. Volume IX is divided into three sections: Chapter 1, Reactor Systems; Chapter 2, Fuel-Cycle Systems; and the Appendixes. Chapter 1 contains the characterizations of the following 12 reactor types: light-water reactor; heavy-water reactor; water-cooled breeder reactor; high-temperature gas-cooled reactor; gas-cooled fast reactor; liquid-metal fast breeder reactor; spectral-shift-controlled reactor; accelerator-driven reactor; molten-salt reactor; gaseous-core reactor; tokamak fusion-fisson hybrid reactor; and fast mixed-spectrum reactor. Chapter 2 contains similar information developed for fuel-cycle facilities in the following categories: mining and milling; conversion and enrichment; fuel fabrication; spent fuel reprocessing; waste handling and disposal; and transportation of nuclear materials.

  8. Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle descriptions

    International Nuclear Information System (INIS)

    1979-12-01

    The Nonproliferation Alternative Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates. Volume IX is divided into three sections: Chapter 1, Reactor Systems; Chapter 2, Fuel-Cycle Systems; and the Appendixes. Chapter 1 contains the characterizations of the following 12 reactor types: light-water reactor; heavy-water reactor; water-cooled breeder reactor; high-temperature gas-cooled reactor; gas-cooled fast reactor; liquid-metal fast breeder reactor; spectral-shift-controlled reactor; accelerator-driven reactor; molten-salt reactor; gaseous-core reactor; tokamak fusion-fisson hybrid reactor; and fast mixed-spectrum reactor. Chapter 2 contains similar information developed for fuel-cycle facilities in the following categories: mining and milling; conversion and enrichment; fuel fabrication; spent fuel reprocessing; waste handling and disposal; and transportation of nuclear materials

  9. Preliminary analysis of basic reactor physics of the Dual Fluid Reactor - 15270

    International Nuclear Information System (INIS)

    Wang, X.; Macian-Juan, R.; Seidl, M.

    2015-01-01

    The Dual Fluid Reactor (DFR) is a novel fast nuclear reactor concept invented by the IFK based on the Generation IV Molten Salt Reactor and the Liquid Metal Cooled Reactor. The DFR uses a chloride based molten fuel salt in order to harden the neutron spectrum. The molten fuel salt is cooled with a separated liquid lead loop, which in principle allows for higher power densities and better breeding performance. The DFR does not combine heat removal and breeding into a single circuit but separates the two functions into two independent circuits. Since there are attractive features mentioned in this design, the main task of this paper is to verify the model of the whole reactor based on this concept. For this purpose several calculations are presented, including steady state calculations, sensitivity calculations with regard to the nuclide cross sections, the temperature and geometry coefficient of k eff as well as the burnup calculation. The Monte Carlo calculation codes MCNP, SERPENT and SCALE are used for the analysis. As expected the study shows a significant negative reactivity feedback with temperature in the overall fission zone. For the coupled coolant and reflector design the temperature feedback is rather small for practical purposes such as reactor control during normal operation. In the view of these results the DFR in principle can be self-regulated totally by the temperature change of its own fuel salt and consequently can rely on fully passive safety systems for accident management

  10. PRE design of a molten salt thorium reactor loop

    International Nuclear Information System (INIS)

    Caire, Jean-Pierre; Roure, Anthony

    2007-01-01

    This study is a contribution to the 2004 PCR-RSF program of the Centre National de la Recherche Scientifique (CNRS) devoted to research on high temperature thorium molten salt reactors. A major issue of high temperature molten salt reactors is the very large heat duty to be transferred from primary to secondary loop of the reactor with minimal thermal losses. A possible inner loop made of a series of conventional graphite filter plate exchangers, pipes and pumps was investigated. The loop was assumed to use two counter current flows of the same LiF, BeF 2 , ZrF 4 , UF 4 molten salt flowing through the reactor. The 3D model used the coupling of k-ε turbulent Navier-Stokes equations and thermal applications of the Heat Transfer module of COMSOL Multiphysics. For a reactor delivering 2700 MWth, the model required a set of 114 identical exchangers. Each one was optimized to limit the heat losses to 2882 W. The pipes made of a succession of graphite, ceramics, Hastelloy-N alloy and insulating Microtherm layers led to a thermal loss limited to 550 W per linear meter. In such conditions, the global thermal losses represent only 0.013% of the reactor thermal power for elements covered with an insulator only 3 cm thick. (author)

  11. Quantities of actinides in nuclear reactor fuel cycles

    International Nuclear Information System (INIS)

    Ang, K.P.

    1975-01-01

    The quantities of plutonium and other fuel actinides have been calculated for equilibrium fuel cycles for 1000 MW reactors of the following types: water reactors fueled with slightly enriched uranium, water reactors fueled with plutonium and natural uranium, fast-breeder reactors, gas-cooled reactors fueled with thorium and highly enriched uranium, and gas-cooled reactors fueled with thorium, plutonium, and recycled uranium. The radioactivity levels of plutonium, americium, and curium processed yearly in these fuel cycles are greatest for the water reactors fueled with natural uranium and recycled plutonium. The total amount of actinides processed is calculated for the predicted future growth of the United States nuclear power industry. For the same total installed nuclear power capacity, the introduction of the plutonium breeder has little effect upon the total amount of plutonium processed in this century. The estimated amount of plutonium in the low-level process wastes in the plutonium fuel cycles is comparable to the amount of plutonium in the high-level fission product wastes. The amount of plutonium processed in the nuclear fuel cycles can be considerably reduced by using gas-cooled reactors to consume plutonium produced in uranium-fueled water reactors. These, and other reactors dedicated for plutonium utilization, could be co-located with facilities for fuel reprocessing and fuel fabrication to eliminate the off-site transport of separated plutonium. (U.S.)

  12. Fuel exchange device for FBR type reactor

    International Nuclear Information System (INIS)

    Onuki, Koji.

    1993-01-01

    The device of the present invention can provide fresh fuels with a rotational angle aligned with the direction in the reactor core, so that the fresh fuels can be inserted being aligned with apertures of the reactor core even if a self orientation mechanism should fail to operate. That is, a rotational angle detection means (1) detects the rotational angle of fresh fuels before insertion to the reactor core. A fuel rotational angle control means (2) controls the rotational angle of the fresh fuels by comparing the detection result of the means (1) and the data for the insertion position of the reactor core. A fuel rotation means (3) compensates the rotational angel of the fresh fuels based on the control signal from the means (2). In this way, when the fresh fuels are inserted to the reactor core, the fresh fuels set at the same angle as that for the aperture of the reactor core. Accordingly, even if the self orientation mechanism should not operate, the fresh fuels can be inserted smoothly. As a result, it is possible to save loss time upon fuel exchange and mitigate operator's burden during operation. (I.S.)

  13. Molten salt reactors. Synthesis of studies realized between 1973 and 1983. General synthesis

    International Nuclear Information System (INIS)

    Hery, M.; Lecocq, A.

    1983-03-01

    After a brief recall of the MSBR project, French studies on molten salt reactors are summed up. Theoretical and experimental studies for a graphite moderated 1000 MWe reactor using molten Li, Be, Th and U fluorides cooled by salt-lead direct contact are given. These studies concern the core, molten salt chemistry, graphite, metals (molybdenum, alloy TZM), corrosion, reactor components [fr

  14. Challenges Related to the Use of Liquid Metal and Molten Salt Coolants in Advanced Reactors. Report of the collaborative project COOL of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    2013-05-01

    The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was launched in 2000, based on a resolution by the IAEA General Conference (GC(44)/RES/21). INPRO aims at helping to ensure that nuclear energy is available in the twenty-first century in a sustainable manner, and seeks to bring together all interested Member States, both technology holders and technology users, to jointly consider actions to achieve desired innovations. INPRO is taking care of the specific needs of developing countries. One of the aims of INPRO is to develop options for enhanced sustainability through promotion of technical and institutional innovations in nuclear energy technology through collaborative projects among IAEA Member States. Collaboration among INPRO members is fostered on selected innovative nuclear technologies to bridge technology gaps. Collaborative projects have been selected so that they complement other national and international R and D activities. The INPRO Collaborative Project COOL on Investigation of Technological Challenges Related to the Removal of Heat by Liquid Metal and Molten Salt Coolants from Reactor Cores Operating at High Temperatures investigated the technological challenges of cooling reactor cores that operate at high temperatures in advanced fast reactors, high temperature reactors and accelerator driven systems by using liquid metals and molten salts as coolants. The project was initiated in 2008 and was led by India; experts from Brazil, China, Germany, India, Italy and the Republic of Korea participated and provided chapters of this report. The INPRO Collaborative Project COOL addressed the following fields of research regarding liquid metal and molten salt coolants: (i) survey of thermophysical properties; (ii) experimental investigations and computational fluid dynamics studies on thermohydraulics, specifically pressure drop and heat transfer under different operating conditions; (iii) monitoring and control of coolant

  15. United States Domestic Research Reactor Infrastructure TRIGA Reactor Fuel Support

    International Nuclear Information System (INIS)

    Morrell, Douglas

    2011-01-01

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  16. Economic evaluation of fast reactor fuel cycling

    International Nuclear Information System (INIS)

    Hu Ping; Zhao Fuyu; Yan Zhou; Li Chong

    2012-01-01

    Economic calculation and analysis of two kinds of nuclear fuel cycle are conducted by check off method, based on the nuclear fuel cycling process and model for fast reactor power plant, and comparison is carried out for the economy of fast reactor fuel cycle and PWR once-through fuel cycle. Calculated based on the current price level, the economy of PWR one-through fuel cycle is better than that of the fast reactor fuel cycle. However, in the long term considering the rising of the natural uranium's price and the development of the post treatment technology for nuclear fuels, the cost of the fast reactor fuel cycle is expected to match or lower than that of the PWR once-through fuel cycle. (authors)

  17. Mechanical structure and problem of thorium molten salt reactor

    International Nuclear Information System (INIS)

    Kamei, Takashi

    2011-01-01

    After Fukushima Daiichi accident, there became great interest in Thorium Molten Salt Reactor (MSR) for the safety as station blackout leading to auto drainage of molten salts with freeze valve. This article described mechanical structure of MSR and problems of materials and pipes. Material corrosion problem by molten salts would be solved using modified Hastelloy N with Ti and Nb added, which should be confirmed by operation of an experimental reactor. Trends in international activities of MSR were also referred including China declaring MSR development in January 2011 to solve thorium contamination issues at rare earth production and India rich in thorium resources. (T. Tanaka)

  18. Fast reactor fuel reprocessing in the UK

    International Nuclear Information System (INIS)

    Allardice, R.H.; Williams, J.; Buck, C.

    1977-01-01

    Enriched uranium metal fuel irradiated in the Dounreay Fast Reactor has been reprocessed and refabricated in plants specifically designed for the purpose in the U.K. since 1961. Efficient and reliable fuel recycle is essential to the development of a plutonium based fast reactor system and the importance of establishing at an early stage fast reactor fuel reprocessing has been reinforced by current world difficulties in reprocessing high burn-up thermal reactor oxide fuel. In consequence, the U.K. has decided to reprocess irradiated fuel from the 250 MW(E) Prototype Fast Reactor as an integral part of the fast reactor development programme. Flowsheet and equipment development work for the small scale fully active demonstration plant have been carried out over the past 5 years and the plant will be commissioned and ready for active operation during 1977. In parallel, a comprehensive waste management system has been developed and installed. Based on this development work and the information which will arise from active operation of the plant a parallel development programme has been initiated to provide the basis for the design of a large scale fast reactor fuel reprocessing plant to come into operation in the late 1980s to support the projected U.K. fast reactor installation programme. The paper identifies the important differences between fast reactor and thermal reactor fuel reprocessing technologies and describes some of the development work carried out in these areas for the small scale P.F.R. fuel reprocessing operation. In addition, the development programme in aid of the design of a larger scale fast reactor fuel reprocessing plant is outlined and the current design philosophy is discussed

  19. Fuel assembly for FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki.

    1995-01-01

    Ordinary sodium bond-type fuel pins using nitride fuels, carbide fuels or metal fuels and pins incorporated with hydride moderators are loaded in a wrapper tube at a ratio of from 2 to 10% based on the total number of fuel pins. The hydride moderators are sealed in the hydride moderator incorporated pins at the position only for a range from the upper end to a reactor core upper position of substantially 1/4 of the height of the reactor core from the upper end of the reactor core as a center. Then, even upon occurrence of ULOF (loss of flow rate scram failure phenomenon), it gives characteristic of reducing the power only by a doppler coefficient and not causing boiling of coolant sodium but providing stable cooling to the reactor core. Therefore, a way of thinking on the assurance of passive safety is simplified to make a verification including on the reactor structure unnecessary. In an LMFBR type reactor using the fuel assembly, a critical experiment for confirming accuracy of nuclear design is sufficient for the item required for study and development, which provides a great economical effect. (N.H.)

  20. Molt salts reactors capacity for wastes incineration and energy production

    International Nuclear Information System (INIS)

    David, S.; Nuttin, A.

    2005-01-01

    The molten salt reactors present many advantages in the framework of the IV generation systems development for the energy production and/or the wastes incineration. After a recall of the main studies realized on the molten salt reactors, this document presents the new concepts and the identified research axis: the MSRE project and experience, the incinerators concepts, the thorium cycle. (A.L.B.)

  1. Engineering development studies for molten-salt breeder reactor processing No. 18

    International Nuclear Information System (INIS)

    Hightower, J.R. Jr.

    1975-03-01

    A water--mercury system was used to study the effect of geometric variations on mass transfer rates in rectangular contractors similar to those proposed for the molten-salt breeder reactor (MSBR) fuel reprocessing scheme. Since mass transfer rates were not accurately predicted by the Lewis correlation, other correlations were investigated. A correlation which was found to fit the experimental results is given. Mass transfer rates are being measured in a fluoride salt--bismuth contactor. Experimental results indicate that the mass transfer rates in the salt--bismuth system fall between the Lewis correlation and the modified correlation given above. Autoresistance heating tests were continued in the fluorinator mock-up using LiF--BeF 2 --ThF 4 (72-16-12 mole percent) salt. The equipment was returned to operating condition, and five experiments were run. Although correct steady-state operation was not achieved, the results were encouraging. A two-dimensional electrical analog was constructed to study current flow through the electrode sidearm and other critical areas of the test vessel. These studies indicate that no regions of abnormally high current density existed in the first nine runs with the present autoresistance heating equipment. Localized heating had previously been the suspected cause for the failure to achieve proper operation of this equipment. (U.S.)

  2. Russian research reactor fuel return program starts shipping fuel to Russia

    International Nuclear Information System (INIS)

    Dedik, T.; Bolshinsky, I.; Krass, A.

    2003-01-01

    For almost four years the United States (U.S), the Russian Federation (R.F.), and the International Atomic Energy Agency (IAEA) have been discussing an initiative to return Soviet/Russian-origin research reactor fuel to the Russian Federation. In a series of bilateral and trilateral meetings in Vienna and Moscow, considerable progress has been made toward defining the Russian Research Reactor Fuel Return Program as well as obtaining the necessary technical data to facilitate the return. More than 20 research reactors in 17 countries that have Soviet- or Russian-supplied fuel have identified. Most of these reactors have stocks of both fresh and irradiated HEU fuel that must be carefully stored and managed for many years to come. On September 21, 2003 the Russian Research Reactor Fuel Return program shipped 14 kg of fresh Russian-origin HEU fuel from Romania to the nuclear fuel fabrication facility in Russia, which represented the beginning of the practical implementation of the program. (author)

  3. Fuel assembly and reactor core

    International Nuclear Information System (INIS)

    Yuchi, Yoko; Aoyama, Motoo; Haikawa, Katsumasa; Yamanaka, Akihiro; Koyama, Jun-ichi.

    1996-01-01

    In a fuel assembly of a BWR type reactor, a region substantially containing burnable poison is divided into an upper region and a lower region having different average concentrations of burnable poison along a transverse cross section perpendicular to the axial direction. The ratio of burnable poison contents of both regions is determined to not more than 80%, and the average concentration of the burnable poison in the lower region is determined to not less than 9% by weight. An infinite multiplication factor at an initial stage of the burning of the fuel assembly is controlled effectively by the burnable poisons. Namely, the ratio of the axial power can be controlled by the distribution of the enrichment degree of uranium fuels and the distribution of the burnable poison concentration in the axial direction. Since the average enrichment degree of the reactor core has to be increased in order to provide an initially loaded reactor core at high burnup degree. Distortion of the power distribution in the axial direction of the reactor core to which fuel assemblies at high enrichment degree are loaded is flattened to improve thermal margin, to extend continuous operation period and increase a burnup degree upon take-out thereby improving fuel economy without worsening the reactor core characteristics of the initially loaded reactor core. (N.H.)

  4. Facilities of fuel transfer for nuclear reactors

    International Nuclear Information System (INIS)

    Wade, E.E.

    1977-01-01

    This invention relates to sodium cooled fast breeder reactors. It particularly concerns facilities for the transfer of fuel assemblies between the reactor core and a fuel transfer area. The installation is simple in construction and enables a relatively small vessel to be used. In greater detail, the invention includes a vessel with a head, fuel assemblies housed in this vessel, and an inlet and outlet for the coolant covering these fuel assemblies. The reactor has a fuel transfer area in communication with this vessel and gear inside the vessel for the transfer of these fuel assemblies. These facilities are borne by the vessel head and serve to transfer the fuel assemblies from the vessel to the transfer area; whilst leaving the fuel assemblies completely immersed in a continuous mass of coolant. A passageway is provided between the vessel and this transfer area for the fuel assemblies. Facilities are provided for closing off this passageway so that the inside of the reactor vessel may be isolated as desired from this fuel transfer area whilst the reactor is operating [fr

  5. BR2 Reactor: Irradiation of fuels

    International Nuclear Information System (INIS)

    Verwimp, A.

    2005-01-01

    Safe, reliable and economical operation of reactor fuels, both UO 2 and MOX types, requires in-pile testing and qualification up to high target burn-up levels. In-pile testing of advanced fuels for improved performance is also mandatory. The objectives of research performed at SCK-CEN are to perform Neutron irradiation of LWR (Light Water Reactor) fuels in the BR2 reactor under relevant operating and monitoring conditions, as specified by the experimenter's requirements and to improve the on-line measurements on the fuel rods themselves

  6. Probability safety assessment of LOOP accident to molten salt reactor

    International Nuclear Information System (INIS)

    Mei Mudan; Shao Shiwei; Yu Zhizhen; Chen Kun; Zuo Jiaxu

    2013-01-01

    Background: Loss of offsite power (LOOP) is a possible accident to any type of reactor, and this accident can reflect the main idea of reactor safety design. Therefore, it is very important to conduct a study on probabilistic safety assessment (PSA) of the molten salt reactor that is under LOOP circumstance. Purpose: The aim is to calculate the release frequency of molten salt radioactive material to the core caused by LOOP, and find out the biggest contributor to causing the radioactive release frequency. Methods: We carried out the PSA analysis of the LOOP using the PSA process risk spectrum, and assumed that the primary circuit had no valve and equipment reliability data based on the existing mature power plant equipment reliability data. Results: Through the PSA analysis, we got the accident sequences of the release of radioactive material to the core caused by LOOP and its frequency. The results show that the release frequency of molten salt radioactive material to the core caused by LOOP is about 2×10 -11 /(reactor ·year), which is far below that of the AP1000 LOOP. In addition, through the quantitative analysis, we obtained the point estimation and interval estimation of uncertainty analysis, and found that the biggest contributor to cause the release frequency of radioactive material to the core is the reactor cavity cooling function failure. Conclusion: This study provides effective help for the design and improvement of the following molten salt reactor system. (authors)

  7. Mechanism study of freeze-valve for molten salt reactor (MSR)

    International Nuclear Information System (INIS)

    Qinhua, Zhang

    2014-01-01

    Molten salt reactor (MSR) is one of the fourth generation nuclear reactor, ordinary nuclear grade valve is unsuitable for MSR due to its special coolant and extraordinary working temperature. Freeze-valve is proposed as the most appropriate valve for MSR, but the technology issue about freeze-valve has not been report in recent decades. Its significance to test the comprehensive property of freeze-valve for the application in MSR. A high temperature molten salt test loop was built which the physics property of salt is similar to the coolant of MSR. The results indicate that freeze-valve has a good performance use in the molten salt circumstances of high temperature (max 700 deg. C) and strong corrosion (authors)

  8. Build-up and decay of fuel actinides in the fuel cycle of nuclear reactors

    International Nuclear Information System (INIS)

    Tasaka, Kanji; Kikuchi, Yasuyuki; Shindo, Ryuichi; Yoshida, Hiroyuki; Yasukawa, Shigeru

    1976-05-01

    For boiling water reactors, pressurized light-water reactors, pressure-tube-type heavy water reactors, high-temperature gas-cooled reactors, and sodium-cooled fast breeder reactors, uranium fueled and mixed-oxide fueled, each of 1000 MWe, the following have been studied: (1) quantities of plutonium and other fuel actinides built up in the reactor, (2) cooling behaviors of activities of plutonium and other fuel actinides in the spent fuels, and (3) activities of plutonium and other fuel actinides in the high-level reprocessing wastes as a function of storage time. The neutron cross section and decay data of respective actinide nuclides are presented, with their evaluations. For effective utilization of the uranium resources and easy reprocessing and high-level waste management, a thermal reactor must be fueled with uranium; the plutonium produced in a thermal reactor should be used in a fast reactor; and the plutonium produced in the blanket of a fast reactor is more appropriate for a fast reactor than that from a thermal reactor. (auth.)

