Heat transfer and fluid flow aspects of fuel--coolant interactions

International Nuclear Information System (INIS)

Corradini, M.L.

1978-09-01

A major portion of the safety analysis effort for the LMFBR is involved in assessing the consequences of a Hypothetical Core Disruptive Accident (HCDA). The thermal interaction of the hot fuel and the sodium coolant during the HCDA is investigated in two areas. A postulated loss of flow transient may produce a two-phase fuel at high pressures. The thermal interaction phenomena between fuel and coolant as the fuel is ejected into the upper plenum are investigated. A postulated transient overpower accident may produce molten fuel being released into sodium coolant in the core region. An energetic coolant vapor explosion for these reactor materials does not seem likely. However, experiments using other materials (e.g., Freon/water, tin/water) have demonstrated the possibility of this phenomenon

FFTF metal fuel pin sodium bond quality verification

International Nuclear Information System (INIS)

Pitner, A.L.; Dittmer, J.O.

1988-12-01

The Fast Flux Test Facility (FFTF) Series III driver fuel design consists of U-10Zr fuel slugs contained in a ferritic alloy cladding. A liquid metal, sodium bond between the fuel and cladding is required to prevent unacceptable temperatures during operation. Excessive voiding or porosity in the sodium thermal bond could result in localized fuel melting during irradiation. It is therefore imperative that bond quality be verified during fabrication of these metal fuel pins prior to irradiation. This document discusses this verification

Thermodynamic Data to Model the Interaction Between Coolant and Fuel in Gen IV Sodium Cooled Fast Reactors

International Nuclear Information System (INIS)

Dinsdale, Alan; Gisby, John; Davies, Hugh; Konings, Rudy; Benes, Ondrej

2013-06-01

Understanding the behaviour of nuclear fuels in various environments is vital to the design and safe operation of nuclear reactors. While this is true if the reactor is operating within its design specification, it is even more so if accidents occur and the fuel is exposed to unexpected temperatures, pressures or chemical environments. It is clearly hazardous and costly to explore all such scenarios experimentally and therefore it is necessary to undertake modelling where possible using well-grounded theoretical approaches. This paper will show examples of where calculations of chemical and phase equilibria have been applied successfully to the long term storage of nuclear waste, phase formation during core meltdown and prediction of fission product release into the atmosphere. It will also highlight the development of thermodynamic data carried out during the European Metrology Research Project Metrofission required to model the potential interaction between the coolant, nuclear fuel, containment materials and atmosphere of a sodium cooled fast reactor. (authors)

Fuel Coolant Interaction Results in the Fuel Pins Melting Facility (PMF)

International Nuclear Information System (INIS)

Urunashi, H.; Hirabayashi, T.; Mizuta, H.

1976-01-01

The experimental work related to FCI at PNC has been concentrated into the molten UO 2 dropping test. After the completion of molten UO 2 drop experiments, emphasis is directed toward the FCI phenomena of the initiating conditions of the accident under the more realistic geometry. The experiments are conducted within the Pin Melt Facility (PMF) in which UO 2 pellets clad in stainless steel are melted by direct electric heating under the stagnant or flowing sodium. The primary objectives of the PMF test are to: - obtain detail experimental results (heat-input, clad temperature, sodium temperature, etc.) on the FCI under TOP and LOF conditions; - observe the movement of the fuel before and after the pin failure by the X-ray cinematography; - observe the degree of coherence of the pin failures; - accumulate the experience of the FCI experiment which is applicable to the subassembly or more larger scale; - simulate the fuel behavior of the in-pile test (GETR, CABRI). The preliminary conclusions can be drawn from the foregoing observations are as follows: - Although the fuel motion and FCI of the closed test section appeared to be different from those of the open test section, the conclusion of the effect of the inside pressure on FCI needs more experimental data. - The best heating condition of the UO 2 pellet for the FCI study with PMF is established as 40 w/cm at the steady state and 1680 J/g of UO 2 during the additional transient state. The total energy deposition of the UO 2 pellet is thus estimated in the range of 2400 J/g of UO 2 -2600 J/g of UO 2 . The analytical model of the fuel pin failure and the subsequent FCI are suggested to count the following parameters: - The fuel pin failure due to the fuel vaporization due to the rapid energy deposition; - Molten fuel, clad and sodium interaction in the fuel pin after the pin failure; - The upward flow of molten fuel with molten clad or vapor sodium, as well as the slumping of molten fuel

Biological behavior of mixed LMFBR-fuel-sodium aerosols

International Nuclear Information System (INIS)

Mahlum, D.D.; Hackett, P.L.; Hess, J.O.; Allen, M.D.

1979-01-01

Immediately after exposure of rats to mixed aerosols of sodium-LMFBR fuel, about 80 to 90% of the body burden of 239 Pu is in the gastrointestinal tract; 1.5 to 4% is in the lungs. With fuel-only aerosols, less of the body burden was in the GI tract and more in the lung and the head. Blood and urine values suggest an increased absorption of 239 Pu from sodium-fuel than from fuel-only aerosols

Deposition of inhaled LMFBR-fuel-sodium aerosols in beagle dogs

International Nuclear Information System (INIS)

Hackett, P.L.; Mahlum, D.D.; Briant, J.K.; Catt, D.L.; Peters, L.R.; Clary, A.J.

1980-01-01

Initial alveolar deposition of LMFBR-fuel aerosols in beagle dogs amounted to 30% of the inhaled activity, but only 5% of the total inhaled activity was deposited in dogs exposed to sodium-fuel aerosols. Aerosol deposition in the gastrointestinal tract amounted to 4% of the initial body burden of fuel-aerosol exposed dogs and 24% of the burden of animals receiving sodium-fuel aerosols. Preliminary analytical data for the dog exposures appear to agree with rodent data for deposition and distribution patterns of aerosols of similar sodium: fuel ratios

Interactions of hydrazine, ferrous sulfamate, sodium nitrite, and nitric acid in nuclear fuel processing solutions

International Nuclear Information System (INIS)

Gray, L.W.

1977-03-01

Hydrazine and ferrous sulfamate are used as reductants in a variety of nuclear fuel processing solutions. An oxidant, normally sodium nitrite, must frequently be added to these nitric acid solutions before additional processing can proceed. The interactions of these four chemicals have been studied under a wide variety of conditions using a 2/sup p/ factorial experimental design to determine relative reaction rates for desired reactions and side reactions. Evidence for a hydrazine-stabilized, sulfamic acid--nitrous acid intermediate was obtained; this intermediate can hydrolyze to ammonia or decompose to nitrogen. The oxidation of Fe 2+ by NO 2 - was shown to proceed at about the same rate as the scavenging of NO 2 - by sulfamic acid. Various side reactions are discussed

Design and performance of sodium-bonded uranium--plutonium carbide fuel elements

International Nuclear Information System (INIS)

Kerrisk, J.F.; DeMuth, N.S.; Petty, R.L.; Latimer, T.W.; Vitti, J.A.; Jones, L.J.

1979-01-01

Recent results from irradiation tests indicate that sodium-bonded elements provide a practical advanced fuel element design for use in LMFBRs. Shroud tubes have effectively controlled fuel-cladding mechanical interaction; thicker and stronger claddings have also been effective in this respect. Burnups to 11 at.% have been achieved under typical operating conditions. A hetrogeneous core with a breeding ratio of 1.55 and a compound system doubling time of less than 13 years has been designed using these element designs

Current Status of Experimental and Theoretical Work on Sodium/Fuel Interaction (SFI) at Karlsruhe 'Code Developments'

International Nuclear Information System (INIS)

Beutel, H.; Bojarsky, E.; Reiser, H.; Caldarola, L.; Jacobs, H.; Zyszkowski, W.

1976-01-01

The theoretical work follows two main lines: A. Code development; B. Theoretical work on fragmentation. Two computer codes have been developed. The first code contains a heat transfer model (during the vaporization phase) based on the inverse Leidenfrost phenomenon (which has been observed experimentally in water). The exact solution of the heat diffusion equation in a sphere is included in the code. The code accounts for the time history of each fuel particle by means of specially averaged temperature values. The presence of fission gases can also be taken into account. A size distribution of fuel particles has also been incorporated in the code as well as the effect of the friction due to the channel walls and that of the pressure losses at channel outlet. An extensive parametric study has been carried out with this code. The main conclusions are the following: 1. Total mechanical work strongly decreases with the fragmentation and/or mixing time constants. 2. Vapour blanketing during the vaporization phase is effective only if accompanied by a relatively slow process of fragmentation and mixing. In this case total mechanical work strongly decreased with degree of vapour blanketing. 3. Total mechanical work rises with initial length of sodium piston. 4. Time to empty the 120 cm long channel is 15-20 msecs. for values of the fragmentation and/or mixing time constants of the order of 5-10 msecs. 5. Effects due to particle size distribution and gas content are important only fora rapid fragmentation and mixing process. It must be painted out that (as far as the gas is concerned) this conclusion is valid only within the limits of the effects (due to the gas) which have been considered in the model. Propagation effects can be analysed by using the second code. The interaction region can be subdivided into an arbitrary number of sections, each containing fuel and coolant. The thermal conductivity of the liquid sodium has also been taken into account, as well as the

Determining Bond Sodium Remaining in Plenum Region of Spent Nuclear Driver Fuel

International Nuclear Information System (INIS)

Vaden, D.; Li, S.X.

2008-01-01

The Fuel Conditioning Facility (FCF) at the Idaho National Laboratory (INL) treats spent nuclear fuel using an electro-chemical process that separates the uranium from the fission products, sodium thermal bond, and cladding materials (REF 1). Upon immersion into the ER electrolyte, the sodium used to thermally bond the fuel to the clad jacket chemically reacts with the UCl3 in the electrolyte producing NaCl and uranium metal. The uranium in the spent fuel is separated from the cladding and fission products by taking advantage of the electro-chemical potential differences between uranium and the other fuel components. Assuming all the sodium in the thermal bond is converted to NaCl in the ER, the difference between the cumulative bond sodium mass in the fuel elements and the cumulative sodium mass found in the driver ER electrolyte inventory provides an upper mass limit for the sodium that migrated to the upper gas region, or plenum section, of the fuel element during irradiation in the reactor. The plenums are to be processed as metal waste via melting and metal consolidation operations. However, depending on the amount of sodium in the plenums, additional processing may be required to remove the sodium before metal waste processing

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

Fuel-coolant interactions in a jet contact mode

International Nuclear Information System (INIS)

Konishi, K.; Isozaki, M.; Imahori, S.; Kondo, S.; Furutani, A.; Brear, D.J.

1994-01-01

Molten fuel-coolant interactions in a jet contact mode was studied with respect to the safety of liquid-metal-cooled fast reactors (LMFRs). From a series of molten Wood's metal (melting point: 79 deg. C, density: -8400 kg/m 3 ) jet-water interaction experiments, several distinct modes of interaction behaviors were observed for various combinations of initial temperature conditions of the two fluids. A semi-empirical model for a minimum film boiling temperature criterion was developed and used to reasonably explain the different interaction modes. It was concluded that energetic jet-water interactions are only possible under relatively narrow initial thermal conditions. Preliminary extrapolation of the present results in an oxide fuel-sodium system suggests that mild interactions with short breakup length and coolable debris formation should be most likely in LMFRs. (author)

Preliminary Analysis of the Bundle-Duct Interaction for the fuel of SFR

Energy Technology Data Exchange (ETDEWEB)

Lee, Byoung Oon; Cheon, Jin Sik; Hahn, Do Hee; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2008-10-15

BDI (Bundle-Duct Interaction) occurs in the fuel of SFR (Sodium-cooled Fast Reactor) due to the radial expansion and bowing of a fuel pin bundle. Under the BDI condition, excess cladding strain and hot spots would occur. Therefore, BDI, which is the dominant deformation mechanisms in a fuel pin bundle, should be considered to evaluate the FBR fuel integrity. The analysis codes such as ETOILE and BMBOO, have been developed to evaluate the BDI behavior. The bundle duct interaction model is also being developed for SFR in Korea. This model is based on ANSYS. In this paper, the fuel pin configuration model for the BDI calculation was established. The preliminary analysis of the bundle-duct interaction was performed to evaluate the fuel design concept.

Subchannel analysis of sodium-cooled reactor fuel assemblies with annular fuel pins

International Nuclear Information System (INIS)

Memmott, Matthew; Buongiorno, Jacopo; Hejzlar, Pavel

2009-01-01

Using a RELAP5-3D subchannel analysis model, the thermal-hydraulic behavior of sodium-cooled fuel assemblies with internally and externally cooled annular fuel rods was investigated, in an effort to enhance the economic performance of sodium-fast reactors by increasing the core power density, decreasing the core pressure drop, and extending the fuel discharge burnup. Both metal and oxide fuels at high and low conversion ratios (CR=0.25 and CR=1.00) were investigated. The externally and internally cooled annular fuel design is most beneficial when applied to the low CR core, as clad temperatures are reduced by up to 62.3degC for the oxide fuel, and up to 18.5degC for the metal fuel. This could result in a power uprates of up to ∼44% for the oxide fuel, and up to ∼43% for the metal fuel. The use of duct ribs was explored to flatten the temperature distribution at the core outlet. Subchannel analyses revealed that no fuel melting would occur in the case of complete blockage of the hot interior-annular channel for both metal and oxide fuels. Also, clad damage would not occur for the metal fuel if the power uprate is 38% or less, but would indeed occur for the oxide fuel. (author)

Calculations of the possible consequences of molten fuel sodium interactions in subassembly and whole core geometries

International Nuclear Information System (INIS)

Coddington, P.; Fishlock, T.P.; Jakeman, D.

1976-01-01

The possible consequences of molten fuel sodium interactions are calculated using various modelling assumptions and key parameters. And the significance of the choice of assumptions and parameters are discussed. As for subassembly geometry, the results of one-dimensional code EXPEL are compared with the solutions of the one-dimensional Lagrangian equations of a compressible fluid (TOPAL was used). The adequacy of acoustic approximation used in EXPEL is discussed here. The effects of heat transfer time constant on the behaviour of peak pressure are also analyzed by parametric surveys. Other items investigated are the length and position of the interacting zone, the existence of a non-condensable gas volume, and the vapour condensation on cold clad. As for whole core geometry, a simple dynamical model of arc expanding spherical interacting zone immersed in a semi-infinite sea of cold liquid was used (SHORE code). Within the interacting zone a simple heat transfer model (including a heat transfer time and a fragmentation time) was adopted. Vapour blanketing was considered in a number of ways. Representative results of the calculations are given in a table. Containment studies were also performed for ''ducted'' design and ''open pool'' design. The development of new codes in the U.K. for these analysis are also briefly described. (Aoki, K.)

Innovative technologies on fuel assemblies cleaning for sodium fast reactors: First considerations on cleaning process

International Nuclear Information System (INIS)

Simon, N.; Lorcet, H.; Beauchamp, F.; Guigues, E.; Lovera, P.; Fleche, J. L.; Lacroix, M.; Carra, O.; Dechelette, F.; Prele, G.; Rodriguez, G.

2012-01-01

Within the framework of Sodium Fast Reactor development, innovative fuel assembly cleaning operations are investigated to meet the GEN IV goals of safety and of process development. One of the challenges is to mitigate the Sodium Water Reaction currently used in these processes. The potential applications of aqueous solutions of mineral salts (including the possibility of using redox chemical reactions) to mitigate the Sodium Water Reaction are considered in a first part and a new experimental bench, dedicated to this study, is described. Anhydrous alternative options based on Na/CO 2 interaction are also presented. Then, in a second part, a functional study conducted on the cleaning pit is proposed. Based on experimental feedback, some calculations are carried out to estimate the sodium inventory on the fuel elements, and physical methods like hot inert gas sweeping to reduce this inventory are also presented. Finally, the implementation of these innovative solutions in cleaning pits is studied in regard to the expected performances. (authors)

Innovative technologies on fuel assemblies cleaning for sodium fast reactors: First considerations on cleaning process

Energy Technology Data Exchange (ETDEWEB)

Simon, N.; Lorcet, H.; Beauchamp, F.; Guigues, E. [CEA, DEN, DTN Cadarache, F-13108 Saint-Paul-lez-Durance (France); Lovera, P.; Fleche, J. L. [CEA, DEN, DPC Saclay, F-91191 Gif-sur-Yvette (France); Lacroix, M. [CEA, DEN, DTN Cadarache, F-13108 Saint-Paul-lez-Durance (France); Carra, O. [AREVA / NP, 10 Rue Juliette Recamier, 69003 Lyon (France); Dechelette, F. [CEA, DEN, DTN Cadarache, F-13108 Saint-Paul-lez-Durance (France); Prele, G. [EDF/SEPTEN, 12-14 avenue Dutrievoz, 69628 Villeurbane Cedex (France); Rodriguez, G. [CEA, DEN, DTN Cadarache, F-13108 Saint-Paul-lez-Durance (France)

2012-07-01

Within the framework of Sodium Fast Reactor development, innovative fuel assembly cleaning operations are investigated to meet the GEN IV goals of safety and of process development. One of the challenges is to mitigate the Sodium Water Reaction currently used in these processes. The potential applications of aqueous solutions of mineral salts (including the possibility of using redox chemical reactions) to mitigate the Sodium Water Reaction are considered in a first part and a new experimental bench, dedicated to this study, is described. Anhydrous alternative options based on Na/CO{sub 2} interaction are also presented. Then, in a second part, a functional study conducted on the cleaning pit is proposed. Based on experimental feedback, some calculations are carried out to estimate the sodium inventory on the fuel elements, and physical methods like hot inert gas sweeping to reduce this inventory are also presented. Finally, the implementation of these innovative solutions in cleaning pits is studied in regard to the expected performances. (authors)

Remote, under-sodium fuel handling experience at EBR-II

International Nuclear Information System (INIS)

King, R.W.; Planchon, H.P.

1995-01-01

The EBR-II is a pool-type design; the reactor fuel handling components and entire primary-sodium coolant system are submerged in the primary tank, which is 26 feet in diameter, 26 feet high, and contains 86,000 gallons of sodium. Since the reactor is submerged in sodium, fuel handling operations must be performed blind, making exact positioning and precision control of the fuel handling system components essential. EBR-II operated for 30 years, and the fuel handling system has performed approximately 25,000 fuel transfer operations in that time. Due to termination of the IFR program, EBR-II was shut down on September 30, 1994. In preparation for decommissioning, all fuel in the reactor will be transferred out of EBR-II to interim storage. This intensive fuel handling campaign will last approximately two years, and the number of transfers will be equivalent to the fuel handling done over about nine years of normal reactor operation. With this demand on the system, system reliability will be extremely important. Because of this increased demand, and considering that the system has been operating for about 32 years, system upgrades to increase reliability and efficiency are proceeding. Upgrades to the system to install new digital, solid state controls, and to take advantage of new visualization technology, are underway. Future reactor designs using liquid metal coolant will be able to incorporate imaging technology now being investigated, such as ultraviolet laser imaging and ultrasonic imaging

Fuel-sodium reaction product formation in breached mixed-oxide fuel

International Nuclear Information System (INIS)

Bottcher, J.H.; Lambert, J.D.B.; Strain, R.V.; Ukai, S.; Shibahara, S.

1988-01-01

The run-beyond-cladding-breach (RBCB) operation of mixed-oxide LMR fuel pins has been studied for six years in the Experimental Breeder Reactor-II (EBR-II) as part of a joint program between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan. The formation of fuel-sodium reaction product (FSRP), Na 3 MO 4 , where M = U/sub 1-y/Pu/sub y/, in the outer fuel regions is the major phenomenon governing RBCB behavior. It increases fuel volume, decreases fuel stoichiometry, modifies fission-product distributions, and alters thermal performance of a pin. This paper describes the morphology of Na 3 MO 4 observed in 5.84-mm diameter pins covering a variety of conditions and RBCB times up to 150 EFPD's. 8 refs., 1 fig

A study of atomic interaction between suspended nanoparticles and sodium atoms in liquid sodium

International Nuclear Information System (INIS)

Saito, Jun-ichi; Ara, Kuniaki

2010-01-01

A feasibility study of suppression of the chemical reactivity of sodium itself using an atomic interaction between nanoparticles and sodium atoms has been carried out. We expected that the atomic interaction strengthens when the nanoparticle metal is the transition element which has a major difference in electronegativity from sodium. We also calculated the atomic interaction between nanoparticle and sodium atoms. It became clear that the atomic bond between the nanoparticle atom and the sodium atom is larger than that between sodium atoms, and the charge transfer takes place to the nanoparticle atom from the sodium atom. Using sodium with suspended nanoparticles, the fundamental physical properties related to the atomic interaction were investigated to verify the atomic bond. The surface tension of sodium with suspended nanoparticles increased, and the evaporation rate of sodium with suspended nanoparticles also decreased compared with that of sodium. Therefore the presence of the atomic interaction between nanoparticles and sodium was verified from these experiments. Because the fundamental physical property changes by the atomic interaction, we expected changes in the chemical reactivity characteristics. The chemical reaction properties of sodium with suspended nanoparticles with water were investigated experimentally. The released reaction heat and the reaction rate of sodium with suspended nanoparticles were reduced than those of sodium. The influence of the charge state of nanoparticle on the chemical process with water was theoretically investigated to speculate on the cause of reaction suppression. The potential energy in both primary and side reactions changed by the charge transfer, and the free energy of activation of the reaction with water increased. Accordingly, the reaction barrier also increased. This suggests there is a possibility of the reduction in the reaction of sodium by the suspension of nanoparticles. Consequently the possibility of the

Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

Energy Technology Data Exchange (ETDEWEB)

Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr

2014-12-15

Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

Electrometallurgical treatment of sodium-bonded spent nuclear fuel

International Nuclear Information System (INIS)

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

2001-01-01

For 20 years Argonne National Laboratory has been developing electrometallurgical technology for application to spent nuclear fuel. Progress has been rapid during the past 5 years as 1,6 tonnes spent fuel from the Experimental Breeder Reactor-II was treated and preparations were made for processing the remaining 25 tonnes of sodium-bonded fuel from the shutdown reactor. Two high level waste forms are being qualified for geologic disposal. Extension of the technology to oxide fuels or to actinide recycling has been on hold because of US policy on reprocessing. (author)

Comparative sodium void effects for different advanced liquid metal reactor fuel and core designs

International Nuclear Information System (INIS)

Dobbin, K.D.; Kessler, S.F.; Nelson, J.V.; Gedeon, S.R.; Omberg, R.P.

1991-01-01

An analysis of metal-, oxide-, and nitride-fueled advanced liquid metal reactor cores was performed to investigate the calculated differences in sodium void reactivity, and to determine the relationship between sodium void reactivity and burnup reactivity swing using the three fuel types. The results of this analysis indicate that nitride fuel has the least positive sodium void reactivity for any given burnup reactivity swing. Thus, it appears that a good design compromise between transient overpower and loss of flow response is obtained using nitride fuel. Additional studies were made to understand these and other nitride advantages. (author)

Final Environmental Impact Statement for the Treatment and Management of Sodium-Bonded Spent Nuclear Fuel

International Nuclear Information System (INIS)

2000-01-01

DOE is responsible for the safe and efficient management of its sodium-bonded spent nuclear fuel. This fuel contains metallic sodium, a highly reactive material; metallic uranium, which is also reactive; and in some cases, highly enriched uranium. The presence of reactive materials could complicate the process of qualifying and licensing DOE's sodium-bonded spent nuclear fuel inventory for disposal in a geologic repository. Currently, more than 98 percent of this inventory is located at the Idaho National Engineering and Environmental Laboratory (INEEL), near Idaho Falls, Idaho. In addition, in a 1995 agreement with the State of Idaho, DOE committed to remove all spent nuclear fuel from Idaho by 2035. This EIS evaluates the potential environmental impacts associated with the treatment and management of sodium-bonded spent nuclear fuel in one or more facilities located at Argonne National Laboratory-West (ANL-W) at INEEL and either the F-Canyon or Building 105-L at the Savannah River Site (SRS) near Aiken, South Carolina. DOE has identified and assessed six proposed action alternatives in this EIS. These are: (1) electrometallurgical treatment of all fuel at ANL-W, (2) direct disposal of blanket fuel in high-integrity cans with the sodium removed at ANL-W, (3) plutonium-uranium extraction (PUREX) processing of blanket fuel at SRS, (4) melt and dilute processing of blanket fuel at ANL-W, (5) melt and dilute processing of blanket fuel at SRS, and (6) melt and dilute processing of all fuel at ANL-W. In addition, Alternatives 2 through 5 include the electrometallurgical treatment of driver fuel at ANL-W. Under the No Action Alternative, the EIS evaluates both the continued storage of sodium-bonded spent nuclear fuel until the development of a new treatment technology or direct disposal without treatment. Under all of the alternatives, the affected environment is primarily within 80 kilometers (50 miles) of spent nuclear fuel treatment facilities. Analyses indicate

Out-of-pile simulation experiments and theoretical analysis on sodium fuel interaction

International Nuclear Information System (INIS)

Conti, M.; Luigi, G. Di; Federico, A.; Mennini, G.; Scarano, G.; Tavano, F.

1978-01-01

Activities on fuel coolant interaction are being carried out since many years at C.N.E.N. in the frame of the Italian Fast Reactor Program. This paper describes the experimental and theoretical results recently obtained. (author)

Multi-criteria methodology to design a sodium-cooled carbide-fueled Gen-IV reactor

International Nuclear Information System (INIS)

Stauff, N.