  9. Efficiency of an LBE spallation target in an accelerator-driven molten salt subcritical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bak, Sang-In [Sungkyunkwan University, Suwon (Korea, Republic of); Hong, Seung-Woo [Sungkyunkwan University, Suwon (Korea, Republic of); Kadi, Yacine [CERN, Geneva (Switzerland)

    2016-10-15

    An Accelerator-Driven System (ADS) combined with a subcritical Molten Salt Reactor (MSR) is a type of hybrid reactor originally designed to breed uranium from thorium or to incinerate long-lived minor actinides in nuclear wastes. In an MSR, the salt material is used not only as a nuclear fuel but also as a primary coolant. In addition, this material is used as a target for inducing spallation neutrons in most AD-MSR concepts. A high energy proton beam impinges on a heavy metal target to induce spallation reactions and produces neutrons. Accordingly, a reliable proton accelerator is needed to feed the source neutrons. As ADSs have been criticized for requiring high power accelerators, minimization of beam power is an important aspect of ADS design. A primary concern associated with ADS development is stable high-power accelerators. We therefore studied the neutron source efficiencies of an AD-MSR involving chloride fuels by including a Pb-Bi eutectic (LBE) spallation target. The proton source efficiency and the accelerator beam power required have been studied for an AD-MSR. Adoption of an LBE spallation target induces an increase in proton source efficiencies in comparison to the case without a spallation target. Thus the presence of an efficient spallation target is useful in the reduction of the beam power of an accelerator. Almost 33 % of the beam power can be reduced in comparison to the case without the target for NaCl-Th/{sup 233}U fuel, and about 16 % for NaCl-U/TRU fuel. The beam power amplifications increase by 1.5 times for NaCl-Th/{sup 233}U and 1.2 times for NaCl-U/TRU in comparison with the no target AD-MSR.

  10. Fuel Fabrication and Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-02

    The uranium from the enrichment plant is still in the form of UF6. UF6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF6 is converted into UO2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

  11. CANDU reactor experience: fuel performance

    International Nuclear Information System (INIS)

    Truant, P.T.; Hastings, I.J.

    1985-07-01

    Ontario Hydro has more than 126 reactor-years experience in operating CANDU reactors. Fuel performance has been excellent with 47 000 channel fuelling operations successfully completed and 99.9 percent of the more than 380 000 bundles irradiated operating as designed. Fuel performance limits and fuel defects have had a negligible effect on station safety, reliability, the environment and cost. The actual incapability charged to fuel is less than 0.1 percent over the stations' lifetimes, and more recently has been zero

  12. Estimates of relative areas for the disposal in bedded salt of LWR wastes from alternative fuel cycles

    International Nuclear Information System (INIS)

    Lincoln, R.C.; Larson, D.W.; Sisson, C.E.

    1978-01-01

    The relative mine-level areas (land use requirements) which would be required for the disposal of light-water reactor (LWR) radioactive wastes in a hypothetical bedded-salt formation have been estimated. Five waste types from alternative fuel cycles have been considered. The relative thermal response of each of five different site conditions to each waste type has been determined. The fuel cycles considered are the once-through (no recycle), the uranium-only recycle, and the uranium and plutonium recycle. The waste types which were considered include (1) unreprocessed spent reactor fuel, (2) solidified waste derived from reprocessing uranium oxide fuel, (3) plutonium recovered from reprocessing spent reactor fuel and doped with 1.5% of the accompanying waste from reprocessing uranium oxide fuel, (4) waste derived from reprocessing mixed uranium/plutonium oxide fuel in the third recycle, and (5) unreprocessed spent fuel after three recycles of mixed uranium/plutonium oxide fuels. The relative waste-disposal areas were determined from a calculated value of maximum thermal energy (MTE) content of the geologic formations. Results are presented for each geologic site condition in terms of area ratios. Disposal area requirements for each waste type are expressed as ratios relative to the smallest area requirement (for waste type No. 2 above). For the reference geologic site condition, the estimated mine-level disposal area ratios are 4.9 for waste type No. 1, 4.3 for No. 3, 2.6 for No. 4, and 11 for No. 5

  13. Fuel handling system of nuclear reactor plants

    International Nuclear Information System (INIS)

    Faulstich, D.L.

    1991-01-01

    This patent describes a fuel handing system for nuclear reactor plants comprising a reactor vessel having an openable top and removable cover for refueling and containing therein, submerged in coolant water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units. It comprises a fuel bundle handing platform moveable over the open top of the reactor vessel; a fuel bundle handing mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grappling hook means for attaching to and transporting fuel bundles into and out from the fuel core; and a camera with a prismatic viewing head surrounded by a radioactive resisting quartz cylinder and enclosed within the grapple head which is provided with at least three windows with at least two windows provided with an angled surface for aiming the camera prismatic viewing head in different directions and thereby viewing the fuel bundles of the fuel core from different perspectives, and having a cable connecting the camera with a viewing monitor located above the reactor vessel for observing the fuel bundles of the fuel core and for enabling aiming of the camera prismatic viewing head through the windows by an operator

  14. The uranium-plutonium breeder reactor fuel cycle

    International Nuclear Information System (INIS)

    Salmon, A.; Allardice, R.H.

    1979-01-01

    All power-producing systems have an associated fuel cycle covering the history of the fuel from its source to its eventual sink. Most, if not all, of the processes of extraction, preparation, generation, reprocessing, waste treatment and transportation are involved. With thermal nuclear reactors more than one fuel cycle is possible, however it is probable that the uranium-plutonium fuel cycle will become predominant; in this cycle the fuel is mined, usually enriched, fabricated, used and then reprocessed. The useful components of the fuel, the uranium and the plutonium, are then available for further use, the waste products are treated and disposed of safely. This particular thermal reactor fuel cycle is essential if the fast breeder reactor (FBR) using plutonium as its major fuel is to be used in a power-producing system, because it provides the necessary initial plutonium to get the system started. In this paper the authors only consider the FBR using plutonium as its major fuel, at present it is the type envisaged in all, current national plans for FBR power systems. The corresponding fuel cycle, the uranium-plutonium breeder reactor fuel cycle, is basically the same as the thermal reactor fuel cycle - the fuel is used and then reprocessed to separate the useful components from the waste products, the useful uranium and plutonium are used again and the waste disposed of safely. However the details of the cycle are significantly different from those of the thermal reactor cycle. (Auth.)

  15. Transient core characteristics of small molten salt reactor coupling problem between heat transfer/flow and nuclear fission reaction

    International Nuclear Information System (INIS)

    Yamamoto, Takahisa; Mitachi, Koshi

    2004-01-01

    This paper performed the transient core analysis of a small Molten Salt Reactor (MSR). The emphasis is that the numerical model employed in this paper takes into account the interaction among fuel salt flow, nuclear reaction and heat transfer. The model consists of two group diffusion equations for fast and thermal neutron fluexs, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The results of transient analysis are that (1) fission reaction (heat generation) rate significantly increases soon after step reactivity insertion, e.g., the peak of fission reaction rate achieves about 2.7 times larger than the rated power 350 MW when the reactivity of 0.15% Δk/k 0 is inserted to the rated state, and (2) the self-control performance of the small MSR effectively works under the step reactivity insertion of 0.56% Δk/k 0 , putting the fission reaction rate back on the rated state. (author)

  16. The Canadian research reactor spent fuel situation

    International Nuclear Information System (INIS)

    Ernst, P.C.

    1996-01-01

    This paper summarizes the present research reactor spent fuel situation in Canada. The research reactors currently operating are listed along with the types of fuel that they utilize. Other shut down research reactors contributing to the storage volume are included for completeness. The spent fuel storage facilities associated with these reactors and the methods used to determine criticality safety are described. Finally the current inventory of spent fuel and where it is stored is presented along with concerns for future storage. (author). 3 figs

  17. Candu reactors with thorium fuel cycles

    International Nuclear Information System (INIS)

    Hopwood, J.M.; Fehrenbach, P.; Duffey, R.; Kuran, S.; Ivanco, M.; Dyck, G.R.; Chan, P.S.W.; Tyagi, A.K.; Mancuso, C.

    2006-01-01

    Over the last decade and a half AECL has established a strong record of delivering CANDU 6 nuclear power plants on time and at budget. Inherently flexible features of the CANDU type reactors, such as on-power fuelling, high neutron economy, fuel channel based heat transport system, simple fuel bundle configuration, two independent shut down systems, a cool moderator and a defence-in-depth based safety philosophy provides an evolutionary path to further improvements in design. The immediate milestone on this path is the Advanced CANDU ReactorTM** (ACRTM**), in the form of the ACR-1000TM**. This effort is being followed by the Super Critical Water Reactor (SCWR) design that will allow water-cooled reactors to attain high efficiencies by increasing the coolant temperature above 550 0 C. Adaptability of the CANDU design to different fuel cycles is another technology advantage that offers an additional avenue for design evolution. Thorium is one of the potential fuels for future reactors due to relative abundance, neutronics advantage as a fertile material in thermal reactors and proliferation resistance. The Thorium fuel cycle is also of interest to China, India, and Turkey due to local abundance that can ensure sustainable energy independence over the long term. AECL has performed an assessment of both CANDU 6 and ACR-1000 designs to identify systems, components, safety features and operational processes that may need to be modified to replace the NU or SEU fuel cycles with one based on Thorium. The paper reviews some of these requirements and the associated practical design solutions. These modifications can either be incorporated into the design prior to construction or, for currently operational reactors, during a refurbishment outage. In parallel with reactor modifications, various Thorium fuel cycles, either based on mixed bundles (homogeneous) or mixed channels (heterogeneous) have been assessed for technical and economic viability. Potential applications of a

  18. Immobilization of IFR salt wastes in mortar

    International Nuclear Information System (INIS)

    Fischer, D.F.; Johnson, T.R.

    1988-01-01

    Portland cement-base mortars are being considered for immobilizing chloride salt wastes produced by the fuel cycles of Integral Fast Reactors (IFR). The IFR is a sodium-cooled fast reactor with metal alloy fuels. It has a close-coupled fuel cycle in which fission products are separated from the actinides in an electrochemical cell operating at 500/degree/C. This cell has a liquid cadmium anode in which the fuels are dissolved and a liquid salt electrolyte. The salt will be a mixture of either lithium, potassium, and sodium chlorides or lithium, calcium, barium, and sodium chlorides. One method being considered for immobilizing the treated nontransuranic salt waste is to disperse the salt in a portland cement-base mortar that will be sealed in corrosion-resistant containers. For this application, the grout must be sufficiently fluid that it can be pumped into canister-molds where it will solidify into a strong, leach-resistant material. The set times must be longer than a few hours to allow sufficient time for processing, and the mortar must reach a reasonable compressive strength (/approximately/7 MPa) within three days to permit handling. Because fission product heating will be high, about 0.6 W/kg for a mortar containing 10% waste salt, the effects of elevated temperatures during curing and storage on mortar properties must be considered

  19. Proposed fuel cycle for the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Burris, L.; Walters, L.C.

    1985-01-01

    One of the key features of ANL's Integral Fast Reactor (IFR) concept is a close-coupled fuel cycle. The proposed fuel cycle is similar to that demonstrated over the first five to six years of operation of EBR-II, when a fuel cycle facility adjacent to EBR-II was operated to reprocess and refabricate rapidly fuel discharged from the EBR-II. Locating the IFR and its fuel cycle facility on the same site makes the IFR a self-contained system. Because the reactor fuel and the uranium blanket are metals, pyrometallurgical processes (shortned to ''pyroprocesses'') have been chosen. The objectives of the IFR processes for the reactor fuel and blanket materials are to (1) recover fissionable materials in high yield; (2) remove fission products adequately from the reactor fuel, e.g., a decontamination factor of 10 to 100; and (3) upgrade the concentration of plutonium in uranium sufficiently to replenish the fissile-material content of the reactor fuel. After the fuel has been reconstituted, new fuel elements will be fabricated for recycle to the reactor

  20. Fuel-management simulations for once-through thorium fuel cycle in CANDU reactors

    International Nuclear Information System (INIS)

    Chan, P.S.W.; Boczar, P.G.; Ellis, R.J.; Ardeshiri, F.

    1999-01-01

    High neutron economy, on-power refuelling and a simple fuel bundle design result in unsurpassed fuel cycle flexibility for CANDU reactors. These features facilitate the introduction and exploitation of thorium fuel cycles in existing CANDU reactors in an evolutionary fashion. Detailed full-core fuel-management simulations concluded that a once-through thorium fuel cycle can be successfully implemented in an existing CANDU reactor without requiring major modifications. (author)

  1. Analysis of molten salt thermal-hydraulics using computational fluid dynamics

    International Nuclear Information System (INIS)

    Yamaji, B.; Csom, G.; Aszodi, A.

    2003-01-01

    To give a good solution for the problem of high level radioactive waste partitioning and transmutation is expected to be a pro missing option. Application of this technology also could extend the possibilities of nuclear energy. Large number of liquid-fuelled reactor concepts or accelerator driven subcritical systems was proposed as transmutors. Several of these consider fluoride based molten salts as the liquid fuel and coolant medium. The thermal-hydraulic behaviour of these systems is expected to be fundamentally different than the behaviour of widely used water-cooled reactors with solid fuel. Considering large flow domains three-dimensional thermal-hydraulic analysis is the method seeming to be applicable. Since the fuel is the coolant medium as well, one can expect a strong coupling between neutronics and thermal-hydraulics too. In the present paper the application of Computational Fluid Dynamics for three-dimensional thermal-hydraulics simulations of molten salt reactor concepts is introduced. In our past and recent works several calculations were carried out to investigate the capabilities of Computational Fluid Dynamics through the analysis of different molten salt reactor concepts. Homogenous single region molten salt reactor concept is studied and optimised. Another single region reactor concept is introduced also. This concept has internal heat exchanges in the flow domain and the molten salt is circulated by natural convection. The analysis of the MSRE experiment is also a part of our work since it may form a good background from the validation point of view. In the paper the results of the Computational Fluid Dynamics calculations with these concepts are presented. In the further work our objective is to investigate the thermal-hydraulics of the multi-region molten salt reactor (Authors)

  2. Electrorefining open-quotes Nclose quotes reactor fuel

    International Nuclear Information System (INIS)

    Gay, E.C.; Miller, W.E.

    1995-01-01

    Principles of purifying of uranium metal by electrorefining are reviewed. Metal reactor fuel after irradiation is a form of impure uranium. Dissolution and deposition electrorefining processes were developed for spent metal fuel under the Integral Fast Reactor Program. Application of these processes to the conditioning of spent N-reactor fuel slugs is examined

  3. A review on manufacturing technology for long-lived radionuclide fuel compounds

    International Nuclear Information System (INIS)

    Hwang, Doo Seong; Park, Jin Ho; Kim, Eung Ho; Chung, Won Myung; Lee, Kui Ill; Woo, Moon Sik; Kim, Yeon Ku; Yoo, Jae Hyung

    1998-03-01

    Thermal neutron reactor (LWR), fast neutron reactor (FBR), accelerator-driven subcritical system have been studied as the potential transmutation devices. The fuel types can be classified according to the concept of each reactor. Oxide fuel is considered in LWR and metal, oxide, and nitride fuels are studied in FBR. In accelerator-driven subcritical system molten salt, metal, and oxide fuels are considered. This review focused on characteristics according to transmutation system, and manufacturing technologies of each fuels. Accelerator-driven system is being proposed as the most reasonable concept in recent, since it has merits in terms of stability and free control of nuclides composition rate in charge of long-lived nuclides. Fluorides molten salt fuel is better chemically stable and corrosion resistant, and lower vapor pressure than chloride molten salt and metal in the fuel type of accelerator-driven system. And then the detail manufacturing technology of fluorides molten salt were reviewed. (author). 62 refs., 23 tabs., 37 figs

  4. Nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    2014-01-01

    The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100 th nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U 3 O 8 were replaced by U 3 Si 2 -based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to fulfill its mission that is

  5. Nuclear reactors and fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-01

    The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100{sup th} nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U{sub 3}O{sub 8} were replaced by U{sub 3}Si{sub 2}-based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to

  6. Salt Selection for the LS-VHTR

    International Nuclear Information System (INIS)

    Williams, D.F.; Clarno, K.T.

    2006-01-01

    Molten fluorides were initially developed for use in the nuclear industry as the high temperature fluid-fuel for a Molten Salt Reactor (MSR). The Office of Nuclear Energy is exploring the use of molten fluorides as a primary coolant (rather than helium) in an Advanced High Temperature Reactor (AHTR) design, also know as the Liquid-Salt cooled Very High Temperature Reactor (LS-VHTR). This paper provides a review of relevant properties for use in evaluation and ranking of candidate coolants for the LS-VHTR. Nuclear, physical, and chemical properties were reviewed and metrics for evaluation are recommended. Chemical properties of the salt were examined for the purpose of identifying factors that effect materials compatibility (i.e., corrosion). Some preliminary consideration of economic factors for the candidate salts is also presented. (authors)

  7. Status of research reactor spent fuel world-wide

    International Nuclear Information System (INIS)

    Ritchie, I.G.

    2004-01-01

    Results compiled in the research reactor spent fuel database are used to assess the status of research reactor spent fuel world-wide. Fuel assemblies, their types, enrichment, origin of enrichment and geological distribution among the industrialised and developed countries of the world are discussed. Fuel management practices in wet and dry storage facilities and the concerns of reactor operators about long-term storage of their spent fuel are presented and some of the activities carried out by the International Atomic Energy Agency to address the issues associated with research reactor spent fuel are outlined. (author)

  8. United States Domestic Research Reactor Infrastructure - TRIGA Reactor Fuel Support

    International Nuclear Information System (INIS)

    Morrell, Douglas

    2008-01-01

    The purpose of the United State Domestic Research Reactor Infrastructure Program is to provide fresh nuclear reactor fuel to United States universities at no, or low, cost to the university. The title of the fuel remains with the United States government and when universities are finished with the fuel, the fuel is returned to the United States government. The program is funded by the United States Department of Energy - Nuclear Energy division, managed by Department of Energy - Idaho Field Office, and contracted to the Idaho National Laboratory's Management and Operations Contractor - Battelle Energy Alliance. Program has been at Idaho since 1977 and INL subcontracts with 26 United States domestic reactor facilities (13 TRIGA facilities, 9 plate fuel facilities, 2 AGN facilities, 1 Pulstar fuel facility, 1 Critical facility). University has not shipped fuel since 1968 and as such, we have no present procedures for shipping spent fuel. In addition: floor loading rate is unknown, many interferences must be removed to allow direct access to the reactor tank, floor space in the reactor cell is very limited, pavement ends inside our fence; some of the surface is not finished. The whole approach is narrow, curving and downhill. A truck large enough to transport the cask cannot pull into the lot and then back out (nearly impossible / refused by drivers); a large capacity (100 ton), long boom crane would have to be used due to loading dock obstructions. Access to the entrance door is on a sidewalk. The campus uses it as a road for construction equipment, deliveries and security response. Large trees are on both sides of sidewalk. Spent fuel shipments have never been done, no procedures approved or in place, no approved casks, no accident or safety analysis for spent fuel loading. Any cask assembly used in this facility will have to be removed from one crane, moved on the floor and then attached to another crane to get from the staging area to the reactor room. Reactor

  9. Recent BWR fuel management reactor physics advances

    International Nuclear Information System (INIS)

    Crowther, R.L.; Congdon, S.P.; Crawford, B.W.; Kang, C.M.; Martin, C.L.; Reese, A.P.; Savoia, P.J.; Specker, S.R.; Welchly, R.

    1982-01-01

    Improvements in BWR fuel management have been under development to reduce uranium and separative work (SWU) requirements and reduce fuel cycle costs, while also maintaining maximal capacity factors and high fuel reliability. Improved reactor physics methods are playing an increasingly important role in making such advances feasible. The improved design, process computer and analysis methods both increase knowledge of the thermal margins which are available to implement fuel management advance, and improve the capability to reliably and efficiently analyze and design for fuel management advances. Gamma scan measurements of the power distributions of advanced fuel assembly and advanced reactor core designs, and improved in-core instruments also are important contributors to improving 3-d predictive methods and to increasing thermal margins. This paper is an overview of the recent advances in BWR reactor physics fuel management methods, coupled with fuel management and core design advances. The reactor physics measurements which are required to confirm the predictions of performance fo fuel management advances also are summarized

  10. Main Experimental Results of ISTC-1606 for Recycling and Transmutation in Molten Salt Systems

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, Victor; Feynberg, Olga; Merzlyakov, Aleksandr; Surenkov, Aleksandr [Russian Research Center - Kurchatov Institute, Kurchatov sq. 1, Moscow, RF, 123182 (Russian Federation); Subbotin, Vladimir; Zakirov, Raul; Toropov, Andrey; Panov, Aleksandr [Russian Federal Nuclear Center - Institute of Technical Physics, Snezhinsk (Russian Federation); Afonichkin, Valery [Institute of High-Temperature Electrochemistry, Ekaterinburg (Russian Federation)

    2008-07-01

    To examine and demonstrate the feasibility of molten salt reactors (MSR) to reduce long lived waste toxicity and to produce efficiently electricity in closed fuel cycle, some national and international studies were initiated last years. In this paper main focus is placed on experimental evaluation of single stream Molten Salt Actinide Recycler and Transmuter (MOSART) system fuelled with different compositions of plutonium plus minor actinide trifluoride (AnF{sub 3}) from LWR spent nuclear fuel without U-Th support. This paper summarizes main experimental results of ISTC-1606 related to physical and chemical properties of fuel salt, container materials for fuel circuit, and fuel salt clean up of MOSART system. As result of ISTC-1606 studies claim is made, that the {sup 7}Li,Na,Be/F and {sup 7}Li,Be/F solvents selected for primary system appear to resolve main reactor physics, thermal hydraulics, materials compatibility, fuel salt clean up and safety problems as applied to the MOSART concept development. The created experimental facilities and the database on properties of fuel salt mixtures and container materials are used for a choice and improvement fuel salts and coolants for new applications of this high temperature technology for sustainable nuclear power development. (authors)

  11. Main Experimental Results of ISTC-1606 for Recycling and Transmutation in Molten Salt Systems

    International Nuclear Information System (INIS)

    Ignatiev, Victor; Feynberg, Olga; Merzlyakov, Aleksandr; Surenkov, Aleksandr; Subbotin, Vladimir; Zakirov, Raul; Toropov, Andrey; Panov, Aleksandr; Afonichkin, Valery

    2008-01-01

    To examine and demonstrate the feasibility of molten salt reactors (MSR) to reduce long lived waste toxicity and to produce efficiently electricity in closed fuel cycle, some national and international studies were initiated last years. In this paper main focus is placed on experimental evaluation of single stream Molten Salt Actinide Recycler and Transmuter (MOSART) system fuelled with different compositions of plutonium plus minor actinide trifluoride (AnF 3 ) from LWR spent nuclear fuel without U-Th support. This paper summarizes main experimental results of ISTC-1606 related to physical and chemical properties of fuel salt, container materials for fuel circuit, and fuel salt clean up of MOSART system. As result of ISTC-1606 studies claim is made, that the 7 Li,Na,Be/F and 7 Li,Be/F solvents selected for primary system appear to resolve main reactor physics, thermal hydraulics, materials compatibility, fuel salt clean up and safety problems as applied to the MOSART concept development. The created experimental facilities and the database on properties of fuel salt mixtures and container materials are used for a choice and improvement fuel salts and coolants for new applications of this high temperature technology for sustainable nuclear power development. (authors)

  12. Concept of the demonstration molten salt unit for the transuranium elements transmutations

    International Nuclear Information System (INIS)

    Alekseev, P.; Dudnikov, A.; Prusakov, V.; Subbotin, S.; Zakirov, R.; Lelek, V.; Peka, I.

    1999-01-01

    Fluorine reprocessing is discussed of spent fuel and of fluoride molten salt reactor in critical and subcritical modes for plutonium and minor actinides burning. International collaboration for creation of such system is proposed. Additional neutron source in the core will have positive influence on the transmutation processes in the reactor. Demonstration critical molten salt reactor of small power capacity will permit to decide the most part of problems inherent to large critical reactors and subcritical drivers. It could be expected that fluoride molten salt transmuter can work without accelerator as a critical reactor. (author)

  13. Reactor-specific spent fuel discharge projections, 1987-2020

    International Nuclear Information System (INIS)

    Walling, R.C.; Heeb, C.M.; Purcell, W.L.

    1988-03-01

    The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from U.S. commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water eactors (BWR), and one existing high temperature gas reactor (HTGR). The projections are based on individual reactor information supplied by the U.S. reactor owners. The basic information is adjusted to conform to Energy Information Administration (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: No New Orders (assumes increasing burnup), No New Orders with No Increased Burnup, Upper Reference (assumes increasing burnup), Upper Reference with No Increased Burnup, and Lower Reference (assumes increasing burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum at-reactor storage, and for storage requirements assuming maximum at-reactor storage plus intra-utility transshipment of spent fuel. 8 refs., 8 figs., 10 tabs

  14. Water Reactor Fuel Performance Meeting 2008

    International Nuclear Information System (INIS)

    2008-10-01

    This meeting contains articles of the Water Reactor Fuel Performance Meeting 2008 of Korean Nuclear Society, Atomic Energy Society of Japan, Chinese Nuclear Society, European Nuclear Society and American Nuclear Society. It was held on Oct. 19-23, 2008 in Seoul, Korea and subject of Meeting is 'New Clear' Fuel - A green energy solution. This proceedings is comprised of 5 tracks. The main topic titles of track are as follows: Advances in water reactor fuel technology, Fuel performance and operational experience, Transient fuel behavior and safety-related issues, Fuel cycle, spent fuel storage and transportations and Fuel modeling and analysis. (Yi, J. H.)

  15. Neutron shielding studies on an advanced molten salt fast reactor design

    International Nuclear Information System (INIS)

    Merk, Bruno; Konheiser, Jörg

    2014-01-01

    Highlights: • Material damage due to irradiation has already been discovered at the MSRE. • Neutronic analysis of MSFR with curved blanket wall geometry. • Neutron fluence limit at the wall of the outer vessel can be kept for 80 years. • Shielded MSFR core will be of same dimension than a SFR core. - Abstract: The molten salt reactor technology has gained some new interest. In contrast to the historic molten salt reactors, the current projects are based on designing a molten salt fast reactor. Thus the shielding becomes significantly more challenging than in historic concepts. One very interesting and innovative result of the most recent EURATOM project on molten salt reactors – EVOL – is the fluid flow optimized design of the inner reactor vessel using curved blanket walls. The developed structure leads to a very uniform flow distribution. The design avoids all internal structures. Based on this new geometry a model for neutron physics calculation is presented. The major steps are: the modeling of the curved geometry in the unstructured mesh neutron transport code HELIOS and the determination of the real neutron flux and power distribution for this new geometry. The developed model is then used for the determination of the neutron fluence distribution in the inner and outer wall of the system. Based on these results an optimized shielding strategy is developed for the molten salt fast reactor to keep the fluence in the safety related outer vessel below expected limit values. A lifetime of 80 years can be assured, but the size of the core/blanket system will be comparable to a sodium cooled fast reactor. The HELIOS results are verified against Monte-Carlo calculations with very satisfactory agreement for a deep penetration problem

  16. Reproduction of the RA reactor fuel element fabrication

    International Nuclear Information System (INIS)

    Novakovic, M.

    1961-12-01

    This document includes the following nine reports: Final report on task 08/12 - testing the Ra reactor fuel element; design concept for fabrication of RA reactor fuel element; investigation of the microstructure of the Ra reactor fuel element; Final report on task 08/13 producing binary alloys with Al, Mo, Zr, Nb and B additions; fabrication of U-Al alloy; final report on tasks 08/14 and 08/16; final report on task 08/32 diffusion bond between the fuel and the cladding of the Ra reactor fuel element; Final report on task 08/33, fabrication of the RA reactor fuel element cladding; and final report on task 08/36, diffusion of solid state metals [sr

  17. IAEA Activities in the Area of Fast Reactors and Related Fuels and Fuel Cycles

    International Nuclear Information System (INIS)

    Monti, S.; Basak, U.; Dyck, G.; Inozemtsev, V.; Toti, A.; Zeman, A.

    2013-01-01

    Summary: • The IAEA role to support fast reactors and associated fuel cycle development programmes; • Main IAEA activities on fast reactors and related fuel and fuel cycle technology; • Main IAEA deliverables on fast reactors and related fuel and fuel cycle technology

  18. Research and development of thorium fuel cycle

    International Nuclear Information System (INIS)

    Oishi, Jun.

    1994-01-01

    Nuclear properties of thorium are summarized and present status of research and development of the use of thorium as nuclear fuel is reviewed. Thorium may be used for nuclear fuel in forms of metal, oxide, carbide and nitride independently, alloy with uranium or plutonium or mixture of the compound. Their use in reactors is described. The reprocessing of the spent oxide fuel in thorium fuel cycle is called the thorex process and similar to the purex process. A concept of a molten salt fuel reactor and chemical processing of the molten salt fuel are explained. The required future research on thorium fuel cycle is commented briefly. (T.H.)

  19. Space reactor fuels performance and development issues

    International Nuclear Information System (INIS)

    Wewerka, E.M.

    1984-01-01

    Three compact reactor concepts are now under consideration by the US Space Nuclear Power Program (the SP-100 Program) as candidates for the first 100-kWe-class space reactor. Each of these reactor designs puts unique constraints and requirements on the fuels system, and raises issues of fuel systems feasibility and performance. This paper presents a brief overview of the fuel requirements for the proposed space reactor designs, a delineation of the technical feasibility issues that each raises, and a description of the fuel systems development and testing program that has been established to address key technical issues

  20. Future fuel cycle development for CANDU reactors

    International Nuclear Information System (INIS)

    Hatcher, S.R.; McDonnell, F.N.; Griffiths, J.; Boczar, P.G.

    1987-01-01

    The CANDU reactor has proven to be safe and economical and has demonstrated outstanding performance with natural uranium fuel. The use of on-power fuelling, coupled with excellent neutron economy, leads to a very flexible reactor system with can utilize a wide variety of fuels. The spectrum of fuel cycles ranges from natural uranium, through slightly enriched uranium, to plutonium and ultimately thorium fuels which offer many of the advantages of the fast breeder reactor system. CANDU can also burn the recycled uranium and/or the plutonium from fuel discharged from light water reactors. This synergistic relationship could obviate the need to re-enrich the reprocessed uranium and allow a simpler reprocessing scheme. Fule management strategies that will permit future fuel cycles to be used in existing CANDU reactors have been identified. Evolutionary design changes will lead to an even greater flexibility, which will guarantee the continued success of the CANDU system. (author)

  1. Method of fueling for a nuclear reactor

    International Nuclear Information System (INIS)

    Igarashi, Takao.

    1983-01-01

    Purpose: To enable the monitoring of reactor power with sufficient accuracy, upon starting even without existence of neutron source in case of a low average burnup degree in the reactor core. Constitution: Each of fuel assemblies is charged such that neutron source region monitors for the start-up system in a reactor core neutron instrumentation system having nuclear fuel assemblies and a neutron instrumentation system are surrounded with 4 or 16 fuel assemblies of a low burnup degree. Then, the average burnup degree of the fuel assemblies surrounding the neutron source region monitors are increased than the reactor core burnup degree, whereby neutrons released from the peripheral fuels are increased, sufficient number of neutron counts can be obtained even with no neutron sources upon start-up and the reactor power can be monitored at a sufficient accuracy. (Sekiya, K.)

  2. Gaseous fuel reactors for power systems

    International Nuclear Information System (INIS)

    Helmick, H.H.; Schwenk, F.C.

    1978-01-01

    The Los Alamos Scientific Laboratory is participating in a NASA-sponsored program to demonstrate the feasibility of a gaseous uranium fueled reactor. The work is aimed at acquiring experimental and theoretical information for the design of a prototype plasma core reactor which will test heat removal by optical radiation. The basic goal of this work is for space applications, however, other NASA-sponsored work suggests several attractive applications to help meet earth-bound energy needs. Such potential benefits are small critical mass, on-site fuel processing, high fuel burnup, low fission fragment inventory in reactor core, high temperature for process heat, optical radiation for photochemistry and space power transmission, and high temperature for advanced propulsion systems. Low power reactor experiments using uranium hexafluoride gas as fuel demonstrated performance in accordance with reactor physics predictions. The final phase of experimental activity now in progress is the fabrication and testing of a buffer gas vortex confinement system

  3. Lanthanide fission product separation from the transuranics in the integral fast reactor fuel cycle demonstration

    International Nuclear Information System (INIS)

    Goff, K.M.; Mariani, R.D.; Benedict, R.W.; Ackerman, J.P.

    1993-01-01

    The Integral Fast Reactor (IFR) is an innovative reactor concept being developed by Argonne National Laboratory. This reactor uses liquid-metal cooling and metallic fuel. Its spent fuel will be reprocessed using a pyrochemical method employing molten salts and liquid metals in an electrofining operation. The lanthanide fission products are a concern during reprocessing because of heating and fuel performance issues, so they must be removed periodically from the system to lessen their impact. The actinides must first be removed form the system before the lanthanides are removed as a waste stream. This operation requires a relatively good lanthanide-actinide separation to minimize both the amount of transuranic material lost in the waste stream and the amount of lanthanides collected when the actinides are first removed. A computer code, PYRO, that models these operations using thermodynamic and empirical data was developed at Argonne and has been used to model the removal of the lanthanides from the electrorefiner after a normal operating campaign. Data from this model are presented. The results demonstrate that greater that 75% of the lanthanides can be separated from the actinides at the end of the first fuel reprocessing campaign using only the electrorefiner vessel

  4. Fuel salt and container material studies for MOSART transforming system

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V.; Feynberg, O.; Merzlyakov, A.; Surenkov, A.; Zagnitko, A. [National Research Center, Kurchatov Institute, Moscow (Russian Federation); Afonichkin, V.; Bovet, A.; Khokhlov, V. [Institute of High Temperature Electrochemisty, Ekaterinburg (Russian Federation); Subbotin, V.; Gordeev, M.; Panov, A.; Toropov, A. [Institute of Technical Physics, Snezhinsk (Russian Federation)

    2013-07-01

    A study is under progress to examine the feasibility of single stream Molten Salt Actinide Recycling and Transmuting system without and with Th support (MOSART) fuelled with different compositions of actinide tri-fluorides (AnF{sub 3}) from used LWR fuel. New fast-spectrum design options with homogeneous core and fuel salts with high enough solubility for AnF{sub 3} are being examined because of new goals. The flexibility of single fluid MOSART concept with Th support is underlined, particularly, possibility of its operation in self-sustainable mode (Conversion Ratio: CR=1) using different loadings and make up. The paper summarizes the most current status of fuel salt and container material data for the MOSART concept received within ISTC-3749 and ROSATOM-MARS projects. Key physical and chemical properties of various fluoride fuel salts are reported. The issues like salt purification, the electroreduction of U(IV) to U(III) in LiF-ThF{sub 4} and the electroreduction of Yb(III) to Yb(II) in LiF-NaF are detailed.

  5. Fuel assemblies for nuclear reactor

    International Nuclear Information System (INIS)

    Nishi, Akihito.

    1987-01-01

    Purpose: To control power-up rate at the initial burning stage of new fuel assemblies due to fuel exchange in a pressure tube type power reactor. Constitution: Burnable poisons are disposed to a most portion of fuel pellets in a fuel assembly to such a low concentration as the burn-up rate changes with time at the initial stage of the burning. The most portion means substantially more than one-half part of the pellets and gadolinia is used as burn-up poisons to be dispersed and the concentration is set to less than about 0.2 %. Upon elapse of about 15 days after the charging, the burnable poisons are eliminated and the infinite multiplication factors are about at 1.2 to attain a predetermined power state. Since the power-up rate of the nuclear reactor fuel assembly is about 0.1 % power/hour and the power-up rate of the fuel assembly around the exchanged channel is lower than that, it can be lowered sufficiently than the limit for the power-up rate practiced upon reactor start-up thereby enabling to replace fuels during power operation. (Horiuchi, T.)

  6. Inter renewal travelling wave reactor with rotary fuel columns

    International Nuclear Information System (INIS)

    Terai, Yuzo

    2016-01-01

    To realize the COP21 decision, this paper proposes Inter Renewal Travelling Wave Reactor that bear high burn-up rate 50% and product TRU fuel efficiently. The reactor is based on 4S Fast Reactor and has Reactor Fuel Columns as fuel assemblies that equalize temperature in the fuel assembly so that fewer structure is need to restrain thermal transformation. To equalize burn-up rate of all fuel assemblies in the reactor, each rotary fuel column has each motor-lifter. The rotary fuel column has two types (Cylinder type and Heat Pipe type using natrium at 15 kPa which supply high temperature energy for Ultra Super Critical power plant). At 4 years cycle all rotary fuel columns of the reactor are renewed by the metallurgy method (vacuum re-smelting) and TRU fuel is gotten from the water fuel. (author)

  7. Cost aspects of the research reactor fuel cycle

    International Nuclear Information System (INIS)

    2010-01-01

    Research reactors have made valuable contributions to the development of nuclear power, basic science, materials development, radioisotope production for medicine and industry, and education and training. In doing so, they have provided an invaluable service to humanity. Research reactors are expected to make important contributions in the coming decades to further development of the peaceful uses of nuclear technology, in particular for advanced nuclear fission reactors and fuel cycles, fusion, high energy physics, basic research, materials science, nuclear medicine, and biological sciences. However, in the context of decreased public sector support, research reactors are increasingly faced with financial constraints. It is therefore of great importance that their operations are based on a sound understanding of the costs of the complete research reactor fuel cycle, and that they are managed according to sound financial and economic principles. This publication is targeted at individuals and organizations involved with research reactor operations, with the aim of providing both information and an analytical framework for assessing and determining the cost structure of fuel cycle related activities. Efficient management of fuel cycle expenditures is an important component in developing strategies for sustainable future operation of a research reactor. The elements of the fuel cycle are presented with a description of how they can affect the cost efficient operation of a research reactor. A systematic review of fuel cycle choices is particularly important when a new reactor is being planned or when an existing reactor is facing major changes in its fuel cycle structure, for example because of conversion of the core from high enriched uranium (HEU) to low enriched uranium (LEU) fuel, or the changes in spent fuel management provision. Review and optimization of fuel cycle issues is also recommended for existing research reactors, even in cases where research reactor

  8. Removal of spent fuel from the TVR reactor for reprocessing and proposals for the RA reactor spent fuel handling

    International Nuclear Information System (INIS)

    Volkov, E.B.; Konev, V.N.; Shvedov, O.V.; Bulkin, S.Yu; Sokolov, A.V.

    2002-01-01

    The 2,5 MW heavy-water moderated and cooled research reactor TVR was located at the Moscow Institute for Theoretical and Experimental Physics site. In 1990 the final batch of spent nuclear fuel (SNF) from the TVR reactor was transported for reprocessing to Production Association (PA) 'Mayak'. This transportation of the SNF was a part of TVR reactor decommissioning. The special technology and equipment was developed in order to fulfill the preparation of TVR SNF for transportation. The design of the TVR reactor and the fuel elements used are similar to the design and fuel elements of the RA reactor. Two different ways of RA spent fuel elements for transportation to reprocessing plant are considered: in aluminum barrels, and in additional cans. The experience and equipment used for the preparing TVR fuel elements for transportation can help the staff of RA reactor to find the optimal way for these technical operations. (author)

  9. Performance of metallic fuels in liquid-metal fast reactors

    International Nuclear Information System (INIS)

    Seidel, B.R.; Walters, L.C.; Kittel, J.H.

    1984-01-01

    Interest in metallic fuels for liquid-metal fast reactors has come full circle. Metallic fuels are once again a viable alternative for fast reactors because reactor outlet temperature of interest to industry are well within the range where metallic fuels have demonstrated high burnup and reliable performance. In addition, metallic fuel is very tolerant of off-normal events of its high thermal conductivity and fuel behavior. Futhermore, metallic fuels lend themselves to compact and simplified reprocessing and refabrication technologies, a key feature in a new concept for deployment of fast reactors called the Integral Fast Reactor (IFR). The IFR concept is a metallic-fueled pool reactor(s) coupled to an integral-remote reprocessing and fabrication facility. The purpose of this paper is to review recent metallic fuel performance, much of which was tested and proven during the twenty years of EBR-II operation

  10. Fast-reactor fuel reprocessing in the United Kingdom

    International Nuclear Information System (INIS)

    Allardice, R.H.; Buck, C.; Williams, J.

    1977-01-01

    Enriched uranium metal fuel irradiated in the Dounreay Fast Reactor has been reprocessed and refabricated in plants specifically designed for the purpose in the United Kingdom since 1961. Efficient and reliable fuel recycle is essential to the development of a plutonium-based fast-reactor system, and the importance of establishing at an early stage fast-reactor fuel reprocessing has been reinforced by current world difficulties in reprocessing high-burnup thermal-reactor oxide fuel. The United Kingdom therefore decided to reprocess irradiated fuel from the 250MW(e) Prototype Fast Reactor (PFR) as an integral part of the fast reactor development programme. Flowsheet and equipment development work for the small-scale fully active demonstration plant has been carried out since 1972, and the plant will be commissioned and ready for active operation during 1977. In parallel, a comprehensive waste-management system has been developed and installed. Based on this development work and the information which will arise from active operation of the plant, a parallel development programme has been initiated to provide the basis for the design of a large-scale fast-reactor fuel-reprocessing plant to come into operation in the late 1980s to support the projected UK fast-reactor installation programme. The paper identifies the important differences between fast-reactor and thermal-reactor fuel-reprocessing technologies and describes some of the development work carried out in these areas for the small-scale PFR fuel-reprocessing operation. In addition, the development programme in aid of the design of a larger scale fast-reactor fuel-reprocessing plant is outlined and the current design philosophy discussed. (author)

  11. Advanced fuel in the Budapest research reactor

    International Nuclear Information System (INIS)

    Hargitai, T.; Vidovsky, I.

    1997-01-01

    The Budapest Research Reactor, the first nuclear facility of Hungary, started to operate in 1959. The main goal of the reactor is to serve neutron research, but applications as neutron radiography, radioisotope production, pressure vessel surveillance test, etc. are important as well. The Budapest Research Reactor is a tank type reactor, moderated and cooled by light water. After a reconstruction and upgrading in 1967 the VVR-SM type fuel elements were used in it. These fuel elements provided a thermal power of 5 MW in the period 1967-1986 and 10 MW after the reconstruction from 1992. In the late eighties the Russian vendor changed the fuel elements slightly, i.e. the main parameters of the fuel remained unchanged, however a higher uranium content was reached. This new fuel is called VVR-M2. The geometry of VVR-SM and VVR-M2 are identical, allowing the use to load old and new fuel assemblies together to the active core. The first new type fuel assemblies were loaded to the Budapest Research Reactor in 1996. The present paper describes the operational experience with the new type of fuel elements in Hungary. (author)

  12. WATER BOILER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1960-11-22

    As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

  13. Molten salt reactor technology for long-range and wide-scale nuclear energy system

    International Nuclear Information System (INIS)

    Ignatiev, V.; Alexseev, P.; Menshikov, L.; Prusakov, V.; Subbotine, S.

    1997-01-01

    A possibility of creation of multi-component nuclear power system in which alongside with thermal and fast reactors, molten salt burner reactors, for incineration of weapon grade plutonium, some minor actinides and transmutation of some fission products will be presented. The purposes of this work are to review the present status of the molten salt reactor technology and innovative non-aqueous chemical processing methods, to indicate the importance of the uncertainties remaining, to identify the additional work needed, and to evaluate the probability of success in obtaining improved safety characteristics for new concept of molten salt - burner reactor with external neutron source. 8 refs., 3 figs., 2 tabs

  14. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1984-01-01

    The fuel elements for a pressurised water reactor comprise arrays of rods of zirconium alloy sheathed nuclear fuel pellets. The helium gas pressure within each rod differs substantially from that of its closest neighbours

  15. Evaluation of Metal-Fueled Surface Reactor Concepts

    International Nuclear Information System (INIS)

    Poston, David I.; Marcille, Thomas F.; Kapernick, Richard J.; Hiatt, Matthew T.; Amiri, Benjamin W.

    2007-01-01

    Surface fission power systems for use on the Moon and Mars may provide the first use of near-term reactor technology in space. Most near-term surface reactor concepts specify reactor temperatures <1000 K to allow the use of established material and power conversion technology and minimize the impact of the in-situ environment. Metal alloy fuels (e.g. U-10Zr and U-10Mo) have not traditionally been considered for space reactors because of high-temperature requirements, but they might be an attractive option for these lower temperature surface power missions. In addition to temperature limitations, metal fuels are also known to swell significantly at rather low fuel burnups (∼1 a/o), but near-term surface missions can mitigate this concern as well, because power and lifetime requirements generally keep fuel burnups <1 a/o. If temperature and swelling issues are not a concern, then a surface reactor concept may be able to benefit from the high uranium density and relative ease of manufacture of metal fuels. This paper investigates two reactor concepts that utilize metal fuels. It is found that these concepts compare very well to concepts that utilize other fuels (UN, UO2, UZrH) on a mass basis, while also providing the potential to simplify material safeguards issues

  16. Innovative nuclear system based on liquid fuel

    International Nuclear Information System (INIS)

    Delpech, S.; Jaskierowicz, S.; Picard, G.; Merle-Lucotte, E.; Heuer, D.; Doligez, X.