2011-01-01

Compared with earlier plant designs (Phenix, Super-Phenix, EFR), Gen IV Sodium-cooled Fast Reactor requires improved economics while meeting safety and non-proliferation criteria. Mixed Oxide (U-Pu)O 2 fuels are considered as the reference fuels due to their important and satisfactory feedback experience. However, innovative carbide (U-Pu)C fuels can be considered as serious competitors for a prospective SFR fleet since carbide-fueled SFRs can offer another type of optimization which might overtake on some aspects the oxide fuel technology. The goal of this thesis is to reveal the potentials of carbide by designing an optimum carbide-fueled SFR with competitive features and a naturally safe behavior during transients. For a French nuclear fleet, a 1500 MW(e) break-even core is considered. To do so, a multi-physic approach was developed taking into account neutronics, fuel thermo-mechanics and thermal-hydraulic at a pre-design stage. Simplified modeling with the calculation of global neutronic feedback coefficients and a quasi-static evaluation was developed to estimate the behavior of a core during overpower transients, loss of flow and/or loss of heat removal transients. The breakthrough of this approach is to provide the designer with an overall view of the iterative process, emphasizing the well-suited innovations and the most efficient directions that can improve the SFR design project.This methodology was used to design a core that benefits from the favorable features of carbide fuels. The core developed is a large carbide-fueled SFR with high power density, low fissile inventory, break-even capability and forgiving behaviors during the un-scrammed transients studied that should prevent using expensive mitigate systems. However, the core-peak burnup is unlikely to significantly exceed 100 MWd/kg because of the large swelling of the carbide fuel leading to quick pellet-clad mechanical interaction and the low creep capacity of carbide. Moderate linear power fuel

Interaction of theobromine with sodium benzoate

Energy Technology Data Exchange (ETDEWEB)

Nishijo, J.; Yonetani, I.

1982-03-01

The interaction of theobromine with sodium benzoate was investigated by PMR spectroscopy. The interaction of theobromine with pentadeuterated benzoic acid (benzoic acid-d5) was examined in the same manner but to a lesser degree. Chemical shifts of theobromine protons were determined as a function of sodium benzoate concentration in deuterium oxide at 30 and 15 degrees. Signals of both methyl groups of theobromine underwent significant upfield shifts when sodium benzoate was added to a theobromine solution. This fact suggests that a complex is formed by vertical stacking or plane-to-plane stacking. The same results were obtained for benzoic acid-d5.

Experimental investigations and modelling of sodium-concrete interaction

International Nuclear Information System (INIS)

Schultheiss, G.F.; Deeg, H.J.

1990-01-01

The use of sodium as a coolant in liquid metal fast breeder reactors, fusion reactors, and solar plants requires special consideration of its chemical reactivity and related safety problems in the case of sodium leckage. On contact between hot sodium and concrete an interaction takes place resulting in energy release and hydrogen generation, which may contribute to containment loading by pressurization in a hypothetical accident situation. For this reason, sodium-concrete interactions were investigated experimentally and theoretically. The experiments revealed important effects of quartzitic material within the concrete and of the sodium temperature on the interaction mechanisms, the energy release and the consequent hydrogen production. The numerical model shows good agreement with the experimental results. (orig.) [de

Investigation of Plugging of Narrow Sodium Channels by Sodium and Carbon Dioxide Interaction

Energy Technology Data Exchange (ETDEWEB)

Park, Sun Hee; Wi, Myung-Hwan; Min, Jae Hong; Kim, Tae-joon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

The supercritical CO{sub 2} Brayton cycle system is known to be a promising power conversion system for improving the efficiency and preventing the sodium water reaction (SWR) of the current SFR concept using a Rankine steam cycle. PCHEs are known to have potential for reducing the volume occupied by the sodium-to-CO{sub 2} exchangers as well as the heat exchanger mass relative to traditional shell-and-tube heat exchangers. Here, we report a study on a plugging test by the interaction of sodium and CO{sub 2} to investigate design parameters of sodium channels in the realistic operating conditions. We investigated a plugging test by an interaction of sodium and CO{sub 2} with different cross sectional areas of the sodium channels. It was found that the flow rate of sodium decreased earlier and faster with a narrower cross sectional area compared to a wider one. Our experimental results are expected to be used for determining the sodium channel areas of PCHEs.

Test of fuel handling machine for Monju in sodium

International Nuclear Information System (INIS)

Ishii, Yoichiro; Masuda, Yoichi; Kataoka, Hajime

1980-01-01

Various types of fuel handling machines were studied, and under-the-plug method of fuel exchange and the fuel handling machine of single turning plug, fixed arm type were selected for the prototype reactor ''Monju'', because the turning plug is relatively small, and the rate of operation, safety, operational ability, maintainability and reliability required for the reactor are satisfied, moreover, the extrapolation to the demonstration reactor was considered. Attention must be paid to the points that the fuel handling machine is very long and invisible from outside, and the smooth operation and endurance in sodium are required for it. The full mock-up testing facility of single turning plug, fixed arm type was installed in 1974, and the full mock-up test has been carried out since 1975 in Oarai. Fuel exchange is carried out at about 6 months intervals in Monju, and about 20 to 30% of core and blanket fuels are exchanged for about one month period. The functions required for the fuel handling machine for Monju, the outline of the testing facility, the schedule of the testing, the items of testing and the results, and the matters to be specially written are described. The full mock-up test in sodium has been carried out for 5 years, and the functions and the endurance have been proved sufficiently. (Kako, I.)

Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

International Nuclear Information System (INIS)

Grabaskas, David; Bucknor, Matthew; Jerden, James

2016-01-01

The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.

Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

Energy Technology Data Exchange (ETDEWEB)

Grabaskas, David [Argonne National Lab. (ANL), Argonne, IL (United States); Bucknor, Matthew [Argonne National Lab. (ANL), Argonne, IL (United States); Jerden, James [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-02-01

The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.

Sodium flow distribution in test fuel assembly P-23B

International Nuclear Information System (INIS)

Taylor, J.P.S.

1978-08-01

Relatively large cladding diametral increases in the exterior fuel pins of HEDL's test fuel subassembly P-23B were successfully explained by a thermal-hydraulic/solid mechanics analysis. This analysis indicates that while at power, the subassembly flow was less than planned and that the fuel pins were considerably displaced and bowed from their nominal position. In accomplishing this analysis, a method was developed to estimate the sodium flow distribution and pin distortions in a fuel subassembly at power

Sodium removal of fuel elements by vacuum distillation

International Nuclear Information System (INIS)

Buescher, E.; Haubold, W.; Jansing, W.; Kirchner, G.

1978-01-01

Cleaning of sodium-wetted core components can be performed by using either lead, moist nitrogen, or alcohol. The advantages of these methods for cleaning fuel elements without causing damage are well known. The disadvantage is that large amounts of radioactive liquids are formed during handling in the latter two cases. In this paper a new method to clean components is described. The main idea is to remove all liquid metal from the core components within a comparatively short period of time. Fuel elements removed from the reactor must be cooled because of high decay heat release. To date, vacuum distillation of fuel elements has not yet been applied

In-pile study of the reaction between breeder fuel and sodium

International Nuclear Information System (INIS)

Hugot, J.P.

1982-10-01

Studies carried out until now show that the determinant parameter of fuel can failure evolution is the development of the reaction between mixed uranium and plutonium dioxide and sodium. The parameters of the reaction are presented from results of an out of pile study, as also results obtained from examination on pins failed in reactors. The best way to study in pile the development of the reaction was to irradiate at a constant power a fuel pin containing sodium. In the experiment, the pin was equipped with a central thermocouple. It shows, that the reaction is developing intergranularly, from cracks and interpellet spaces, in an internal fringe of the fuel before spreading to the periphery. An overheating of the pin is associated to the development of the reaction as also a modification of the fuel pin geometry and a reduction of the oxide [fr

A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

Energy Technology Data Exchange (ETDEWEB)

Karahan, Aydin, E-mail: karahan@mit.ed [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology (United States); Buongiorno, Jacopo [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology (United States)

2010-01-31

An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO{sub 2}-PuO{sub 2} mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium

A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

International Nuclear Information System (INIS)

Karahan, Aydin; Buongiorno, Jacopo

2010-01-01

An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO 2 -PuO 2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium reactors

Development of failed fuel detection and location system in sodium-cooled large reactor. Sampling method of failed fuels under the slit

International Nuclear Information System (INIS)

Aizawa, Kousuke; Fujita, Kaoru; Kamide, Hideki; Kasahara, Naoto

2010-01-01

A conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR) is in progress as an issue of the 'Fast Reactor Cycle Technology Development (FaCT)' project in Japan. JSFR adopts a Selector-Valve mechanism for the failed fuel detection and location (FFDL) system. The Selector-Valve FFDL system identifies failed fuel subassemblies by sampling sodium from each fuel subassembly outlet and detecting fission product. One of the JSFR design features is employing an upper internal structure (UIS) with a radial slit, in which an arm of fuel handling machine can move and access the fuel assemblies under the UIS. Thus, JSFR cannot place sampling nozzles right above the fuel subassemblies located under the slit. In this study, the sampling method for indentifying under-slit failed fuel subassemblies has been demonstrated by water experiments. (author)

Analysis of the sodium concrete interactions with the NABE code

International Nuclear Information System (INIS)

Soule, N.

1989-01-01

Experimental studies have been performed in France to investigate sodium-concrete interactions: thermal decomposition of concrete, specific chemical reactions, experimentation in liquid and vapour phase, sodium-concrete interaction without liner protection. Simultaneously computer codes have been developed in order to study the response of the containment building of a liquid metal fast breeder reactor to a sodium pool fire worsened by a sodium-concrete interaction: the NABE code. This code takes into account: a) sodium combustion; b) thermal decomposition of concrete with associated chemical reactions: (liquid sodium-vapour water reaction, liquid sodium-carbon dioxide reaction, liquid sodium-solid compounds of concrete, hydrogen combustion); c) chemical reactions in vapour phase; d) decay heat; e) gas aerosol inlets/outlets; f) aerosol behaviour (sedimentation, diffusion, leak); g) thermal exchanges. An example of a situation, typical of assessment of beyond design basis situations in LMFBR, is given. (author)

Sodium-cooled fast reactor core designs for transmutation of MHR spent fuel

International Nuclear Information System (INIS)

Hong, S. G.; Kim, Y. H.; Venneri, F.

2010-01-01

In this paper, the core design analyses of sodium cooled fast reactors (SFR) are performed for the effective transmutation of the DB (Deep Burn)-MHR (Modular Helium Reactor). In this concept, the spent fuels of DB-MHR are transmuted in SFRs with a closed fuel cycle after TRUs from LWR are first incinerated in a DB-MHR. We introduced two different type SFR core designs for this purpose, and evaluated their core performance parameters including the safety-related parameters. In particular, the cores are designed to have lower transmutation rate relatively to our previous work so as to make the fuel characteristics more feasible. The first type cores which consist of two enrichment regions are typical homogeneous annular cores and they rate 900 MWt power. On the other hand, the second type cores which consist of a central non-fuel region and a single enrichment fuel region rate relatively higher power of 1500 MWt. For these cores, the moderator rods (YH 1.8 ) are used to achieve less positive sodium void worth and the more negative Doppler coefficient because the loading of DB-MHR spent fuel leads to the degradation of these safety parameters. The analysis results show that these cores have low sodium void worth and negative reactivity coefficients except for the one related with the coolant expansion but the coolant expansion reactivity coefficient is within the typical range of the typical SFR cores. (authors)

Fast reactor fuel failures and steam generator leaks: Transient and accident analysis approaches

International Nuclear Information System (INIS)

1996-10-01

This report consists of a survey of activities on transient and accident analysis for the LMFR. It is focused on the following subjects: Fuel transient tests and analyses in hypothetical incident/accident situations; sodium-water interaction in steam generators, and sodium fires: test and analyses. There are also sections dealing with the experimental and analytical studies of: fuel subassembly failures; sodium boiling, molten fuel-coolant interaction; molten material movement and relocation in fuel bundles; heat removal after an accident or incident; sodium-water reaction in steam generator; steam generator protection systems; sodium-water contact in steam generator building; fire-fighting methods and systems to deal with sodium fires. Refs, figs, tabs

A design study of high breeding ratio sodium cooled metal fuel core without blanket fuels

International Nuclear Information System (INIS)

Kobayashi, Noboru; Ogawa, Takashi; Ohki, Shigeo; Mizuno, Tomoyasu; Ogata, Takanari

2009-01-01

The metal fuel core is superior to the mixed oxide fuel core because of its high breeding ratio and compact core size resulting from hard neutron spectrum and high heavy metal densities. Utilizing these characteristics, a conceptual design for a high breeding ratio was performed without blanket fuels. The design conditions were set so a sodium void worth of less than 8 $, a core height of less than 150 cm, the maximum cladding temperature of 650degC, and the maximum fuel pin bundle pressure drop of 0.4 MPa. The breeding ratio of the resultant core was 1.34 with 6wt% zirconium content fuel. Applying 3wt% zirconium content fuel enhanced the breeding ratio up to 1.40. (author)

Mechanical energy yields and pressure volume and pressure time curves for whole core fuel-coolant interactions

Energy Technology Data Exchange (ETDEWEB)

Coddington, P [United Kingdom Atomic Energy Authority, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

1979-10-15

In determining the damage consequences of a whole core Fuel-Coolant Interaction (FCI), one measure of the strength of a FCI that can be used and is independent of the system geometry is the constant volume mixing mechanical yield (often referred to as the Hicks-Menzies yield), which represents a near upper limit to the mechanical work of a FCI. This paper presents a recalculation of the Hicks-Menzies yields for UO{sub 2} and sodium for a range of initial fuel temperatures and fuel to coolant mass ratios, using recently published UO{sub 2} and sodium equation of state data. The work presented here takes a small number of postulated FCIs with as wide range as possible of thermal interaction parameters and determines their pressure-volume P(V) and pressure-time P(t) relations, using geometrical constraints representative of the reactor. Then by examining these P(V) and P(t) curves a representative pressure-relative volume curve or range of possible curves, for use in containment analysis, is recommended

Sodium-cooled Fast Reactor Cores using Uranium-Free Metallic Fuels for Maximizing TRU Support Ratio

International Nuclear Information System (INIS)

You, WuSeung; Hong, Ser Gi

2014-01-01

The depleted uranium plays important roles in the SFR burner cores because it substantially contributes to the inherent safety of the core through the negative Doppler coefficient and large delayed neutron. However, the use of depleted uranium as a diluent nuclide leads to a limited value of TRU support ratio due to the generation of TRUs through the breeding. In this paper, we designed sodium cooled fast reactor (SFR) cores having uranium-free fuels 3,4 for maximization of TRU consumption rate. However, the uranium-free fuelled burner cores can be penalized by unacceptably small values of the Doppler coefficient and small delayed neutron fraction. In this work, metallic fuels of TRU-(W or Ni)-Zr are considered to improve the performances of the uranium-free cores. The objective of this work is to consistently compare the neutronic performances of uranium-free sodium cooled fast reactor cores having TRU-Zr metallic fuels added with Ni or W and also to clarify what are the problematic features to be resolved. In this paper, a consistent comparative study of 400MWe sodium cooled burner cores having uranium-based fuels and uranium-free fuels was done to analyze the relative core neutronic features. Also, we proposed a uranium-free metallic fuel based on Nickel. From the results, it is found that tungsten-based uranium-free metallic fuel gives large negative Doppler coefficient due to high resonance of tungsten isotopes but this core has large sodium void worth and small effective delayed neutron fraction while the nickel-based uranium-free metallic fuelled core has less negative Doppler coefficient but smaller sodium void worth and larger effective delayed neutron fraction than the tungsten-based one. On the other hand, the core having TRU-Zr has very high burnup reactivity swing which may be problematic in compensating it using control rods and the least negative Doppler coefficient

Sodium removal from the grapples of the fuel handling facility of Joyo

Energy Technology Data Exchange (ETDEWEB)

Mukaibo, R; Matsuno, Y; Sato, I; Yoneda, Y; Sato, H [O-arai Engineering Centre, PNC, Ibaraki-ken, Tokio (Japan)

1978-08-01

Sodium removal from the grapples of the fuel handling facility of 'JOYO' is done in alcohol. The operations of the cleaning facility started as the functional tests of the fuel handling facility began. Since then, criticality test and low power tests had been done and during this period, sodium removal from the grapples, after a certain amount of time in use, were done. In order to lessen the time for the cleaning process for the grapples of the machines inside the containment vessel, demineralized water concentration in the alcohol was gained to as much as 10% and good results were obtained. On the other hand, there were very small amounts of sodium on the grapples of the machine used outside the containment vessel and direct charging of demineralized water into the cleaning pot was done experimentally, also with good results. In this report, the sodium removal experience of the grapples before power up tests and some remarks on the improvements of the facility for the future are presented. (author)

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.

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

Large scale sodium interactions. Part 2. Preliminary test results for limestone concrete

International Nuclear Information System (INIS)

Smaardyk, J.E.; Sutherland, H.J.; King, D.L.; Dahlgren, D.A.

1977-01-01

Any sodium cooled reactor system must consider the interaction of hot sodium with cell liners, and given either a failed liner or a hypothetical core disruptive accident, the interaction of hot sodium with concrete. The data base available for safety assessments involving these interactions is limited, especially for the concrete and failed liner interactions. To better understand what happens when hot sodium comes in contact with concrete, a series of tests is being carried out to investigate sodium-concrete reactions under conditions which are similar to actual reactor accident conditions. Tests cover the cases of sodium spills on bare concrete and on cells with defective steel liners. Specific objectives have been to obtain a complete description of the sodium/concrete interaction including heat balance, gas evolution and flow, movement and heat generation of the reaction zone, reaction product formation, and the layering or movement of the products

Behaviour of conductivity improvers in jet fuel

Energy Technology Data Exchange (ETDEWEB)

Dacre, B.; Hetherington, J.I. [Cranfield Univ., Wiltshire (United Kingdom)

1995-05-01

Dangerous accumulation of electrostatic charge can occur due to high speed pumping and microfiltration of fuel. This can be avoided by increasing the electrical conductivity of the fuel using conductivity improver additives. However, marked variations occur in the conductivity response of different fuels when doped to the same level with conductivity improver. This has been attributed to interactions of the conductivity improver with other fuel additives or fuel contaminants. The present work concentrates on the effects of fuel contaminants, in particular polar compounds, on the performance of the conductivity improver. Conductivity is the fuel property of prime interest. The conductivity response of model systems of the conductivity improver STADIS 450 in dodecane has been measured and the effect on this conductivity of additions of model polar contaminants sodium naphthenate, sodium dodecyl benzene sulphonate, and sodium phenate have been measured. The sodium salts have been found to have a complex effect on the performance of STADIS 450, reducing the conductivity at low concentrations to a minimum value and then increasing the conductivity at high concentrations of sodium salts. This work has focused on characterising this minimum in the conductivity values and on understanding the reason for its occurrence. The effects on the minimum conductivity value of the following parameters are investigated: (a) time, (b) STADIS 450 concentration, (c) sodium salt concentration, (d) mixed sodium salts, (e) experimental method, (f) a phenol, (g) individual components of STADIS 450. The complex conductivity response of the STADIS 450 to sodium salt impurities is discussed in terms of possible inter-molecular interactions.

Advanced sodium fast reactor accident source terms :

Energy Technology Data Exchange (ETDEWEB)

Powers, Dana Auburn; Clement, Bernard; Denning, Richard; Ohno, Shuji; Zeyen, Roland

2010-09-01

An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic event Energetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolant Entrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached cladding Rates of radionuclide leaching from fuel by liquid sodium Surface enrichment of sodium pools by dissolved and suspended radionuclides Thermal decomposition of sodium iodide in the containment atmosphere Reactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

Impact of reducing sodium void worth on the severe accident response of metallic-fueled sodium-cooled reactors

International Nuclear Information System (INIS)

Wigeland, R.A.; Turski, R.B.; Pizzica, P.A.

1994-01-01

Analyses have performed on the severe accident response of four 90 MWth reactor cores, all designed using the metallic fuel of the Integrated Fast Reactor (IFR) concept. The four core designs have different sodium void worth, in the range of -3$ to 5$. The purpose of the investigation is to determine the improvement in safety, as measured by the severe accident consequences, that can be achieved from a reduction in the sodium void worth for reactor cores designed using the IFR concept

Experimental and analytical studies of sodium interactions with various concretes

International Nuclear Information System (INIS)

Suo-Anttila, A.; Smaardyk, J.E.

1982-01-01

Mechanistic models of sodium/concrete interactions are described. The SCAM model of interactions with basaltic concrete is being verified by experiments. Modelling of sodium interactions with limestone concrete is still at a preliminary stage but shows promise of being able to predict quantitatively the experimental data. Comparisons with experimental data are presented

Molecular modeling studies of interactions between sodium polyacrylate polymer and calcite surface

Energy Technology Data Exchange (ETDEWEB)

Ylikantola, A. [University of Jyväskylä, Department of Chemistry, P.O. Box 35, University of Jyväskylä, FI-40014 (Finland); Linnanto, J., E-mail: juha.m.linnanto@gmail.com [University of Jyväskylä, Department of Chemistry, P.O. Box 35, University of Jyväskylä, FI-40014 (Finland); University of Tartu, Institute of Physics, Riia 142, EE-51014 Tartu (Estonia); Knuutinen, J.; Oravilahti, A. [University of Jyväskylä, Department of Chemistry, P.O. Box 35, University of Jyväskylä, FI-40014 (Finland); Toivakka, M. [Åbo Akademi University, Laboratory of Paper Coating and Converting and Center for Functional Materials, FI-20500 Turku/Åbo (Finland)

2013-07-01

The interactions between calcite pigment and sodium polyacrylate dispersing agent, widely used in papermaking as paper coating components, were investigated using classical force field and quantum chemical approaches. The objective was to understand interactions between the calcite surface and sodium polyacrylate polymer at 300 K using molecular dynamics simulations. A quantum mechanical ab initio Hartree–Fock method was also used to obtain detailed information about the sodium polyacrylate polymer structure. The effect of water molecules (moisture) on the interactions was also examined. Calculations showed that molecular weight, branching and the orientation of sodium polyacrylate polymers influence the interactions between the calcite surface and the polymer. The force field applied, and also water molecules, were found to have an impact on all systems studied. Ab initio Hartree–Fock calculations indicated that there are two types of coordination between sodium atoms and carboxylate groups of the sodium polyacrylate polymer, inter- and intra-carboxylate group coordination. In addition, ab initio Hartree–Fock calculations of the structure of the sodium polyacrylate polymer produced important information regarding interactions between the polymers and carboxylated styrene-butadiene latex particles.

Pyroprocessing of oxidized sodium-bonded fast reactor fuel - An experimental study of treatment options for degraded EBR-II fuel

Energy Technology Data Exchange (ETDEWEB)

Hermann, S.D.; Gese, N.J. [Separations Department, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States); Wurth, L.A. [Zinc Air Inc., 5314-A US Hwy 2 West, Columbia Falls, MT 59912 (United States)

2013-07-01

An experimental study was conducted to assess pyrochemical treatment options for degraded EBR-II fuel. As oxidized material, the degraded fuel would need to be converted back to metal to enable electrorefining within an existing electro-metallurgical treatment process. A lithium-based electrolytic reduction process was studied to assess the efficacy of converting oxide materials to metal with a particular focus on the impact of zirconium oxide and sodium oxide on this process. Bench-scale electrolytic reduction experiments were performed in LiCl-Li{sub 2}O at 650 C. degrees with combinations of manganese oxide (used as a surrogate for uranium oxide), zirconium oxide, and sodium oxide. In the absence of zirconium or sodium oxide, the electrolytic reduction of MnO showed nearly complete conversion to metal. The electrolytic reduction of a blend of MnO-ZrO{sub 2} in LiCl - 1 wt% Li{sub 2}O showed substantial reduction of manganese, but only 8.5% of the zirconium was found in the metal phase. The electrolytic reduction of the same blend of MnO-ZrO{sub 2} in LiCl - 1 wt% Li{sub 2}O - 6.2 wt% Na{sub 2}O showed substantial reduction of manganese, but zirconium reduction was even less at 2.4%. This study concluded that ZrO{sub 2} cannot be substantially reduced to metal in an electrolytic reduction system with LiCl - 1 wt% Li{sub 2}O at 650 C. degrees due to the perceived preferential formation of lithium zirconate. This study also identified a possible interference that sodium oxide may have on the same system by introducing a parasitic and cyclic reaction of dissolved sodium metal between oxidation at the anode and reduction at the cathode. When applied to oxidized sodium-bonded EBR-II fuel (e.g., U-10Zr), the prescribed electrolytic reduction system would not be expected to substantially reduce zirconium oxide, and the accumulation of sodium in the electrolyte could interfere with the reduction of uranium oxide, or at least render it less efficient.

Tradeoff of sodium void worth and burnup reactivity swing: Impacts on balance safety position in metallic-fueled cores

International Nuclear Information System (INIS)

Wigeland, R.A.; Turski, R.B.; Pizzica, P.A.

1994-01-01

A study has been conducted to investigate the effect of a lower sodium void worth on the consequences of severe accidents in metallic-fueled sodium-cooled reactors. Four 900 MWth designs were used for the study, where all of the reactor cores were designed based on the metallic fuel of the Integral Fast Reactor (IFR) concept. The four core designs each have different sodium void worth, in the range of -3$ to 5$. The purpose of the investigation was to determine the differences in severe accident response for the four core designs, in order to estimate the improvement in overall safety that could be achieved from a reduction in the sodium void worth for reactor cores which use a metallic fuel form

Alkaline Sodium Hypochlorite Irrigant and Its Chemical Interactions

Directory of Open Access Journals (Sweden)

Patricia P. Wright

2017-09-01

Full Text Available Endodontic irrigating solutions may interact chemically with one another. This is important, because even when solutions are not admixed, they will come into contact with one another during an alternating irrigation technique, forming unwanted by-products, which may be toxic or irritant. Mixing or alternating irrigants can also reduce their ability to clean and disinfect the root canal system of teeth by changing their chemical structure with subsequent loss of the active agent, or by inducing precipitate formation in the root canal system. Precipitates occlude dental tubules, resulting in less penetration of antimicrobials and a loss of disinfection efficacy. Sodium hypochlorite is not only a very reactive oxidizing agent, but is also the most commonly used endodontic irrigant. As such, many interactions occurring between it and other irrigants, chelators and other antimicrobials, may occur. Of particular interest is the interaction between sodium hypochlorite and the chelators EDTA, citric acid and etidronate and between sodium hypochlorite and the antimicrobials chlorhexidine, alexidine, MTAD and octenisept.