    2009-01-01

    The aim of this paper is to present the physical properties and characteristics of the innovative concept of Molten Salt Fast Reactor (MSFR) developed by CNRS (France) and the corresponding fuel salt reprocessing proposed to clean up the fuel salt based on an analytical approach of lanthanides and actinides extraction. (author)

  17. Molten salt reactors. Synthesis of studies realized between 1973 and 1983. Experimental loop file

    International Nuclear Information System (INIS)

    1983-03-01

    Four test loops were developed for the experimental study of a molten salt reactor with lead salt direct contact. A molten salt loop, completely in graphite, including the pump, showed that this material is convenient for salt containment and circulation. Reactor components like flowmeters, electromagnetic pumps, pressure gauge, valves developed for liquid sodium, were tested with liquid lead. A water-mercury loop was built for lead-molten salt simulation studies. Finally a lead-salt loop (COMPARSE) was built to study the behaviour of salt particles carried by lead in the heat exchanger. [fr

  18. Irradiation behavior of metallic fast reactor fuels

    International Nuclear Information System (INIS)

    Pahl, R.G.; Porter, D.L.; Crawford, D.C.; Walters, L.C.

    1991-01-01

    Metallic fuels were the first fuels chosen for liquid metal cooled fast reactors (LMR's). In the late 1960's world-wide interest turned toward ceramic LMR fuels before the full potential of metallic fuel was realized. However, during the 1970's the performance limitations of metallic fuel were resolved in order to achieve a high plant factor at the Argonne National Laboratory's Experimental Breeder Reactor II. The 1980's spawned renewed interest in metallic fuel when the Integral Fast Reactor (IFR) concept emerged at Argonne National Laboratory. A fuel performance demonstration program was put into place to obtain the data needed for the eventual licensing of metallic fuel. This paper will summarize the results of the irradiation program carried out since 1985

  19. Advanced and sustainable fuel cycles for innovative reactor systems

    International Nuclear Information System (INIS)

    Glatz, J. P.; Malmbeck, R.; Purroy, D. S.; Soucek, P.; Inoue, T.; Uozumi, K.

    2007-01-01

    reprocessing facility with improved economics and the higher radiation stability of the molten salt media are some of the arguments in favour of pyro-reprocessing. Adaptations of this technology exist for the treatment of both oxide and nitride fuels. The flowsheet for the treatment of nitride fuels is similar to that of metal fuel. In the case of oxides a head-end reduction step is needed. It can be performed by direct electroreduction, where the heat generating fission products are removed and the fissile materials are recovered as an alloy, which can be again directly reprocessed by electrorefining. The present paper describes the progress made at ITU - mainly in the frame of the network projects mentioned above - in developing the grouped actinide recycling concept with a view to the sustainability goals described above for innovative reactor systems. In the frame of these projects, reprocessing of EBRII type metallic alloy fuel with 2% of Am and 5% of lanthanides (U 6 0Pu 2 0-Zr 1 0Am 2 Nd 3 .5Y 0 .5Ce 0 .5Gd 0 .5) is being carried out by electrorefining at ITU. An excellent grouped separation of actinides from lanthanides (An/Ln mass ratio = 2400) had been obtained. The high neutron capture of lanthanides and their possibly detrimental interaction with the cladding material implies that they must be separated. In this sense the choice of the cathode material for the actinide recovery is essential and it could be shown that aluminium is an excellent material for a pyrochemical partitioning process. The results are confirmed in conditions simulating the scaling up (multiple run) of the process, with an accumulation of Ln in the salt. One of the major goals is the minimization of actinide losses and to thereby reduce significantly the radiotoxicity of the waste produced. The results shown here represent the first demonstration of an efficient grouped actinide recovery from realistic metallic fuels and are therefore an important step in achieving the sustainability goals of

  20. Residual salt separation from simulated spent nuclear fuel reduced in a LiCl-Li2O salt

    International Nuclear Information System (INIS)

    Hur, Jin-Mok; Hong, Sun-Seok; Seo, Chung-Seok

    2006-01-01

    The electrochemical reduction of spent nuclear fuel in LiCl-Li 2 O molten salt for the conditioning of spent nuclear fuel requires the separation of the residual salts from a reduced metal product after the reduction process. Considering the behavior of spent nuclear fuel during the electrochemical reduction process, a surrogate material matrix was constructed and inactive tests on a salt separation were carried out to produce the data required for active tests. Fresh uranium metal prepared from the electrochemical reduction of U 3 O 8 powder was used as the surrogates of the spent nuclear fuel Atomic Energy Society of Japan, Tokyo, Japan, All rights reservedopyriprocess. LiCl, Li 2 O, Y 2 O 3 and SrCl 2 were selected as the components of the residual salts. Interactions between the salts and their influence on the separation of the residual salts were analyzed by differential scanning calorimetry (DSC) and thermogravimetry (TG). Eutectic melting of LiCl-Li 2 O and LiCl-SrCl 2 led to a melting point which was lower than that of the LiCl molten salt was observed. Residual salts were separated by a vaporization method. Co-vaporization of LiCl-Li 2 O and LiCl-SrCl 2 was achieved below the temperatures which could make the uranium metal oxidation by Li 2 O possible. The salt vaporization rates at 950degC were measured as follows: LiCl-8 wt% Li 2 O>LiCl>LiCl-8 wt% SrCl 2 >SrCl 2 . (author)

  1. Thorium Molten Salt Nuclear Energy Synergetic System (THORIMS-NES)

    International Nuclear Information System (INIS)

    Yoshioka, Ritsuo; Mitachi, Koshi

    2013-01-01

    The authors have been promoting nuclear energy technology based on thorium molten salt as Thorium Molten Salt Nuclear Energy Synergetic System (THORIMS-NES). This system is a combination of fission power reactor of Molten Salt Reactor (MSR), and Accelerator Molten Salt Breeder (AMSB) for production of fissile 233 U with connecting chemical processing facility. In this paper, concept of THORIMS-NES, advantages of thorium and molten salt recent MSR design results such as FUJI-U3 using 233 U fuel, FUJI-Pu, large sized super-FUJI, pilot plant miniFUJI, AMSB, and chemical processing facility are described. (author)

  2. Fissile fuel doubling time characteristics for reactor lifetime fuel logistics

    International Nuclear Information System (INIS)

    Heindler, M.; Harms, A.A.

    1978-01-01

    The establishment of nuclear fuel requirements and their efficient utilization requires a detailed knowledge of some aspects of fuel dynamics and processing during the reactor lifetime. It is shown here that the use of the fuel stockpile inventory concept can serve effectively for this fuel management purpose. The temporal variation of the fissile fuel doubling time as well as nonequilibrium core conditions are among the characteristics which thus become more evident. These characteristics - rather than a single figure-of-merit - clearly provide an improved description of the expansion capacity and/or fuel requirements of a nuclear reactor energy system

  3. Fuel Development For Gas-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    M. K. Meyer

    2006-06-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  4. A method for predicting fuel maintenance in once-through MSRs

    International Nuclear Information System (INIS)

    Ridley, Gavin; Chvala, Ondrej

    2017-01-01

    Highlights: •Novel method to control LEU-fueled MSR reactivity and chemistry is developed. •The method relies on adjusting material flows only. •Ten year full power depletion of a FLiBe core demonstrates the method. •Material flows and reactor physics data are presented. •Graphite lifetime and trifluoride actinide limits are satisfied. -- Abstract: Liquid fuel molten salt reactors allow reactivity control by material addition. This paper presents a method to adjust material flows in a molten salt reactor to keep the core critical, and to maintain desired reduction-oxidation potential in the core salt melt. The method is aimed at low-enriched uranium fueled thermal systems. It is developed as a Python library and uses Serpent2 Monte-Carlo transport and depletion code. A toy 300 MW(th) reactor with a FLiBe carrier salt is employed to demonstrate the performance of the method over 10 full power years. Results of the calculation are presented, including material flows, conversion ratio, effective delayed neutron fraction, and expected limits on trifluoride concentrations and graphite lifetime are investigated. This method lays a foundation for future studies including fuel cycle performance of molten salt reactors and dynamic behavior of the core during depletion.

  5. Thermochemical investigation of molten fluoride salts for Generation IV nuclear applications - an equilibrium exercise

    NARCIS (Netherlands)

    van der Meer, J.P.M.

    2006-01-01

    The concept of the Molten Salt Reactor, one of the so-called Generation IV future reactors, is that the fuel, a fissile material, which is dissolved in a molten fluoride salt, circulates through a closed circuit. The heat of fission is transferred to a second molten salt coolant loop, the heat of

  6. Future fuel cycle and reactor strategies

    International Nuclear Information System (INIS)

    Meneley, D.A.

    1999-01-01

    Within the framework of the 1997 IAEA Symposium 'Future Fuel Cycle and Reactor Strategies Adjusting to New Realities', Working Group No.3 produced a Key Issues paper addressing the title of the symposium. The scope of the Key Issues paper included those factors that are expected to remain or become important in the time period from 2015 to 2050, considering all facets of nuclear energy utilization from ore extraction to final disposal of waste products. The paper addressed the factors influencing the choice of reactor and fuel cycle. It then addressed the quantitatively largest category of reactor types expected to be important during the period; that is, thermal reactors burning uranium and plutonium fuel. The fast reactor then was discussed both as a stand-alone technology and as might be used in combination with thermal reactors. Thorium fuel use was discussed briefly. The present paper includes of a digest of the Key Issues Paper. Some comparisons arc made between the directions suggested in that paper and those indicated by the Abstracts of this Technical Committee Meeting- Recommendations are made for work which might be undertaken in the short and medium time frames, to ensure that fuel cycle technologies and processes established by the year 2050 will support the continuation of nuclear energy applications in the long term. (author)

  7. Fuel transfer system for a nuclear reactor

    International Nuclear Information System (INIS)

    Katz, L.R.; Marshall, J.R.; Desmarchais, W.E.

    1977-01-01

    Disclosed is a fuel transfer system for moving nuclear reactor fuel assemblies from a new fuel storage pit to a containment area containing the nuclear reactor, and for transferring spent fuel assemblies under water from the reactor to a spent fuel storage area. The system includes an underwater track which extends through a wall dividing the fuel building from the reactor containment and a car on the track serves as the vehicle for moving fuel assemblies between these two areas. The car is driven by a motor and linkage extending from an operating deck to a chain belt drive on the car. A housing pivotally mounted at its center on the car is hydraulically actuated to vertically receive a fuel assembly which then is rotated to a horizontal position to permit movement through the wall between the containment and fuel building areas. Return to the vertical position provides for fuel assembly removal and the reverse process is repeated when transferring an assembly in the opposite direction. Limit switches used in controlling operation of the system are designed to be replaced from the operating deck when necessary by tools designed for this purpose. 5 claims, 8 figures

  8. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bahri, Che Nor Aniza Che Zainul, E-mail: anizazainul@gmail.com; Majid, Amran Ab.; Al-Areqi, Wadeeah M. [Nuclear Science Program, School of Applied Physics, Faculty of Science and Technology, Universiti Kebangsaan Malaysia, 43600 UKM Bangi, Selangor (Malaysia)

    2015-04-29

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

  9. Aspects regarding the fuel management for PHWR nuclear reactors

    International Nuclear Information System (INIS)

    Dragusin, O.; Bobolea, A.; Voicu, A.

    2001-01-01

    Fuel management for PHWR nuclear reactors is completely different from the PWR reactors fuel management. PHWR reactor fuel loading procedures are repeated after an interval of time, as defined and specified in the project documentation, using a fuel machine that can be attached to the terminal fittings of horizontal pressure tubes while the reactor is a full power. Another aspect of fuel management policy is related to the possibility of bi-directional loading of the reactor, with the primary advantage of uniform and symmetrical characteristics. (authors)

  10. Method of monitoring fuel-rod vibrations in a nuclear fuel reactor

    International Nuclear Information System (INIS)

    Kawamura, Makoto; Takai, Katsuaki.

    1985-01-01

    Purpose: To monitor the vibration modes of fuel rods continuously and on real time during operation of a PWR type nuclear reactor. Method: Vibrations of fuel rods during reactor operation are mainly caused by the lateral flow of coolants flowing through the gaps at the joints of reactor core buffle plates into a reactor core and fretting damages may possibly be caused to the fuel rod support portions due to the vibrations. In view of the above, self-powered detectors are disposed at a plurality of axial positions for the respective peripheral fuel assemblies in adjacent with the buffle plates and the detection signals from neutron detectors, that is, the fluctuations in neutrons are subjected to a frequency analysis during the operation period. The neutron detectors are disposed at the periphery of the reactor core, because the fuel assemblies disposed at the peripheral portion directly undergo the lateral flow from the joints of the buffle plates and vibrates most violently. Thus, the vibration situations can be monitored continuously, in a three demensional manner and on real time. (Moriyama, K.)

  11. Progress of the DUPIC fuel compatibility analysis (I) - reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Jeong, Chang Joon; Roh, Gyu Hong; Rhee, Bo Wook; Park, Jee Won

    2003-12-01

    Since 1992, the direct use of spent pressurized water reactor fuel in CANada Deuterium Uranium (CANDU) reactors (DUPIC) has been studied as an alternative to the once-through fuel cycle. The DUPIC fuel cycle study is focused on the technical feasibility analysis, the fabrication of DUPIC fuels for irradiation tests and the demonstration of the DUPIC fuel performance. The feasibility analysis was conducted for the compatibility of the DUPIC fuel with existing CANDU-6 reactors from the viewpoints of reactor physics, reactor safety, fuel cycle economics, etc. This study has summarized the intermediate results of the DUPIC fuel compatibility analysis, which includes the CANDU reactor physics design requirements, DUPIC fuel core physics design method, performance of the DUPIC fuel core, regional overpower trip setpoint, and the CANDU primary shielding. The physics analysis showed that the CANDU-6 reactor can accommodate the DUPIC fuel without deteriorating the physics design requirements by adjusting the fuel management scheme if the fissile content of the DUPIC fuel is tightly controlled.

  12. 04 - Sodium cooled fast breeder fourth-generation reactors - The experimental reactor ALLEGRO, the other ways for fast breeder fourth-generation reactors

    International Nuclear Information System (INIS)

    2012-12-01

    The authors first present the technology of gas-cooled fast breeder reactors (basic principles, specific innovations, feasibility studies, fuel element, safety) and notably the ALLEGRO project (design options and expected performances, preliminary safety demonstration). Then, they present the lead-cooled fast-breeder reactor technology: interests and obstacles, return on experience, the issue of lead density, neutron assessment, transmutation potential, dosimetry, safety chemical properties and compatibility with the fuel, water, air and steels. The next part addresses the technology of molten-salt fast-breeder reactors: choice of the liquid fuel and geometry, reactor concept (difficulties, lack of past R and D), demonstration and demonstrators, international context

  13. Reactor fueling system

    International Nuclear Information System (INIS)

    Hattori, Noriaki; Hirano, Haruyoshi.

    1983-01-01

    Purpose: To optimally position a fuel catcher by mounting a television camera to a fuel catching portion and judging video images by the use of a computer or the like. Constitution: A television camera is mounted to the lower end of a fuel catching mechanism for handling nuclear fuels and a fuel assembly disposed within a reactor core or a fuel storage pool is observed directly from above to judge the position for the fuel assembly by means of video signals. Then, the relative deviation between the actual position of the fuel catcher and that set in a memory device is determined and the positional correction is carried out automatically so as to reduce the determined deviation to zero. This enables to catch the fuel assembly without failure and improves the efficiency for the fuel exchange operation. (Moriyama, K.)

  14. Ultimate storage of spent fuel elements of the AVR test power station in the Asse salt mine

    International Nuclear Information System (INIS)

    Wolf, J.

    1975-02-01

    With regard to the ultimate storage of irradiated AVR pebble-bed reactor carbide fuel elements in the saline of Asse, a number of tests and calculations are presented to demonstrate that there is no credible possibility of the MCA (maximum credible accident) defined for the saline. The safety of persons is not threatened during the operation of spent fuel storage nor at any later time (extrapolation up to approx. 1,000 years after storage). 1,000 fuel elements at a time are packed up in gas-tight containers which are stacked in boreholes. The boreholes are then sealed with concrete. Lay-out and functions of the special airlock and transportation systems - from the packing of the containers in a hot cell to the final storage in the borehole - are described with special reference to aspects of the safety of the overall procedure. The possible accidents in the mine are discussed in detail. 85 Kr and T release rates are determined in laboratory tests by heating of the spherical fuel elements. Tests with fuel elements embedded in salt or stagnant brine were carried out at varies temperatures to investigate their behaviour in final storage. Kr and T release, extraction of fission products, mechanical resistance and corrosion were examined in these tests. Finally, the permeability of salt and salt concrete to radioactive gases were investigated in a special experimental arrangement. The diffusion and permeability coefficients obtained for 85 Kr, HT and HTO allow an estimation of the gas discharge of the stored fuel element. (RB/AK) [de

  15. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1981-01-01

    An array of rods comprising zirconium alloy sheathed nuclear fuel pellets assembled to form a fuel element for a pressurised water reactor is claimed. The helium gas pressure within each rod differs substantially from that of its closest neighbours

  16. Fuels for Canadian research reactors

    International Nuclear Information System (INIS)

    Feraday, M.A.

    1993-01-01

    This paper includes some statements and remarks concerning the uranium silicide fuels for which there is significant fabrication in AECL, irradiation and defect performance experience; description of two Canadian high flux research reactors which use high enrichment uranium (HEU) and the fuels currently used in these reactors; limited fabrication work done on Al-U alloys to uranium contents as high as 40 wt%. The latter concerns work aimed at AECL fast neutron program. This experience in general terms is applied to the NRX and NRU designs of fuel

  17. Fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

    Gjertsen, R.K.; Tower, S.N.; Huckestein, E.A.

    1982-01-01

    A fuel assembly for a nuclear reactor comprises a 5x5 array of guide tubes in a generally 20x20 array of fuel elements, the guide tubes being arranged to accommodate either control rods or water displacer rods. The fuel assembly has top and bottom Inconel (Registered Trade Mark) grids and intermediate Zircaloy grids in engagement with the guide tubes and supporting the fuel elements and guide tubes while allowing flow of reactor coolant through the assembly. (author)

  18. Systematic study on Thorium fuel

    International Nuclear Information System (INIS)

    Shibata, Toshikazu; Kimura, Itsuro; Iwata, Shiro; Furuya, Hirotaka; Suzuki, Susumu.

    1988-01-01

    Introduced is the activities of the Joint Research Project Team on Thorium Fuel organized by mainly university researchers in Japan and supported by the Ministry of Education, Science and Culture for seven years since 1980. Four major groups were organized; (1) nuclear data, reactor physics and design, (2) nuclear fuel, (3) down stream and (4) biological effects of thorium. The first group covered measurements and analysis on nuclear data of thorium related nuclides, experiment and analysis on nuclear characteristics of thorium containing cores, basic engineering on a thorium molten salt reactor, and designs of several types of reactors. Fabrication and irradiation tests of thorium oxide fuel, and basic studies on new type thorium fuels (e.g. carbide and nitride) were studied by the second group. The third group covered the use of solutions in reprocessing of spent fuel, behavior of fission products, immobilization of high level radioactive waste, and continuous reprocessing for a molten salt reactor. The fourth group performed the trace study for patients who had been intravascularly injected with thorotrast for diagnosis of war injuries during the Second World War. (author)

  19. Research reactor spent fuel in Ukraine

    International Nuclear Information System (INIS)

    Trofimenko, A.P.

    1996-01-01

    This paper describes the research reactors in Ukraine, their spent fuel facilities and spent fuel management problems. Nuclear sciences, technology and industry are highly developed in Ukraine. There are 5 NPPs in the country with 14 operating reactors which have total power capacity of 12,800 MW

  20. Influence of remaining fission products in low-decontaminated fuel on reactor core characteristics

    International Nuclear Information System (INIS)

    Ohki, Shigeo

    2002-07-01

    Design study of core, fuel and related fuel cycle system with low-decontaminated fuel has been performed in the framework of the feasibility study (F/S) on commercialized fast reactor cycle systems. This report summarizes the influence on core characteristics of remaining fission products (FPs) in low-decontaminated fuel related to the reprocessing systems nominated in F/S phase I. For simple treatment of the remaining FPs in core neutronics calculation the representative nuclide method parameterized by the FP equivalent coefficient and the FP volume fraction was developed, which enabled an efficient evaluation procedure. As a result of the investigation on the sodium cooled fast reactor with MOX fuel designed in fiscal year 1999, it was found that the pyrochemical reprocessing with molten salt (the RIAR method) brought the largest influence. Nevertheless, it was still within the allowable range. Assuming an infinite-times recycling, the alternations in core characteristics were evaluated as follows: increment of burnup reactivity by 0.5%Δk/kk', decrement of breeding ratio by 0.04, increment of sodium void reactivity by 0.1x10 -2 Δk/kk' and decrement of Doppler constant (in absolute value) by 0.7x10 -3 Tdk/dT. (author)

  1. Optimization of temperature coefficient and breeding ratio for a graphite-moderated molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zou, C.Y.; Cai, X.Z.; Jiang, D.Z.; Yu, C.G.; Li, X.X.; Ma, Y.W.; Han, J.L. [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China); Chen, J.G., E-mail: chenjg@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China)

    2015-01-15

    Highlights: • The temperature feedback coefficient with different moderation ratios for TMSR in thermal neutron region is optimized. • The breeding ratio and doubling time of a thermal TMSR with three different reprocessing schemes are analyzed. • The smaller hexagon size and larger salt fraction with more negative feedback coefficient can better satisfy the safety demands. • A shorter reprocessing time can achieve a better breeding ratio in a thermal TMSR. • The graphite moderator lifespan is compared with other MSRs and discussed. - Abstract: Molten salt reactor (MSR) has fascinating features: inherent safety, no fuel fabrication, online fuel reprocessing, etc. However, the graphite moderated MSR may present positive feedback coefficient which has severe implications for the transient behavior during operation. In this paper, the feedback coefficient and the breeding ratio are optimized based on the fuel-to-graphite ratio variation for a thorium based MSR (TMSR). A certain thermal core with negative feedback coefficient and relative high initial breeding ratio is chosen for the reprocessing scheme analysis. The breeding performances for the TMSR under different online fuel reprocessing efficiencies and frequencies are evaluated and compared with other MSR concepts. The results indicate that the thermal TMSR can get a breeding ratio greater than 1.0 with appropriate reprocessing scheme. The low fissile inventory in thermal TMSR leads to a short doubling time and low transuranic (TRU) inventory. The lifetime of graphite used for the TMSR is also discussed.

  2. Reactor-specific spent fuel discharge projections: 1985 to 2020

    International Nuclear Information System (INIS)

    Heeb, C.M.; Libby, R.A.; Walling, R.C.; Purcell, W.L.

    1986-09-01

    The creation of four spent-fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No New Orders with Extended Burnup, (2) No New Orders with Constant Burnup, (3) Middle Case with Extended Burnup, and (4) Middle Case with Constant Burnup. Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel

  3. Micro-Reactor Physics of MOX-Fueled Core

    International Nuclear Information System (INIS)

    Takeda, T.

    2001-01-01

    Recently, fuel assemblies of light water reactors have become complicated because of the extension of fuel burnup and the use of high-enriched Gd and mixed-oxide (MOX) fuel, etc. In conventional assembly calculations, the detailed flux distribution, spectrum distribution, and space dependence of self-shielding within a fuel pellet are not directly taken into account. The experimental and theoretical study of investigating these microscopic properties is named micro-reactor physics. The purpose of this work is to show the importance of micro-reactor physics in the analysis of MOX fuel assemblies. Several authors have done related studies; however, their studies are limited to fuel pin cells, and they are never mentioned with regard to burnup effect, which is important for actual core design

  4. Overview of the recovery and processing of 233U from the Oak Ridge molten salt reactor experiment (MSRE) remediation activities

    International Nuclear Information System (INIS)

    Del Cul, G.D.; Icenhour, A.S.; Simmons, D.W.; Trowbridge, L.D.; Williams, D.F.; Toth, L.M.; Dai, S.

    2001-01-01

    The Molten Salt Reactor Experiment (MSRE) was operated at Oak Ridge National Laboratory (ORNL) from 1965 to 1969 to test the concept of a high-temperature, homogeneous, fluid-fueled reactor. The discovery that UF 6 and F 2 migrated from the storage tanks into distant pipes and a charcoal bed resulted in significant activities to remove and recover the 233 U and to decommission the reactor. The recovered fissile uranium will be converted into uranium oxide (U 3 O 8 ), which is a suitable form for long-term storage. This publication reports the research and several new developments that were needed to carry out these unique activities. (author)

  5. Fuel handling grapple for nuclear reactor plants

    International Nuclear Information System (INIS)

    Rousar, D.L.

    1992-01-01

    This patent describes a fuel handling system for nuclear reactor plants. It comprises: a reactor vessel having an openable top and removable cover and containing therein, submerged in water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units, the fuel handling system consisting essentially of the combination of: a fuel bundle handling platform movable over the open top of the reactor vessel; a fuel bundle handling mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grapple means comprising complementary hooks which pivot inward toward each other to securely grasp a bail handle of a nuclear reactor fuel bundle and pivot backward away from each other to release a bail handle; the grapple means having a hollow cylindrical support shaft fixed within the grapple head with hollow cylindrical sleeves rotatably mounted and fixed in longitudinal axial position on the support shaft and each sleeve having complementary hooks secured thereto whereby each hook pivots with the rotation of the sleeve secured thereto; and the hollow cylindrical support shaft being provided with complementary orifices on opposite sides of its hollow cylindrical and intermediate to the sleeves mounted thereon whereby the orifices on both sides of the hollow cylindrical support shaft are vertically aligned providing a direct in-line optical viewing path downward there-through and a remote operator positioned above the grapple means can observe from overhead the area immediately below the grapple hooks

  6. Reactor-specific spent fuel discharge projections: 1986 to 2020

    International Nuclear Information System (INIS)

    Heeb, C.M.; Walling, R.C.; Purcell, W.L.

    1987-03-01

    The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No new orders with extended burnup, (2) No new orders with constant burnup, (3) Upper reference (which assumes extended burnup), (4) Upper reference with constant burnup, and (5) Lower reference (which assumes extended burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel. 6 refs., 8 figs., 8 tabs

  7. Neutronics study on hybrid reactor cooled by helium, water and molten salt

    International Nuclear Information System (INIS)

    Li Zaixin; Feng Kaiming; Zhang Guoshu; Zheng Guoyao; Zhao Fengchao

    2009-01-01

    There is no serious magnetohydrodynamics (MHD) problem when helium,water or molten salt of Flibe flows in high magnetic field. Thus helium, water and Flibe were proposed as candidate of coolant for fusion-fission hybrid reactor based on magnetic confinement. The effect on neutronics of hybrid reactor due to coolant was investigated. The analyses of neutron spectra and fuel breeding of blanket with different coolants were performed. Variations of tritium breeding ratio (TBR), blanket energy multiplication (M) and keff with operating time were also studied. MCNP code was used for neutron transport simulation. It is shown that spectra change greatly with different coolants. The blanket with helium exhibits very hard spectrum and good tritium breeding ability. And fission reactions are mainly from fast neutron. The blanket with water has soft spectrum and high energy multiplication factor. However, it needs to improve TBR. The blanket with Flibe has hard spectrum and less energy release. (authors)

  8. Research reactors for power reactor fuel and materials testing - Studsvik's experience

    International Nuclear Information System (INIS)

    Grounes, M.

    1998-01-01

    Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

  9. Hydriding failure in water reactor fuel elements

    International Nuclear Information System (INIS)

    Sah, D.N.; Ramadasan, E.; Unnikrishnan, K.

    1980-01-01

    Hydriding of the zircaloy cladding has been one of the important causes of failure in water reactor fuel elements. This report reviews the causes, the mechanisms and the methods for prevention of hydriding failure in zircaloy clad water reactor fuel elements. The different types of hydriding of zircaloy cladding have been classified. Various factors influencing zircaloy hydriding from internal and external sources in an operating fuel element have been brought out. The findings of post-irradiation examination of fuel elements from Indian reactors, with respect to clad hydriding and features of hydriding failure are included. (author)

  10. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Sasaki, Y.; Tashima, J.

    1975-01-01

    A description is given of nuclear reactor fuel assemblies arranged in the form of a lattice wherein there is attached to the interface of one of two adjacent fuel assemblies a plate spring having a concave portion curved toward said interface and to the interface of the other fuel assembly a plate spring having a convex portion curved away from said interface

  11. Target-fueled nuclear reactor for medical isotope production

    Science.gov (United States)

    Coats, Richard L.; Parma, Edward J.

    2017-06-27

    A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7 to 21 days.

  12. Power generation costs for alternate reactor fuel cycles

    International Nuclear Information System (INIS)

    Smolen, G.R.; Delene, J.G.

    1980-09-01

    The total electric generating costs at the power plant busbar are estimated for various nuclear reactor fuel cycles which may be considered for power generation in the future. The reactor systems include pressurized water reactors (PWR), heavy-water reactors (HWR), high-temperature gas cooled reactors (HTGR), liquid-metal fast breeder reactors (LMFBR), light-water pre-breeder and breeder reactors (LWPR, LWBR), and a fast mixed spectrum reactor (FMSR). Fuel cycles include once-through, uranium-only recycle, and full recycle of the uranium and plutonium in the spent fuel assemblies. The U 3 O 8 price for economic transition from once-through LWR fuel cycles to both PWR recycle and LMFBR systems is estimated. Electric power generation costs were determined both for a reference set of unit cost parameters and for a range of uncertainty in these parameters. In addition, cost sensitivity parameters are provided so that independent estimations can be made for alternate cost assumptions

  13. Homogeneous Thorium Fuel Cycles in Candu Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hyland, B.; Dyck, G.R.; Edwards, G.W.R.; Magill, M. [Chalk River Laboratories, Atomic Energy of Canada Limited (Canada)

    2009-06-15

    The CANDU{sup R} reactor has an unsurpassed degree of fuel-cycle flexibility, as a consequence of its fuel-channel design, excellent neutron economy, on-power refueling, and simple fuel bundle [1]. These features facilitate the introduction and full exploitation of thorium fuel cycles in Candu reactors in an evolutionary fashion. Because thorium itself does not contain a fissile isotope, neutrons must be provided by adding a fissile material, either within or outside of the thorium-based fuel. Those same Candu features that provide fuel-cycle flexibility also make possible many thorium fuel-cycle options. Various thorium fuel cycles can be categorized by the type and geometry of the added fissile material. The simplest of these fuel cycles are based on homogeneous thorium fuel designs, where the fissile material is mixed uniformly with the fertile thorium. These fuel cycles can be competitive in resource utilization with the best uranium-based fuel cycles, while building up a 'mine' of U-233 in the spent fuel, for possible recycle in thermal reactors. When U-233 is recycled from the spent fuel, thorium-based fuel cycles in Candu reactors can provide substantial improvements in the efficiency of energy production from existing fissile resources. The fissile component driving the initial fuel could be enriched uranium, plutonium, or uranium-233. Many different thorium fuel cycle options have been studied at AECL [2,3]. This paper presents the results of recent homogeneous thorium fuel cycle calculations using plutonium and enriched uranium as driver fuels, with and without U-233 recycle. High and low burnup cases have been investigated for both the once-through and U-233 recycle cases. CANDU{sup R} is a registered trademark of Atomic Energy of Canada Limited (AECL). 1. Boczar, P.G. 'Candu Fuel-Cycle Vision', Presented at IAEA Technical Committee Meeting on 'Fuel Cycle Options for LWRs and HWRs', 1998 April 28 - May 01, also Atomic Energy

  14. Solar gasification of biomass: design and characterization of a molten salt gasification reactor

    Science.gov (United States)

    Hathaway, Brandon Jay

    The design and implementation of a prototype molten salt solar reactor for gasification of biomass is a significant milestone in the development of a solar gasification process. The reactor developed in this work allows for 3 kWth operation with an average aperture flux of 1530 suns at salt temperatures of 1200 K with pneumatic injection of ground or powdered dry biomass feedstocks directly into the salt melt. Laboratory scale experiments in an electrically heated reactor demonstrate the benefits of molten salt and the data was evaluated to determine the kinetics of pyrolysis and gasification of biomass or carbon in molten salt. In the presence of molten salt overall gas yields are increased by up to 22%; pyrolysis rates double due to improved heat transfer, while carbon gasification rates increase by an order of magnitude. Existing kinetic models for cellulose pyrolysis fit the data well, while carbon gasification in molten salt follows kinetics modeled with a 2/3 order shrinking-grain model with a pre-exponential factor of 1.5*106 min-1 and activation energy of 158 kJ/mol. A reactor concept is developed based around a concentric cylinder geometry with a cavity-style solar receiver immersed within a volume of molten carbonate salt. Concentrated radiation delivered to the cavity is absorbed in the cavity walls and transferred via convection to the salt volume. Feedstock is delivered into the molten salt volume where biomass gasification reactions will be carried out producing the desired product gas. The features of the cavity receiver/reactor concept are optimized based on modeling of the key physical processes. The cavity absorber geometry is optimized according to a parametric survey of radiative exchange using a Monte Carlo ray tracing model, resulting in a cavity design that achieves absorption efficiencies of 80%-90%. A parametric survey coupling the radiative exchange simulations to a CFD model of molten salt natural convection is used to size the annulus

  15. Fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

    Gjertsen, R.K.

    1982-01-01

    A fuel assembly in a nuclear reactor comprises a locking mechanism that is capable of locking the fuel assembly to the core plate of a nuclear reactor to prevent inadvertent movement of the fuel assembly. The locking mechanism comprises a ratchet mechanism 108 that allows the fuel assembly to be easily locked to the core plate but prevents unlocking except when the ratchet is disengaged. The ratchet mechanism is coupled to the locking mechanism by a rotatable guide tube for a control rod or water displacer rod. (author)

  16. Nuclear data covariances and sensitivity analysis, validation of a methodology based on the perturbation theory; application to an innovative concept: the molten thorium salt fueled reactor

    International Nuclear Information System (INIS)

    Bidaud, A.

    2005-10-01

    Neutron transport simulation of nuclear reactors is based on the knowledge of the neutron-nucleus interaction (cross-sections, fission neutron yields and spectra...) for the dozens of nuclei present in the core over a very large energy range (fractions of eV to several MeV). To obtain the goal of the sustainable development of nuclear power, future reactors must have new and more strict constraints to their design: optimization of ore materials will necessitate breeding (generation of fissile material from fertile material), and waste management will require transmutation. Innovative reactors that could achieve such objectives - generation IV or ADS (accelerator driven system) - are loaded with new fuels (thorium, heavy actinides) and function with neutron spectra for which nuclear data do not benefit from 50 years of industrial experience, and thus present particular challenges. After validation on an experimental reactor using an international benchmark, we take classical reactor physics tools along with available nuclear data uncertainties to calculate the sensitivities and uncertainties of the criticality and temperature coefficient of a thorium molten salt reactor. In addition, a study based on the important reaction rates for the calculation of cycle's equilibrium allows us to estimate the efficiency of different reprocessing strategies and the contribution of these reaction rates on the uncertainty of the breeding and then on the uncertainty of the size of the reprocessing plant. Finally, we use this work to propose an improvement of the high priority experimental request list. (author)

  17. Review on the current status of molten chloride reactor and its future prospect

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Seok Bin; Shin, Yukyung; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2016-10-15

    This paper has summarized and reviewed the current status of MCR as an online pyroprocessing reactor, and introduced the related works in UNIST. As the developments of the next generation nuclear energy systems require the fuel sustainability, passive operation safety, nuclear proliferation, and reduction of highly radioactive waste, only several types of nuclear reactor systems survive to the last. Among these, molten salt reactor (MSR) is one of the most promising concepts of next generation nuclear reactor system that deliver on these requirements. MSR have great advantages in the fuel cycle and reduction of nuclear waste, since MSR can serve the online reprocessing system for the reprocessing of spent fuel. Especially, MSR utilizing chloride-based fuel, called molten chloride reactor (MCR) has been recently highlighted in USA under the DOE’s Gateway for Accelerated Innovation in Nuclear (GAIN) program. Recently, the interests in the molten chloride salt have arisen. The use of chloride-based salt gives great advantages to the reactor operating in a fast spectrum. Then MCR can serve waste management functions or fuel cycle sustainability functions, which can solve the current issues in nuclear field. Thus, research plan was established in UNIST which includes the investigation of thermal-hydraulic characteristics of chloride salt and optimization of heat transport system of MCR, using both numerical method and experimental method.

  18. Review on the current status of molten chloride reactor and its future prospect

    International Nuclear Information System (INIS)

    Seo, Seok Bin; Shin, Yukyung; Bang, In Cheol

    2016-01-01

    This paper has summarized and reviewed the current status of MCR as an online pyroprocessing reactor, and introduced the related works in UNIST. As the developments of the next generation nuclear energy systems require the fuel sustainability, passive operation safety, nuclear proliferation, and reduction of highly radioactive waste, only several types of nuclear reactor systems survive to the last. Among these, molten salt reactor (MSR) is one of the most promising concepts of next generation nuclear reactor system that deliver on these requirements. MSR have great advantages in the fuel cycle and reduction of nuclear waste, since MSR can serve the online reprocessing system for the reprocessing of spent fuel. Especially, MSR utilizing chloride-based fuel, called molten chloride reactor (MCR) has been recently highlighted in USA under the DOE’s Gateway for Accelerated Innovation in Nuclear (GAIN) program. Recently, the interests in the molten chloride salt have arisen. The use of chloride-based salt gives great advantages to the reactor operating in a fast spectrum. Then MCR can serve waste management functions or fuel cycle sustainability functions, which can solve the current issues in nuclear field. Thus, research plan was established in UNIST which includes the investigation of thermal-hydraulic characteristics of chloride salt and optimization of heat transport system of MCR, using both numerical method and experimental method

  19. Heat Transfer Salts for Nuclear Reactor Systems - Chemistry Control, Corrosion Mitigation, and Modeling

    International Nuclear Information System (INIS)

    Anderson, Mark; Sridharan, Kumar; Morgan, Dane; Peterson, Per; Calderoni, Pattrick; Scheele, Randall; Casekka, Andrew; McNamara, Bruce

    2015-01-01

    The concept of a molten salt reactor has existed for nearly sixty years. Previously all work was done during a large collaborative effort at Oak Ridge National Laboratory, culminating in a research reactor which operated for 15,000 hours without major error. This technical success has garnished interest in modern, high temperature, reactor schemes. Research using molten fluoride salts for nuclear applications requires a steady supply of high grade molten salts. There is no bulk supplier of research grade fluoride salts in the world, so a facility which could provide all the salt needed for testing at the University of Wisconsin had to be produced. Two salt purification devices were made for this purpose, a large scale purifier, and a small scale purifier, each designed to clean the salts from impurities and reduce their corrosion potential. As of now, the small scale has performed with flibe salt, hydrogen, and hydrogen fluoride, yielding clean salt. This salt is currently being used in corrosion testing facilities at the Massachusetts Institute of Technology and the University of Wisconsin. Working with the beryllium based salts requires extensive safety measures and health monitoring to prevent the development of acute or chronic beryllium disease, two pulmonary diseases created by an allergic reaction to beryllium in the lungs. Extensive health monitoring, engineering controls, and environment monitoring had to be set up with the University of Wisconsin department of Environment, Health and Safety. The hydrogen fluoride required for purification was also an extreme health hazard requiring thoughtful planning and execution. These dangers have made research a slow and tedious process. Simple processes, such as chemical handling and clean-up, can take large amounts of ingenuity and time. Other work has complemented the experimental research at Wisconsin to advance high temperature reactor goals. Modeling work has been performed in house to re

  20. Heat Transfer Salts for Nuclear Reactor Systems - Chemistry Control, Corrosion Mitigation, and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, Mark [Univ. of Wisconsin, Madison, WI (United States); Sridharan, Kumar [Univ. of Wisconsin, Madison, WI (United States); Morgan, Dane [Univ. of Wisconsin, Madison, WI (United States); Peterson, Per [Univ. of Wisconsin, Madison, WI (United States); Calderoni, Pattrick [Univ. of Wisconsin, Madison, WI (United States); Scheele, Randall [Univ. of Wisconsin, Madison, WI (United States); Casekka, Andrew [Univ. of Wisconsin, Madison, WI (United States); McNamara, Bruce [Univ. of Wisconsin, Madison, WI (United States)

    2015-01-22

    The concept of a molten salt reactor has existed for nearly sixty years. Previously all work was done during a large collaborative effort at Oak Ridge National Laboratory, culminating in a research reactor which operated for 15,000 hours without major error. This technical success has garnished interest in modern, high temperature, reactor schemes. Research using molten fluoride salts for nuclear applications requires a steady supply of high grade molten salts. There is no bulk supplier of research grade fluoride salts in the world, so a facility which could provide all the salt needed for testing at the University of Wisconsin had to be produced. Two salt purification devices were made for this purpose, a large scale purifier, and a small scale purifier, each designed to clean the salts from impurities and reduce their corrosion potential. As of now, the small scale has performed with flibe salt, hydrogen, and hydrogen fluoride, yielding clean salt. This salt is currently being used in corrosion testing facilities at the Massachusetts Institute of Technology and the University of Wisconsin. Working with the beryllium based salts requires extensive safety measures and health monitoring to prevent the development of acute or chronic beryllium disease, two pulmonary diseases created by an allergic reaction to beryllium in the lungs. Extensive health monitoring, engineering controls, and environment monitoring had to be set up with the University of Wisconsin department of Environment, Health and Safety. The hydrogen fluoride required for purification was also an extreme health hazard requiring thoughtful planning and execution. These dangers have made research a slow and tedious process. Simple processes, such as chemical handling and clean-up, can take large amounts of ingenuity and time. Other work has complemented the experimental research at Wisconsin to advance high temperature reactor goals. Modeling work has been performed in house to re

  1. Growing dimensions. Spent fuel management at research reactors

    International Nuclear Information System (INIS)

    Ritchie, I.G.

    1998-01-01

    More than 550 nuclear research reactors are operating or shout down around the world. At many of these reactors, spent fuel from their operations is stored, pending decisions on its final disposition. In recent years, problems associated with this spent fuel storage have loomed larger in the international nuclear community. In efforts to determine the overall scope of problems and to develop a database on the subject, the IAEA has surveyed research reactor operators in its Member States. Information for the Research Reactor Spent Fuel Database (RRSFDB) so far has been obtained from a limited but representative number of research reactors. It supplements data already on hand in the Agency's more established Research Reactor Database (RRDB). Drawing upon these database resources, this article presents an overall picture of spent fuel management and storage at the world's research reactors, in the context of associated national and international programmes in the field

  2. Treatment of spent fuels from research reactors and reactor development programs in Germany

    International Nuclear Information System (INIS)

    Closs, K.D.