Experimental investigations of heat transfer during sodium boiling in fuel assembly model in justification of advanced fast reactor safety

International Nuclear Information System (INIS)

Khafizov, R.R.; Poplavskij, V.M.; Rachkov, V.I.; Sorokin, A.P.; Ashurko, Yu.M.; Volkov, A.V.; Ivanov, E.F.; Privezentsev, V.V.

2015-01-01

The experimental facility is built up and investigation of heat exchange during sodium boiling in simulated fast reactor core assembly in conditions of natural and forced circulation with sodium plenum and upper end shield model are conducted. It is shown that in the presence of sodium plenum there is possibility to provide long-term cooling of fuel assembly when heat flux density on the surface of fuel element simulator up to 140 and 170 kW/m 2 in conditions of natural and forced circulation, respectively. The obtained data is used for improving calculational model of sodium boiling process in fuel assembly and calculational code COREMELT verification. It is pointed out that heat transfer coefficients in the case of liquid metal boiling in fuel assemblies are slightly over the ones in the case of liquid metals boiling in pipes and pool boiling [ru

Investigation of sodium - carbon dioxide interactions with calorimetric studies

International Nuclear Information System (INIS)

Simon, N.; Latge, C.; Gicquel, L.

2007-01-01

The supercritical CO 2 Brayton cycle could be a promising option to enhance the competitiveness of future Sodium fast reactors but it is highly necessary to get thermodynamic and kinetics information on potential sodium-CO 2 chemical reactions and their consequences. We have studied the interaction between Na and CO 2 via calorimetric methods. These methods are able to point out exothermic/endothermic phenomena and to measure heat of chemical reactions. The main feature of the Na/CO 2 interaction seems to be its sharp dependence on temperature. At low temperature, below 500 C degrees, CO 2 and sodium react and exhibit an induction time which decreases when temperature increases. Above 500 C degrees, we observe a global phenomenon with a fast and instantaneous chemical reaction which may be understood as an auto-combustion of CO 2 in sodium. We clearly demonstrated that Na/CO 2 interaction does not proceed as an auto-catalytic process and is more satisfactorily explained by the occurring of an auto-combustion phenomenon

Investigation of vapor explosions with alumina droplets in sodium

International Nuclear Information System (INIS)

Zimmer, H.J.

1991-02-01

Within the analysis of severe hypothetical fast breeder accidents the consequence of a fuel-coolant interaction has to be considered i.e. the thermal interaction between hot molten fuel and sodium. Experiments have been performed to study the thermal fragmentation of a molten alumina droplet in sodium. Alumina temperatures up to 3100 K and sodium temperatures up to 1143 K were used. For the first time film boiling of alumina drops in sodium was achieved. With some droplets undergoing film boiling, the fragmentation was triggered by an externally applied pressure wave. The trigger was followed promptly by a strong reaction pressure wave if and only if a contact temperature threshold of T I =2060±160 K was exceeded. In agreement with similar experiments in which other materials were studied this threshold corresponds to an interfacial temperature close to the homogeneous nucleation temperature of the vaporising liquid. Based on the present and previous experimental results a model concept of thermal fragmentation is developed. (orig.) [de

Fission product concentration evolution in sodium pool following a fuel subassembly failure in an LMFBR

International Nuclear Information System (INIS)

Natesan, K.; Velusamy, K.; Selvaraj, P.; Kasinathan, N.; Chellapandi, P.; Chetal, S.; Bhoje, S.

2003-01-01

During a fuel element failure in a liquid metal cooled fast breeder reactor, the fission products originating from the failed pins mix into the sodium pool. Delayed Neutron Detectors (DND) are provided in the sodium pool to detect such failures by way of detection of delayed neutrons emitted by the fission products. The transient evolution of fission product concentration is governed by the sodium flow distribution in the pool. Transient hydraulic analysis has been carried out using the CFD code PHOENICS to estimate fission product concentration evolution in hot pool. k- ε turbulence model and zero laminar diffusivity for the fission product concentration have been considered in the analysis. Times at which the failures of various fuel subassemblies (SA) are detected by the DND are obtained. It has been found that in order to effectively detect the failure of every fuel SA, a minimum of 8 DND in hot pool are essential

Temperature noise analysis and sodium boiling detection in the fuel failure mockup

International Nuclear Information System (INIS)

Sides, W.H. Jr.; Fry, D.N.; Leavell, W.H.; Mathis, M.V.; Saxe, R.F.

1976-01-01

Sodium temperature noise was measured at the exit of simulated, fast-reactor fuel subassemblies in the Fuel Failure Mockup (FFM) to determine the feasibility of using temperature noise monitors to detect flow blockages in fast reactors. Also, acoustic noise was measured to determine whether sodium boiling in the FFM could be detected acoustically and whether noncondensable gas entrained in the sodium coolant would affect the sensitivity of the acoustic noise detection system. Information from these studies would be applied to the design of safety systems for operating liquid-metal fast breeder reactors (LMFBRs). It was determined that the statistical properties of temperature noise are dependent on the shape of temperature profiles across the subassemblies, and that a blockage upstream of a thermocouple that increases the gradient of the profile near the blockage will also increase the temperature noise at the thermocouple. Amplitude probability analysis of temperature noise shows a skewed amplitude density function about the mean temperature that varies with the location of the thermocouple with respect to the blockage location. It was concluded that sodium boiling in the FFM could be detected acoustically. However, entrained noncondensable gas in the sodium coolant at void fractions greater than 0.4 percent attenuated the acoustic signals sufficiently that boiling was not detected. At a void fraction of 0.1 percent, boiling was indicated only by the two acoustic detectors closest to the boiling site

Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Choi, Jin Ha; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

2016-05-15

The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O{sub 2} and (U,TRU)O{sub 2} which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O{sub 2}, (Th,Pu)O{sub 2} and (Th,TRU)O{sub 2}, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

International Nuclear Information System (INIS)

Choi, Jin Ha; Kim, Myung Hyun

2016-01-01

The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O_2 and (U,TRU)O_2 which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O_2, (Th,Pu)O_2 and (Th,TRU)O_2, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

Toxicology of plutonium-sodium

International Nuclear Information System (INIS)

Hackett, P.L.

1982-01-01

Scenarios for liquid-metal fast breeder reactor (LMFBR) accidents predict the loss of sodium coolant, with subsequent core melt-down and release of mixed sodium-fuel aerosols [Na-(PuU)O 2 ] into the environment. Studies in other laboratories demonstrated that mixed aerosols of Na 2 O-PuO 2 were more readily transported from the lung than PuO 2 aerosols. We therefore devised a continuous aerosol-generating system for animal exposures in which laser-generated fuel aerosols were swept through sodium vapor to form sodium-fuel aerosols. These fuel and sodium-fuel aerosols were compared with regard to their physicochemical properties and their biological behavior following inhalation studies in rats and dogs

Review of Sodium and Plutonium related Technical Standards in Trans-Uranium Fuel Fabrication Facilities

Energy Technology Data Exchange (ETDEWEB)

Jang, Misuk; Jeon, Jong Seon; Kang, Hyun Sik; Kim, Seoung Rae [NESS, Daejeon (Korea, Republic of)

2016-10-15

In this paper, we would introduce and review technical standards related to sodium fire and plutonium criticality safety. This paper may be helpful to identify considerations in the development of equipment, standards, and etc., to meet the safety requirements in the design, construction and operating of TFFF, KAPF and SFR. The feasibility and conceptual designs are being examined on related facilities, for example, TRU Fuel Fabrication Facilities (TFFF), Korea Advanced Pyro-process Facility (KAPF), and Sodium Cooled Fast Reactor (SFR), in Korea. However, the safety concerns of these facilities have been controversial in part because of the Sodium fire accident and Plutonium related radiation safety caused by transport and handling accident. Thus, many researches have been performed to ensure safety and various documents including safety requirements have been developed. In separating and reducing the long-lived radioactive transuranic(TRU) in the spent nuclear fuel, reusing as the potential energy of uranium fuel resources and reducing the high level wastes, TFFF would be receiving the attention of many people. Thus, people would wonder whether compliance with technical standards that ensures safety. For new facility design, one of the important tasks is to review of technical standards, especially for sodium and Plutonium because of water related highly reactive characteristics and criticality hazard respectively. We have introduced and reviewed two important technical standards for TFFF, which are sodium fire and plutonium criticality safety, in this paper. This paper would provide a brief guidance, about how to start and what is important, to people who are responsible for the initial design to operation of TFFF.

Review of Sodium and Plutonium related Technical Standards in Trans-Uranium Fuel Fabrication Facilities

International Nuclear Information System (INIS)

Jang, Misuk; Jeon, Jong Seon; Kang, Hyun Sik; Kim, Seoung Rae

2016-01-01

In this paper, we would introduce and review technical standards related to sodium fire and plutonium criticality safety. This paper may be helpful to identify considerations in the development of equipment, standards, and etc., to meet the safety requirements in the design, construction and operating of TFFF, KAPF and SFR. The feasibility and conceptual designs are being examined on related facilities, for example, TRU Fuel Fabrication Facilities (TFFF), Korea Advanced Pyro-process Facility (KAPF), and Sodium Cooled Fast Reactor (SFR), in Korea. However, the safety concerns of these facilities have been controversial in part because of the Sodium fire accident and Plutonium related radiation safety caused by transport and handling accident. Thus, many researches have been performed to ensure safety and various documents including safety requirements have been developed. In separating and reducing the long-lived radioactive transuranic(TRU) in the spent nuclear fuel, reusing as the potential energy of uranium fuel resources and reducing the high level wastes, TFFF would be receiving the attention of many people. Thus, people would wonder whether compliance with technical standards that ensures safety. For new facility design, one of the important tasks is to review of technical standards, especially for sodium and Plutonium because of water related highly reactive characteristics and criticality hazard respectively. We have introduced and reviewed two important technical standards for TFFF, which are sodium fire and plutonium criticality safety, in this paper. This paper would provide a brief guidance, about how to start and what is important, to people who are responsible for the initial design to operation of TFFF

Dispersion and thermal interactions of molten metal fuel settling on a horizontal steel plate through a sodium pool

International Nuclear Information System (INIS)

Gabor, J.D.; Purviance, R.T.; Aeschlimann, R.W.; Spencer, B.W.

1989-01-01

Although the Integral Fast Reactor (IFR) possesses inherent safety features, an assessment of the consequences of melting of the metal fuel is necessary for risk analysis. As part of this effort an experimental study was conducted to determine the depths of sodium at 600 C required for pour streams of various molten uranium alloys (U, U-5 wt % Zr, U-10 wt % Zr, and U-10 wt % Fe) to break up and solidify. The quenched particulate material, which was in the shape of filaments and sheets, formed coolable beds because of the high voidage (∼0.9) and large particle size (∼10 mm). In a test with a 0.15-m sodium depth, the fragments from a pure uranium pour stream did not completely solidify but formed an agglomerated mass which did not fuse to the base plate. However, the agglomerated fragments of U-10 wt % Fe eutectic fused to the stainless steel base plate. An analysis of the temperature response of a 25-mm thick base plate was made by volume averaging the properties of the sodium and metal particle phases and assuming two semi-infinite solids coming into contact. Good agreement was obtained with the data during the initial 5 to 10 s of the contact period. 16 refs., 5 figs., 1 tab

Performance of the diffusion barrier in the metallic fuel in sodium-cooled fast reactor

International Nuclear Information System (INIS)

Kim, Jun Hwan; Ryu, Ho Jin; Yang, Seong Woo; Lee, Byoung Oon; Oh, Seok Jin; Lee, Chan Bock; Hahn, Dohee

2009-01-01

The objectives in this study are to propose several kinds of barrier materials and to evaluate their performance to prevent a fuel-clad interaction situation between the metallic fuel and the clad material in the Sodium-cooled Fast Reactor (SFR). Metallic foil made from refractory element, electrodeposition of the Cr on the clad surface, and the vapor deposition of the Zr were used as the barrier layers. The diffusion couple test was performed at the temperature of 800degC for 25 hour. The results showed that considerable amount of reaction occurred at the specimen without barrier, whereas excellent performance was observed in that neither reaction nor inter-diffusion occurred in the case of metallic foil made of Cr or V. Electrodeposition was revealed to be excellent provided that optimum deposition condition can be found. Similar to the electro-deposition result, excellent performance observed in the case of vapor deposition condition. (author)

Study and Evaluation of Innovative Fuel Handling Systems for Sodium-Cooled Fast Reactors: Fuel Handling Route Optimization

Directory of Open Access Journals (Sweden)

Franck Dechelette

2014-01-01

Full Text Available The research for technological improvement and innovation in sodium-cooled fast reactor is a matter of concern in fuel handling systems in a view to perform a better load factor of the reactor thanks to a quicker fuelling/defueling process. An optimized fuel handling route will also limit its investment cost. In that field, CEA has engaged some innovation study either of complete FHR or on the optimization of some specific components. This paper presents the study of three SFR fuel handling route fully described and compared to a reference FHR option. In those three FHR, two use a gas corridor to transfer spent and fresh fuel assembly and the third uses two casks with a sodium pot to evacuate and load an assembly in parallel. All of them are designed for the ASTRID reactor (1500 MWth but can be extrapolated to power reactors and are compatible with the mutualisation of one FHS coupled with two reactors. These three concepts are then intercompared and evaluated with the reference FHR according to four criteria: performances, risk assessment, investment cost, and qualification time. This analysis reveals that the “mixed way” FHR presents interesting solutions mainly in terms of design simplicity and time reduction. Therefore its study will be pursued for ASTRID as an alternative option.

Two dimensional, two fluid model for sodium boiling in LMFBR fuel assemblies

International Nuclear Information System (INIS)

Granziera, M.R.; Kazimi, M.S.

1980-05-01

A two dimensional numerical model for the simulation of sodium boiling transient was developed using the two fluid set of conservation equations. A semiimplicit numerical differencing scheme capable of handling the problems associated with the ill-posedness implied by the complex characteristic roots of the two fluid problems was used, which took advantage of the dumping effect of the exchange terms. Of particular interest in the development of the model was the identification of the numerical problems caused by the strong disparity between the axial and radial dimensions of fuel assemblies. A solution to this problem was found which uses the particular geometry of fuel assemblies to accelerate the convergence of the iterative technique used in the model. Three sodium boiling experiments were simulated with the model, with good agreement between the experimental results and the model predictions

Effects of sodium salt types on the intermolecular interaction of sodium alginate/antarctic krill protein composite fibers.

Science.gov (United States)

Zhang, Rui; Guo, Jing; Liu, Yuanfa; Chen, Shuang; Zhang, Sen; Yu, Yue

2018-06-01

Sodium alginate (SA) and antarctic krill protein (AKP) were blended to fabricate the SA/AKP composite fibers by the conventional wet spinning method using 5% CaCl 2 as coagulation solution. The sodium salt was added to the SA/AKP solution to adjust the ionization degree and intermolecular interaction of composite system. The main purpose of this study is to investigate the influences of sodium salt types (NaCl, CH 3 COONa, Na 2 SO 4 ) on the intermolecular interaction of SA/AKP composite fibers. The intermolecular interaction, morphology, crystallinity, thermal stability and mechanical properties of SA/AKP composite fibers were analyzed by fourier transform infrared spectroscopy (FT-IR), scanning electron microscope (SEM), x-ray diffraction (XRD), thermogravimetric analysis (TGA). The results show that the types of sodium salt have obvious influences on the content of both β-sheet, intermolecular hydrogen bond, breaking strength and surface morphology in SA/AKP composite fibers, but have a negligible effect on the crystallinity and thermal stability. Copyright © 2018 Elsevier Ltd. All rights reserved.

Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

International Nuclear Information System (INIS)

Karahan, Aydin; Buongiorno, Jacopo

2010-01-01

An engineering code to model the irradiation behavior of UO 2 -PuO 2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

Energy Technology Data Exchange (ETDEWEB)

Karahan, Aydin, E-mail: karahan@mit.ed [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, MA (United States); Buongiorno, Jacopo [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, MA (United States)

2010-01-31

An engineering code to model the irradiation behavior of UO{sub 2}-PuO{sub 2} mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

SIMMER-III applications to fuel-coolant interactions

Energy Technology Data Exchange (ETDEWEB)

Morita, K.; Kondo, Sa.; Tobita, Y.; Brear, D.J. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1998-01-01

The main purpose of the SIMMER-III code is to provide a numerical simulation of complex multiphase, multicomponent flow problems essential to investigate core disruptive accidents in liquid-metal fast reactors (LMFRs). However, the code is designed to be sufficiently flexible to be applied to a variety of multiphase flows, in addition to LMFR safety issues. In the present study, some typical experiments relating to fuel-coolant interactions (FCIs) have been analyzed by SIMMER-III to demonstrate that the code is applicable to such complex and highly transient multiphase flow situations. It is shown that SIMMER-III can reproduce the premixing phase both in water and sodium systems as well as the propagation of steam explosion. It is thus demonstrated the code is basically capable of simulating integral multiphase thermal-hydraulic problems included in FCI experiments. (author)

Growth of the interaction layer around fuel particles in dispersion fuel

International Nuclear Information System (INIS)

Olander, D.

2009-01-01

Corrosion of uranium particles in dispersion fuel by the aluminum matrix produces interaction layers (an intermetallic-compound corrosion product) around the shrinking fuel spheres. The rate of this process was modeled as series resistances due to Al diffusion through the interaction layer and reaction of aluminum with uranium in the fuel particle to produce UAl x . The overall kinetics are governed by the relative rates of these two steps, the slowest of which is reaction at the interface between Al in the interaction layer and U in the fuel particle. The substantial volume change as uranium is transferred from the fuel to the interaction layer was accounted for. The model was compared to literature data on in-reactor growth of the interaction layer and the Al/U gradient in this layer, the latter measured in ex-reactor experiments. The rate constant of the Al-U interface reaction and the diffusivity of Al in the interaction layer were obtained from this fitting procedure. The second feature of the corrosion process is the transfer of fission products from the fuel particle to the interaction layer due to the reaction. It is commonly assumed that the observed swelling of irradiated fuel elements of this type is due to release of fission gas in the interaction layer to form large bubbles. This hypothesis was tested by using the model to compute the quantity of fission gas available from this source and comparing the pressure of the resulting gas with the observed swelling of fuel plates. It was determined that the gas pressure so generated is too small to account for the observed delamination of the fuel

Sodium Loop Safety Facility W-2 experiment fuel pin rupture detection system

International Nuclear Information System (INIS)

Hoffman, M.A.; Kirchner, T.L.; Meyers, S.C.

1980-05-01

The objective of the Sodium Loop Safety Facility (SLSF) W-2 experiment is to characterize the combined effects of a preconditioned full-length fuel column and slow transient overpower (TOP) conditions on breeder reactor (BR) fuel pin cladding failures. The W-2 experiment will meet this objective by providing data in two technological areas: (1) time and location of cladding failure, and (2) early post-failure test fuel behavior. The test involves a seven pin, prototypic full-length fast test reactor (FTR) fuel pin bundle which will be subjected to a simulated unprotected 5 cents/s reactivity transient overpower event. The outer six pins will provide the necessary prototypic thermal-hydraulic environment for the center pin

Large scale sodium interactions. Part 1. Test facility design

International Nuclear Information System (INIS)

King, D.L.; Smaardyk, J.E.; Sallach, R.A.

1977-01-01

During the design of the test facility for large scale sodium interaction testing, an attempt was made to keep the system as simple and yet versatile as possible; therefore, a once through design was employed as opposed to any type of conventional sodium ''loop.'' The initial series of tests conducted at the facility call for rapidly dropping from 20 kg to 225 kg of sodium at temperatures from 825 0 K to 1125 0 K into concrete crucibles. The basic system layout is described. A commercial drum heater is used to melt the sodium which is in 55 gallon drums and then a slight argon pressurization is used to force the liquid sodium through a metallic filter and into a dump tank. Then the sodium dump tank is heated to the desired temperature. A diaphragm is mechanically ruptured and the sodium is dumped into a crucible that is housed inside a large steel test chamber

Basic concept of fuel safety design and assessment for sodium-cooled fast reactor

International Nuclear Information System (INIS)

Nakae, Nobuo; Baba, Toshikazu; Kamimura, Katsuichiro

2013-03-01

'Philosophy in Safety Evaluation of Fast Breeder Reactors' was published as a guideline for safety design and safety evaluation of Sodium-Cooled Fast Reactor in Japan. This guideline points out that cladding creep and swelling due to internal pressure should be taken into account since the fuel is used under high temperature and high burnup, and that fuel assembly deformation and the prevention from coolant channel blockage should be taken into account in viewpoints of nuclear and thermal hydraulic design. However, the requirements including their criteria and evaluation items are not described. Two other domestic guidelines related to core design are applied for fuel design of fast reactor, but the description is considered to not be enough to practically use. In addition, technical standard for nuclear fuel used in power reactors is also applied for fuel inspection. Therefore, the technical standard and guideline for fuel design and safety evaluation are considered to be very important issue for nuclear safety regulation. This document has been developed according to the following steps: The guidelines and the technical standards, which are prepared in foreign countries and international organization, were reviewed. The technical background concerning fuel design and safety evaluation for fast reactor was collected and summarized in the world wide scale. The basic concept of fuel safety design and assessment for sodium-cooled fast reactor was developed by considering a wide range of views of the specialists in Japan. In order to discuss the content with foreign specialists IAEA Consultancy Meetings have been held on January, 2011 and January, 2012. The participants of the meeting came from USA, UK, EC, India, China and South Korea. The specialists of IAEA and JNES were also joined. Although this document is prepared for application to 'Monju'(prototype LMFR), it may be applied to experimental, demonstration and commercial types of LMFR after revising it by taking

Recovery of sodium hydroxide and silica from zirconium oxide plant effluent of Nuclear Fuel Complex

International Nuclear Information System (INIS)

Bajpai, M.B.; Shenoi, M.R.K.; Keni, V.S.

1994-01-01

Sodium hydroxide (lye) and silica can be recovered in pure form from the alkaline sodium silicate waste of Nuclear Fuel Complex, Hyderabad. Electrolytic method was used to amalgamate the sodium present in an electrolyser with flowing mercury as cathode and nickel as anode. The amalgam is then denuded with water in a graphite packed tower to recover mercury for recycling to the electrolyser and sodium hydroxide lye. Sodium hydroxide lye can be recycled in the zirconium oxide plant. Silica is recovered from the spent electrolyte by ion exchange method using cation exchange resin. Both the process details are described in this paper, with experimental data useful for the scale up. The process converts waste to value products. (author)

Interaction between Single Nucleotide Polymorphism and Urinary Sodium, Potassium, and Sodium-Potassium Ratio on the Risk of Hypertension in Korean Adults

Directory of Open Access Journals (Sweden)

Yeong Mi Park

2017-03-01

Full Text Available Hypertension is a complex disease explained with diverse factors including environmental factors and genetic factors. The objectives of this study were to determine the interaction effects between gene variants and 24 h estimated urinary sodium and potassium excretion and sodium-potassium excretion ratios on the risk of hypertension. A total of 8839 participants were included in the genome-wide association study (GWAS to find genetic factors associated with hypertension. Tanaka and Kawasaki formulas were applied to estimate 24 h urinary sodium and potassium excretion. A total of 4414 participants were included in interaction analyses to identify the interaction effects of gene variants according to 24 h estimated urinary factors on the risk of hypertension. CSK rs1378942 and CSK-MIR4513 rs3784789 were significantly modified by urinary sodium-potassium excretion ratio. In addition, MKLN rs1643270 with urinary potassium excretion, LOC101929750 rs7554672 with urinary sodium and potassium excretion, and TENM4 rs10466739 with urinary sodium-potassium excretion ratio showed significant interaction effects. The present study results indicated that the mutant alleles of CSK rs1378942 and CSK-MIR4513 rs3784789 had the strongest protective effects against hypertension in the middle group of 24 h estimated urinary sodium-potassium excretion ratio. Further studies are needed to replicate these analyses in other populations.

Task Group E: fuel-cladding interface reactions. Second quarterly report

International Nuclear Information System (INIS)

Kangilaski, M.; Adamson, M.G.

1974-01-01

An interim assessment of possible interactions and their consequences in the various fuel systems was completed. The assessment discusses the interactions of advanced cladding alloys with: (1) helium bonded mixed oxides; (2) helium and sodium bonded mixed carbides; and (3) helium and sodium bonded mixed nitrides

Chemical interaction of fuel and cladding tubes

International Nuclear Information System (INIS)

Kirihara, Tomoo; Yamawaki, Michio; Obata, Naomi; Handa, Muneo.

1983-01-01

It was attempted to take up the behavior of nuclear fuel in cores and summarize it by the expert committee on the irradiation behavior of nuclear fuel from fiscal 1978 to fiscal 1980 from the following viewpoints. The behavior of nuclear fuel in cores has been treated separately according to each reactor type, accordingly this point is reconsidered. The clearly understood points and the uncertain points are discriminated. It is made more easily understandable for people in other fields of atomic energy. This report is that of the group on the chemical interaction, and the first report of this committee. The chemical interaction as the behavior of fuel in cores is in the unseparable relation to the mechanical interaction, but this relation is not included in this report. The chemical interaction of fuel and cladding tubes under irradiation shows different phenomena in LWRs and FBRs, and is called SCC and FCC, respectively. But this point of causing the difference must be understood to grasp the behavior of fuel. The mutual comparison of oxide fuels for FBRs and LWRs, the stress corrosion cracking of zircaloy tubes, and fuel-cladding chemical interaction in FBRs are reported. (Kako, I.)

Recovery of sodium hydroxide and silica from zirconium oxide plant effluent of Nuclear Fuel Complex

Energy Technology Data Exchange (ETDEWEB)

Bajpai, M B; Shenoi, M R.K.; Keni, V S [Chemical Engineering Division, Bhabha Atomic Research Centre, Mumbai (India)

1994-06-01

Sodium hydroxide (lye) and silica can be recovered in pure form from the alkaline sodium silicate waste of Nuclear Fuel Complex, Hyderabad. Electrolytic method was used to amalgamate the sodium present in an electrolyser with flowing mercury as cathode and nickel as anode. The amalgam is then denuded with water in a graphite packed tower to recover mercury for recycling to the electrolyser and sodium hydroxide lye. Sodium hydroxide lye can be recycled in the zirconium oxide plant. Silica is recovered from the spent electrolyte by ion exchange method using cation exchange resin. Both the process details are described in this paper, with experimental data useful for the scale up. The process converts waste to value products. (author). 3 figs., 2 tabs.

Fuel burn analysis of a sodium fast reactor with KANEXT and Serpent

International Nuclear Information System (INIS)

Lopez S, R. C.; Francois L, J. L.

2015-09-01

The fast reactors cooled by sodium are one of the options considered in the Generation IV. Since most of the reactors of Fourth Generation are still in development stage, is necessary to have efficient and reliable computational tools, this in order to obtain accurate results in reasonable computational times. In this paper is introduced and describes the deterministic code KANEXT (KArlsruhe Neutronic EXtended Tool) and is compared against a Monte Carlo code of more diffusion: Serpent. KANEXT, being a modular code requires the interaction of different modules to perform a job, this interaction of modules is described in this article. The parameters to be compared are the results of the neutron multiplication effective factor and the evolution of isotopes during the burning. The mentioned comparison is carried out for a fast reactor cooled by sodium of relatively small size compared to commercial size reactors. In this paper the particularities of the reactor are described, important for the analysis such as geometry, enrichments, reflector, etc. The considerations in the implementation in both codes are also described, as are simplifications, length of the burning steps, possible solutions of the Bateman equations for the burning fuel in Serpent and the solution options for transport (P3) and diffusion (P1) in KANEXT. The results show good correspondence between Serpent and KANEXT, which give confidence to continue using KANEXT as the main tool. Respect to computation time, time saving is evident with the use of deterministic codes instead of Monte Carlo codes, in this particular case, the time savings using KANEXT is about 98.5% of the time used by Serpent. (Author)

UK experience on fuel and cladding interaction in oxide fuels

Energy Technology Data Exchange (ETDEWEB)

Batey, W [Dounreay Experimental Reactor Establishment, Thurso, Caithness (United Kingdom); Findlay, J R [AERE, Harwell, Didcot, Oxon (United Kingdom)

1977-04-01

The occurrence of fuel cladding interactions in fast reactor fuels has been observed in UK irradiations over a period of years. Chemical incompatibility between fuel and clad represents a potential source of failure and has, on this account, been studied using a variety of techniques. The principal fuel of interest to the UK for fast reactor application is mixed uranium plutonium oxide clad in stainless steel and it is in this field that the majority of work has been concentrated. Some consideration has been given to carbide fuels, because of their application as an advanced fuel. This experience is described in the accompanying paper. Several complementary initiatives have been followed to investigate the interactions in oxide fuel. The principal source of experimental information is from the experimental fuel irradiation programme in the Dounreay Fast Reactor (DFR). Supporting information has been obtained from irradiation programmes in Materials Testing Reactors (MTR). Conditions approaching those in a fast reactor are obtained and the effects of specific variables have been examined in specifically designed experiments. Out-of-reactor experiments have been used to determine the limits of fuel and cladding compatibility and also to give indications of corrosion The observations from all experiments have been examined in the light of thermo-dynamic predictions of fuel behaviour to assess the relative significance of various observations and operating conditions. An experimental programme to control and limit the interactions in oxide fuel is being followed.

UK experience on fuel and cladding interaction in oxide fuels

International Nuclear Information System (INIS)

Batey, W.; Findlay, J.R.

1977-01-01

The occurrence of fuel cladding interactions in fast reactor fuels has been observed in UK irradiations over a period of years. Chemical incompatibility between fuel and clad represents a potential source of failure and has, on this account, been studied using a variety of techniques. The principal fuel of interest to the UK for fast reactor application is mixed uranium plutonium oxide clad in stainless steel and it is in this field that the majority of work has been concentrated. Some consideration has been given to carbide fuels, because of their application as an advanced fuel. This experience is described in the accompanying paper. Several complementary initiatives have been followed to investigate the interactions in oxide fuel. The principal source of experimental information is from the experimental fuel irradiation programme in the Dounreay Fast Reactor (DFR). Supporting information has been obtained from irradiation programmes in Materials Testing Reactors (MTR). Conditions approaching those in a fast reactor are obtained and the effects of specific variables have been examined in specifically designed experiments. Out-of-reactor experiments have been used to determine the limits of fuel and cladding compatibility and also to give indications of corrosion The observations from all experiments have been examined in the light of thermo-dynamic predictions of fuel behaviour to assess the relative significance of various observations and operating conditions. An experimental programme to control and limit the interactions in oxide fuel is being followed

Sodium Loop Safety Facility W-2 experiment fuel pin rupture detection system. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Hoffman, M.A.; Kirchner, T.L.; Meyers, S.C.

1980-05-01

The objective of the Sodium Loop Safety Facility (SLSF) W-2 experiment is to characterize the combined effects of a preconditioned full-length fuel column and slow transient overpower (TOP) conditions on breeder reactor (BR) fuel pin cladding failures. The W-2 experiment will meet this objective by providing data in two technological areas: (1) time and location of cladding failure, and (2) early post-failure test fuel behavior. The test involves a seven pin, prototypic full-length fast test reactor (FTR) fuel pin bundle which will be subjected to a simulated unprotected 5 cents/s reactivity transient overpower event. The outer six pins will provide the necessary prototypic thermal-hydraulic environment for the center pin.

Disagregation of (U, Pu)O2 fuels in molten sodium nitrate and oxides system

International Nuclear Information System (INIS)

Chou, T.S.

1976-01-01

An oxidation process based on the use of an alkali-nitrate melt has been considered as a possible head end step for the reprocessing of FBR spent fuels. The total alkali solubility in the nitrate melt was examined. It is influenced by the temperature. At 500 degC the alkali solubility in the sodium nitrate melt is about 17 mol %. Examining solidified mixture of sodium and nitrate or sodium oxides and nitrite by X-ray diffraction has revealed five unknown lattices. NaNO 3 .xNa 2 O 2 is cubic (a=8.71A), NaNO 2 .xNa 2 O 2 is tetragonal (a=5.939A, c=9.997A), NaNO 2 .xNa 2 O is cubic (a=10.586A). The structure of NaNO 3 .xNa 2 O and NaNO 3 .xNaO 2 could not be determined. The solubility of barium and ruthenium was briefly investigated. The reaction (U,Pu)O 2 with the alkaline sodium nitrate melt proceeds along the grain boundaries of the solid solution. Two steps have been recognized. First (U,Pu)O 2 is oxidized to (U,Pu)Osub(2+x) and in a subsequent step (U,Pu)Osub(2+x) reacts with sodium peroxide to form (U,Pu) 2 O 5 .xNa 2 O 2 . Disaggregation efficiency is a function of temperature, alkali concentration and physical properties of the pellets. High temperature and low alkali concentration lead to high efficiency. The structure of the reaction products (U,Pu)O 2 with alkaline NaNO 3 melt was shown to depend mainly on the alkali concentration. As the alkali concentration is lower than 2 mole % (U,Pu) 2 O 5 . Na 2 O 2 is the dominate phase. (U,Pu) 2 O 5 .3Na 2 O 2 corresponds to 6 mole % and over 11 mole % alkali, (U,Pu) 2 O 5 .xNa 2 O 2 becomes the main product. The solubility of the fuel (U,Pu) in the alkali sodium nitrate melt increases with the alkali concentration up to 6000-8000 ppm for uranium and 1200-1700 ppm for plutonium at 500 degC with only 5 mole % alkali. As a result of high losses of fissile material in the salt bath molten salt process must regarded as uneligible for a general head end step in fuel reprocessing. Nevertheless its application can still be

Metal fuel development and verification for prototype generation- IV Sodium- Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Chan Bock; Cheon, Jin Sik; Kim, Sung Ho; Park, Jeong Yong; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U -transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

Metal Fuel Development and Verification for Prototype Generation IV Sodium-Cooled Fast Reactor

Directory of Open Access Journals (Sweden)

Chan Bock Lee

2016-10-01

Full Text Available Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR to be built by 2028. U–Zr fuel is a driver for the initial core of the PGSFR, and U–transuranics (TRU–Zr fuel will gradually replace U–Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U–Zr fuel, work on U–Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U–TRU–Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic–martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

Void reactivity decomposition for the Sodium-cooled Fast Reactor in equilibrium fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Sun Kaichao, E-mail: kaichao.sun@psi.ch [Paul Scherrer Institut (PSI), 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), 1015 Lausanne (Switzerland); Krepel, Jiri; Mikityuk, Konstantin; Pelloni, Sandro [Paul Scherrer Institut (PSI), 5232 Villigen PSI (Switzerland); Chawla, Rakesh [Paul Scherrer Institut (PSI), 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), 1015 Lausanne (Switzerland)

2011-07-15

Highlights: > We analyze the void reactivity effect for three ESFR core fuel cycle states. > The void reactivity effect is decomposed by neutron balance method. > Novelly, the normalization to the integral flux in the active core is applied. > The decomposition is compared with the perturbation theory based results. > The mechanism and the differences of the void reactivity effect are explained. - Abstract: The Sodium-cooled Fast Reactor (SFR) is one of the most promising Generation IV systems with many advantages, but has one dominating neutronic drawback - a positive sodium void reactivity. The aim of this study is to develop and apply a methodology, which should help better understand the causes and consequences of the sodium void effect. It focuses not only on the beginning-of-life (BOL) state of the core, but also on the beginning of open and closed equilibrium (BOC and BEC, respectively) fuel cycle conditions. The deeper understanding of the principal phenomena involved may subsequently lead to appropriate optimization studies. Various voiding scenarios, corresponding to different spatial zones, e.g. node or assembly, have been analyzed, and the most conservative case - the voiding of both inner and outer fuel zones - has been selected as the reference scenario. On the basis of the neutron balance method, the corresponding SFR void reactivity has been decomposed reaction-, isotope-, and energy-group-wise. Complementary results, based on generalized perturbation theory and sensitivity analysis, are also presented. The numerical analysis for both neutron balance and perturbation theory methods has been carried out using appropriate modules of the ERANOS code system. A strong correlation between the flux worth, i.e. the product of flux and adjoint flux, and the void reactivity importance distributions has been found for the node- and assembly-wise voiding scenarios. The neutron balance based decomposition has shown that the void effect is caused mainly by the

Void reactivity decomposition for the Sodium-cooled Fast Reactor in equilibrium fuel cycle

International Nuclear Information System (INIS)

Sun Kaichao; Krepel, Jiri; Mikityuk, Konstantin; Pelloni, Sandro; Chawla, Rakesh

2011-01-01

Highlights: → We analyze the void reactivity effect for three ESFR core fuel cycle states. → The void reactivity effect is decomposed by neutron balance method. → Novelly, the normalization to the integral flux in the active core is applied. → The decomposition is compared with the perturbation theory based results. → The mechanism and the differences of the void reactivity effect are explained. - Abstract: The Sodium-cooled Fast Reactor (SFR) is one of the most promising Generation IV systems with many advantages, but has one dominating neutronic drawback - a positive sodium void reactivity. The aim of this study is to develop and apply a methodology, which should help better understand the causes and consequences of the sodium void effect. It focuses not only on the beginning-of-life (BOL) state of the core, but also on the beginning of open and closed equilibrium (BOC and BEC, respectively) fuel cycle conditions. The deeper understanding of the principal phenomena involved may subsequently lead to appropriate optimization studies. Various voiding scenarios, corresponding to different spatial zones, e.g. node or assembly, have been analyzed, and the most conservative case - the voiding of both inner and outer fuel zones - has been selected as the reference scenario. On the basis of the neutron balance method, the corresponding SFR void reactivity has been decomposed reaction-, isotope-, and energy-group-wise. Complementary results, based on generalized perturbation theory and sensitivity analysis, are also presented. The numerical analysis for both neutron balance and perturbation theory methods has been carried out using appropriate modules of the ERANOS code system. A strong correlation between the flux worth, i.e. the product of flux and adjoint flux, and the void reactivity importance distributions has been found for the node- and assembly-wise voiding scenarios. The neutron balance based decomposition has shown that the void effect is caused mainly

Health and Safety Considerations Associated with Sodium-Cooled Experimental Nuclear Fuel Dismantlement

Energy Technology Data Exchange (ETDEWEB)

Carvo, Alan E. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-04-01

Between the mid-1970s and the mid-1980s Sandia National Laboratory constructed eleven experimental assemblies to simulate debris beds formed in a sodium-cooled fast breeder reactor. All but one of the assemblies were irradiated. The experimental assemblies were transferred to the Idaho National Laboratory (INL) in 2007 and 2008 for storage, dismantlement, recovery of the uranium for reuse in the nuclear fuel cycle, and disposal of unneeded materials. This paper addresses the effort to dismantle the assemblies down to the primary containment vessel and repackage them for temporary storage until such time as equipment necessary for sodium separation is in place.

Thermal-hydraulic numerical simulation of fuel sub-assembly for Sodium-cooled Fast Reactor

International Nuclear Information System (INIS)

Saxena, Aakanksha

2014-01-01

The thesis focuses on the numerical simulation of sodium flow in wire wrapped sub-assembly of Sodium-cooled Fast Reactor (SFR). First calculations were carried out by a time averaging approach called RANS (Reynolds- Averaged Navier-Stokes equations) using industrial code STAR-CCM+. This study gives a clear understanding of heat transfer between the fuel pin and sodium. The main variables of the macroscopic flow are in agreement with correlations used hitherto. However, to obtain a detailed description of temperature fluctuations around the spacer wire, more accurate approaches like LES (Large Eddy Simulation) and DNS (Direct Numerical Simulation) are clearly needed. For LES approach, the code TRIO U was used and for the DNS approach, a research code was used. These approaches require a considerable long calculation time which leads to the need of representative but simplified geometry. The DNS approach enables us to study the thermal hydraulics of sodium that has very low Prandtl number inducing a very different behavior of thermal field in comparison to the hydraulic field. The LES approach is used to study the local region of sub-assembly. This study shows that spacer wire generates the local hot spots (∼20 C) on the wake side of spacer wire with respect to the sodium flow at the region of contact with the fuel pin. Temperature fluctuations around the spacer wire are low (∼1 C-2 C). Under nominal operation, the spectral analysis shows the absence of any dominant peak for temperature oscillations at low frequency (2-10 Hz). The obtained spectra of temperature oscillations can be used as an input for further mechanical studies to determine its impact on the solid structures. (author) [fr

Fuel-cladding chemical interaction

International Nuclear Information System (INIS)

Gueneau, C.; Piron, J.P.; Dumas, J.C.; Bouineau, V.; Iglesias, F.C.; Lewis, B.J.

2015-01-01

The chemistry of the nuclear fuel is very complex. Its chemical composition changes with time due to the formation of fission products and depends on the temperature level history within the fuel pellet and the clad during operation. Firstly, in thermal reactors, zircaloy oxidation from reaction with UO 2 fuel under high-temperature conditions will be addressed. Then other fuel-cladding interaction phenomena occurring in fast reactors will be described. Large thermal gradients existing between the centre and the periphery of the pellet induce the radial redistribution of the fuel constituents. The fuel pellet can react with the clad by different corrosion processes which can involve actinide and/or fission product transport via gas, liquid or/and solid phases. All these phenomena are briefly described in the case of different kinds of fuels (oxide, carbide, nitride, metallic) to be used in fast reactors. The way these phenomena are taken into account in fuel performance codes is presented. (authors)

High Temperature Fuel Cladding Chemical Interactions Between TRIGA Fuels and 304 Stainless Steel

Energy Technology Data Exchange (ETDEWEB)

Perez, Emmanuel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Keiser, Jr., Dennis D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Forsmann, Bryan [Boise State Univ., ID (United States); Janney, Dawn E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Henley, Jody [Idaho National Lab. (INL), Idaho Falls, ID (United States); Woolstenhulme, Eric C. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-02-01

High-temperature fuel-cladding chemical interactions (FCCI) between TRIGA (Training, Research, Isotopes, General Atomics) fuel elements and the 304 stainless steel (304SS) are of interest to develop an understanding of the fuel behavior during transient reactor scenarios. TRIGA fuels are composed of uranium (U) particles dispersed in a zirconium-hydride (Zr-H) matrix. In reactor, the fuel is encased in 304-stainless-steel (304SS) or Incoloy 800 clad tubes. At high temperatures, the fuel can readily interact with the cladding, resulting in FCCI. A number of FCCI can take place in this system. Interactions can be expected between the cladding and the Zr-H matrix, and/or between the cladding and the U-particles. Other interactions may be expected between the Zr-H matrix and the U-particles. Furthermore, the fuel contains erbium-oxide (Er-O) additions. Interactions can also be expected between the Er-O, the cladding, the Zr-H and the U-particles. The overall result is that very complex interactions may take place as a result of fuel and cladding exposures to high temperatures. This report discusses the characterization of the baseline fuel microstructure in the as-received state (prior to exposure to high temperature), characterization of the fuel after annealing at 950C for 24 hours and the results from diffusion couple experiments carries out at 1000C for 5 and 24 hours. Characterization was carried out via scanning electron microscopy (SEM) and transmission electron microscopy (TEM) with sample preparation via focused ion beam in situ-liftout-technique.

High Temperature Fuel Cladding Chemical Interactions Between TRIGA Fuels and 304 Stainless Steel

International Nuclear Information System (INIS)

Perez, Emmanuel; Keiser Jr, Dennis D.; Forsmann, Bryan; Janney, Dawn E.; Henley, Jody; Woolstenhulme, Eric C.

2016-01-01

High-temperature fuel-cladding chemical interactions (FCCI) between TRIGA (Training, Research, Isotopes, General Atomics) fuel elements and the 304 stainless steel (304SS) are of interest to develop an understanding of the fuel behavior during transient reactor scenarios. TRIGA fuels are composed of uranium (U) particles dispersed in a zirconium-hydride (Zr-H) matrix. In reactor, the fuel is encased in 304-stainless-steel (304SS) or Incoloy 800 clad tubes. At high temperatures, the fuel can readily interact with the cladding, resulting in FCCI. A number of FCCI can take place in this system. Interactions can be expected between the cladding and the Zr-H matrix, and/or between the cladding and the U-particles. Other interactions may be expected between the Zr-H matrix and the U-particles. Furthermore, the fuel contains erbium-oxide (Er-O) additions. Interactions can also be expected between the Er-O, the cladding, the Zr-H and the U-particles. The overall result is that very complex interactions may take place as a result of fuel and cladding exposures to high temperatures. This report discusses the characterization of the baseline fuel microstructure in the as-received state (prior to exposure to high temperature), characterization of the fuel after annealing at 950C for 24 hours and the results from diffusion couple experiments carries out at 1000C for 5 and 24 hours. Characterization was carried out via scanning electron microscopy (SEM) and transmission electron microscopy (TEM) with sample preparation via focused ion beam in situ-liftout-technique.

Design of FCI Experiments to Understand Fuel Out-Pin Phenomena in the SFR

Energy Technology Data Exchange (ETDEWEB)

Heo, Hyo; Park, Seong Dae [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Jerng, Dong Wook; Bang, In Cheol [Chungang Univ., Seoul (Korea, Republic of)

2014-05-15

It is important to guarantee a passive nuclear safety regarding enhanced negative reactivity by fragmenting the molten fuel. In the SFR, it has a strong point that the negative reactivity is immediately introduced when the metal fuel is melted by the UTOP or ULOP accident. These characteristics of the metal fuel can prevent from progressing in severe accidents such as core disruptive accidents (CDA). As key phenomena in the accidents, fuel-coolant interaction (FCI) phenomena have been studied over the last few decades. Especially, several previous researches focused on instability and fragmentation of a core melt jet in water. However, the studies showed too limited phenomena to fully understand. In the domestic SFR technology development, researches for severe accidents tend to lag behind ones of other countries. Or, South Korea has a very basic level of the research such as literature survey. Recently, the SAS4A code, which was developed at Argonne National Laboratory (ANL) for thermal-hydraulic and neutronic analyses of power and flow transients in liquid-metal-cooled nuclear reactors (LMRs), is still under development to consider for a metal fuel. The other countries carried out basic experiments for molten fuel and coolant interactions. However, in a high temperature condition, methods for analysis of structural interaction between molten fuel and fuel cladding are very limited. The ultimate objective of the study is to evaluate the possibility of recriticality accident induced by fuel-coolant interaction in the SFR adopting metal fuel. It is a key point to analyze the molten-fuel behavior based on the experimental results which show fuel-coolant interaction with the simulant materials. It is necessary to establish the test facility, to build database, and to develop physical models to understand the FCI phenomena in the SFR; molten fuel-coolant interaction as soon as the molten fuel is ejected to the sodium coolant channel and molten fuel-coolant interaction

Preliminary Results on a Contact between 4 kg of Molten UO2 and Liquid Sodium

International Nuclear Information System (INIS)

Amblard, M.

1976-01-01

The CORECT II Experiment consists in simulating the penetration of sodium into an assembly when the fuel is molten. In other words, it is a shock-tube type of experiment with dimensions representative of a full-scale assembly. the experiment consists in dropping a 100 litre column of sodium onto partially molten UO 2 . The following measurements are carried out in transient regime: - sodium velocity in the column; - pressure in the interaction chamber; - pressures at the bottom and at the top of a 5 m tube; - pressure in the argon blanket. The experimental parameters are: - the mass of UO 2 involved (about 4 or 7 kg of 80% molten UO 2 ); - the initial temperature of the sodium (up to 700 deg. C); - the pressure of the residual gas in the interaction chamber during the fall of the sodium; - the dimensions of the interaction chamber and the sodium supply tube; - the form of contact between the UO 2 and the sodium (the sodium may fall on partially liquid and settled UO 2 or on UO 2 pre-dispersed by forced trapping of sodium). To date, 6 tests have been performed. These tests have always resulted in fine fragmentation without any violent interaction. Since no knowledge is available on the change of grain size distribution with time, on the temperature of grain formation, and on the grain movement in the sodium, it is very difficult to interpret these UO 2 -Na tests. We intend to carry out more severe interaction tests on this experimental set-up, by eliminating as much as possible the non-condensable gas which cushions the mechanical impact of the sodium on the UO 2 (tests have shown that by strongly de-pressurizing the liquid UO 2 the fuel could be dispersed by boiling, and this effect should also improve the possibilities of a liquid/liquid contact). - by injecting a little sodium into the UO 2 to facilitate its dispersion in the coolant

Molten core debris-sodium interactions: M-Series experiments

International Nuclear Information System (INIS)

Sowa, E.S.; Gabor, J.D.; Pavlik, J.R.; Cassulo, J.C.; Cook, C.J.; Baker, L. Jr.

1979-01-01

Five new kilogram-scale experiments have been carried out. Four of the experiments simulated the situation where molten core debris flows from a breached reactor vessel into a dry reactor cavity and is followed by a flow of sodium (Ex-vessel case) and one experiment simulated the flow of core debris into an existing pool of sodium (In-vessel case). The core debris was closely simulated by a thermite reaction which produced a molten mixture of UO 2 , ZrO 2 , and stainless steel. There was efficient fragmentation of the debris in all experiments with no explosive interactions observed

Sodium borohydride as an additive to enhance the performance of direct ethanol fuel cells

Energy Technology Data Exchange (ETDEWEB)

Wang, Lianqin; Fang, Xiang; Shen, Pei Kang [The Key Laboratory of Low-carbon Chemistry and Energy Conservation of Guangdong Province, The State Key Laboratory of Optoelectronic Materials and Technologies, Sun Yat-sen University, Guangzhou 510275 (China); Bambagioni, Valentina; Bevilacqua, Manuela; Bianchini, Claudio; Filippi, Jonathan; Lavacchi, Alessandro; Marchionni, Andrea; Vizza, Francesco [Istituto di Chimica dei Composti Organometallici (ICCOM-CNR), via Madonna del Piano 10, 50019 Sesto Fiorentino, Florence (Italy)

2010-12-15

The effect of adding small quantities (0.1-1 wt.%) of sodium borohydride (NaBH{sub 4}) to the anolyte solution of direct ethanol fuel cells (DEFCs) with membrane-electrode assemblies constituted by nanosized Pd/C anode, Fe-Co cathode and anion-exchange membrane (Tokuyama A006) was investigated by means of various techniques. These include cyclic voltammetry, in situ FTIR spectroelectrochemistry, a study of the performance of monoplanar fuel cells and an analysis of the ethanol oxidation products. A comparison with fuel cells fed with aqueous solutions of ethanol proved unambiguously the existence of a promoting effect of NaBH{sub 4} on the ethanol oxidation. Indeed, the potentiodynamic curves of the ethanol-NaBH{sub 4} mixtures showed higher power and current densities, accompanied by a remarkable increase in the fuel consumption at comparable working time of the cell. A {sup 13}C and {sup 11}B {l_brace}{sup 1}H{r_brace}NMR analysis of the cell exhausts and an in situ FTIR spectroelectrochemical study showed that ethanol is converted selectively to acetate while the oxidation product of NaBH{sub 4} is sodium metaborate (NaBO{sub 2}). The enhancement of the overall cell performance has been explained in terms of the ability of NaBH{sub 4} to reduce the PdO layer on the catalyst surface. (author)

Fabrication of uranium alloy fuel slug for sodium-cooled fast reactor by injection casting

International Nuclear Information System (INIS)

Jong Hwan Kim; Hoon Song; Ki Hwan Kim; Chan Bock Lee

2014-01-01

Metal fuel slugs of U-Zr alloys for a sodium-cooled fast reactor (SFR) have been fabricated using an injection casting method. However, casting alloys containing volatile radioactive constituents such as Am can cause problems in a conventional injection casting method. Therefore, in this study, several injection-casting methods were applied to evaluate the volatility of the metal-fuel elements and control the transport of volatile elements. Mn was selected as a volatile surrogate alloy since it possesses a total vapor pressure equivalent to that of minor actinide-bearing fuels for SFRs. U-10 wt% Zr and U-10 wt% Zr-5 wt% Mn metal fuels were prepared, and the casting processes were evaluated. The casting soundness of the fuel slugs was characterized by gamma-ray radiography and immersion density measurements. Inductively coupled plasma atomic emission spectroscopy was used to determine the chemical composition of fuel slugs. Fuel losses after casting were also evaluated according to the casting conditions. (author)

New interpretation on formation of UO2 Post-Accident Heat Removal particulate in sodium

International Nuclear Information System (INIS)

Schins, H.