    1999-01-01

    Quite a great number of different types of spent fuel from research reactors and development programs exists in Germany. The general policy is to send back to the USA as long as possible fuel from MTRs and TRIGAs of USA origin. An option is reprocessing in Great Britain or France. This option is pursued as long as reprocessing and reuse of the recovered material is economically justifiable. For those fuels which cannot be returned to the USA or which will not be reprocessed, a domestic back-up solution of spent fuel management has been developed in Germany, compatible with the management of spent fuel from power reactors. It consists in dry storage in special casks and, later on, direct disposal. Preliminary results from experimental R and D investigations with research reactor fuel and experience from LWR fuel lead to the conclusion that the direct disposal option even for research reactor fuel or exotic fuel does not impose major technical difficulties for the German waste management and disposal concept. (author)

  3. Reactor TRIGA PUSPATI (RTP) spent fuel pool conceptual design

    International Nuclear Information System (INIS)

    Mohd Fazli Zakaria; Tonny Lanyau; Ahmad Nabil Ab Rahim

    2010-01-01

    Reactor TRIGA PUSPATI (RTP) is the one and only research reactor in Malaysia that has been safely operated and maintained since 1982. In order to enhance technical capabilities and competencies especially in nuclear reactor engineering a feasibility study on RTP power upgrading was proposed to serve future needs for advance nuclear science and technology in the country with the capability of designing and develop reactor system. The need of a Spent Fuel Pool begins with the discharge of spent fuel elements from RTP for temporary storage that includes all activities related to the storage of fuel until it is either sent for reprocessed or sent for final disposal. To support RTP power upgrading there will be major RTP systems replacement such as reactor components and a new temporary storage pool for fuel elements. The spent fuel pool is needed for temporarily store the irradiated fuel elements to accommodate a new reactor core structure. Spent fuel management has always been one of the most important stages in the nuclear fuel cycle and considered among the most common problems to all countries with nuclear reactors. The output of this paper will provide sufficient information to show the Spent Fuel Pool can be design and build with the adequate and reasonable safety assurance to support newly upgraded TRIGA PUSPATI TRIGA Research Reactor. (author)

  4. Study of trans-uranian incineration in molten salt reactor; Etude de l'incineration des transuraniens en reacteur a sel fondu

    Energy Technology Data Exchange (ETDEWEB)

    Valade, M

    2000-10-27

    With the emergence of new options for nuclear power, molten salt reactors are envisaged for waste management. The aim of this thesis is to show how molten salt reactors can help to solve the transuranics issue. Their high versatility regarding to isotopic vector allows to accommodate large fractions of minor actinides as compared to solid fuel system. In this thesis, a neutronics study of molten salt reactors, MSR, has been conducted. For this purpose, two reference systems were considered, TIER1 and AMSTER. In the case of TIER1, an optimisation was made to reach an equilibrium. The analysis of both systems showed the main characteristics of MSR: their link to chemistry and on line reprocessing. In this work, several methods to drive the system to a state of equilibrium have been implemented and compared. During this process the isotopic composition and neutron spectrum, thus the nuclear reaction cross sections, vary tremendously. It is essential to take these evolutions into account in order to accurately estimate the equilibrium state. This has been accomplished inside the multi-recycling procedure we set with ERANOS. A dedicated calculation schema has been realized to simulate superthermal systems with this computation code. These results were checked through a benchmark against other computer codes. Then, with multi-recycling method, several molten salt systems have been compared in order to define the optimal reactor for transuranics incineration. Nevertheless, a final choice can not only be done using only neutronics characteristics since chemistry and thermal-hydraulics constraints are really important for MSR. Moreover, a complete safety study would be required. (author)

  5. Gel-sphere-pac reactor fuel fabrication and its application to a variety of fuels

    International Nuclear Information System (INIS)

    Olsen, A.R.; Judkins, R.R.

    1979-12-01

    The gel-sphere-pac fuel fabrication option was evaluated for its possible application to commercial scale fuel fabrication for 19 fuel element designs that use oxide fuel in metal clad rods. The dry gel spheres are prepared at the reprocessing plant and are then calcined, sintered, inspected, and loaded into fuel rods and packed by low-energy vibration. A fuel smear density of 83 to 88% theoretical can be obtained. All fuel fabrication process steps were defined and evaluated from fuel receiving to finished fuel element shipping. The evaluation also covers the feasibility of the process, the current status of technology, estimates of the required time and cost to develop the technology to commercial status, and the safety and licensability of commercial scale plants. The primary evaluation was for a Light-Water Reactor fuel element containing (U,Pu)O 2 fuel. The other 18 fuel element types - 3 for Light-Water Reactors, 1 for a Heavy-Water Reactor, 1 for a Gas-Cooled Fast Reactor, 7 for Liquid-Metal-Cooled Fast Breeder Reactors, and 3 pairs for Light-Water Prebreeder and Breeder Reactors - were compared with the Light-Water Reactor. The gel-sphere-pac option was found applicable to 17 of the 19 element types; the characteristics of a commercial scale plant were defined for these for making cost estimates for such plants. The evaluation clearly shows the gel-sphere-pac process to be a viable fuel fabrication option. Estimates indicate a significant potential fabrication cost advantage for the gel-sphere-pac process if a remotely operated and remotely maintained fuel fabrication plant is required

  6. Plasma-gun fueling for tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.

    1980-11-01

    In light of the uncertain extrapolation of gas puffing for reactor fueling and certain limitations to pellet injection, the snowplow plasma gun has been studied as a fueling device. Based on current understanding of gun and plasma behavior a design is proposed, and its performance is predicted in a tokamak reactor environment

  7. Fuel elements of research reactors in China

    International Nuclear Information System (INIS)

    Zhou Yongmao; Chen Dianshan; Tan Jiaqiu

    1987-01-01

    This paper describes the current status of design, fabrication of fuel elements for research reactors in China, emphasis is placed on the technology of fuel elements for the High Flux Engineering Test Reactor (HFETR). (author)

  8. A way to limit the corrosion in the Molten Salt Reactor concept: the salt redox potential control

    International Nuclear Information System (INIS)

    Gibilaro, M.; Massot, L.; Chamelot, P.

    2015-01-01

    The possibility of controlling the salt redox potential thanks to a redox buffer in the Molten Salt Fast Reactor was investigated, the goal was to limit the oxidation of the reactor structural material. Tests were performed in LiF-CaF 2 at 850 °C on two different redox couples to fix the salt potential, Eu(III)/Eu(II) and U(IV)/U(III), where the first one was used as inactive system to validate the methodology to be applied on the uranium system. A metallic reducing agent (Gd plate for Eu, and U plate for U system) was inserted in the salt, leading to a spontaneous reaction: Eu(III) and U(IV) were then reduced. Eu(III) was fully converted into Eu(II) with metallic Gd, validating the approach. On the U system, the U(IV)/U(III) ratio has to be set between 10 and 100 to limit the core material oxidation: addition of metallic U decreased the concentration ratio from the infinite to 1, showing the feasibility of the salt redox potential control with the U system

  9. Micro-structural study and Rietveld analysis of fast reactor fuels: U-Mo fuels

    Science.gov (United States)

    Chakraborty, S.; Choudhuri, G.; Banerjee, J.; Agarwal, Renu; Khan, K. B.; Kumar, Arun

    2015-12-01

    U-Mo alloys are the candidate fuels for both research reactors and fast breeder reactors. In-reactor performance of the fuel depends on the microstructural stability and thermal properties of the fuel. To improve the fuel performance, alloying elements viz. Zr, Mo, Nb, Ti and fissium are added in the fuel. The first reactor fuels are normally prepared by injection casting. The objective of this work is to compare microstructure, phase-fields and hardness of as-cast four different U-Mo alloy (2, 5, 10 and 33 at.% Mo) fuels with the equilibrium microstructure of the alloys. Scanning electron microscope with energy dispersive spectrometer and optical microscope have been used to characterize the morphology of the as-cast and annealed alloys. The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. A comparison of metallographic and Rietveld analysis of as-cast (dendritic microstructure) and annealed U-33 at.% Mo alloy, corresponding to intermetallic compound, has been reported here for the first time. This study will provide in depth understanding of microstructural and phase evolution of U-Mo alloys as fast reactor fuel.

  10. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    1975-01-01

    A description is given of a nuclear reactor fuel assembly comprising a cluster of fuel elements supported by transversal grids so that their axes are parallel to and at a distance from each other, in order to establish interstices for the axial flow of a coolant. At least one of the interstices is occupied by an axial duct reserved for an auxiliary cooling fluid and is fitted with side holes through which the auxiliary cooling fluid is sprayed into the cluster. Deflectors extend as from a transversal grid in a position opposite the holes to deflect the cooling fluid jet towards those parts of the fuel elements that are not accessible to the auxiliary coolant. This assembly is intended for reactors cooled by light or heavy water [fr

  11. Fast breeder reactor fuel reprocessing in France

    International Nuclear Information System (INIS)

    Bourgeois, M.; Le Bouhellec, J.; Eymery, R.; Viala, M.

    1984-08-01

    Simultaneous with the effort on fast breeder reactors launched several years ago in France, equivalent investigations have been conducted on the fuel cycle, and in particular on reprocessing, which is an indispensable operation for this reactor. The Rapsodie experimental reactor was associated with the La Hague reprocessing plant AT1 (1 kg/day), which has reprocessed about one ton of fuel. The fuel from the Phenix demonstration reactor is reprocessed partly at the La Hague UP2 plant and partly at the Marcoule pilot facility, undergoing transformation to reprocess all the fuel (TOR project, 5 t/y). The fuel from the Creys Malville prototype power plant will be reprocessed in a specific plant, which is in the design stage. The preliminary project, named MAR 600 (50 t/y), will mobilize a growing share of the CEA's R and D resources, as the engineering needs of the UP3 ''light water'' plant begins to decline. Nearly 20 tonnes of heavy metals irradiated in fast breeder reactors have been processed in France, 17 of which came from Phenix. The plutonium recovered during this reprocessing allowed the power plant cycle to be closed. This power plant now contains approximately 140 fuel asemblies made up with recycled plutonium, that is, more than 75% of the fuel assemblies in the Phenix core

  12. Nuclear reactor, fuel assembly and neutron measuring system

    International Nuclear Information System (INIS)

    Chaki, Masao; Murase, Michio; Zukeran, Atsushi; Moriya, Kimiaki

    1998-01-01

    The present invention provides a BWR type reactor improved with the efficiency of used fuels and fuel economy by increasing a rated power and reducing exchange fuels. Namely, in a BWR type reactor at present, a thermal limit value is determined by conducting nuclear calculation of the reactor core based on data of reactor flow rate measurement and data of neutron flux measurement. However, since the neutron calculation of the reactor core is based on fuel assemblies while the points for the neutron measurement are present at the outside of the fuel assemblies, errors are caused. A margin including the errors has been used as a thermal limit value during operation. In the present invention, neutron fluxes in the fuel assembly as a base of the nuclear calculation can be measured by the same number of neutron detector tubes, but the number of the measuring points is increased to four times. With such procedures, errors caused by the difference of the neutron calculation and values at neutron measuring points can be reduced. As a result, a margin of the thermal limit value is reduced to increase the degree of freedom of reactor operation. Then, the economical property of the reactor operation can be improved. (N.H.)

  13. REACTOR FUEL ELEMENTS TESTING CONTAINER

    Science.gov (United States)

    Whitham, G.K.; Smith, R.R.

    1963-01-15

    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  14. Breeder reactor fuel fabrication system development

    International Nuclear Information System (INIS)

    Bennett, D.W.; Fritz, R.L.; McLemore, D.R.; Yatabe, J.M.

    1981-01-01

    Significant progress has been made in the design and development of remotely operated breeder reactor fuel fabrication and support systems (e.g., analytical chemistry). These activities are focused by the Secure Automated Fabrication (SAF) Program sponsored by the Department of Energy to provide: a reliable supply of fuel pins to support US liquid metal cooled breeder reactors and at the same time demonstrate the fabrication of mixed uranium/plutonium fuel by remotely operated and automated methods

  15. A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guigon, B. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, Reacteur Jules Horowitz, 13 - Saint-Paul-lez-Durance (France); Vacelet, H. [Compagnie pour l' Etude et la Realisation de Combustibles Atomiques, CERCA, Etablissement de Romans, 26 (France); Dornbusch, D. [Technicatome, Service d' Architecture Generale, 13 - Aix-en-Provence (France)

    2003-07-01

    Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs: from activation analysis to power reactor fuel qualification. In this paper will be presented the main characteristics of the Jules Horowitz Reactor: its total power, neutron flux, fuel element... Safety criteria will be explained. Finally merits and disadvantages of UMo compared to the standard U{sub 3}Si{sub 2} fuel will be discussed. (authors)

  16. Fuel assemblies for use in nuclear reactors

    International Nuclear Information System (INIS)

    Schluderberg, D.C.

    1981-01-01

    A fuel assembly for use in pressurized water cooled nuclear fast breeder reactors is described in which moderator to fuel ratios, conducive to a high Pu-U-D 2 O reactor breeding ratio, are obtained whilst at the same time ensuring accurate spacing of fuel pins without the parasitic losses associated with the use of spacer grids. (U.K.)

  17. Conceptual design of Indian molten salt breeder reactor

    Indian Academy of Sciences (India)

    A crucial part for achieving reasonable breeding in such reactors ... lization of India's nuclear resource profiles of modest uranium and abundant thorium. The ..... mass flow rate at different powers for various salts and compared with water,.

  18. Catalyzed deuterium fueled reversed-field pinch reactor assessment

    International Nuclear Information System (INIS)

    Dobrott, D.

    1985-01-01

    This study is part of a Department of Energy supported alternate fusion fuels program at Science Applications International Corporation. The purpose of this portion of the study is to perform an assessment of a conceptual compact reversed-field pinch reactor (CRFPR) that is fueled by the catalyzed-deuterium (Cat-d) fuel cycle with respect to physics, technology, safety, and cost. The Cat-d CRFPR is compared to a d-t fueled fusion reactor with respect to several issues in this study. The comparison includes cost, reactor performance, and technology requirements for a Cat-d fueled CRFPR and a comparable cost-optimized d-t fueled conceptual design developed by LANL

  19. IAEA activities related to research reactor fuel conversion and spent fuel return programs

    International Nuclear Information System (INIS)

    Goldman, Ira N.; Adelfang, Pablo; Ritchie, Iain G.

    2005-01-01

    The IAEA has been involved for more than twenty years in supporting international nuclear non-proliferation efforts associated with reducing the amount of highly enriched uranium (HEU) in international commerce. IAEA projects and activities have directly supported the Reduced Enrichment for Research and Test Reactors (RERTR) programme, as well as directly associated efforts to return research reactor fuel to the country where it was originally enriched. IAEA efforts have included the development and maintenance of several data bases with information related to research reactors and research reactor spent fuel inventories that have been essential in planning and managing both RERTR and spent fuel return programmes. Other IAEA regular budget programs have supported research reactor fuel conversion from HEU to low enriched uranium (LEU), and in addressing issues common to many member states with spent fuel management problems and concerns. The paper briefly describes IAEA involvement since the early 1980's in these areas, including regular budget and Technical Co-operation programme activities, and focuses on efforts in the past five years to continue to support and accelerate U.S. and Russian research reactor spent fuel return programmes. (author)

  20. Tests of the RBMK-1500 reactor fuel assemblies in the Leningrad reactor

    International Nuclear Information System (INIS)

    Aden, V.C.; Varovin, I.A.; Vorontsov, B.A.

    1981-01-01

    Test of fuel assemblies of the RBMK-1500 reactor is conducted in the reactor of the Leningrad NPP unit 2 for proving the calculational values of critical power of the RBMK-1500 reactor fuel assemblies adopted in design. The experiment presupposes the maximal approximation of the fuel assembly operation parameters to the calculational critical parameters without bringing into the mode of heat transfer crisis. The experiments are carried out at 500, 850 and 900 MW(el) of the reactor. The maximal channel power made up 472 kW at 20.5 t/h coolant flow rate and 49% mass steam content at the outlet of the channel. It was concluded that there was supply up to the heat transfer crisis in all the investigated modes. Data of temperature measurings of the fuel element cans, readings of the devices of the failure control system of the fuel element cans and external inspection of the assemblies after the tests testify to it [ru

  1. Fuel assemblies for BWR type reactors

    International Nuclear Information System (INIS)

    Ishizuka, Takao.

    1981-01-01

    Purpose: To enable effective failed fuel detection by the provision of water rod formed with a connecting section connected to a warmed water feed pipe of a sipping device at the lower portion and with a warmed water jetting port in the lower portion in a fuel assembly of a BWR type reactor to thereby carry out rapid sipping. Constitution: Fuel rods and water rods are contained in the channel box of a fuel assembly, and the water rod is provided at its upper portion with a connecting section connected to the warmed water feed pipe of the sipping device and formed at its lower portion with a warmed water jetting port for jetting warmed water fed from the warmed water feed pipe. Upon detection of failed fuels, the reactor operation is shut down and the reactor core is immersed in water. The cover for the reactor container is removed and the cap of the sipping device is inserted to connect the warmed water feed pipe to the connecting section of the water rod. Then, warmed water is fed to the water rod and jetted out from the warmed water jetting port to cause convection and unify the water of the channel box in a short time. Thereafter, specimen is sampled and analyzed for the detection of failed fuels. (Moriyama, K.)

  2. Dynamic tritium inventory of a NET/ITER fuel cycle with lithium salt solution blanket

    International Nuclear Information System (INIS)

    Spannagel, G.; Gierszewski, P.

    1991-01-01

    At the Karlsruhe Nuclear Research Center (KfK) a flexible tool is being developed to simulate the dynamics of tritium inventories. This tool can be applied to any tritium handling system, especially to the fuel cycle components of future nuclear fusion devices. This instrument of simulation will be validated in equipment to be operated at the Karlsruhe Tritium Laboratory. In this study tritium inventories in a NET/ITER type fuel cycle involving a lithium salt solution blanket are investigated. The salt solution blanket serves as an example because it offers technological properties which are attractive in modeling the process; the example does not impair the general validity of the tool. Usually, the operation strategy of complex structures will deteriorate due to failures of the subsystems involved. These failures together with the reduced availability ensuing from them will be simulated. The example of this study is restricted to reduced availabilities of two subsystems, namely the reactor and the tritium recovery system. For these subsystems the influence of statistically varying intervals of operation is considered. Strategies we selected which are representative of expected modes of operation. In the design of a fuel cycle, care will be taken that prescribed availabilities of the subsystems can be achieved; however, the description of reactor operation is a complex task since operation breaks down into several campaigns for which rules have been specified which enable determination of whether a campaign has been successful and can be stopped. Thus, it is difficult to predict the overall behavior prior to a simulation which includes stochastic elements. Using the example mentioned above the capabilities of the tool will be illustrated; besides the presentation of results of inventory simulation, the applicability of these data will be discussed. (orig.)

  3. Gas-cooled Fast Reactor (GFR) fuel and In-Core Fuel Management

    International Nuclear Information System (INIS)

    Weaver, K.D.; Sterbentz, J.; Meyer, M.; Lowden, R.; Hoffman, E.; Wei, T.Y.C.

    2004-01-01

    The Gas-Cooled Fast Reactor (GCFR) has been chosen as one of six candidates for development as a Generation IV nuclear reactor based on: its ability to fully utilize fuel resources; minimize or reduce its own (and other systems) actinide inventory; produce high efficiency electricity; and the possibility to utilize high temperature process heat. Current design approaches include a high temperature (2 850 C) helium cooled reactor using a direct Brayton cycle, and a moderate temperature (550 C - 650 C) helium or supercritical carbon dioxide (S-CO 2 ) cooled reactor using direct or indirect Brayton cycles. These design choices have thermal efficiencies that approach 45% to 50%, and have turbomachinery sizes that are much more compact compared to steam plants. However, there are challenges associated with the GCFR, which are the focus of current research. This includes safety system design for decay heat removal, development of high temperature/high fluence fuels and materials, and development of fuel cycle strategies. The work presented here focuses on the fuel and preliminary in-core fuel management, where advanced ceramic-ceramic (cercer) dispersion fuels are the main focus, and average burnups to 266 M Wd/kg appear achievable for the reference Si C/(U,TRU)C block/plate fuel. Solid solution (pellet) fuel in composite ceramic clad (Si C/Si C) is also being considered, but remains as a backup due to cladding fabrication challenges, and high centerline temperatures in the fuel. (Author)

  4. Fuel assemblies for PWR type reactors: fuel rods, fuel plates. CEA work presentation

    International Nuclear Information System (INIS)

    Delafosse, Jacques.

    1976-01-01

    French work on PWR type reactors is reported: basic knowledge on Zr and its alloys and on uranium oxide; experience gained on other programs (fast neutron and heavy water reactors); zircaloy-2 or zircaloy-4 clad UO 2 fuel rods; fuel plates consisting of zircaloy-2 clad UO 2 squares of thickness varying between 2 and 4mm [fr

  5. Nuclear fuels for material test reactors

    International Nuclear Information System (INIS)

    Ramanathan, L.V.; Durazzo, M.; Freitas, C.T. de

    1982-01-01

    Experimental results related do the development of nuclear fuels for reactors cooled and moderated by water have been presented cylindrical and plate type fuels have been described in which the core consists of U compouns dispersed in an Al matrix and is clad with aluminium. Fabrication details involving rollmilling, swaging or hot pressing have been described. Corrosion and irradiation test results are also discussed. The performance of the different types of fuels indicates that it is possible to locally fabricate fuel plates with U 3 O 8 +Al cores (20% enriched U) for use in operating Brazilian research reactors. (Author) [pt

  6. Impacts of reactor. Induced cladding defects on spent fuel storage

    International Nuclear Information System (INIS)

    Johnson, A.B.

    1978-01-01

    Defects arise in the fuel cladding on a small fraction of fuel rods during irradiation in water-cooled power reactors. Defects from mechanical damage in fuel handling and shipping have been almost negligible. No commercial water reactor fuel has yet been observed to develop defects while stored in spent fuel pools. In some pools, defective fuel is placed in closed canisters as it is removed from the reactor. However, hundreds of defective fuel bundles are stored in numerous pools on the same basis as intact fuel. Radioactive species carried into the pool from the reactor coolant must be dealt with by the pool purification system. However, additional radiation releases from the defective fuel during storage appear tu be minimal, with the possible exception of fuel discharged while the reactor is operating (CANDU fuel). Over approximately two decades, defective commercial fuel has been handled, stored, shipped and reprocessed. (author)

  7. Fuel management codes for fast reactors

    International Nuclear Information System (INIS)

    Sicard, B.; Coulon, P.; Mougniot, J.C.; Gouriou, A.; Pontier, M.; Skok, J.; Carnoy, M.; Martin, J.

    The CAPHE code is used for managing and following up fuel subassemblies in the Phenix fast neutron reactor; the principal experimental results obtained since this reactor was commissioned are analyzed with this code. They are mainly concerned with following up fuel subassembly powers and core reactivity variations observed up to the beginning of the fifth Phenix working cycle (3/75). Characteristics of Phenix irradiated fuel subassemblies calculated by the CAPHE code are detailed as at April 1, 1975 (burn-up steel damage)

  8. Thermal diffusivity measurement of molten fluoride salt containing ThF4 (improvement of the simple ceramic cell)

    International Nuclear Information System (INIS)

    Kato, Y.; Araki, N.; Kobayashi, K.; Makino, A.