1986-01-01

A comparative experimental study on quenching in sodium of four molten fuel materials, UO 2 Al 2 P 3 , Cu and stainless steel, is presented. Experimental results like temperatures, pressures, particle shapes, particle size distributions, crack patterns and crystal grain sizes are given and interpreted. These fuel-coolant interactions (FCI) can be understood as all being characterized by transition boiling of sodium. The fuel is first fragmented by the sodium vapor bubble growth and collapse process. These particulates have smooth surfaces. The two materials, UO 2 and Al 2 O 3 , are fragmented further by a delayed mechanism which is thermal stress shrinkage cracking. Delayed particles are fragments of larger ones. Furthermore, attention is drawn to the theoretical results which show that pure FCI-particulate is significantly finer

Development of multi-dimensional thermal-hydraulic modeling using mixing factors for wire wrapped fuel pin bundles in fast reactors. Validation through a sodium experiment of 169-pin fuel subassembly

International Nuclear Information System (INIS)

Nishimura, M.; Kamide, H.; Miyake, Y.

1997-04-01

Temperature distributions in fuel subassemblies of fast reactors interactively affect heat transfer from center to outer region of the core (inter-subassembly heat transfer) and cooling capability of an inter-wrapper flow, as well as maximum cladding temperature. The prediction of temperature distribution in the subassembly is, therefore one of the important issues for the reactor safety assessment. Mixing factors were applied to multi-dimensional thermal-hydraulic code AQUA to enhance the predictive capability of simulating maximum cladding temperature in the fuel subassemblies. In the previous studies, this analytical method had been validated through the calculations of the sodium experiments using driver subassembly test rig PLANDTL-DHX with 37-pin bundle and blanket subassembly test rig CCTL-CFR with 61-pin bundle. The error of the analyses were comparable to the error of instrumentation's. Thus the modeling was capable of predicting thermal-hydraulic field in the middle scale subassemblies. Before the application to large scale real subassemblies with more than 217 pins, accuracy of the analytical method have to be inspected through calculations of sodium tests in a large scale pin bundle. Therefore, computations were performed on sodium experiments in the relatively large 169-pin subassembly which had heater pins sparsely within the bundle. The analysis succeeded to predict the experimental temperature distributions. The errors of temperature rise from inlet to maximum values were reduced to half magnitudes by using mixing factors, compared to those of analyses without mixing factors. Thus the modeling is capable of predicting the large scale real subassemblies. (author)

Scoping studies of vapor behavior during a severe accident in a metal-fueled reactor

International Nuclear Information System (INIS)

Spencer, B.W.; Marchaterre, J.F.

1985-01-01

Scoping calculations have been performed examining the consequences of fuel melting and pin failures for a reactivity-insertion type accident in a sodium-cooled, pool-type reactor fueled with a metal alloy fuel. The principal gas and vapor species released are shown to be Xe, Cs,and bond sodium contained within the fuel porosity. Fuel vapor pressure is insignificant, and there is no energetic fuel-coolant interaction for the conditions considered. Condensation of sodium vapor as it expands into the upper sodium pool in a jet mixing regime may occur as rapidly as the vapor emerges from the disrupted core (although reactor-material experiments are needed to confirm these high condensation rates). If the predictions of rapid direct-contact condensation can be verified experimentally for the sodium system, the implication is that the ability of vapor expansion to perform appreciable work on the system is largely eliminated. Furthermore, the ability of an expanding vapor bubble to transport fuel and fission product species to the cover gas region where they may be released to the containment is also largely eliminated. The radionuclide species except for fission gas are largely retained within the core and sodium pool

W-1 Sodium Loop Safety Facility experiment centerline fuel thermocouple performance

International Nuclear Information System (INIS)

Meyers, S.C.; Henderson, J.M.

1980-05-01

The W-1 Sodium Loop Safety Facility (SLSF) experiment is the fifth in a series of experiments sponsored by the Department of Energy (DOE) as part of the National Fast Breeder Reactor (FBR) Safety Assurance Program. The experiments are being conducted under the direction of Argonne National Laboratory (ANL) and Hanford Engineering Development Laboratory (HEDL). The irradiation phase of the W-1 SLSF experiment was conducted between May 27 and July 20, 1979, and terminated with incipient fuel pin cladding failure during the final boiling transient. Experimental hardware and facility performed as designed, allowing completion of all planned tests and test objectives. This paper focuses on high temperature in-fuel thermocouples and discusses their development, fabrication, and performance in the W-1 experiment

Nuclear fuel element

International Nuclear Information System (INIS)

Hirayama, Satoshi; Kawada, Toshiyuki; Matsuzaki, Masayoshi.

1980-01-01

Purpose: To provide a fuel element for reducing the mechanical interactions between a fuel-cladding tube and the fuel element and for alleviating the limits of the operating conditions of a reactor. Constitution: A fuel element having mainly uranium dioxide consists of a cylindrical outer pellet and cylindrical inner pellet inserted into the outer pellet. The outer pellet contains two or more additives selected from aluminium oxide, beryllium oxide, magnesium oxide, silicon oxide, sodium oxide, phosphorus oxide, calcium oxide and iron oxide, and the inner pellet contains nuclear fuel substance solely or one additive selected from calcium oxide, silicon oxide, aluminium oxide, magnesium oxide, zirconium oxide and iron oxide. The outer pellet of the fuel thus constituted is reduced in mechanical strength and also in the mechanical interactions with the cladding tube, and the plastic fluidity of the entire pellet is prevented by the inner pellet increased in the mechanical strength. (Kamimura, M.)

Influence of Fuel-Matrix Interaction on the Deformation of U-Mo Dispersion Fuel

Energy Technology Data Exchange (ETDEWEB)

Ryu, Ho Jin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, Chicago (United States)

2014-05-15

In order to predict the fuel plate failure leading to breakaway swelling in the meat, an understanding of the effects of the fuel-matrix interaction behavior on the deformation of fuel meat is necessary. However, the effects of IL formation on the development of breakaway swelling have not been studied thoroughly. A mechanism that explains large pore growth that leads to breakaway swelling has not been included in the existing fuel performance models. In this study, the effect of the fuel-matrix interaction on large interfacial porosity development at the IL-Al interface is analyzed using both mechanistic correlations and observations from the post-irradiation examination results of U-Mo Dispersion fuels. The effects of fuel-matrix interaction on the fuel performance of U-Mo/Al Dispersion fuel were investigated. Fuel-matrix interaction bears the causes for breakaway swelling that can lead to a fuel failure under a high-power irradiation condition. Fission gas atoms are released from U-Mo particles to the interaction layer via diffusion and recoil. The fission gases released from the U-Mo and produced in the ILs are further released to the IL-Al interface by diffusion in the IL and recoil. Large pore formation at the IL-Al interface is attributed to the active diffusion of fission gas atoms in the ILs and coalescence between the small bubbles there. A model calculation showed that IL growth increases the probability of forming a breakaway swelling condition. ILs are connected to each other and the Al matrix decreases as ILs grow. When more ILs are interconnected, breakaway swelling can occur when the effective stress from the fission gas pressure in the IL-Al interfacial pore becomes larger than the yield strength of the Al matrix.

CAPRICORN subchannel code for sodium boiling in LMFBR fuel bundles

International Nuclear Information System (INIS)

Padilla, A. Jr.; Smith, D.E.; O'Dell, L.D.

1983-01-01

The CAPRICORN computer code analyzes steady-state and transient, single-phase and boiling problems in LMFBR fuel bundles. CAPRICORN uses the same type of subchannel geometry as the COBRA family of codes and solves a similar system of conservation equations for mass, momentum, and energy. However, CAPRICORN uses a different numerical solution method which allows it to handle the full liquid-to-vapor density change for sodium boiling. Results of the initial comparison with data (the W-1 SLSF pipe rupture experiment) are very promising and provide an optimistic basis for proceeding with further development

State-of-the-technology review of fuel-cladding interaction

International Nuclear Information System (INIS)

Bailey, W.J.; Wilson, C.L.; MacGowan, L.J.; Pankaskie, P.J.

1977-12-01

A literature survey and a summarization of postulated fuel-cladding-interaction mechanisms and associated supportive data are reported. The results of that activity are described in the report and include comments on experience with power-ramped fuel, fuel-cladding mechanical interaction, stress-corrosion cracking and fission-product embrittlement, potential remedial actions, fuel-cladding-interaction mechanistic considerations, other ongoing programs, and related patents of interest. An assessment of the candidate fuel concepts to be evaluated as part of this program is provided

Transformation and fragmentation behavior of molten metal drop in sodium pool

International Nuclear Information System (INIS)

Nishimura, Satoshi; Kinoshita, Izumi; Zhang, Zhi-gang; Sugiyama, Ken-ichiro

2006-01-01

In order to clarify the fragmentation mechanism of a metallic alloy (U-Pu-Zr) fuel on liquid phase formed by metallurgical reactions (liquefaction temperature =650degC), which is important in evaluating the sequence of core disruptive accidents for metallic fuel fast reactors, a series of experiments was carried out using molten aluminum (m.p.=660degC) and sodium mainly under the condition that the boiling of sodium does not occur. When the instantaneous contact interface temperature (T i ) between molten aluminum drop and sodium is lower than the boiling point of sodium (T c,bp ), the molten aluminum drop can be fragmented and the mass median diameter (D m ) of aluminum fragments becomes small with increasing T i . When T i is roughly equivalent to or higher than T c,bp , the fragmentation of aluminum drop is promoted by thermal interaction caused by the boiling of sodium on the surface of the drop. Furthermore, even under the condition that the boiling of sodium does not occur and the solid crust is formed on the surface of the drop, it is found from an analytical evaluation that the thermal fragmentation of molten aluminum drop with solid crust is caused by the transient pressurization within the melt confined by the crust. These results indicate the possibility that the metallic alloy fuel on liquid phase formed by the metallurgical reactions can be fragmented without occurring the boiling of sodium on the surface of the melt. (author)

Effects of duct configuration on flow and temperature structure in sodium-cooled 19-rod simulated LMFBR fuel bundles with helical wire-wrap spacers

International Nuclear Information System (INIS)

Wantland, J.L.; Fontana, M.H.; Gnadt, P.A.; Hanus, N.; MacPherson, R.E.; Smith, C.M.

1976-01-01

Thermal-hydrodynamic testing of sodium-cooled 19-rod simulated LMFBR fuel bundles is being conducted at the O ak Ridge National Laboratory in the Fuel Failure Mockup (FFM), an engineering-scale high-temperature sodium facility which provides prototypic flows, temperatures and power densities. Electrically heated bundles have been tested with two scalloped and two hexagonal duct configurations. Peripheral helical flows, attributed to the spacers, have been observed with strengths dependent upon the evenness and relative sizes of the peripheral flow areas. Diametral sodium temperature profiles are more uniform with smaller peripheral flow areas

Measurement and analysis of vibrational behavior of an SNR-fuel element in sodium flow

International Nuclear Information System (INIS)

Hess, B.F.H.; Ruppert, E.; Schmidt, H.; Vinzens, K.

1975-01-01

Within the framework of SNR-300 fuel element development programme a complete full size fuel element dummy has been tested thoroughly for nearly 3000 hours at 650 deg C system temperature in the AKB sodium loop at Interatom, Bensberg. It is known that the coolant flow through a subassembly can induce flutter or vibrations of structural parts such as single pins, the wrapper and the total pin bundle all of which have been of interest during this test. To detect these vibrations of different structural parts simultaneously with a minimum of instrumentation only 3 weldable high temperature strain gauges were employed. These strain gauges were especially prepared and bent in such a way as to form a bridge between the inner wrapper and a fuel pin top and spot-welded to both the wrapper and the fuel pin. Although this arrangement seems to be a rather unusual one, the simultaneous-measurement of bundle, wrapper and pin vibrations was possible and periodic flow fluctuations were also detected. The presented results are only relative due to calibration difficulties with these deformed strain gauges which were first used during this test. It is, however, believed that this arrangement, in connection with the proposed anlytical approach, leads to a simple and technical representation of the vibrational behavior of core elements during sodium tests. Detailed information needed for check and calibration of computer codes are however displayed by the respective power spectral density functions

Thermal expansion of the nuclear fuel-sodium reaction product Na{sub 3}(U{sub 0.84(2)},Na{sub 0.16(2)})O{sub 4} - Structural mechanism and comparison with related sodium-metal ternary oxides

Energy Technology Data Exchange (ETDEWEB)

Illy, Marie-Claire [Chimie ParisTech, PSL Research University, CNRS, Institut de Recherche de Chimie Paris (IRCP), F-75005 Paris (France); European Commission, Joint Research Centre, P.O. Box 2340, 76125 Karlsruhe (Germany); Smith, Anna L. [European Commission, Joint Research Centre, P.O. Box 2340, 76125 Karlsruhe (Germany); Delft University of Technology, Faculty of Applied Sciences, Department of Radiation Science & Technology, Nuclear Energy and Radiation Applications (NERA), Mekelweg 15, 2629 JB, Delft (Netherlands); Wallez, Gilles, E-mail: gilles.wallez@upmc.fr [Chimie ParisTech, PSL Research University, CNRS, Institut de Recherche de Chimie Paris (IRCP), F-75005 Paris (France); Sorbonne University, UPMC Université, Paris 06, 75005 Paris (France); Raison, Philippe E.; Caciuffo, Roberto; Konings, Rudy J.M. [European Commission, Joint Research Centre, P.O. Box 2340, 76125 Karlsruhe (Germany)

2017-07-15

Na{sub 3.16(2)}U{sup V,VI}{sub 0.84(2)}O{sub 4} is obtained from the reaction of sodium with uranium dioxide under oxygen potential conditions typical of a sodium-cooled fast nuclear reactor. In the event of a breach of the steel cladding, it would be the dominant reaction product forming at the rim of the mixed (U,Pu)O{sub 2} fuel pellets. High-temperature X-ray diffraction measurements show that a distortion of the uranium environment in Na{sub 3.16(2)}U{sup V,VI}{sub 0.84(2)}O{sub 4} results in a strongly anisotropic thermal expansion. A comparison with several related sodium metallates Na{sub n-2}M{sup n+}O{sub n-1} - including Na{sub 3}SbO{sub 4} and Na{sub 3}TaO{sub 4}, whose crystal structures are reported for the first time - has allowed us to assess the role played in the lattice expansion by the M{sup n+} cation radius and the Na/M ratio. On this basis, the thermomechanical behavior of the title compound is discussed, along with those of several related double oxides of sodium and actinide elements, surrogate elements, or fission products. - Highlights: •Thermal expansion and structural mechanism of Na{sub 3}(U{sub 0.84(2)},Na{sub 0.16(2)})O{sub 4}, main product of the reaction of sodium with nuclear fuel. •Thermomechanical behavior of sodium uranate suggests possible strains on the fuel cladding and risks of de-cohesion with the fuel pin. •Effect of homo- and aliovalent cation substitutions allows to predict the thermomechanical behavior of sodium metallates involving fission products or minor actinide elements. •Crystal structure of new compounds Na{sub 3}SbO{sub 4} and Na{sub 3}TaO{sub 4}.

Fuel cladding mechanical interaction during power ramps

International Nuclear Information System (INIS)

Guerin, Y.

1985-01-01

Mechanical interaction between fuel and cladding may occur as a consequence of two types of phenomenon: i) fuel swelling especially at levels of caesium accumulation, and ii) thermal differential expansion during power changes. Slow overpower ramps which may occur during incidental events are of course one of the circumstances responsible for this second type of fuel cladding mechanical interaction (FCMI). Experiments and analysis of this problem that have been done at C.E.A. allow to determine the main parameters which will fix the level of stress and the risk of damage induced by the fuel in the cladding during overpower transients

Characterization of IFR metal fuel fragmentation

International Nuclear Information System (INIS)

Gabor, J.D.; Purviance, R.T.; Aeschlimann, R.W.; Spencer, B.W.

1987-01-01

The integral fast reactor (IFR) employs a reactor design that has inherent safety features. An important safety advantage is derived from its pool configuration, which facilitates passive decay heat removal and isolates the core from accidents that might occur elsewhere in the plant. The metal-alloy fuel has superior heat transfer properties compared to oxide fuels. While the IFR design has these inherent safety features, a complete analysis of reactor safety requires assessment of the consequences of the melting of the uranium alloy fuel in the core and the contact of molten core materials with sodium. A series of eight tests was conducted in which the fragmentation and interaction behavior of kilogram quantities of uranium-zirconium alloy in sodium was studied

Thermally decarboxylated sodium bicarbonate: Interactions with water vapour, calorimetric study

Directory of Open Access Journals (Sweden)

Natalia Volkova

2013-06-01

Full Text Available Isothermal titration calorimetry (ITC was used to study interactions between water vapour and the surface of thermally converted sodium bicarbonate (NaHCO3. The decarboxylation degree of the samples was varied from 3% to 35% and the humidity range was 54–100%. The obtained enthalpy values were all exothermic and showed a positive linear correlation with decarboxylation degrees for each humidity studied. The critical humidity, 75% (RHo, was determined as the inflection point on a plot of the mean−ΔHkJ/mole Na2CO3 against RH. Humidities above the critical humidity lead to complete surface dissolution. The water uptake (m was determined after each calorimetric experiment, complementing the enthalpy data. A mechanism of water vapour interaction with decarboxylated samples, including the formation of trona and Wegscheider’s salt on the bicarbonate surface is proposed for humidities below RHo. Keywords: Isothermal titration calorimetry, Sodium bicarbonate, Sodium carbonate, Trona salt, Wegscheider’s salt, Enthalpy, Relative humidity, Pyrolytic decarboxylation

UO$sub 2$/sodium thermal interaction experiments at A.W.R.E. Foulness since January 1972

Energy Technology Data Exchange (ETDEWEB)

Darby, K. I.; Pottinger, R. C.; Turner, R. G.; Rees, N. J.M.

1974-01-15

From 2nd specialist meeting on sodium fuel interaction in fast reactors; Ispra-Varese, Italy (21 Nov 1973). Work has continued to refine the experimental techniques. The unwanted gas content of the pyrotechnic charge has been identified as absorbed water vapor and a dry glove box assembly system is being constructed to remove it from the pyrotechnic charges. Some bomb calorimeter firings of both burst and unburst charges have confirmed the total theoretical energy release, and recent chemical analyses of the debris recovered from sodium (by chemical extraction with alcohol) showed the pyrotechnic reaction had substantially burnt to completion. Experiments to correlate rate of charge burning with the pressure nise in the charge container are proceeding using a transparent quartz charge holder and photography together with attempted thermocouple measurements. Charges with UO/sub 2/ diluent have been fired under sodium and water producing effects identical to previously reported experiments using Al/sub 2/O/sub 3/ diluent. Steps are currently being taken to separate debris from sodium by the purely physical process of distillation. Pyrotechnic charges diluted with finely divided aluminium metal and separately with silver have been fired under water in a viewing vessel to see if a metal/water thermal interaction could be demonstrated in this rig. This new vessel retains the original internal cylindrical geometry with optical correction on the viewing window for the cylindrical lens effect. These experiments produced very much higher gas blanket pressures on the first bubble expansion but this is thought to be a vapor pressure effect from the charge. The main debris engulfment is seen to take place shortly after the first peak gas blanket pressure and there is some evidence of a variation in the gas blanket pressure decay at this point in time. This may be evidence of a thermal interaction. One frictional plate has been constructed and used (50% void area) wtth a

Experiments on the behaviour of thermite melt injected into sodium: Final report on the THINA test results

International Nuclear Information System (INIS)

Huber, F.; Kaiser, A.; Peppler, W.

1994-01-01

During hypothetical accidents of fast breeder reactors the core melts and part of the core material inventory is ejected into the upper coolant plenum. As a consequence, a fuel to coolant thermal interaction occurs between the melt and the sodium. A series of simulating experiments was carried out in KfK/IRS to improve the knowledge about the phenomenology of molten fuel/coolant interactions and to support theoretical work on the safety of fast breeder reactors. In the tests, a thermite melt of up to 3270 K is injected from below into a sodium pool the temperature of which is between 770 and 820 K. The masses of the melt and the sodium are about five and 150 kg, respectively. Thermal interactions have been observed to occur as a sequence of small local pressure events mainly during the melt injection. Large-scale vapour explosions have not been observed. Generally, the conversion ratios of thermal to mechanical energy have been low. (author)

Metrological certification of systems to monitor the seal integrity of fuel-element cladding based on exposed fuel in sodium-cooled fast reactors

International Nuclear Information System (INIS)

Eliseev, A.V.; Filonov, V.S.; Ushakov, V.M.; Belov, S.P.; Pedyash, B.V.; Zemtsev, B.V.; Skorikov, N.V.

1992-01-01

In sodium-cooled fast reactors, the clad monitoring system for seal integrity of the fuel element cladding is practically the only source of operator information on the serviceability of fuel elements in the core. The monitoring system can be used as the basis for critical decisions whether the reactor must be shut down of whether operation can continue, but only if the meterologically provided measurements are reliable. This article describes a method developed for certifying working rods on the basis of the domestic standard. The method includes a combined irradiation of the sample and the rod to be certified in an arbitrary field of a plutonium-beryllium neutron source with an output rate greater than 10 8 sec -1 , which is mounted in a paraffin moderator. The positive results of the metrological certification of the system to monitor cladding seal integrity leads the authors to recommend this method for other current and planned sodium-cooled fast reactors. 6 refs., 2 tabs

Interactions between selected bile salts and Triton X-100 or sodium lauryl ether sulfate

Directory of Open Access Journals (Sweden)

Ćirin Dejan M

2011-12-01

Full Text Available Abstract Background In order to develop colloidal drug carriers with desired properties, it is important to determine physico-chemical characteristics of these systems. Bile salt mixed micelles are extensively studied as novel drug delivery systems. The objective of the present investigation is to develop and characterize mixed micelles of nonionic (Triton X-100 or anionic (sodium lauryl ether sulfate surfactant having oxyethylene groups in the polar head and following bile salts: cholate, deoxycholate and 7-oxodeoxycholate. Results The micellization behaviour of binary anionic-nonionic and anionic-anionic surfactant mixtures was investigated by conductivity and surface tension measurements. The results of the study have been analyzed using Clint's, Rubingh's, and Motomura's theories for mixed binary systems. The negative values of the interaction parameter indicate synergism between micelle building units. It was noticed that Triton X-100 and sodium lauryl ether sulfate generate the weakest synergistic interactions with sodium deoxycholate, while 7-oxodeoxycholate creates the strongest attractive interaction with investigated co-surfactants. Conclusion It was concluded that increased synergistic interactions can be attributed to the larger number of hydrophilic groups at α side of the bile salts. Additionally, 7-oxo group of 7-oxodeoxycholate enhance attractive interactions with selected co-surfactants more than 7-hydroxyl group of sodium cholate.

Interactions between selected bile salts and Triton X-100 or sodium lauryl ether sulfate.

Science.gov (United States)

Cirin, Dejan M; Poša, Mihalj M; Krstonošić, Veljko S

2011-12-29

In order to develop colloidal drug carriers with desired properties, it is important to determine physico-chemical characteristics of these systems. Bile salt mixed micelles are extensively studied as novel drug delivery systems. The objective of the present investigation is to develop and characterize mixed micelles of nonionic (Triton X-100) or anionic (sodium lauryl ether sulfate) surfactant having oxyethylene groups in the polar head and following bile salts: cholate, deoxycholate and 7-oxodeoxycholate. The micellization behaviour of binary anionic-nonionic and anionic-anionic surfactant mixtures was investigated by conductivity and surface tension measurements. The results of the study have been analyzed using Clint's, Rubingh's, and Motomura's theories for mixed binary systems. The negative values of the interaction parameter indicate synergism between micelle building units. It was noticed that Triton X-100 and sodium lauryl ether sulfate generate the weakest synergistic interactions with sodium deoxycholate, while 7-oxodeoxycholate creates the strongest attractive interaction with investigated co-surfactants. It was concluded that increased synergistic interactions can be attributed to the larger number of hydrophilic groups at α side of the bile salts. Additionally, 7-oxo group of 7-oxodeoxycholate enhance attractive interactions with selected co-surfactants more than 7-hydroxyl group of sodium cholate.

An Advanced Sodium-Cooled Fast Reactor Core Concept Using Uranium-Free Metallic Fuels for Maximizing TRU Burning Rate

Directory of Open Access Journals (Sweden)

Wuseong You

2017-12-01

Full Text Available In this paper, we designed and analyzed advanced sodium-cooled fast reactor cores using uranium-free metallic fuels for maximizing burning rate of transuranics (TRU nuclides from PWR spent fuels. It is well known that the removal of fertile nuclides such as 238U from fuels in liquid metal cooled fast reactor leads to the degradation of important safety parameters such as the Doppler coefficient, coolant void worth, and delayed neutron fraction. To resolve the degradation of the Doppler coefficient, we considered adding resonant nuclides to the uranium-free metallic fuels. The analysis results showed that the cores using uranium-free fuels loaded with tungsten instead of uranium have a significantly lower burnup reactivity swing and more negative Doppler coefficients than the core using uranium-free fuels without resonant nuclides. In addition, we considered the use of axially central B4C absorber region and moderator rods to further improve safety parameters such as sodium void worth, burnup reactivity swing, and the Doppler coefficient. The results of the analysis showed that the final design core can consume ~353 kg per cycle and satisfies self-controllability under unprotected accidents. The fuel cycle analysis showed that the PWR–SFR coupling fuel cycle option drastically reduces the amount of waste going to repository and the SFR burner can consume the amount of TRUs discharged from 3.72 PWRs generating the same electricity.

Transformation and fragmentation behavior of molten metal drop in sodium pool

International Nuclear Information System (INIS)

Nishimura, Satoshi; Zhang Zhigang; Sugiyama, Ken-Ichiro; Kinoshita, Izumi

2007-01-01

In order to clarify the fragmentation mechanism of a metallic alloy (U-Pu-Zr) fuel on liquid phase formed by metallurgical reactions (liquefaction temperature = 650 deg. C), which is important in evaluating the sequence of core disruptive accidents for metallic fuel fast reactors, a series of experiments was carried out using molten aluminum (melting point 660 deg. C) and sodium mainly under the condition that the boiling of sodium does not occur. When the instantaneous contact interface temperature (T i ) between molten aluminum drop and sodium is lower than the boiling point of sodium (T c,bp ), the molten aluminum drop can be fragmented and the mass median diameter (D m ) of aluminum fragments becomes small with increasing T i . When T i is roughly equivalent to or higher than T c,bp , the fragmentation of aluminum drop is promoted by thermal interaction caused by the boiling of sodium on the surface of the drop. Furthermore, even under the condition that the boiling of sodium does not occur and the solid crust is formed on the surface of the drop, it is confirmed from an analytical evaluation that the thermal fragmentation of molten aluminum drop with solid crust has a potential to be caused by the transient pressurization within the melt confined by the crust. These results indicate the possibility that the metallic alloy fuel on liquid phase formed by the metallurgical reactions can be fragmented without occurring the boiling of sodium on the surface of the melt

Synthesis Report on the understanding of failed LMFBR fuel element performance

International Nuclear Information System (INIS)

Plitz, H.; Bagley, K.; Harbourne, B.

1990-07-01

In the coarse of LMFBR operation fuel element failures cannot entirely be avoided as experienced during the operation of PFR, PHENIX and KNK II, where 44 failed fuel elements have been registered between 1978 and 1989. In earlier irradiations, post irradiation examinations showed mixed oxide pin diameter increases up to pin pitch distance, urging to stress reactor safety questions on the potential of fuel pin failure propagation within pin bundles. The chemical interaction of sodium with mixed oxide fuel is regarded to be the key for the understanding of failed fuel behavior. Valuable results on the failed fuel pin behavior during operation were obtained from the SILOE sodium loop test. Based on the bulk of experience with the detection of fuel pin failures, with the continued operation and with the handling of failed pins respectively elements, one can state: 1. All fuel pin failures have been detected securely in time and have been located. 2. Small defects are developing slowly. 3. Even large defects at end-of-life pins resulted in limited fuel loss. 4. Clad failures behave benign in main aspects. 5. The chemical interaction of sodium with mixed oxide is an important factor in the behavior of failed fuel pins, especially at high burnup. 6. Despite different pin designs and different operation conditions, on the basis of 44 failed elements in PFR, PHENIX and KNK II no pin-to-pin propagation was observed and fuel release was rather low, often not detectable. 7. In no case hazard conditions affecting reactor safety have been experienced

Investigation of sodium concrete interaction and the effect of different by-products

International Nuclear Information System (INIS)

Schultheiss, G.F.; Fritzke, H.W.

1983-01-01

For heat transfer at temperatures of more than 770 K sodium or sodium-potassium alloy is used in fast breeder reactors or solar power plants. In case of leakage of hot liquid metal concrete is loaded thermally and chemically. The interaction of sodium with several concretes of different composition has been investigated experimentally and theoretically. Especially the quartz content of concrete has a significant influence on reaction behavior. Quartz-containing concrete specimens were severely damaged at temperatures of more than 600 0 C. Computer code modelling shows good agreement with experiments. (orig.) [de

Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

2015-12-31

A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

Fuel-Coolant Interactions - some Basic Studies at the UKAEA Culham Laboratory

International Nuclear Information System (INIS)

Reynolds, J.A.; Dullforce, T.A.; Peckover, R.S.; Vaughan, G.J.

1976-01-01

In a hypothetical fault sequence important effects of fuel-coolant interactions include voiding and dispersion of core debris as well as the pressure damage usually discussed. The development of the fuel-coolant interaction probably depends on any pre-mixing Weber break-up that may occur, and is therefore a function of the way the fuel and coolant come together. Four contact modes are identified: jetting, shock tube, drops and static, and Culham's experiments have been mainly concerned with simulating the falling drop mode by using molten tin in water. It was observed that the fuel-coolant interaction is a short series of violent coolant oscillations centred at a localized position on the drop, generating a spray of submillimeter sized debris. The interaction started spontaneously at a specific time after the drop first contacted the water. There was a definite limited fuel-coolant interaction zone on a plot of initial coolant temperature versus initial fuel temperature outside which interactions never occurred. The. interaction time was a function of the initial temperatures. Theoretical scaling formulae are given which describe the fuel-coolant interaction zone and dwell time. Bounds of fuel and coolant temperature below which fuel-coolant interactions do not occur are explained by freezing. Upper bounds of fuel and coolant temperatures above which there were no fuel-coolant interactions are interpreted in terms of heat transfer through vapour films of various thicknesses. In conclusion: We have considered the effects of fuel-coolant interactions in a hypothetical fault sequence, emphasising that debris and vapour production as well as the pressure pulse can be important factors. The fuel-coolant interaction has been classified into types, according to possible modes of mixing in the fault sequence. Culham has been studying one type, the self-triggering of falling drops, by simulant experiments. It is found that there is a definite zone of interaction on a plot

Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

Energy Technology Data Exchange (ETDEWEB)

Karahan, Aydin, E-mail: karahan@mit.edu [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, 77 Massachusetts Avenue, Cambridge MA 24-204 (United States)

2011-07-15

Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other

Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

Science.gov (United States)

Karahan, Aydın

2011-07-01

Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other

Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

International Nuclear Information System (INIS)

Karahan, Aydin

2011-01-01

Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other

Pellet-clad interaction in water reactor fuels

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

The aim of this seminar is was to draw up a comprehensive picture of the pellet clad interaction and its impact on the fuel rod. This document is a detailed abstract of the papers presented during the following five sessions: industrial goals, fuel material behaviour in PCI situation, cladding behaviour relevant to PCI, in pile rod behaviour and modelling of the mechanical interaction between pellet and cladding. (A.L.B.)

Pellet-clad interaction in water reactor fuels

International Nuclear Information System (INIS)

2004-01-01

The aim of this seminar is was to draw up a comprehensive picture of the pellet clad interaction and its impact on the fuel rod. This document is a detailed abstract of the papers presented during the following five sessions: industrial goals, fuel material behaviour in PCI situation, cladding behaviour relevant to PCI, in pile rod behaviour and modelling of the mechanical interaction between pellet and cladding. (A.L.B.)

Some aspects of the chemistry of fast reactor fuel, structural material and decontamination

International Nuclear Information System (INIS)

Ganesan, V.

2012-01-01

The chemistry of materials pertaining to fast reactors is both fascinating and challenging considering the nature of materials involved such as the fuel, coolant, control and shielding materials in addition to the interactions between the structural materials and the fuel/coolant depending on the nature and conditions involved. The different chemical forms of fuel materials, the need to operate up to high burnups with consequent interactions of the fuel with clad materials, the need to close the fuel cycle by recovery of the fuel materials from spent fuels for refabrication and the necessity to manage the waste, throw a host of challenges which make their study scientifically interesting and technologically important. The use of liquid sodium as coolant in fast reactor heat transport systems combined with its inherent chemical reactivity opens up an interesting branch of chemistry involving liquid sodium especially in contact with structural materials during normal operation of the reactor and with fuels in the event of fuel pin failure. The phenomenon of sodium wetting and the associated corrosion of structural materials in contact with it combined with the need to carryout decontamination of such materials make it interesting to examine and evaluate their suitability for reuse without compromising on their structural integrity. Boron being the material of choice for control and shielding applications in fast reactors with varying isotopic enrichment and the technological challenge to produce large quantities of boron carbide makes it unique. Some of these aspects are addressed in this paper. (author)

Corrosion phenomena in sodium-potassium coolant resulting from solute interaction in multicomponent solution

Science.gov (United States)

Krasin, V. P.; Soyustova, S. I.

2018-03-01

The solubility of Fe, Cr, Ni, V, Mn and Mo in sodium-potassium melt has been calculated using the mathematical framework of pseudo-regular solution model. The calculation results are compared with available published experimental data on mass transfer of components of austenitic stainless steel in sodium-potassium loop under non-isothermal conditions. It is shown that the parameters of pair interaction of oxygen with transition metal can be used to predict the corrosion behavior of structural materials in sodium-potassium melt in the presence of oxygen impurity. The results of calculation of threshold concentration of oxygen of ternary oxide formation of sodium with transitional metals (Fe, Cr, Ni, V, Mn, Mo) are given in conditions when pure solid metal comes in contact with sodium-potassium melt.

Hydrophobic interactions between polymethacrylic acid and sodium laureth sulfate in aqueous solutions

Science.gov (United States)

Yaremko, Z. M.; Fedushinskaya, L. B.; Burka, O. A.; Soltys, M. N.

2014-09-01

The role of hydrophobic interaction in the development of associative processes is demonstrated, based on the concentration dependences of the viscosity and pH of binary solutions of polymethacrylic acid as an anionic polyelectrolyte and sodium laureth sulfate as an anionic surfactant. It is found that the inflection point on the dependence of the difference between the pH values of binary solutions of polymethacrylic acid and sodium laureth sulfate on the polyelectrolyte concentration is a criterion for determining the predominant contribution from hydrophobic interaction, as is the inflection point on the dependence of pH of individual solutions of polymethacrylic acid on the polyelectrolyte concentration.

Influence of fuel-matrix interaction on the breakaway swelling of U-Mo dispersion fuel in Al

Energy Technology Data Exchange (ETDEWEB)

Ryu, Ho Jin [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Kim, Yeon Soo [Nuclear Engineering Division, Argonne National Laboratory, Arogonne (United States)

2014-04-15

In order to advance understanding of the breakaway swelling behavior of U-Mo/Al dispersion fuel under a high-power irradiation condition, the effects of fuel-matrix interaction on the fuel performance of U-Mo/Al dispersion fuel were investigated. Fission gas release into large interfacial pores between interaction layers and the Al matrix was analyzed using both mechanistic models and observations of the post-irradiation examination results of U-Mo dispersion fuels. Using the model predictions, advantageous fuel design parameters are recommended to prevent breakaway swelling.

Results of thermal interaction tests for various materials performed in the Ispra tank facility

International Nuclear Information System (INIS)

Fasoli-Stella, P.; Holtbecker, H.; Jorzik, E.; Schlittenhardt, P.; Thoma, U.

A test facility for fuel/coolant thermal interaction measurements is described together with recent improvements of the melting oven design, the instrumentation and the collection and cleaning of the debris. The formation of a UO 2 crust on the melting crucible is investigated theoretically taking into account the heat losses during transport of the crucible from the oven to the reaction chamber. Experimental results for the systems steel-sodium, steel-water and UO 2 -sodium are presented and discussed with respect to particle size distribution and appearence of the debris. A sodium/fuel interaction model is introduced in the hydrodynamic REXCO-H-code. The results of test calculations are dealt with

Method of storing the fuel storage pot in a fuel storage tank for away-from-reactor-storage

International Nuclear Information System (INIS)

Ishiguro, Jun-ichi.

1980-01-01

Purpose: To prevent the contact of sodium in the away-from-reactor-storage fuel storage tank with sodium in a fuel storage pool having radioactivity ana always retain clean state therein. Method: Sodium is filled in a container body of the away-from-reactor-storage fuel storage tank, and a conduit, a cycling pump, and cooling means are disposed to form a sodium coolant cycling loop. The fuel storage pool is so stored in the container body that the heat of the pool is projected from the liquid surface of the sodium in the container. Therefore, the sodium in the container is isolated from the sodium in the pool containing strong radioactivity to prevent contact of the former sodium from the latter sodium. (Sekiya, K.)

Analysis of fuel cladding chemical interaction in mixed oxide fuel pins

International Nuclear Information System (INIS)

Weber, J.W.; Dutt, D.S.

1976-01-01

An analysis is presented of the observed interaction between mixed oxide 75 wt percent UO 2 --25 wt percent PuO 2 fuel and 316--20 percent CW stainless steel cladding in LMFBR type fuel pins irradiated in EBR-II. A description is given of the test pins and their operating conditions together with, metallographic observations and measurements of the fuel/cladding reaction, and a correlation equation is developed relating depth of cladding attack to temperature and burnup. Some recent data on cladding reaction in fuel pins with low initial O/M in the fuel are given and compared with the correlation equation curves

INFLUENCE OF FUEL-MATRIX INTERACTION ON THE BREAKAWAY SWELLING OF U-MO DISPERSION FUEL IN AL

OpenAIRE

HO JIN RYU; YEON SOO KIM

2014-01-01

In order to advance understanding of the breakaway swelling behavior of U-Mo/Al dispersion fuel under a high-power irradiation condition, the effects of fuel-matrix interaction on the fuel performance of U-Mo/Al dispersion fuel were investigated. Fission gas release into large interfacial pores between interaction layers and the Al matrix was analyzed using both mechanistic models and observations of the post-irradiation examination results of U-Mo dispersion fuels. Using the model prediction...

Fuel-cladding chemical interaction in mixed-oxide fuels

International Nuclear Information System (INIS)

Lawrence, L.A.; Weber, J.W.; Devary, J.L.

1978-10-01

The character and extent of fuel-cladding chemical interaction (FCCI) was established for UO 2 -25 wt% PuO 2 clad with 20% cold worked Type 316 stainless steel irradiated at high cladding temperatures to peak burnups greater than 8 atom %. The data base consists of 153 data sets from fuel pins irradiated in EBR-II with peak burnups to 9.5 atom %, local cladding inner surface temperatures to 725 0 C, and exposure times to 415 equivalent full power days. As-fabricated oxygen-to-metal ratios (O/M) ranged from 1.938 to 1.984 with the bulk of the data in the range 1.96 to 1.98. HEDL P-15 pins provided data at low heat rates, approx. 200 W/cm, and P-23 series pins provided data at higher heat rates, approx. 400 W/cm. A design practice for breeder reactors is to consider an initial reduction of 50 microns in cladding thickness to compensate for possible FCCI. This approach was considered to be a conservative approximation in the absence of a comprehensive design correlation for extent of interaction. This work provides to the designer a statistically based correlation for depth of FCCI which reflects the influences of the major fuel and operating parameters on FCCI

Interactions between kaolinite Al−OH surface and sodium hexametaphosphate

Energy Technology Data Exchange (ETDEWEB)

Han, Yonghua, E-mail: hyh19891102@163.com [School of Chemical and Environmental Engineering, China University of Mining and Technology, Beijing 100083 (China); Liu, Wenli; Zhou, Jia [School of Chemical and Environmental Engineering, China University of Mining and Technology, Beijing 100083 (China); Chen, Jianhua [College of Resources and Metallurgy, Guangxi University, Nanning 530004 (China)

2016-11-30

Highlights: • Sodium hexametaphosphate (NaHMP) can adsorb on kaolinite Al−OH terminated (001) surface easily. • The oxygen atoms of hexametaphosphate form strong hydrogen bonds with the hydrogen atoms of kaolinite Al−OH surface. • The electrostatic force is the main interaction between NaHMP and Al−OH surface. • The linear hexaphosphate −[PO{sub 3}]{sub m}− chains adsorb stably than −[HPO{sub 3}]{sub m}− chains. - Abstract: To investigate the dispersion mechanism of sodium hexametaphosphate on kaolinite particles, we simulated the interaction between linear polyphosphate chains and kaolinite Al−OH terminated surface by molecular dynamics, as well as the interaction between the [HPO{sub 4}]{sup 2−} anion and kaolinite Al−OH surface by density functional theory (DFT). The calculated results demonstrate that hexametaphosphate can be adsorbed by the kaolinite Al−OH surface. The oxygen atoms of hexametaphosphate anions may receive many electrons from the Al−OH surface and form hydrogen bonds with the hydrogen atoms of surface hydroxyl groups. Moreover, electrostatic force dominates the interactions between hexametaphosphate anions and kaolinite Al−OH surface. Therefore, after the adsorption of hexametaphosphate on kaolinite Al−OH surface, the kaolinite particles carry more negative charge and the electrostatic repulsion between particles increases. In addition, the adsorption of −[PO{sub 3}]{sub m}− species on the Al−OH surface should be more stable than the adsorption of −[HPO{sub 3}]{sub m}− species.

Experimental assessment of fuel-cladding interactions

Energy Technology Data Exchange (ETDEWEB)

Wood, Elizabeth Sooby [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-06-29

A range of fuel concepts designed to better tolerate accident scenarios and reactor transients are currently undergoing fundamental development at national laboratories as well as university and industrial partners. Pellet-clad mechanical and chemical interaction can be expected to affect fuel failure rates experienced during steady state operation, as well as dramatically impact the response of the fuel form under loss of coolant and other accident scenarios. The importance of this aspect of fuel design prompted research initiated by AFC in FY14 to begin exploratory efforts to characterize this phenomenon for candidate fuelcladding systems of immediate interest. Continued efforts in FY15 and FY17 aimed to better understand and simulate initial pellet-clad interaction with little-to-no pressure on the pellet-clad interface. Reported here are the results from 1000 h heat treatments at 400, 500, and 600°C of diffusion couples pairing UN with a FeCrAl alloy, SiC, and Zr-based cladding candidate sealed in evacuated quartz ampoules. No gross reactions were observed, though trace elemental contaminants were identified.

Performance comparison of metallic, actinide burning fuel in lead-bismuth and sodium cooled fast reactors

International Nuclear Information System (INIS)

Weaver, K.D.; Herring, J.S.; Macdonald, P.E.

2001-01-01

Various methods have been proposed to ''incinerate'' or ''transmute'' the current inventory of transuranic waste (TRU) that exits in spent light-water-reactor (LWR) fuel, and weapons plutonium. These methods include both critical (e.g., fast reactors) and non-critical (e.g., accelerator transmutation) systems. The work discussed here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead and lead-alloy cooled fast reactors for producing low-cost electricity as well as for actinide burning. The neutronics of non fertile fuel loaded with 20 or 30-wt% light water reactor (LWR) plutonium plus minor actinides for use in a lead-bismuth cooled fast reactor are discussed in this paper, with an emphasis on the fuel cycle life and isotopic content. Calculations show that the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from -1.02 to -1.16 g/MWd, compared to a typical LWR actinide generation rate of 0.303 g/MWd. However, when using the same parameters, the sodium-cooled case went subcritical after 0.2 to 0.8 effective full power years, and the lead-bismuth cooled case ranged from 1.5 to 4.5 effective full power years. (author)

Thermochemical aspects of fuel-cladding and fuel-coolant interactions in LMFBR oxide fuel pins

International Nuclear Information System (INIS)

Adamson, M.G.; Aitken, E.A.; Caputi, R.W.; Potter, P.E.; Mignanelli, M.A.

1979-01-01

This paper examines several thermochemical aspects of the fuel-cladding, fuel-coolant and fuel-fission product interactions that occur in LMFBR austenitic stainless steel-clad mixed (U,Pu)-oxide fuel pins during irradiation under normal operating conditions. Results are reported from a variety of high temperature EMF cell experiments in which continuous oxygen activity measurements on reacting and equilibrium mixtures of metal oxides and (excess) liquid alkali metal (Na, K, Cs) were performed. Oxygen potential and 0:M thresholds for Na-fuel reactions are re-evaluated in the light of new measurements and newly-assessed thermochemical data, and the influence on oxygen potential of possible U-Pu segregation between oxide and urano-plutonate (equilibrium) phases has been analyzed. (orig./RW) [de

Start-up fuel and power flattening of sodium-cooled candle core

International Nuclear Information System (INIS)

Takaki, Naoyuki; Sagawa, Yu; Umino, Akitake; Sekimoto, Hiroshi

2013-01-01

The hard neutron spectrum and unique power shape of CANDLE enable its distinctive performances such as achieving high burnup more than 30% and exempting necessity of both enrichment and reprocessing. On the other hand, they also cause several challenging problems. One is how the initial fuel can be prepared to start up the first CANDLE reactor because the equilibrium fuel composition that enables stable CANDLE burning is complex both in axial and radial directions. Another prominent problem is high radial power peaking factor that worsens averaged burnup, namely resource utilization factor in once-through mode and shorten the life time of structure materials. The purposes of this study are to solve these two problems. Several ideas for core configurations and startup fuel using single enrichment uranium and iron as a substitute of fission products are studied. As a result, it is found that low enriched uranium is applicable to ignite the core but all concepts examined here exceeded heat limits. Adjustment in enrichment and height of active and burnt zone is opened for future work. Sodium duct assemblies and thorium fuel assemblies loaded in the center region are studied as measures to reduce radial power peaking factor. Replacing 37 fuels by thorium fuel assemblies in the zeroth to third row provides well-balanced performance with flattened radial power distribution. The CANDLE core loaded with natural uranium in the outer and thorium in the center region achieved 35.6% of averaged burnup and 7.0 years of cladding life time owing to mitigated local fast neutron irradiation at the center. Using thorium with natural or depleted uranium in CANDLE reactor is also beneficial to diversifying fission resource and extending available term of fission energy without expansion of needs for enrichment and reprocessing

Measurement and analysis of vibrational behaviour of an SNR-fuel element in sodium flow

International Nuclear Information System (INIS)

Hess, B.F.H.; Ruppert, E.; Schmidt, H.; Vinzens, K.

1975-01-01

Within the framework of SNR-300 fuel element development programme a complete full size fuel element dummy has been tested thoroughly for nearly 3000 hours at 650 0 C system temperature in the AKB sodium loop at Interatom, Bensberg. Investigations of the hydraulic characteristics by measurements of specific pressure losses, flow velocities, leakage flow through the piston rings and investigations of its vibrational behaviour were part of this endurance test at elevated temperatures. The pressure drop versus flow and the leakage measurement are mentioned briefly to confirm the correctness of the test hydraulics. The vibrational behaviour of the element and the approach to analysis is the main object of this report. (Auth.)

Interactions between California's Low Carbon Fuel Standard and the National Renewable Fuel Standard

International Nuclear Information System (INIS)

Whistance, Jarrett; Thompson, Wyatt; Meyer, Seth

2017-01-01

This study investigates the economic interactions between a national renewable fuel policy, namely the Renewable Fuel Standard (RFS) in the United States, and a sub-national renewable fuel policy, the Low Carbon Fuel Standard (LCFS) in California. The two policies have a similar objective of reducing greenhouse gas emissions, but the policies differ in the manner in which those objectives are met. The RFS imposes a hierarchical mandate of renewable fuel use for each year whereas the LCFS imposes a specific annual carbon-intensity reduction with less of a fuel specific mandate. We model the interactions using a partial-equilibrium structural model of agricultural and energy markets in the US and Rest-of-World regions. Our results suggest the policies are mutually reinforcing in that the compliance costs of meeting one of the requirements is lower in the presence of the other policy. In addition, the two policies combine to create a spatial shift in renewable fuel use toward California even though overall renewable fuel use remains relatively unchanged. - Highlights: • Results suggest the RFS and LCFS are mutually reinforcing. • Overall level of renewable fuel use is similar across scenarios. • Renewable fuel use shifts toward California in the presence of the LCFS. • Higher ethanol blend (e.g. E85) use also shifts toward California.

Mapping of sodium void worth and doppler effect for sodium-cooled fast reactor - 15458

International Nuclear Information System (INIS)

Krepel, J.; Pelloni, S.; Bortot, S.; Panadero, A.L.; Mikityuk, K.

2015-01-01

The sodium-cooled fast reactor (SFR) represents the reference and the most technologically mastered system among the Generation-IV reactors. Nevertheless, the sodium void worth in the fuel regions of SFR is usually positive. To overcome this safety drawback, low-void sodium-cooled fast spectrum core (CFV) was proposed by CEA. Such a CFV core is used in the frame of WP6 'Core safety' of the FP7 Euratom ESNII+ project as a reference SFR design. The overall sodium void effect is negative for the CFV core. Nevertheless, locally it is positive in the fuel region and negative in the sodium plenum. Similarly, also the Doppler effect is spatially dependent and it varies between the inner and outer fuel regions and between the middle and lower blankets. Accordingly, knowledge of the local distributions or actually mappings of the two safety-related parameters will be necessary, before safety assessment and transient analysis can be done. In this study these maps have been produced using the deterministic code ERANOS. The obtained mapping shows strong local dependency of both safety-related effects. A sensitivity of the void effect to the sodium plenum modeling was also demonstrated. The results may serve as an input for the transient analysis of the CFV core or as a cross-check for the Monte Carlo method based maps. (authors)

Implications and control of fuel-cladding chemical interaction for LMFBR fuel pin design

International Nuclear Information System (INIS)

Roake, W.E.