    1985-01-01

    Design conditions of a cylindrical ceramic cell are estimated which can be used to measure the absolute value of thermal diffusivity of molten salts by applying the stepwise heating method. Molten salt is expected to be used in nuclear systems such as the Molten-Salt Reactor, the Accelerator Molten-Salt Breeder, the Fusion Reactor Blanket Coolant, the Fuel Reprocessing System, and so on

  9. Engineering development studies for molten-salt breeder reactor processing No. 21

    International Nuclear Information System (INIS)

    Hightower, J.R. Jr.

    1976-03-01

    The status of the following programs is reported: (1) continuous fluorinator development: autoresistance heating test AHT-4; (2) development of the metal transfer process; (3) salt-metal contactor development: experiments with a mechanically agitated, nondispersing contactor using water and mercury and in the salt-bismuth flowthrough facility; and (4) fuel reconstitution development: installation of equipment for a fuel reconstitution engineering experiment

  10. Aspects of the fast reactors fuel cycle

    International Nuclear Information System (INIS)

    Zouain, D.M.

    1982-06-01

    The fuel cycle for fast reactors, is analysed, regarding the technical aspects of the developing of the reprocessing stages and the fuel fabrication. The environmental impact of LMFBRs and the waste management of this cycle are studied. The economic aspects of the fuel cycle, are studied too. Some coments about the Brazilian fast reactors programs are done. (E.G.) [pt

  11. International topical meeting. Research Reactor Fuel Management (RRFM) and meeting of the International Group on Reactor Research (IGORR)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-07-01

    Nuclear research and test reactors have been in operation for over 60 years, over 270 research reactors are currently operating in more than 50 countries. This meeting is dedicated to different aspects of research reactor fuels: new fuels for new reactors, the conversion to low enriched uranium fuels, spent fuel management and computational tools for core simulation. About 80 contributions are reported in this document, they are organized into 7 sessions: 1) international topics and overview on new projects and fuel, 2) new projects and upgrades, 3) fuel development, 4) optimisation and research reactor utilisation, 5) innovative methods in research reactors physics, 6) safety, operation and research reactor conversion, 7) fuel back-end management, and a poster session. Experience from Australian, Romanian, Libyan, Syrian, Vietnamese, South-African and Ghana research reactors are reported among other things. The Russian program for research reactor spent fuel management is described and the status of the American-driven program for the conversion to low enriched uranium fuels is presented. (A.C.)

  12. IAEA activities related to research reactor fuel conversion and spent fuel return programmes

    International Nuclear Information System (INIS)

    Ritchie, I.G.; Adelfang, P.; Goldman, I.N.

    2004-01-01

    Full text: The IAEA has been involved for more than twenty years in supporting international nuclear non-proliferation efforts associated with reducing the amount of highly enriched uranium (HEU) in international commerce. IAEA projects and activities have directly supported the Reduced Enrichment for Research and Test Reactors (RERTR) programme, as well as directly associated efforts to return research reactor fuel to the country of origin where it was originally enriched. IAEA efforts have included the development and maintenance of several data bases with information related to research reactors and research reactor spent fuel inventories that have been essential in planning and managing both RERTR and spent fuel return programmes. Other IAEA regular budget programmes have supported research reactor fuel conversion from HEU to low enriched uranium, and in addressing issues common to many member states with spent fuel management problems and concerns. The paper briefly describes IAEA involvement since the early 1980's in these areas, including regular budget and Technical Co-operation programme activities, and focuses on efforts in the past five years to continue to support and accelerate U.S. and Russian research reactor spent fuel return programmes. It is hoped that an announcement of the extension of the U.S. Acceptance Programme, which is expected in the very near future, will facilitate the life extensions of many productive TRIGA reactors around the world. (author)

  13. The chemistry of water reactor fuel

    International Nuclear Information System (INIS)

    Potter, P.E.

    1990-01-01

    In this paper, the authors discuss features of the changes in chemical constitution which occur in fuel and fuel rods for water reactors during operation and in fault conditions. The fuel for water reactors consists of pellets of urania (UO 2 ) clad in Zircaloy. An essential step in the prediction of the fate of all the radionuclides in a fault or accident is to possess a detailed knowledge of their chemical behavior at all stages of the development of such incidents. In this paper, the authors consider: the chemical constitution of fuel during operation at temperatures corresponding to rather low ratings, together with a quite detailed discussion of the chemistry within the fuel-clad gap; the behavior of fuel subjected to higher temperatures and ratings than those of contemporary fuel; and the changes in constitution on failure of fuel rods in fault or accident conditions

  14. Fuel designs for VVER reactors

    International Nuclear Information System (INIS)

    Simonov, K.V.; Carbon, P.; Silberstein, A.

    1995-01-01

    That progresses in efficiency and safety through progresses in technology and better prediction with fully benchmarked upgraded computer codes is a common goal for on the one hand the original designer of the VVER reactors and their respective fuels and on the other hand for EVF a western company resulting from a combined force with highly diversified and complementary talents in reactor and fuel design and manufacturing. It can be expected that this new challenge and dialogue between the two Russian and European industrial ventures will be mutually beneficial and yield innovative and high quality products and as a consequence strong return will be produced for the best interest of utilities operating VVER reactors. (orig./HP)

  15. A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

    International Nuclear Information System (INIS)

    Guigon, B.; Vacelet, H.; Dornbusch, D.

    2000-01-01

    Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs from activation analysis to power reactor fuel qualification. In this paper the main characteristics of the Jules Horowitz Reactor are presented. Safety criteria are explained. Finally, merits and disadvantages of UMo compared to the standard U 3 Si 2 fuel are discussed. (author)

  16. Compatibility studies of potential molten-salt breeder reactor materials in molten fluoride salts

    International Nuclear Information System (INIS)

    Keiser, J.R.

    1977-05-01

    The molten fluoride salt compatibility studies carried out during the period 1974--76 in support of the Molten-Salt Reactor Program are summarized. Thermal-convection and forced-circulation loops were used to measure the corrosion rate of selected alloys. Results confirmed the relationship of time, initial chromium concentration, and mass loss developed by previous workers. The corrosion rates of Hastelloy N and Hastelloy N modified by the addition of 1--3 wt percent Nb were well within the acceptable range for use in an MSBR. 13 figures, 3 tables

  17. Study of trans-uranian incineration in molten salt reactor; Etude de l'incineration des transuraniens en reacteur a sel fondu

    Energy Technology Data Exchange (ETDEWEB)

    Valade, M

    2000-10-27

    With the emergence of new options for nuclear power, molten salt reactors are envisaged for waste management. The aim of this thesis is to show how molten salt reactors can help to solve the transuranics issue. Their high versatility regarding to isotopic vector allows to accommodate large fractions of minor actinides as compared to solid fuel system. In this thesis, a neutronics study of molten salt reactors, MSR, has been conducted. For this purpose, two reference systems were considered, TIER1 and AMSTER. In the case of TIER1, an optimisation was made to reach an equilibrium. The analysis of both systems showed the main characteristics of MSR: their link to chemistry and on line reprocessing. In this work, several methods to drive the system to a state of equilibrium have been implemented and compared. During this process the isotopic composition and neutron spectrum, thus the nuclear reaction cross sections, vary tremendously. It is essential to take these evolutions into account in order to accurately estimate the equilibrium state. This has been accomplished inside the multi-recycling procedure we set with ERANOS. A dedicated calculation schema has been realized to simulate superthermal systems with this computation code. These results were checked through a benchmark against other computer codes. Then, with multi-recycling method, several molten salt systems have been compared in order to define the optimal reactor for transuranics incineration. Nevertheless, a final choice can not only be done using only neutronics characteristics since chemistry and thermal-hydraulics constraints are really important for MSR. Moreover, a complete safety study would be required. (author)

  18. Overview of remote technologies applied to research reactor fuel

    International Nuclear Information System (INIS)

    Oerdoegh, M.; Takats, F.

    1999-01-01

    This paper gives a brief overview of the remote technologies applied to research reactor fuels. Due to many reasons, the remote technology utilization to research reactor fuel is not so widespread as it is for power reactor fuels, however, the advantages of the application of such techniques are obvious. (author)

  19. Nuclear reactor using fuel sphere for combustion and fuel spheres for breeding

    International Nuclear Information System (INIS)

    Yamashita, Kiyonobu.

    1995-01-01

    The present invention concerns a pebble bed-type reactor which can efficiently convert parent nuclides to fission nuclides. Fuel spheres for combustion having fission nuclides as main fuels, and fuel spheres for breeding having parent nuclides as main fuels are used separately, in the pebble bed-type reactor. According to the present invention, fuel spheres for breeding can be stayed in a reactor core for a long period of time, so that parent nuclides can be sufficiently converted into fission nuclides. In addition, since fuel spheres for breeding are loaded repeatedly, the amount thereof to be used is reduced. Therefore, the amount of the fuel spheres for breeding is small even when they are re-processed. On the other hand, since the content of the fission nuclides in the fuel spheres for breeding is not great, they can be put to final storage. This is attributable that although the fuel spheres for breeding contain fission nuclides generated by conversion, the fission nuclides are annihilated by nuclear fission reactions at the same time with the generation thereof. (I.S.)

  20. Terrestrial Energy bets on molten salt reactors

    International Nuclear Information System (INIS)

    Anon.

    2015-01-01

    Terrestrial Energy is a Canadian enterprise, founded in 2013, for marketing the integral molten salt reactor (IMSR). A first prototype (called MSRE and with an energy output of 8 MW) was designed and operated between 1965 and 1969 by the Oak Ridge National Laboratory. IMSR is a small, modular reactor with a thermal energy output of 400 MW. According to Terrestrial Energy the technology of conventional power reactors is too complicated and too expensive. On the contrary IMSR's technology appears to be simple, easy to operate and affordable. With a staff of 30 people Terrestrial Energy appears to be a start-up in the nuclear sector. A process of pre-licensing will be launched in 2016 with the Canadian nuclear safety authority. (A.C.)

  1. Sodium fast reactors with closed fuel cycle

    CERN Document Server

    Raj, Baldev; Vasudeva Rao, PR 0

    2015-01-01

    Sodium Fast Reactors with Closed Fuel Cycle delivers a detailed discussion of an important technology that is being harnessed for commercial energy production in many parts of the world. Presenting the state of the art of sodium-cooled fast reactors with closed fuel cycles, this book:Offers in-depth coverage of reactor physics, materials, design, safety analysis, validations, engineering, construction, and commissioning aspectsFeatures a special chapter on allied sciences to highlight advanced reactor core materials, specialized manufacturing technologies, chemical sensors, in-service inspecti

  2. Fissile fuel dynamics of breeder/converter reactors

    International Nuclear Information System (INIS)

    Harms, A.A.

    1978-01-01

    The long-term fissile fuel dynamics for a hierarchy of fission reactors covering the range from pure-burners to super-breeders is examined. It is found that the breeding gains of the core and blanket can be used to identify several distinct fissile fuel histories and elucidate the importance of fuel cycle characteristics such as the time dependence of the fissile fuel doubling time. On this basis, a self-sufficient fission reactor is introduced and its determining characteristics are identified. (author)

  3. Paired replacement fuel assemblies for BWR-type reactor

    International Nuclear Information System (INIS)

    Oguchi, Kazushige.

    1997-01-01

    There are disposed a large-diameter water rod constituting a non-boiling region at a central portion and paired replacement fuel assemblies for two streams having the same average enrichment degree and different amount of burnable poisons. The paired replacement fuel assemblies comprise a first fuel assembly having a less amount of burnable poisons and a second fuel assembly having a larger amount of burnable poisons. A number of burnable poison-containing fuel rods in adjacent with the large diameter water rod is increased in the second fuel assembly than the first fuel assembly. Then, the poison of the paired replacement fuel assemblies for the BWR type reactor can be annihilated simultaneously at the final stage of the cycle. Accordingly, fuels for a BWR type reactor excellent in economical property and safety and facilitating the design of the replacement reactor core can be obtained. (N.H.)

  4. Cermet-fueled reactors for advanced space applications

    International Nuclear Information System (INIS)

    Cowan, C.L.; Palmer, R.S.; Taylor, I.N.; Vaidyanathan, S.; Bhattacharyya, S.K.; Barner, J.O.

    1987-12-01

    Cermet-fueled nuclear reactors are attractive candidates for high-performance advanced space power systems. The cermet consists of a hexagonal matrix of a refractory metal and a ceramic fuel, with multiple tubular flow channels. The high performance characteristics of the fuel matrix come from its high strength at elevated temperatures and its high thermal conductivity. The cermet fuel concept evolved in the 1960s with the objective of developing a reactor design that could be used for a wide range of mobile power generating sytems, including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests for the cermet fuel were carried out by General Electric as part of the 710 Project (General Electric 1966), and by Argonne National Laboratory in the Direct Nuclear Rocket Program (1965). Development programs for cermet fuel are currently under way at Argonne National Laboratory and Pacific Northwest Laboratory. The high temperature qualification tests from the 1960s have provided a base for the incorporation of cermet fuel in advanced space applications. The status of the cermet fuel development activities and descriptions of the key features of the cermet-fueled reactor design are summarized in this paper

  5. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  6. Reactor fuel depletion benchmark of TINDER

    International Nuclear Information System (INIS)

    Martin, W.J.; Oliveira, C.R.E. de; Hecht, A.A.

    2014-01-01

    Highlights: • A reactor burnup benchmark of TINDER, coupling MCNP6 to CINDER2008, was performed. • TINDER is a poor candidate for fuel depletion calculations using its current libraries. • Data library modification is necessary if fuel depletion is desired from TINDER. - Abstract: Accurate burnup calculations are key to proper nuclear reactor design, fuel cycle modeling, and disposal estimations. The TINDER code, originally designed for activation analyses, has been modified to handle full burnup calculations, including the widely used predictor–corrector feature. In order to properly characterize the performance of TINDER for this application, a benchmark calculation was performed. Although the results followed the trends of past benchmarked codes for a UO 2 PWR fuel sample from the Takahama-3 reactor, there were obvious deficiencies in the final result, likely in the nuclear data library that was used. Isotopic comparisons versus experiment and past code benchmarks are given, as well as hypothesized areas of deficiency and future work

  7. MTR fuel plate qualification in OSIRIS reactor

    International Nuclear Information System (INIS)

    Sacristan, P.; Boulcourt, P.; Naury, S.; Marchard, L.; Carcreff, H.; Noirot, J.

    2005-01-01

    Qualification of new MTR fuel needs the irradiation in research reactors under representative neutronic, heat flux and thermohydraulic conditions. The experiments are performed in France in the OSIRIS reactor by irradiating MTR full size fuel plates in the IRIS device located in the reactor core. The fuel plates are easily removed from the device during the shutdown of the reactor for performing thickness measurements along the plates by means of a swelling measurement device. Beside the calculation capabilities, the experimental platform includes: the ISIS neutron mock-up for the measurement of neutron flux distribution along the plates; the γ spectrometry for the purpose of measuring the activities of the radionuclides representative of the power and the burnup and to compare with the neutronic calculation. Owing to the experience feedback, a good agreement is observed between calculation and measurement; destructive post irradiation examinations in the LECA facility (Cadarache). New irradiations with the IRIS device and at higher heat flux are under preparation for qualification of MTR fuels. (author)

  8. Solution of heat removal from nuclear reactors by natural convection

    Directory of Open Access Journals (Sweden)

    Zitek Pavel

    2014-03-01

    Full Text Available This paper summarizes the basis for the solution of heat removal by natural convection from both conventional nuclear reactors and reactors with fuel flowing coolant (such as reactors with molten fluoride salts MSR.The possibility of intensification of heat removal through gas lift is focused on. It might be used in an MSR (Molten Salt Reactor for cleaning the salt mixture of degassed fission products and therefore eliminating problems with iodine pitting. Heat removal by natural convection and its intensification increases significantly the safety of nuclear reactors. Simultaneously the heat removal also solves problems with lifetime of pumps in the primary circuit of high-temperature reactors.

  9. Fuel assembly and nuclear reactor core

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Aoyama, Motoo; Yamashita, Jun-ichi.

    1995-01-01

    The present invention concerns a fuel assembly and a nuclear reactor core capable of improving a transmutation rate of transuranium elements while improving a residual rate of fission products. In a reactor core of a BWR type reactor to which fuel rods with transuranium elements (TRU) enriched are loaded, the enrichment degree of transuranium elements occupying in fuel materials is determined not less than 2wt%, as well as a ratio of number of atoms between hydrogen and fuel heavy metals in an average reactor core under usual operation state (H/HM) is determined not more than 3 times. In addition, a ratio of the volumes between coolant regions and fuel material regions is determined not more than 2 times. A T ratio (TRU/Pu) is lowered as the TRU enrichment degree is higher and the H/HM ratio is lower. In order to reduce the T ratio not more than 1, the TRU enrichment degree is determined as not less than 2wt%, and the H/HM ratio is determined to not more than 3 times. Accordingly, since the H/HM ratio is reduced to not more than 1, and TRU is transmuted while recycling it with plutonium, the transmutation ratio of transuranium elements can be improved while improving the residual rate of fission products. (N.H.)

  10. The effect of salt composition on reductive extraction of some typical elements from molten LiF-BeF2 salt into liquid bismuth

    International Nuclear Information System (INIS)

    Hirotake, M.; Jun, O.; Kimikazu, M.; Kunimitsu, Y.; Yasunobu, T.

    1983-01-01

    The distribution coefficients of thorium and radium between molten LiF-BeF 2 and liquid bismuth solutions were measured at 600 0 C in support of the processing of the molten-salt breeder reactor (MSBR) fuel. The increasing mole fraction of LiF in the salt phase from 40 to 70 mol% resulted in the rapid decrease of the distribution coefficient of thorium and in the slow decrease of that of radium. A comprehensive correlation of distribution behavior with salt composition is given by taking into account the formation of complex ions. The equilibrium distribution data affirm that thorium and radium exist mainly as Li 2 ThF 6 and RaF 2 , respectively, in the salt phase. It is suggested that the lower mole fraction of LiF in the fuel salt is effective in the MSBR fuel processing

  11. Situation of test and research reactors' spent fuels

    International Nuclear Information System (INIS)

    Shimizu, Kenichi; Uchiyama, Junzo; Sato, Hiroshi

    1996-01-01

    The U.S. DOE decided a renewal Off-Site Fuel Policy for stopping to spread a highly enriched uranium which was originally enriched at the U.S., the policy declared that to receive all HEU spent fuels from Test and Research reactors in all the world. In Japan, under bilateral agreement of cooperation between the government of the United States and the government of Japan concerning peaceful uses of nuclear energy, the highly enriched uranium of Test and Research Reactors' fuels was purchased from the U.S. and the fuels had been manufactured in Japan, America, Germany and France. On the other hand, a former president of the U.S. J. Carter proposed that to convert the fuels from HEU to LEU concerning a nonproliferation of nuclear materials in 1978, and Japan absolutely supported this policy. Under this condition, the U.S. stopped to receive the spent fuels from the other countries concerning legal action to the Off-Site Fuels Policy. As a result, the spent fuels are increasing, and to cross to each reactor's storage capacity, and if this policy start, a faced crisis of Test and Research Reactors will be avoided. (author)

  12. Status of research reactor spent fuel world-wide: Database summary

    International Nuclear Information System (INIS)

    Ritchie, I.G.

    1996-01-01

    Results complied in the research reactor spent fuel database are used to assess the status of research reactor spent fuel world-wide. Fuel assemblies, their types, enrichment, origin of enrichment and geological distribution among the industrialized and developed countries of the world are discussed. Fuel management practices in wet and dry storage facilities and the concerns of reactor operators about long-term storage of their spent fuel are presented and some of the activities carried out by the International Atomic Energy Agency to address the issues associated with research reactor spent fuel are outlined. (author). 4 refs, 17 figs, 4 tabs

  13. A new method to evaluate the sealing reliability of the flanged connections for Molten Salt Reactors

    International Nuclear Information System (INIS)

    Li, Qiming; Tian, Jian; Zhou, Chong; Wang, Naxiu

    2015-01-01

    Highlights: • We novelly valuate the sealing reliability of the flanged connections for MSRs. • We focus on the passive decrease of the leak impetus in flanged connections. • The modified flanged connections are acquired a sealing ability of self-adjustment. • Effects of redesigned flange configurations on molten salt leakage are discussed. - Abstract: The Thorium based Molten Salt Reactor (TMSR) project is a future Generation IV nuclear reactor system proposed by the Chinese Academy of Sciences with the strategic goal of meeting the growing energy needs in the Chinese economic development and social progress. It is based on liquid salts served as both fuel and primary coolant and consequently great challenges are brought into the sealing of the flanged connections. In this study, an improved prototype flange assembly is performed on the strength of the Freeze-Flange initially developed by Oak Ridge National Laboratory (ORNL). The calculation results of the finite element model established to analyze the temperature profile of the Freeze-Flange agree well with the experimental data, which indicates that the numerical simulation method is credible. For further consideration, the ideal-gas thermodynamic model, together with the mathematical approximation, is novelly borrowed to theoretically evaluate the sealing performance of the modified Freeze-Flange and the traditional double gaskets bolted flange joint. This study focuses on the passive decrease of the leak driving force due to multiple gaskets introduced in flanged connections for MSR. The effects of the redesigned flange configuration on molten salt leakage resistance are discussed in detail

  14. A new method to evaluate the sealing reliability of the flanged connections for Molten Salt Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Li, Qiming, E-mail: liqiming@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China); Tian, Jian; Zhou, Chong [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China); Wang, Naxiu, E-mail: wangnaxiu@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China)

    2015-06-15

    Highlights: • We novelly valuate the sealing reliability of the flanged connections for MSRs. • We focus on the passive decrease of the leak impetus in flanged connections. • The modified flanged connections are acquired a sealing ability of self-adjustment. • Effects of redesigned flange configurations on molten salt leakage are discussed. - Abstract: The Thorium based Molten Salt Reactor (TMSR) project is a future Generation IV nuclear reactor system proposed by the Chinese Academy of Sciences with the strategic goal of meeting the growing energy needs in the Chinese economic development and social progress. It is based on liquid salts served as both fuel and primary coolant and consequently great challenges are brought into the sealing of the flanged connections. In this study, an improved prototype flange assembly is performed on the strength of the Freeze-Flange initially developed by Oak Ridge National Laboratory (ORNL). The calculation results of the finite element model established to analyze the temperature profile of the Freeze-Flange agree well with the experimental data, which indicates that the numerical simulation method is credible. For further consideration, the ideal-gas thermodynamic model, together with the mathematical approximation, is novelly borrowed to theoretically evaluate the sealing performance of the modified Freeze-Flange and the traditional double gaskets bolted flange joint. This study focuses on the passive decrease of the leak driving force due to multiple gaskets introduced in flanged connections for MSR. The effects of the redesigned flange configuration on molten salt leakage resistance are discussed in detail.