1977-01-01

Fuel-cladding-chemical-interaction (FCCI) is typically incorporated into the design of an LMFBR fuel pin as a wastage allowance. Several interrelated factors are considered during the evolution of an LMFBR fuel pin design. Those which are indirectly affected by FCCI include: allowable pin power, fuel restructuring, fission gas migration and release from the fuel, fuel cracking, fuel swelling, in-reactor cladding creep, cladding swelling, and the cladding mechanical strain. Chemical activity of oxygen is the most readily controlled factor in FCCI. Two methods are being investigated: control of total oxygen inventory by limiting fuel O/M, and control of oxygen activity with buffer metals

Implications and control of fuel-cladding chemical interaction for LMFBR fuel pin design

Energy Technology Data Exchange (ETDEWEB)

Roake, W E [Westinghouse-Hanford Co., Richland, WA (United States)

1977-04-01

Fuel-cladding-chemical-interaction (FCCI) is typically incorporated into the design of an LMFBR fuel pin as a wastage allowance. Several interrelated factors are considered during the evolution of an LMFBR fuel pin design. Those which are indirectly affected by FCCI include: allowable pin power, fuel restructuring, fission gas migration and release from the fuel, fuel cracking, fuel swelling, in-reactor cladding creep, cladding swelling, and the cladding mechanical strain. Chemical activity of oxygen is the most readily controlled factor in FCCI. Two methods are being investigated: control of total oxygen inventory by limiting fuel O/M, and control of oxygen activity with buffer metals.

Prediction of the Sodium Void Reactivity in the Metal-fueled SFR Using the ENDF/B-VII.0 Library

Energy Technology Data Exchange (ETDEWEB)

Yun, Sunghwan; Lim, Jae-Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The SVR (Sodium Void Reactivity) is one of the most important parameters in SFR (Sodium-cooled Fast Reactor) safety analysis. In this paper, to estimate the error of the SVR in metal-fueled SFR, three physics experiments named as BFS-75-1, BFS-109-2A, and BFS-84-1 were examined using recent cross-section library, ENDF/B-VII.0 and the MCNP code. In the MCNP6 calculation, two million histories/generation with 50 inactive/300 active generations are used with the continuous-energy ENDF/B-VII.0 library. We expect that accuracy of total cross-section of the sodium may play a dominant role in errors of SVRs at core peripheral and sodium plenum regions, whereas accuracy of capture cross-section of the sodium may play a dominant role for the results in errors of SVRs at core central region. In addition, capture cross-sections of the sodium in the ENDF/B-VII.0, the JEFF-3.2, and the JENDL-4.0 libraries show significant differences between each other, while total cross-sections of sodium in three libraries show good agreement.

Reuse of waste foundry sand through interaction with sodium silicate binder

International Nuclear Information System (INIS)

Souza, J.C.; Chinelatto, A.S.A.; Chinelatto, A.L.; Oliveira, I.L.

2012-01-01

Green sand molds are used in metal casting process. However, after heating, activated bentonite present in green sand lose the binding properties, and part of the foundry sand has to be discarded from the process. The ABNT NBR 15.984/2011 establishes the management of waste foundry sand (WFS) avoiding disposal in landfills. The objective of this work was to investigate the possibility of reusing the WFS from the study of their interaction with sodium silicate binder. Studies with silica sand and new green sand was performed to compare the results obtained with the WFS. The characterizations of the samples were performed by measures the compressive strength, X-ray diffraction, optical microscopy and scanning electron microscopy. The results showed that there is interaction of the sodium silicate with the WFS as well as with the silica sand and green sand. (author)

Fuel-cladding mechanical interaction effects in fast reactor mixed oxide fuel

Energy Technology Data Exchange (ETDEWEB)

Boltax, A [Westinghouse Electric Corporation, Advanced Reactor Division, Madison, PA (United States); Biancheria, A

1977-04-01

Thermal and fast reactor irradiation experiments on mixed oxide fuel pins under steady-state and power change conditions reveal evidence for significant fuel-cladding mechanical interaction (FCMI) effects. Analytical studies with the LIFE-III fuel performance code indicate that high cladding stresses can be produced by general and local FCMI effects. Also, evidence is presented to show that local cladding strains can be caused by the accumulation of cesium at the fuel-cladding interface. Although it is apparent that steady-state FCMI effects have not given rise to cladding breaches in current fast reactors, it is anticipated that FCMI may become more important in the future because of interest in: higher fuel burnups; increased power ramp rates; load follow operation; and low swelling cladding alloys. (author)

Fuel-cladding mechanical interaction effects in fast reactor mixed oxide fuel

International Nuclear Information System (INIS)

Boltax, A.; Biancheria, A.

1977-01-01

Thermal and fast reactor irradiation experiments on mixed oxide fuel pins under steady-state and power change conditions reveal evidence for significant fuel-cladding mechanical interaction (FCMI) effects. Analytical studies with the LIFE-III fuel performance code indicate that high cladding stresses can be produced by general and local FCMI effects. Also, evidence is presented to show that local cladding strains can be caused by the accumulation of cesium at the fuel-cladding interface. Although it is apparent that steady-state FCMI effects have not given rise to cladding breaches in current fast reactors, it is anticipated that FCMI may become more important in the future because of interest in: higher fuel burnups; increased power ramp rates; load follow operation; and low swelling cladding alloys. (author)

Molecular interactions between selected sodium salts of bile acids and morphine hydrochloride.

Science.gov (United States)

Poša, Mihalj; Csanádi, János; Kövér, Katalin E; Guzsvány, Valéria; Batta, Gyula

2012-06-01

The objective of this study was to understand the prolonged analgesic action of morphine hydrochloride observed in the presence of sodium 12-oxochenodeoxycholanate. Based on literature, this phenomenon may be due to the formation of aggregates in the cell between the molecules of bile acids and morphine. In addition to the sodium 12-oxochenodeoxycholanate, the present investigation also included salts of cholic and 7-oxodeoxycholic acids. Saturation transfer difference NMR experiments showed that morphine binds to the bile acid molecule close to the aromatic protons H1 and H2 provided that the concentration of the bile acid salt approaches the critical micellar concentration (CMC). The spin-lattice relaxation times (T(1)) of the affected protons decrease significantly in the presence of micellar solutions of the bile acid salts, and the most pronounced change in T(1) was observed for sodium 7-oxodeoxycholate. Diffusion-ordered NMR experiments suggested that morphine hydrochloride can interact only with sodium 7-oxochenodeoxycholate. It can be supposed that the molecular ratio of sodium 7-oxodeoxycholate and morphine hydrochloride in the mixed micelle is 2:1. The CMC values of mixed micelles do not differ from the CMC values of the micelle constituents, which suggests that the binding of morphine hydrochloride does not perturb the hydrophobic domain of the bile acid molecule. In the presence of bile acids, the transfer rate constant (k(12)) of morphine hydrochloride from the buffered aqueous solution to chloroform (model of the cell membrane) shows a decrease. A significant decrease of the k(12) was also observed in the presence of micellar solutions. Kinetic measurements indicated that, in addition to micellar interaction between morphine hydrochloride and sodium salts of bile acids, a complex may also be formed in chloroform via hydrogen bonds formed between the drug and bile acid molecules. Copyright © 2012 Elsevier B.V. All rights reserved.

Prompt burst energetics experiments: fresh oxide/sodium series

International Nuclear Information System (INIS)

Reil, K.O.; Young, M.F.

1978-08-01

A series of in-pile experiments has been performed to provide information on thermal energy to work conversion under prompt burst excursion (PBE) conditions. These consisted of single pin tests using fresh uranium oxide or uranium carbide fuel in a capsule geometry, with either stagnant sodium or helium in the coolant channel. The experiments were irradiated with single or double pulses in the Annular Core Pulse Reactor (ACPR) to provide energy depositions up to 2900 J/g. This report covers the seven single and five double pulse UO 2 sodium-in tests. Experimental data includes pressure and linear motion transducer histories, measured work-energy conversion efficiencies, and post-irradiation examination. Analysis includes derived work-energy conversion efficiencies (up to 0.54%), pin failure modeling, hydrodynamic analysis of pressure pulse propagation in the channel, and piston stopping effects. Initial pressure events in the single pulse experiments appear to be dominated by fuel vapor pressure. Definite fuel-coolant interactions were observed in several experiments, including some that were coincident with stopping of the linear motion transducer piston, suggesting a possible triggering effect by the deceleration pressure

Molten fuel-moderator interaction

International Nuclear Information System (INIS)

Lee, J.H.S.; Kynstautas, R.

1987-02-01

A critical review of the current understanding of vapor explosions was carried out. It was concluded that, on the basis of actual industrial accidents and large scale experiments, energetic high yield steam explosion cannot be regarded as an improbable event if large quantities of molten fuel and coolant are mixed together. This study also reviewed a hydrodynamic transient model proposed by Henry and Fauske Associates to assess a molten fuel-moderator interaction event. It was found that the proposed model negates a priori the possibility of a violent event, by introducing two assumptions: 1) fine fragmentation of the molten fuel, and ii) rapid heat transfer from the fine fragments to form steam. Using the Hicks and Menzies thermodynamic model, maximum work potential and pressure rise in the calandria were estimated. However, it is recommended that a more representative upper bound model based on an underwater explosion of a pressurized volume of steam be developed

Fuel coolant interaction experiment by direct electrical heating method

International Nuclear Information System (INIS)

Takeda, Tsuneo; Hirano, Kenmei

1979-01-01

In the PCM (Power Cooling Mismatch) experiments, the FCI (Fuel Coolant Interaction) test is one of necessary tests in order to predict various phenomena that occur during PCM in the core. A direct electrical heating method is used for the FCI tests for fuel pellet temperature of over 1000 0 C. Therefore, preheating is required before initiating the direct electrical heating. The fuel pin used in the FCI tests is typical LWR fuel element, which is surrounded by coolant water. It is undersirable to heat up the coolant water during preheating of the fuel pin. Therefore, a zirconia (ZrO 2 ) pellet which is similar to a UO 2 pellet in physical and chemical properties is used. Electric property (electric conductivity) of ZrO 2 is particularly suitable for direct electrical heating as in the case of UO 2 . In this experiment, ZrO 2 pellet (melting point 2500 0 C) melting was achieved by use of both preheating and direct electrical heating. Temperature changes of coolant and fuel surface, as well as the pressure change of coolant water, were measured. The molten fuel interacted with the coolant and generated shock waves. A portion of this molten fuel fragmented into small particles during this interaction. The peak pressure of the observed shock wave was about 35 bars. The damaged fuel pin was photographed after disassembly. This report shows the measured coolant pressure changes and the coolant temperature changes, as well as photographs of damaged fuel pin and fuel fragments. (author)

Dimensional, microstructural and compositional stability of metal fuels

International Nuclear Information System (INIS)

Solomon, A.A.; Dayananda, M.A.

1993-01-01

The projects undertaken were to address two areas of concern for metal-fueled fast reactors: metallurgical compatibility of fuel and its fission products with the stainless steel cladding, and effects of porosity development in the fuel on fuel/cladding interactions and on sodium penetration in fuel. The following studies are reported on extensively in appendices: hot isostatic pressing of U-10Zr by coupled boundary diffusion/power law creep cavitation, liquid Na intrusion into porous U-10Zr fuel alloy by differential capillarity, interdiffusion between U-Zr fuel and selected Fe-Ni-Cr alloys, interdiffusion between U-Zr fuel vs selected cladding steels, and interdiffusion of Ce in Fe-base alloys with Ni or Cr

On the significance of modeling nuclear fuel behavior with the right representation of physical phenomena

Energy Technology Data Exchange (ETDEWEB)

Karahan, Aydin, E-mail: karahan@mit.ed [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, 77 Massachusetts Avenue, 24-204, Cambridge, MA 02139 (United States); Kazimi, Mujid S. [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, 77 Massachusetts Avenue, 24-204, Cambridge, MA 02139 (United States)

2011-02-15

Research highlights: Essence of more physics based modeling approaches to the fuel behavior problem is emphasized. Demonstrations on modeling of metallic and oxide fuel dimensional changes and fission gas behavior with more physics based and semi-empirical approaches are given. Essence of fuel clad chemical interaction modeling of the metallic fuel in an appropriate way and implications during short and long term transients for sodium fast reactor applications are discussed. - Abstract: This work emphasizes the relevance of representation of appropriate mechanisms for understanding the actual physical behavior of the fuel pin under irradiation. Replacing fully empirical simplified treatments with more rigorous semi-empirical models which include the important pieces of physics, would open the path to more accurately capture the sensitivity to various parameters such as operating conditions, geometry, composition, and enhance the uncertainty quantification process. Steady state and transient fuel behavior demonstration examples and implications are given for sodium fast reactor metallic fuels by using FEAST-METAL. The essence of appropriate modeling of the fuel clad mechanical interaction and fuel clad chemical interaction of the metallic fuels are emphasized. Furthermore, validation efforts for oxide fuel pellet swelling behavior at high temperature and high burnup LWR conditions and comparison with FRAPCON-EP and FRAPCON-3.4 codes will be given. The value of discriminating the oxide fuel swelling modes, instead of applying a linear line, is pointed out. Future directions on fuel performance modeling will be addressed.

Sodium fast reactors energy conversion systems. Na-CO2 interaction. Comparison with Na-water interaction of conventional water Rankine cycle

International Nuclear Information System (INIS)

Latge, Christian; Simon, Nicole

2006-01-01

The Sodium Fast Reactor is a very promising candidate for the development of Fast Neutron Reactors. It is well known owing to its wide development since the 1950's, throughout all countries involved in the development of nuclear power plants. The development of Sodium-cooled fast neutron reactors is possible due to its very attractive sodium, nuclear, physical and even some of its chemical properties. Nevertheless, the operational feedback has shown that the concept has several drawbacks: difficulties for In-Service Inspection and Repair operations due to the sodium opacity and possible detrimental effects of its reactivity with air and water when the heat conversion is performed with a conventional Rankine cycle. Moreover, the various design projects have shown some difficulties in enhancing its competitiveness with regards to existing NPPs without any new innovative options, i.e. the possibility of suppressing the intermediate circuits and/or the development of an optimized energy conversion system. The Supercritical CO 2 Brayton Cycle option for the energy conversion has been widely suggested because of its high thermodynamic efficiency (over 40%), its potential compactness of the Balance Of Plant equipment due to the small-sized turbo machinery system, and for its applicability to both Direct or Indirect Cycle (Na, PbBi, He) assuming the hypothesis that the Supercritical CO 2 -Na interaction has less serious potential consequences than sodium-water consequences in the conventional Rankine cycle. Within the framework of the SMFR (Small Modular Fast Reactor) project, developed jointly by Argonne National Laboratory (ANL-USA), the 'Commissariat a l'Energie Atomique' (CEA) and Japan Atomic Energy Agency (JAEA, formerly Japan Nuclear Cycle development), this option has been selected and investigated. This paper deals with the study of the interaction between Na and CO 2 , based on a literature review: the result of this study will allow the definition of R and D

Phase characteristics of rare earth elements in metallic fuel for a sodium-cooled fast reactor by injection casting

Energy Technology Data Exchange (ETDEWEB)

Kuk, Seoung Woo, E-mail: swkuk@kaeri.re.kr [Next Generation Fuel Development Division, Korea Atomic Energy Research Institute, Daedeok-daero 989-111, Yuseong-gu, Daejeon, 34057 (Korea, Republic of); Kim, Ki Hwan; Kim, Jong Hwan; Song, Hoon; Oh, Seok Jin; Park, Jeong-Yong; Lee, Chan Bock [Next Generation Fuel Development Division, Korea Atomic Energy Research Institute, Daedeok-daero 989-111, Yuseong-gu, Daejeon, 34057 (Korea, Republic of); Youn, Young-Sang [Nuclear Chemistry Research Division, Korea Atomic Energy Research Institute, Daedeok-daero 989-111, Yuseong-gu, Daejeon, 34057 (Korea, Republic of); Kim, Jong-Yun [Nuclear Chemistry Research Division, Korea Atomic Energy Research Institute, Daedeok-daero 989-111, Yuseong-gu, Daejeon, 34057 (Korea, Republic of); Radiochemistry & Nuclear Nonproliferation, University of Science & Technology, Gajeong-ro 217, Yuseong-gu, Daejeon, 34113 (Korea, Republic of)

2017-04-01

Uranium-zirconium-rare earth (U-Zr-RE) fuel slugs for a sodium-cooled fast reactor were manufactured using a modified injection casting method, and investigated with respect to their uniformity, distribution, composition, and phase behavior according to RE content. Nd, Ce, Pr, and La were chosen as four representative lanthanide elements because they are considered to be major RE components of fuel ingots after pyroprocessing. Immiscible layers were found on the top layers of the melt-residue commensurate with higher fuel slug RE content. Scanning electron microscopy-energy-dispersive X-ray spectroscopy (SEM-EDS) data showed that RE elements in the melt-residue were distributed uniformly throughout the fuel slugs. RE element agglomeration did not contaminate the fuel slugs but strongly affected the RE content of the slugs.

Acoustic detection of boiling in the Sodium Loop Safety Facility in-reactor experiment P1

International Nuclear Information System (INIS)

Carey, W.M.; Anderson, T.T.; Bobis, J.P.

1976-06-01

Acoustic data were obtained from two high-temperature lithium niobate microphones on the loop background noise and transient pressure pulses during the Sodium Loop Safety Facility (SLSF) P1 in-reactor experiment. This experiment simulated an LMFBR loss-of-piping-integrity (LOPI) transient on a nineteen element, end-of-life, enriched-UO 2 fuel assembly. The microphones were exposed to liquid sodium at a distance 4.85 meters above the reactor core at temperatures between 315 0 and 590 0 C. The distance and location of the microphones in the P1 Test Train provided an attenuative transmission path which was undesirable for optimum acoustic detection of sodium boiling and fuel failure. The data gathered on the loop background noise was observed to be dominated by pump and electrical noise at frequencies below 1.5 KHz and appeared to be dominated by flow induced local turbulence noise at higher frequencies. During the period of time that the sodium in the fuel assembly was at its saturation temperature 943 0 C (1730 0 F), as indicated by the wire wrap thermocouples, several discrete pulses were observed with peak-to-peak pressure between 3.3 kPa and 7.9 kPa and center frequencies between 360 and 550 Hz. The pulses occurred at two separate gradually increasing repetition rates. These observations appear to be consistent with the result of an impulsive forcing function interacting with a band passed Helmholtz resonator. These data are consistent with the hypothesis that sodium boiling occurred in the P1 fuel assembly, resulting in the formation of individual voids that collapsed upon reaching the subcooled sodium. These data provide pertinent information regarding the feasibility of sodium boiling detection and may provide additional insight into the dynamics of the void behavior

DART model for irradiation-induced swelling of dispersion fuel elements including aluminum-fuel interaction

International Nuclear Information System (INIS)

Rest, J.; Hofman, G.L.

1997-01-01

The Dispersion Analysis Research Tool (DART) contains models for fission-gas-induced fuel swelling, interaction of fuel with the matrix aluminum, for the resultant reaction-product swelling, and for the calculation of the stress gradient within the fuel particle. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by a comparison of DART calculations of fuel swelling of U 3 SiAl-Al and U 3 Si 2 -Al for various dispersion fuel element designs with the data

Mechanisms of the initial stage of fuel elements degradation of BN reactor fuel assemblies

International Nuclear Information System (INIS)

Zagorul'ko, Yu.I.; Kashcheev, M.V.; Ganichev, N.S.

2015-01-01

On the base of developed calculational technique numerical evaluation is carried out to the time of fuel element fracture in conditions of loss of sodium flow through fuel element jacket. Data on mechanical properties of steel EhK-164 is used in calculations. Calculations are carried out for different conditions of jacket outer surface cooling: by sodium of 1073 K temperature, by boiling sodium and by sodium in condition of film boiling. It is shown that time to jacket fracture under considered rupture mechanisms essentially depends on fuel element cooling conditions [ru

23Na-NMR-studies on the detection of the interaction of phospholipids with sodium ions

International Nuclear Information System (INIS)

Arnold, K.; Pausch, R.; Frenzel, J.; Winkler, E.

1975-01-01

The 23 Na-NMR-relaxation times have been measured in different sonicated phospholipid dispersions in dependence on the NaCl concentration. In an egg lecithin dispersion and a DPPC dispersion the relaxation rates are independent of the sodium concentration. In both systems there is no interaction between sodium ions and phospholipids. However, in a phosphatidylethanolamine dispersion a concentration dependence may be observed. Its interpretation is only possible for a stoichiometric ratio of 3:1 of the lecithin-ion-complex. The association constant is found to be k=65,0 l/Mol. For the case of an equimolar egg lecithin/phosphatidylethanolamine dispersion a stronger interaction is measured. The addition of CaCl 2 results in a complete inhibition of the binding of sodium ions at phosphatidylethanolamine

Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

International Nuclear Information System (INIS)

Yoon, K. H.; Lee, C. B.

2014-01-01

The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness

Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Yoon, K. H.; Lee, C. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness.

Triggered fragmentation experiment with sodium, silicone oil and pentane

International Nuclear Information System (INIS)

Morita, T.

1990-12-01

Within the analysis of severe hypothetical fast breeder accidents the consequences of a fuel-coolant-interaction have to be considered, i.e. the thermal interaction between hot molten fuel and sodium. For the detailed understanding of the fragmentation during the thermal interaction of a hot liquid droplet with a cold fluid series of experiments were performed with sodium and solicone oil as hot liquid and pentane as cold easily volatile fluid. For the precise observation of the reaction an efficient high speed camera with a maximum recording frequency of 1x105 f/s was used. So the fragmentation caused by boiling phenomena could be observed. The pictures were used to estimate quantitatively e.g. the volume of the reaction zone and its expansion rate. By a special measuring device for the first time results on the time dependent portion of the liquid within the reaction zone could be gained. Based on the measured results of the experiments the course of a typical reaction, which can be devided into six phases, is presented and physically explained in this report. The influence of experimental parameters, as pressure of the external trigger and temperature of the hot liquid droplet, was investigated and from this the role of the homogeneous nucleation temperature and the external trigger for the reaction was deduced. (orig.) [de

Sodium fire suppression

International Nuclear Information System (INIS)

Malet, J.C.

1979-01-01

Ignition and combustion studies have provided valuable data and guidelines for sodium fire suppression research. The primary necessity is to isolate the oxidant from the fuel, rather than to attempt to cool the sodium below its ignition temperature. Work along these lines has led to the development of smothering tank systems and a dry extinguishing powder. Based on the results obtained, the implementation of these techniques is discussed with regard to sodium fire suppression in the Super-Phenix reactor. (author)

Sodium fire suppression

Energy Technology Data Exchange (ETDEWEB)

Malet, J C [DSN/SESTR, Centre de Cadarache, Saint-Paul-lez-Durance (France)

1979-03-01

Ignition and combustion studies have provided valuable data and guidelines for sodium fire suppression research. The primary necessity is to isolate the oxidant from the fuel, rather than to attempt to cool the sodium below its ignition temperature. Work along these lines has led to the development of smothering tank systems and a dry extinguishing powder. Based on the results obtained, the implementation of these techniques is discussed with regard to sodium fire suppression in the Super-Phenix reactor. (author)

Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor

International Nuclear Information System (INIS)

Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock

2007-09-01

The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor (SFR). From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents (74%) were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number (9) of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean

Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock

2007-09-15

The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor (SFR). From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents (74%) were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number (9) of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean.

Challenges and innovative technologies on fuel handling systems for future sodium-cooled fast reactors

International Nuclear Information System (INIS)

Chassignet, Mathieu; Dumas, Sebastien; Penigot, Christophe; Prele, Gerard; Capitaine, Alain; Rodriguez, Gilles; Sanseigne, Emmanuel; Beauchamp, Francois

2011-01-01

The reactor refuelling system provides the means of transporting, storing, and handling reactor core subassemblies. The system consists of the facilities and equipment needed to accomplish the scheduled refuelling operations. The choice of a FHS impacts directly on the general design of the reactor vessel (primary vessel, storage, and final cooling before going to reprocessing), its construction cost, and its availability factor. Fuel handling design must take into account various items and in particular operating strategies such as core design and management and core configuration. Moreover, the FHS will have to cope with safety assessments: a permanent cooling strategy to prevent fuel clad rupture, plus provisions to handle short-cooled fuel and criteria to ensure safety during handling. In addition, the handling and elimination of residual sodium must be investigated; it implies specific cleaning treatment to prevent chemical risks such as corrosion or excess hydrogen production. The objective of this study is to identify the challenges of a SFR fuel handling system. It will then present the range of technical options incorporating innovative technologies under development to answer the GENERATION IV SFR requirements. (author)

Fuel compliance model for pellet-cladding mechanical interaction

International Nuclear Information System (INIS)

Shah, V.N.; Carlson, E.R.

1985-01-01

This paper describes two aspects of fuel pellet deformation that play significant roles in determining maximum cladding hoop strains during pellet-cladding mechanical interaction: compliance of fragmented fuel pellets and influence of the pellet end-face design on the transmission of axial compressive force in the fuel stack. The latter aspect affects cladding ridge formation and explains several related observations that cannot be explained by the hourglassing model. An empirical model, called the fuel compliance model and representing the above aspects of fuel deformation, has been developed using the results from two Halden experiments and incorporated into the FRAP-T6 fuel performance code

Some techniques for sodium removal in CIAE

International Nuclear Information System (INIS)

Yuan Waimai; Ding Dejun; Guo Huanfang; Hong Shuzhang; Zhou Shuxia; Shen Fenyang; Yang Zhongmin; Xu Yongxing

1997-01-01

In this paper the experiment and application on sodium removal and sodium disposal are presented. Steam-nitrogen process was used in CIAE for cleaning cold traps, sodium vapor traps, a sodium tank. Atomized water-nitrogen process was used for cleaning dummy fuel assembly for CEFR and a sintered stainless steel filter. Sprinkle process was used for cleaning some tubes. Bultylcellosolve was used for cleaning sintered stainless steel filter and sodium flow measurement device. Ethanol alcohol was used for cleaning electromagnetic pump. Paraffin, transformer-oil or their mixture was used for cleaning sodium valves, a sodium vapor trap and sodium-potassium alloy absorber. A small sintered stainless steel filter was distillated in vacuum. A simple sodium disposal device has been served for several years in CIA.E. It can dispose about 10 Kg sodium each time and the disposal process is no-aerosol. It operates in open air for non-radioactive sodium. In recent years a small sodium cleaning plant has been built. It can use atomized water, steam or organic alcohol to removal of sodium. The LAVEL cleaning plant and SLAPSO cleaning plant were introduced from Italy. And CEFR preliminary design on sodium cleaning for spent fuel assembly and on sodium removal-decontamination for large reactor components is introduced. Vapour-nitrogen process is planned to use in them. (author)

Status of sodium cooled fast reactors with closed fuel cycle in India

International Nuclear Information System (INIS)

Raj, B.