  15. Testing of research reactor fuel in the high flux reactor (Petten)

    International Nuclear Information System (INIS)

    Guidez, J.; Markgraf, J.W.; Sordon, G.; Wijtsma, F.J.; Thijssen, P.J.M.; Hendriks, J.A.

    1999-01-01

    The two types of fuel most frequently used by the main research reactors are metallic: highly enriched uranium (>90%) and silicide low enriched uranium ( 3 . However, a need exists for research on new reactor fuel. This would permit some plants to convert without losses in flux or in cycle length and would allow new reactor projects to achieve higher possibilities especially in fluxes. In these cases research is made either on silicide with higher density, or on other types of fuel (UMo, etc.). In all cases when new fuel is proposed, there is a need, for safety reasons, to test it, especially regarding the mechanical evolution due to burn-up (swelling, etc.). Initially, such tests are often made with separate plates, but lately, using entire elements. Destructive examinations are often necessary. For this type of test, the High Flux Reactor, located in Petten (The Netherlands) has many specific advantages: a large core, providing a variety of interesting positions with high fluence rate; a downward coolant flow simplifies the engineering of the device; there exists easy access with all handling possibilities to the hot-cells; the high number of operating days (>280 days/year), together with the high flux, gives a possibility to reach quickly the high burn-up needs; an experienced engineering department capable of translating specific requirements to tailor-made experimental devices; a well equipped hot-cell laboratory on site to perform all necessary measurements (swelling, γ-scanning, profilometry) and all destructive examinations. In conclusion, the HFR reactor readily permits experimental research on specific fuels used for research reactors with all the necessary facilities on the Petten site. (author)

  16. A Thermodynamic Model for the Fuel of a Molten Salt Actinide Burner

    Energy Technology Data Exchange (ETDEWEB)

    Benes, Ondrej [Institute of Chemical Technology, Technicka 5, 16603 Prague (Czech Republic); European Commission - Joint Research Centre - Institute for Transuranium Elements, P.O. BOX 2340, 76125 Karlsruhe (Germany)

    2008-07-01

    In this study the importance of the thermodynamic description of a multi-component system when optimizing the fuel choice for a molten salt reactor is demonstrated. It is shown on the MF-PuF{sub 3} (M=Li,Na,K,Rb) system, one of the fuel alternatives, how properties such as vapour pressure or the solubility of the actinides in the alkali halide matrix can be obtained. Moreover it is shown that much bigger PuF{sub 3} solubility is achieved in the matrix containing only alkali halides than in a matrix that contains some concentrations of BeF{sub 2}. In order to obtain full thermodynamic description of the MF-PuF{sub 3} (M=Li,Na,K,Rb,Cs) system all the binary phase diagrams must be assessed. This is done according to the CALPHAD method including the critical review of all available data followed by an interactive optimization of the phase diagram to achieve the best possible agreement between the measurement and the calculation. A novel approach of obtaining the excess enthalpies of the (Rb,Cs)F solid solution by Ab initio has been used and the results are compared to the experimentally determined phase diagram measured in this study as well. For the measurement of the phase diagrams of the volatile fluoride salts special encapsulation technique has been developed. (authors)

  17. Assessment of the thorium fuel cycle in power reactors

    International Nuclear Information System (INIS)

    Kasten, P.R.; Homan, F.J.; Allen, E.J.

    1977-01-01

    A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled

  18. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2009-06-15

    SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing

  19. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    International Nuclear Information System (INIS)

    2009-06-01

    SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing facilities. - 3. Advances in Water

  20. Burnup performance of OTTO cycle pebble bed reactors with ROX fuel

    International Nuclear Information System (INIS)

    Ho, Hai Quan; Obara, Toru

    2015-01-01

    Highlights: • A 300 MW t Small Pebble Bed Reactor with Rock-like oxide fuel is proposed. • Using ROX fuel can achieve high discharged burnup of spent fuel. • High geological stability can be expected in direct disposal of the spent ROX fuel. • The Pebble Bed Reactor with ROX fuel can be critical at steady state operation. • All the reactor designs have a negative temperature coefficient. - Abstract: A pebble bed high-temperature gas-cooled reactor (PBR) with rock-like oxide (ROX) fuel was designed to achieve high discharged burnup and improve the integrity of the spent fuel in geological disposal. The MCPBR code with a JENDL-4.0 library, which developed the analysis of the Once-Through-Then-Out (OTTO) cycle in PBR, was used to perform the criticality and burnup analysis. Burnup calculations for eight cases were carried out for both ROX fuel and a UO 2 fuel reactor with different heavy-metal loading conditions. The effective multiplication factor of all cases approximately equalled unity in the equilibrium condition. The ROX fuel reactor showed lower FIFA than the UO 2 fuel reactor at the same heavy-metal loading, about 5–15%. However, the power peaking factor and maximum power per fuel ball in the ROX fuel core were lower than that of UO 2 fuel core. This effect makes it possible to compensate for the lower-FIFA disadvantage in a ROX fuel core. All reactor designs had a negative temperature coefficient that is needed for the passive safety features of a pebble bed reactor

  1. Disintegration of graphite matrix from the simulative high temperature gas-cooled reactor fuel element by electrochemical method

    International Nuclear Information System (INIS)

    Tian Lifang; Wen Mingfen; Li Linyan; Chen Jing

    2009-01-01

    Electrochemical method with salt as electrolyte has been studied to disintegrate the graphite matrix from the simulative high temperature gas-cooled reactor fuel elements. Ammonium nitrate was experimentally chosen as the appropriate electrolyte. The volume average diameter of disintegrated graphite fragments is about 100 μm and the maximal value is less than 900 μm. After disintegration, the weight of graphite is found to increase by about 20% without the release of a large amount of CO 2 probably owing to the partial oxidation to graphite in electrochemical process. The present work indicates that the improved electrochemical method has the potential to reduce the secondary nuclear waste and is a promising option to disintegrate graphite matrix from high temperature gas-cooled reactor spent fuel elements in the head-end of reprocessing.

  2. Structural analysis of reactor fuel elements

    International Nuclear Information System (INIS)

    Weeks, R.W.

    1977-01-01

    An overview of fuel-element modeling is presented that traces the development of codes for the prediction of light-water-reactor and fast-breeder-reactor fuel-element performance. It is concluded that although the mathematical analysis is now far advanced, the development and incorporation of mechanistic constitutive equations has not kept pace. The resultant reliance on empirical correlations severely limits the physical insight that can be gained from code extrapolations. Current efforts include modeling of alternate fuel systems, analysis of local fuel-cladding interactions, and development of a predictive capability for off-normal behavior. Future work should help remedy the current constitutive deficiencies and should include the development of deterministic failure criteria for use in design

  3. Fuel salt reprocessing influence on the MSFR behavior and on its associated reprocessing unit

    International Nuclear Information System (INIS)

    Doligez, X.

    2010-10-01

    In order to face with the growing of the energy demand, the nuclear industry has to reach the fourth generation technology. Among those concept, molten salt reactor, and especially the fast neutron spectrum configuration, seems very promising: indeed breeding is achievable while the feedback coefficient are still negative. However, the reprocessing salt scheme is not totally set down yet. A lot of uncertainties remain on chemical properties of the salt. Thanks to numerical simulation we studied the behavior of the molten Salt Fast Reactor coupled to a nominal reprocessing unit. We are now able to determine heat transfer and radiation in each elementary step of the unit and, by this way determine those that need special study for radioprotection. We also studied which elements are fundamental to extract for the reactor operation. Finally, we present a sensibility analysis of the chemical uncertainties to few relevant properties of the reactor behavior. (author)

  4. QUARTERLY PROGRESS REPORT JANUARY, FEBRUARY, MARCH, 1968 REACTOR FUELS AND MATERIALS DEVELOPMENT PROGRAMS FOR FUELS AND MATERIALS BRANCH OF USAEC DIVISION OF REACTOR DEVELOPMENT AND TECHNOLOGY

    Energy Technology Data Exchange (ETDEWEB)

    Cadwell, J. J.; de Halas, D. R.; Nightingale, R. E.; Worlton, D. C.

    1968-06-01

    Progress is reported in these areas: nuclear graphite; fuel development for gas-cooled reactors; HTGR graphite studies; nuclear ceramics; fast-reactor nitrides research; non-destructive testing; metallic fuels; basic swelling studies; ATR gas and water loop operation and maintenance; reactor fuels and materials; fast reactor dosimetry and damage analysis; and irradiation damage to reactor metals.

  5. Preparations for the Integral Fast Reactor fuel cycle demonstration

    International Nuclear Information System (INIS)

    Lineberry, M.J.; Phipps, R.D.

    1989-01-01

    Modifications to the Hot Fuel Examination Facility-South (HFEF/S) have been in progress since mid-1988 to ready the facility for demonstration of the unique Integral Fast Reactor (IFR) pyroprocess fuel cycle. This paper updates the last report on this subject to the American Nuclear Society and describes the progress made in the modifications to the facility and in fabrication of the new process equipment. The IFR is a breeder reactor, which is central to the capability of any reactor concept to contribute to mitigation of environmental impacts of fossil fuel combustion. As a fast breeder, fuel of course must be recycled in order to have any chance of an economical fuel cycle. The pyroprocess fuel cycle, relying on a metal alloy reactor fuel rather than oxide, has the potential to be economical even at small-scale deployment. Establishing this quantitatively is one important goal of the IFR fuel cycle demonstration

  6. Fuel-to-cladding heat transfer coefficient into reactor fuel element

    International Nuclear Information System (INIS)

    Lassmann, K.

    1979-01-01

    Models describing the fuel-to-cladding heat transfer coefficient in a reactor fuel element are reviewed critically. A new model is developed with contributions from solid, fluid and radiation heat transfer components. It provides a consistent description of the transition from an open gap to the contact case. Model parameters are easily available and highly independent of different combinations of material surfaces. There are no restrictions for fast transients. The model parameters are fitted to 388 data points under reactor conditions. For model verification another 274 data points of steel-steel and aluminium-aluminium interfaces, respectively, were used. The fluid component takes into account peak-to-peak surface roughnesses and, approximatively, also the wavelengths of surface roughnesses. For minor surface roughnesses normally prevailing in reactor fuel elements the model asymptotically yields Ross' and Stoute's model for the open gap, which is thus confirmed. Experimental contact data can be interpreted in very different ways. The new model differs greatly from Ross' and Stoute's contact term and results in better correlation coefficients. The numerical algorithm provides an adequate representation for calculating the fuel-to-cladding heat transfer coefficient in large fuel element structural analysis computer systems. (orig.) [de

  7. Fact reactor fuel alloys: Retrospective and prospective views

    International Nuclear Information System (INIS)

    Nevitt, M.V.

    1989-01-01

    The relationship between the physical metallurgy of the EBR-II metallic fuel, U-5% Fs, and its performance in the reactor are described. An understanding of these relationships, along with the optimal matching of fuel properties to fuel-element design, have been essential in the 23 year successful utilization of the fuel. The knowledge and experience gained are being employed in the current development of a new U-Pu-Zr metallic fuel for a proposed advanced reactor (orig./MM)

  8. Calculation of the evolution of molten salt breeder reactor

    International Nuclear Information System (INIS)

    Esteves, Fernando de Avelar

    1999-01-01

    A forecast for the future electrical consumption in Brazil and forecast of the nuclear electrical generation demand are discussed in this paper, which includes also an analysis on advanced nuclear reactors concept to supply that demand. This paper presents a concise description of the Molten Salt Breeder Reactor, considered the most appropriated to meet that demand. This paper also presents the burnup calculation modeling, including the operation modeling of this type of reactor from an initial load o 233 U up to the equilibrium cycle, the results of these calculations and its analysis. (author)

  9. Management of research reactor; dynamic characteristics analysis for reactor structures related with vibration of HANARO fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Chang Kee; Shim, Joo Sup [Shinwa Technology Information, Seoul (Korea)

    2001-04-01

    The objective of this study is to deduce the dynamic correlation between the fuel assembly and the reactor structure. Dynamic characteristics analyses for reactor structure related with vibration of HANARO fuel assembly have been performed For the dynamic characteristic analysis, the in-air models of the round and hexagonal flow tubes, 18-element and 36-element fuel assemblies, and reactor structure were developed. By calculating the hydrodynamic mass and distributing it on the in-air models, the in-water models of the flow tubes, the fuel assemblies, and the reactor structure were developed. Then, modal analyses for developed in-air and in-water models have been performed. Especially, two 18-element fuel assemblies and three 36-element fuel assemblies were included in the in-water reactor models. For the verification of the modal analysis results, the natural frequencies and the mode shapes of the fuel assembly were compared with those obtained from the experiment. Finally the analysis results of the reactor structure were compared with them performed by AECL Based on the reactor model without PCS piping, the in-water reactor model including the fuel assemblies was developed, and its modal analysis was performed. The analysis results demonstrate that there are no resonance between the fuel assembly and the reactor structures. 26 refs., 419 figs., 85 tabs. (Author)

  10. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    STAN, MARIUS [Los Alamos National Laboratory; HECKER, SIEGFRIED S. [Los Alamos National Laboratory

    2007-02-07

    Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low-melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher-melting U-PuZr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.

  11. Development, Fabrication and Characterization of Fuels for Indian Fast Reactor Programme

    International Nuclear Information System (INIS)

    Kumar, Arun

    2013-01-01

    Development of Fast Reactor fuels in India started in early Seventies. The successful development of Mixed Carbide fuels for FBTR and MOX fuel for PFBR have given confidence in manufacture of fuels for Fast Reactors. Effort is being put to develop high Breeding Ratio Metallic fuel (binary/ternary). Few fuel pins have been fabricated and is under test irradiation. However, this is only a beginning and complete fuel cycle activities are under development. Metal fuelled Fast Reactors will provide high growth rate in Indian Fast Reactor programme

  12. Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2009-06-15

    Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

  13. The Integral Molten Salt Reactor (IMSR)

    Energy Technology Data Exchange (ETDEWEB)

    Leblanc, D. [Terrestrial Energy, Mississauga, Ontario (Canada)

    2014-12-15

    The Integral Molten Salt Reactor is a simple burner or converter design that seeks to maximize passive and inherent safety features in order to minimize development time and achieve true cost innovation. Its integration of all primary systems into a unit sealed for the design life of the reactor will be reviewed with focus on the unique design aspects that make this a pragmatic approach. The IMSR is being developed by Terrestrial Energy in a range of power outputs with initial focus on an 80 MWth (32.5 MWe) unit primarily for remote energy needs. Similar units of modestly larger dimension and up to 600 MWth (291 MWe) are planned that remain truck transportable and able to compete in base load electricity markets worldwide. (author)

  14. The Integral Molten Salt Reactor (IMSR)

    Energy Technology Data Exchange (ETDEWEB)

    LeBlanc, D., E-mail: dleblanc@terrestrialenergy.com [Terrestrial Energy, Mississauga, Ontario (Canada)

    2014-07-01

    The Integral Molten Salt Reactor is a simple burner or converter design that seeks to maximize passive and inherent safety features in order to minimize development time and achieve true cost innovation. Its integration of all primary systems into a unit sealed for the design life of the reactor will be reviewed with focus on the unique design aspects that make this a pragmatic approach. The IMSR is being developed by Terrestrial Energy in a range of power outputs with initial focus on an 80 MWth (32.5 MWe) unit primarily for remote energy needs. Similar units of modestly larger dimension and up to 600 MWth (291 MWe) are planned that remain truck transportable and able to compete in base load electricity markets worldwide. (author)

  15. Fuel utilization potential in light water reactors with once-through fuel irradiation (AWBA Development Program)

    International Nuclear Information System (INIS)

    Rampolla, D.S.; Conley, G.H.; Candelore, N.R.; Cowell, G.K.; Estes, G.P.; Flanery, B.K.; Duncombe, E.; Dunyak, J.; Satterwhite, D.G.

    1979-07-01

    Current commercial light water reactor cores operate without recylce of fuel, on a once-through fuel cycle. To help conserve the limited nuclear fuel resources, there is interest in increasing the energy yield and, hence, fuel utilization from once-through fuel irradiation. This report evaluates the potential increase in fuel utilization of light water reactor cores operating on a once-through cycle assuming 0.2% enrichment plant tails assay. This evaluation is based on a large number of survey calculations using techniques which were verified by more detailed calculations of several core concepts. It is concluded that the maximum fuel utilization which could be achieved by practical once-through pressurized light water reactor cores with either uranium or thorium is about 17 MWYth/ST U 3 O 8 (Megawatt Years Thermal per Short Ton of U 3 O 8 ). This is about 50% higher than that of current commercial light water reactor cores. Achievement of this increased fuel utilization would require average fuel burnup beyond 50,000 MWD/MT and incorporation of the following design features to reduce parasitic losses of neutrons: reflector blankets to utilize neutrons that would otherwise leak out of the core; fuel management practices in which a smaller fraction of the core is replaced at each refueling; and neutron economic reactivity control, such as movable fuel control rather than soluble boron control. For a hypothetical situation in which all neutron leakage and parasitic losses are eliminated and fuel depletion is not limited by design considerations, a maximum fuel utilization of about 20 MWYth/ST U 3 O 8 is calculated for either uranium or thorium. It is concluded that fuel utilization for comparable reactor designs is better with uranium fuel than with thorium fuel for average fuel depletions of 30,000 to 35,000 MWD/MT which are characteristic of present light water reactor cores

  16. Coherence of reactor design and fuel element design

    International Nuclear Information System (INIS)

    Vom Scheidt, S.

    1995-01-01

    Its background of more than 25 years of experience makes Framatome the world's leading company in the design and sales of fuel elements for pressurized water reactors (PWR). In 1994, the fuel fabrication units were incorporated as subsidiaries, which further strengthens the company's position. The activities in the fuel sector comprise fuel element design, selection and sourcing of materials, fuel element fabrication, and the services associated with nuclear fuel. Design responsibility lies with the Design and sales Management, which closely cooperates with the engineers of the reactor plant for which the fuel elements are being designed, for fuel elements are inseparable parts of the respective reactors. The Design and Sales Management also has developed a complete line of services associated with fuel element inspection and repair. As far as fuel element sales are concerned, Framatome delivers the first core in order to be able to assume full responsibility vis-a-vis the customer for the performance of the nuclear steam supply system. Reloads are sold through the Fragema Association established by Framatome and Cogema. (orig.) [de

  17. Gas reactor international cooperative program interim report. Pebble bed reactor fuel cycle evaluation

    International Nuclear Information System (INIS)

    1978-09-01

    Nuclear fuel cycles were evaluated for the Pebble Bed Gas Cooled Reactor under development in the Federal Republic of Germany. The basic fuel cycle specified for the HTR-K and PNP is well qualified and will meet the requirements of these reactors. Twenty alternate fuel cycles are described, including high-conversion cycles, net-breeding cycles, and proliferation-resistant cycles. High-conversion cycles, which have a high probability of being successfully developed, promise a significant improvement in resource utilization. Proliferation-resistant cycles, also with a high probability of successful development, compare very favorably with those for other types of reactors. Most of the advanced cycles could be adapted to first-generation pebble bed reactors with no significant modifications

  18. Utilization of particle fuels in different reactor concepts

    International Nuclear Information System (INIS)

    1983-04-01

    To date, particle fuel is only used in high temperature reactors (HTR). In this reactor type the particles exist of oxide fuel with a diameter of about 0.5 mm and are surrounded by various coatings in order to safely enclose fission products and decrease the radioactive release into the primary circuit. However, it is felt that fuel based upon spherical particles could have some advantages compared with pellets both on fabrication and in-core behaviour in several reactor concepts. This fuel is now of general interest and there is a high level of research and development activity in some countries. In order to collect, organize additional information and summarize experience on utilization of particle fuels in different reactor concepts, a questionnaire was prepared by IAEA in 1980 and sent to Member States, which might be involved in relevant developments. This survey has been prepared by a group of consultants and is mainly based on the responses to the IAEA questionnaire

  19. Nuclear reactor fuel cycle technology with pyroelectrochemical processes

    International Nuclear Information System (INIS)

    Skiba, O.V.; Maershin, A.A.; Bychkov, A.V.; Zhdanov, A.N.; Kislyj, V.A.; Vavilov, S.K.; Babikov, L.G.

    1999-01-01

    A group of dry technologies and processes of vibro-packing granulated fuel in combination with unique properties of vibro-packed FEs make it possible to implement a new comprehensive approach to the fuel cycle with plutonium fuel. Testing of a big number of FEs with vibro-packed U-Pu oxide fuel in the BOR-60 reactor, successful testing of experimental FSAs in the BN-600 rector, reliable operation of the experimental and research complex facilities allow to make the conclusion about a real possibility to develop a safe, economically beneficial U-Pu fuel cycle based on the technologies enumerated above and to use both reactor-grade and weapon-grade plutonium in nuclear reactors with a reliable control and accounting system [ru

  20. Fuel motion in overpower tests of metallic integral fast reactor fuel

    International Nuclear Information System (INIS)

    Rhodes, E.A.; Bauer, T.H.; Stanford, G.S.; Regis, J.P.; Dickerman, C.E.

    1992-01-01

    In this paper results from hodoscope data analyses are presented for transient overpower (TOP) tests M5, M6, and M7 at the Transient Reactor Test Facility, with emphasis on transient feedback mechanisms, including prefailure expansion at the tops of the fuel pins, subsequent dispersive axial fuel motion, and losses in relative worth of the fuel pins during the tests. Tests M5 and M6 were the first TOP tests of margin to cladding breach and prefailure elongation of D9-clad ternary (U-Pu-Zr) integral fast reactor-type fuel. Test M7 extended these results to high-burnup fuel and also initiated transient testing of HT-9-clad binary (U-Zr) Fast Flux Test Facility driver fuel. Results show significant prefailure negative reactivity feedback and strongly negative feedback from fuel driven to failure