2007-01-01

Fast reactors form the second stage of India's 3-stage nuclear power programme. The seed for India's fast reactor programme was sown through the construction of the Fast Breeder Test Reactor (FBTR) at IGCAR, Kalpakkam, that was commissioned in 1985. FBTR has operated with an unique, indigenously developed plutonium rich mixed carbide fuel, which has reached a burn up as high as 155 GWd/t without any fuel failure in the core. The sodium systems in the reactor have performed excellently. The availability of the reactor has been as high as 92% in the recent campaigns. The fuel discharged from FBTR up to 100 GWd/t has been reprocessed successfully. The experience gained in the construction, commissioning and operation of FBTR has provided the necessary confidence to launch a Prototype FBR of 500 MWe capacity (PFBR). This reactor will be fuelled by uranium, plutonium mixed oxide. The reactor construction started in 2003 and the reactor is scheduled to be commissioned by 2010. The design of the reactor has incorporated the worldwide operating experience from the FBRs and has addressed various safety issues reported in literature, besides introducing a number of innovative features which have reduced the unit energy cost and contributed to its enhanced safety. Simultaneous with the construction of the reactor, the fuel cycle of the reactor has been addressed in a comprehensive manner and construction of a fuel cycle facility has been initiated. Subsequent to the PFBR, 4 more reactors with identical design are proposed to be constructed. Various elements of reactor design are being carefully analysed with the aim of introducing innovative features towards further reduction in unit energy cost and enhancing safety in these reactors

Sodium boiling and mixed oxide fuel thermal behavior in FBR undercooling transients; W-1 SLSF experiment results

International Nuclear Information System (INIS)

Henderson, J.M.; Wood, S.A.; Knight, D.D.

1981-01-01

The W-1 Sodium Loop Safety Facility (SLSF) Experiment was conducted to study fuel pin heat release characteristics during a series of LMFBR Loss-of-Piping Integrity (LOPI) transients and to investigate a regime of coolant boiling during a second series of transients at low, medium and high bundle power levels. The LOPI transients produced no coolant boiling and showed only small changes in coolant temperatures as the test fuel microstructure changed from a fresh, unrestructured to a low burnup, restructured condition. During the last of seven boiling transients, intense coolant boiling produced inlet flow reversal, cladding dryout and moderate cladding melting

Interactive hypermedia training manual for spent-fuel bundle counters

International Nuclear Information System (INIS)

Basso, R.A.

1990-07-01

Spent-fuel bundle counters, developed by the Canadian Safeguards Support Program for the International Atomic Energy Agency, provide a secure and independent means of counting the number of irradiated fuel bundles discharged into the fuel storage bays at CANDU nuclear power stations. Paper manuals have been traditionally used to familiarize IAEA inspectors with the operation, maintenance and extensive reporting capabilities of the bundle counters. To further assist inspectors, an interactive training manual has been developed on an Apple Macintosh computer using hypermedia software. The manual uses interactive animation and sound, in conjunction with the traditional text and graphics, to simulate the underlying operation and logic of the bundle counters. This paper presents the key features of the interactive manual and highlights the advantages of this new technology for training

Preliminary Analysis of the Fuel Bundle Stiffness by ANSYS for SFR

Energy Technology Data Exchange (ETDEWEB)

Lee, Byoung Oon; Cheon, Jin Sik; Hahn, Do Hee; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2008-05-15

In SFR (Sodium-cooled Fast Reactor) the temperature of the fuel pin is higher than that of the hexagonal duct, so the thermal expansion rate of the fuel bundle is higher than that of the duct. The neutron fluence and the fuel pin pressure are also increased according to the burnup. So the radial expansion and bowing of a fuel pin bundle would occur, and then fuel bundle would interact with a duct. This phenomenon is called bundle-to-duct interaction (BDI). Under the BDI condition, excess cladding strain and hot spots would occur. Therefore BDI as well as the core mechanics should be considered to evaluate the FBR fuel integrity. The analysis codes such as ETOILE, SHADOW, and MARSE, have been developed to evaluate the BDI behavior. The ANSYS based model is also being developed to analysis the bundle duct interaction for SFR in Korea. In this paper, the fuel pin/bundle model for analyzing the bending deflection and oval deformation was described. The preliminary analysis of the fuel bundle stiffness was performed by the developed model.

Advances in Metallic Fuels for High Burnup and Actinide Transmutation

Energy Technology Data Exchange (ETDEWEB)

Hayes, S. L.; Harp, J. M.; Chichester, H. J. M.; Fielding, R. S.; Mariani, R. D.; Carmack, W. J.

2016-10-01

Research and development activities on metallic fuels in the US are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is a desire to demonstrate a multifold increase in burnup potential. A number of metallic fuel design innovations are under investigation with a view toward significantly increasing the burnup potential of metallic fuels, since higher discharge burnups equate to lower potential actinide losses during recycle. Promising innovations under investigation include: 1) lowering the fuel smeared density in order to accommodate the additional swelling expected as burnups increase, 2) utilizing an annular fuel geometry for better geometrical stability at low smeared densities, as well as the potential to eliminate the need for a sodium bond, and 3) minor alloy additions to immobilize lanthanide fission products inside the metallic fuel matrix and prevent their transport to the cladding resulting in fuel-cladding chemical interaction. This paper presents results from these efforts to advance metallic fuel technology in support of high burnup and actinide transmutation objectives. Highlights include examples of fabrication of low smeared density annular metallic fuels, experiments to identify alloy additions effective in immobilizing lanthanide fission products, and early postirradiation examinations of annular metallic fuels having low smeared densities and palladium additions for fission product immobilization.

Chemical interaction between the oxide and the clad in PHENIX fuel at burnup up to 60,000 MWd/t

International Nuclear Information System (INIS)

Conte, M.; Marcon, J.P.

1977-01-01

In every fuel element there is a potential problem of chemical interaction between the fissile portion and the clad. As a matter of fact, even if the choice of materials is made after having established a satisfactory chemical compatibility between the fuel- (UO 2 (U,Pu)O 2 , (U,Pu) C, . . .) and the clad (stainless steel, zircaloy, . . . ) out of pile, it is difficult to guarantee this compatibility after operation in the reactor due, on one hand, to the presence of fission products and, on the other hand, to impurities which are always present in the fuel to a greater or lesser degree. The fuel element currently chosen for the sodium-cooled fast reactors ((U,Pu)O 2 in stainless steel clad) does not avoid this problem, in particular because of the relatively high temperatures envisioned for this type of reactor - the clad temperature is about 650 deg. C. Since it is considered as a demonstration reactor, Phenix should be able to provide additional information on this phenomenon, and one will see that we have been able to shed light on some points which the experiments or irradiations made to date have been unable to explain. However, before presenting the experimental results obtained with Phenix fuel end drawing conclusions, we shall give a brief resume of the expected behavior of this fuel with respect to the phenomenon of interest. (author)

Hazardous Materials Verification and Limited Characterization Report on Sodium and Caustic Residuals in Materials and Fuel Complex Facilities MFC-799/799A

Energy Technology Data Exchange (ETDEWEB)

Gary Mecham

2010-08-01

This report is a companion to the Facilities Condition and Hazard Assessment for Materials and Fuel Complex Sodium Processing Facilities MFC-799/799A and Nuclear Calibration Laboratory MFC-770C (referred to as the Facilities Condition and Hazards Assessment). This report specifically responds to the requirement of Section 9.2, Item 6, of the Facilities Condition and Hazards Assessment to provide an updated assessment and verification of the residual hazardous materials remaining in the Sodium Processing Facilities processing system. The hazardous materials of concern are sodium and sodium hydroxide (caustic). The information supplied in this report supports the end-point objectives identified in the Transition Plan for Multiple Facilities at the Materials and Fuels Complex, Advanced Test Reactor, Central Facilities Area, and Power Burst Facility, as well as the deactivation and decommissioning critical decision milestone 1, as specified in U.S. Department of Energy Guide 413.3-8, “Environmental Management Cleanup Projects.” Using a tailored approach and based on information obtained through a combination of process knowledge, emergency management hazardous assessment documentation, and visual inspection, this report provides sufficient detail regarding the quantity of hazardous materials for the purposes of facility transfer; it also provides that further characterization/verification of these materials is unnecessary.

Chemical and mineralogical aspects of water-bentonite interaction in nuclear fuel disposal conditions

International Nuclear Information System (INIS)

Melamed, A.; Pitkaenen, P.

1996-01-01

In the field of nuclear fuel disposal, bentonite has been selected as the principal sealing and buffer material for placement around waste canisters, forming both a mechanical and chemical barrier between the radioactive waste and the surrounding ground water. Ion exchange and mineral alteration processes were investigated in a laboratory study of the long-term interaction between compacted Na-bentonite (Volclay MX-80) and ground water solutions, conducted under simulated nuclear fuel disposal conditions. The possible alteration of montmorillonite into illite has been a major object of the mineralogical study. However, no analytical evidence was found, that would indicate the formation of this non-expandable clay type. Apparently, the change of montmorillonite from Na- to Ca-rich was found to be the major alteration process in bentonite. In the water, a concentration decrease in Ca, Mg, and K, and an increase in Na, HCO 3 and SO 4 were recorded. The amount of calcium ions available in the water was considered insufficient to account for the recorded formation of Ca-montmorillonite. It is therefore assumed that the accessory Ca-bearing minerals in bentonite provide the fundamental source of these cations, which exchange with sodium during the alteration process. (38 refs.)

Effect of sodium borohydride synthesis on NaBH4-H2 system economics

International Nuclear Information System (INIS)

Tabakoglu, F. oeznur; Kurtulus, Guelbahar

2007-01-01

The hazards and negative impacts of fossil fuel usage on environment and the prospect of fossil fuel depletion in near future have urged scientists to search for and use clean energy sources and alternative fuels. Hydrogen is the best fuel among others, which can minimize the effects of global warming. Although it is currently more expensive than other fuels, it will be cheaper following further developments in hydrogen technologies from production till end-use. Hydrogen storage is a critical issue in terms of safety and economics of hydrogen energy system. Chemical hydrides are an attractive hydrogen storage method due to their potential of achieving high volumetric and gravimetric storage densities. Among chemical hydrides, sodium borohydride (NaBH 4 ) is given a big attention, due to its 10.8% theoretical hydrogen storage capacity. Hydrogen, which can be released by sodium borohydride hydrolysis reaction on-site, can be used in a proton exchange membrane fuel cell (PEMFC) at anode. on the other hand, sodium borohydride solution can be used directly in a borohydride fuel cell (DBFC) at anode. Like the other chemical hydrides, sodium borohydride has been an expensive material up to now, constituting a major obstacle to commercialization of sodium borohydride as a hydrogen storage method. This paper aims to give an approximate estimation process cost of the NaBH 4 -H 2 system by taking into account both the energy and raw material costs, starting with sodium borohydride production till recycling of it. Two different methods to synthesize sodium borohydride are analyzed and their effects on total cost are compared. It was found that the usage of Bayer process to synthesize sodium borohydride makes the overall sodium borohydride - hydrogen system cost higher than the total cost of the alternative process which starts with the production of sodium borohydride from borax decahydrate. (authors)

Under sodium ultrasonic imaging system for PFBR

International Nuclear Information System (INIS)

Patankar, V.H.; Lalwani, S.K.; Agashe, A.A.

2014-01-01

Under Sodium UltraSonic Scanner (USUSS) has been developed to detect the growth and protrusion of fuel sub-assemblies of PFBR, submerged in liquid sodium by using the ultrasonic imaging technique during reactor shut-down when liquid sodium is at 180 ℃. The imaging is carried out prior to every Fuel handling operation. Electronics Division, BARC has designed and developed an 8-Channel Ultrasonic Imaging System (UIS) which consists of 4 downward viewing and 4 side viewing ultrasonic transducers alongwith pulser-receiver, signal processing electronics hardware and software. An automated mechanical scanner developed by IGCAR houses sodium immersible transducers to image the fuel sub assemblies. The system has been successfully tested with dummy protruding and grown FSAs, submerged under liquid sodium. Such ultrasonic imaging systems are not available to India from international market. The USUSS developed indigenously has all the features available in similar systems developed by other countries. After every imaging campaign, the mechanical scanner containing ultrasonic transducers is stored in the Argon filled storage-pit. Before every campaign of USUSS, it is necessary to check the healthiness of the sodium immersible and contaminated ultrasonic transducers, as the under-sodium scanner is decontaminated once in five years. For this purpose, a novel Non Contact Ultrasonic Inspection System (NCUIS) has been designed and developed by Electronics Division, BARC to check the functionality of the high-temperature and contaminated transducers of USUSS, using air-coupled ultrasonic technique. (author)

A Neutronic Feasibility Study on the Recycling of an Oxide Fuel in Sodium-Cooled Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

Roh, Gyu Hong; Choi, Hang Bok

2006-06-15

Neutronic feasibility was implemented for the recycling of a mixed oxide fuel in sodium-cooled fast reactors (SFR) through a thermal/mechanical dry process, which is recognized as one of the most proliferation- resistant recycling processes. In order to assess the applicability of a simple dry process which is not capable of completely removing all the fission products from a spent fuel, sensitivity calculations were performed for the reactor physics parameters with a dependency on the fission product removal rate of the recycled spent fuel. The equilibrium core calculations were performed by the REBUS-3 code for a BN-600 core without blanket fuels and a modified core with an increased fuel volume fraction. The reactor performance parameters such as the transuranic content, breeding ratio, peak linear power, burnup reactivity swing and reactivity coefficients were calculated for an equilibrium core under a fixed fuel management scheme. The results showed that a recycling of the oxide fuel in the SFR is feasible if the fission products are removed by more than 70% through the dry process as far as the material balance is concerned. However the physics analysis also showed that some of the physics design parameters are slightly deteriorated. The results of this study indicate that the recycling characteristics can be improved if the dry process can remove more fission products, and the reactor configuration is further optimized or the spent fuel composition is adjusted.

A Neutronic Feasibility Study on the Recycling of an Oxide Fuel in Sodium-Cooled Fast Reactors

International Nuclear Information System (INIS)

Roh, Gyu Hong; Choi, Hang Bok

2006-06-01

Neutronic feasibility was implemented for the recycling of a mixed oxide fuel in sodium-cooled fast reactors (SFR) through a thermal/mechanical dry process, which is recognized as one of the most proliferation- resistant recycling processes. In order to assess the applicability of a simple dry process which is not capable of completely removing all the fission products from a spent fuel, sensitivity calculations were performed for the reactor physics parameters with a dependency on the fission product removal rate of the recycled spent fuel. The equilibrium core calculations were performed by the REBUS-3 code for a BN-600 core without blanket fuels and a modified core with an increased fuel volume fraction. The reactor performance parameters such as the transuranic content, breeding ratio, peak linear power, burnup reactivity swing and reactivity coefficients were calculated for an equilibrium core under a fixed fuel management scheme. The results showed that a recycling of the oxide fuel in the SFR is feasible if the fission products are removed by more than 70% through the dry process as far as the material balance is concerned. However the physics analysis also showed that some of the physics design parameters are slightly deteriorated. The results of this study indicate that the recycling characteristics can be improved if the dry process can remove more fission products, and the reactor configuration is further optimized or the spent fuel composition is adjusted

Interaction of oil and paraffin with sodium

Energy Technology Data Exchange (ETDEWEB)

Kozlov, F A; Zagorulko, Yu I; Kovalev, Yu P; Lamykin, E V; Alekseev, V V; Sergeev, G P [Institute of Physics and Power Engineering, Obninsk (USSR)

1980-05-01

This report presents the investigation results on the kinetics of a gaseous oil pyrolysis product composition change in the presence of sodium under static conditions at 350 deg. C, 550 deg. C and 700 deg. C as well as the behaviour of oil and paraffin within the sodium circulating loops. The calculated estimations of the critical oil physical state parameters in sodium flows are given. (author)

Interaction of oil and paraffin with sodium

International Nuclear Information System (INIS)

Kozlov, F.A.; Zagorulko, Yu.I.; Kovalev, Yu.P.; Lamykin, E.V.; Alekseev, V.V.; Sergeev, G.P.

1980-01-01

This report presents the investigation results on the kinetics of a gaseous oil pyrolysis product composition change in the presence of sodium under static conditions at 350 deg. C, 550 deg. C and 700 deg. C as well as the behaviour of oil and paraffin within the sodium circulating loops. The calculated estimations of the critical oil physical state parameters in sodium flows are given. (author)

Integrated Fuel-Coolant Interaction (IFCI 6.0) code

International Nuclear Information System (INIS)

Davis, F.J.; Young, M.F.

1994-04-01

The integrated Fuel-Coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, four-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a product of the effort to generate a stand-alone version of IFCI, IFCI 6.0. The User's Manual describes in detail the hydrodynamic method and physical models used in IFCI 6.0. Appendix A is an input manual, provided for the creation of working decks

MONJU fuel pin performance analysis

International Nuclear Information System (INIS)

Kitagawa, H.; Yamanaka, T.; Hayashi, H.

1979-01-01

Monju fuel pin has almost the same properties as other LMFBR fuel pins, i.e. Phenix, PFR, CRBR, but would be irradiated under severe conditions: maximum linear heat rate of 381 watt/cm, hot spot cladding temperature of 675 deg C, peak burnup of 131,000 MWd/t, peak fluence (E greater than 0.1 MeV) of 2.3 10 23 n/cm 2 . In order to understand in-core performance of Monju fuel pin, its thermal and mechanical behaviour was predicted using the fast running performance code SIMPLE. The code takes into account pellet-cladding interaction due to thermal expansion and swelling, gap conductance, structural changes of fuel pellets, fission product gas release with burnup and temperature increase, swelling and creep of fuel pellets, corrosion of cladding due to sodium flow and chemical attack by fission products, and cumulative damage of the cladding due to thermal creep

Development of Diffusion barrier coatings and Deposition Technologies for Mitigating Fuel Cladding Chemical Interactions (FCCI)

Energy Technology Data Exchange (ETDEWEB)

Sridharan, Kumar; Allen, Todd; Cole, James

2013-02-27

The goal of this project is to develop diffusion barrier coatings on the inner cladding surface to mitigate fuel-cladding chemical interaction (FCCI). FCCI occurs due to thermal and radiation enhanced inter-diffusion between the cladding and fuel materials, and can have the detrimental effects of reducing the effective cladding wall thickness and lowering the melting points of the fuel and cladding. The research is aimed at the Advanced Burner Reactor (ABR), a sodium-cooled fast reactor, in which higher burn-ups will exacerbate the FCCI problem. This project will study both diffusion barrier coating materials and deposition technologies. Researchers will investigate pure vanadium, zirconium, and titanium metals, along with their respective oxides, on substrates of HT-9, T91, and oxide dispersion-strengthened (ODS) steels; these materials are leading candidates for ABR fuel cladding. To test the efficacy of the coating materials, the research team will perform high-temperature diffusion couple studies using both a prototypic metallic uranium fuel and a surrogate the rare-earth element lanthanum. Ion irradiation experiments will test the stability of the coating and the coating-cladding interface. A critical technological challenge is the ability to deposit uniform coatings on the inner surface of cladding. The team will develop a promising non-line-of-sight approach that uses nanofluids . Recent research has shown the feasibility of this simple yet novel approach to deposit coatings on test flats and inside small sections of claddings. Two approaches will be investigated: 1) modified electrophoretic deposition (MEPD) and 2) boiling nanofluids. The coatings will be evaluated in the as-deposited condition and after sintering.

Fundamental aspects of nuclear reactor fuel elements

International Nuclear Information System (INIS)

Olander, D.R.

1976-01-01

The book presented is designed to function both as a text for first-year graduate courses in nuclear materials and as a reference for workers involved in the materials design and performance aspects of nuclear power plants. The contents are arranged under the following chapter headings: statistical thermodynamics, thermal properties of solids, crystal structures, cohesive energy of solids, chemical equilibrium, point defects in solids, diffusion in solids, dislocations and grain boundaries, equation of state of UO 2 , fuel element thermal performance, fuel chemistry, behavior of solid fission products in oxide fuel elements, swelling due to fission gases, pore migration and fuel restructuring kinetics, fission gas release, mechanical properties of UO 2 , radiation damage, radiation effects in metals, interaction of sodium and stainless steel, modeling of the structural behavior of fuel elements and assemblies

Final Report - Low Temperature Combustion Chemistry And Fuel Component Interactions

Energy Technology Data Exchange (ETDEWEB)

Wooldridge, Margaret [Univ. of Michigan, Ann Arbor, MI (United States)

2017-02-24

Recent research into combustion chemistry has shown that reactions at “low temperatures” (700 – 1100 K) have a dramatic influence on ignition and combustion of fuels in virtually every practical combustion system. A powerful class of laboratory-scale experimental facilities that can focus on fuel chemistry in this temperature range is the rapid compression facility (RCF), which has proven to be a versatile tool to examine the details of fuel chemistry in this important regime. An RCF was used in this project to advance our understanding of low temperature chemistry of important fuel compounds. We show how factors including fuel molecular structure, the presence of unsaturated C=C bonds, and the presence of alkyl ester groups influence fuel auto-ignition and produce variable amounts of negative temperature coefficient behavior of fuel ignition. We report new discoveries of synergistic ignition interactions between alkane and alcohol fuels, with both experimental and kinetic modeling studies of these complex interactions. The results of this project quantify the effects of molecular structure on combustion chemistry including carbon bond saturation, through low temperature experimental studies of esters, alkanes, alkenes, and alcohols.

Posttest examination results of recent treat tests on metal fuel

International Nuclear Information System (INIS)

Holland, J.W.; Wright, A.E.; Bauer, T.H.; Goldman, A.J.; Klickman, A.E.; Sevy, R.H.

1986-01-01

A series of in-reactor transient tests is underway to study the characteristics of metal-alloy fuel during transient-overpower-without-scam conditions. The initial tests focused on determining the margin to cladding breach and the axial fuel motions that would mitigate the power excursion. The tests were conducted in flowing-sodium loops with uranium - 5% fissium EBR-II Mark-II driver fuel elements in the TREAT facility. Posttest examination of the tests evaluated fuel elongation in intact pins and postfailure fuel motion. Microscopic examination of the intact pins studied the nature and extent of fuel/cladding interaction, fuel melt fraction and mass distribution, and distribution of porosity. Eutectic penetration and failure of the cladding were also examined in the failed pins

Mechanisms of fuel-cladding chemical interaction: US interpretation

International Nuclear Information System (INIS)

Adamson, M.G.

1977-01-01

Proposed mechanisms of fuel-cladding chemical interaction (FCCI) in LMFBR fuel pins are reviewed and examined in terms of in-pile and out-of-pile data. From this examination several factors are identified which may govern the occurrence of localized deep intergranular penetrations of Type-316SS cladding. Using a plausible mechanistic hypothesis for FCCI, first steps have been taken towards developing a quantitative, physically-meaningful, mathematical method of predicting cladding wastage in operating fuel pins. Both kinetic and thermodynamic aspects of FCCI are considered in the development of this prediction method, together with a fuel chemistry model that describes the evolution of thermochemical conditions at the fuel-cladding gap. On the basis of results from recent fuel pin and laboratory tests a thermal transport mechanism has been proposed to explain the thermal gradient-induced migration of Fe, Cr, and Ni from cladding into the fuel. This mechanism involves chemical transport of the metallic cladding components (as tellurides) in liquid Cs-Te. (author)

Mechanisms of fuel-cladding chemical interaction: US interpretation

Energy Technology Data Exchange (ETDEWEB)

Adamson, M G [General Electric Company, Vallecitos Nuclear Center, Pleasanton, CA (United States)

1977-04-01

Proposed mechanisms of fuel-cladding chemical interaction (FCCI) in LMFBR fuel pins are reviewed and examined in terms of in-pile and out-of-pile data. From this examination several factors are identified which may govern the occurrence of localized deep intergranular penetrations of Type-316SS cladding. Using a plausible mechanistic hypothesis for FCCI, first steps have been taken towards developing a quantitative, physically-meaningful, mathematical method of predicting cladding wastage in operating fuel pins. Both kinetic and thermodynamic aspects of FCCI are considered in the development of this prediction method, together with a fuel chemistry model that describes the evolution of thermochemical conditions at the fuel-cladding gap. On the basis of results from recent fuel pin and laboratory tests a thermal transport mechanism has been proposed to explain the thermal gradient-induced migration of Fe, Cr, and Ni from cladding into the fuel. This mechanism involves chemical transport of the metallic cladding components (as tellurides) in liquid Cs-Te. (author)

IFPE/IFA-432, Fission Gas Release, Mechanical Interaction BWR Fuel Rods, Halden

International Nuclear Information System (INIS)

Turnbull, J.A.

1996-01-01

Description: It contains data from experiments that have been performed at the IFE/OECD Halden Reactor Project, available for use in fuel performance studies. It covers experiments on thermal performance, fission product release, clad properties and pellet clad mechanical interaction. It includes also experimental data relevant to high burn-up behaviour. IFA-432: Measurements of fuel temperature response, fission gas release and mechanical interaction on BWR-type fuel rods up to high burn-ups. The assembly featured several variations in rod design parameters, including fuel type, fuel/cladding gap size, fill gas composition (He and Xe) and fuel stability. It contained 6 BWR-type fuel rods with fuel centre thermocouples at two horizontal planes, rods were also equipped with pressure transducers and cladding extensometers. Only data from 6 rods are compiled here

Mechanical modelling of transient- to- failure SFR fuel cladding

Energy Technology Data Exchange (ETDEWEB)

Feria, F.; Herranz, L. E.

2014-07-01

The response of Sodium Fast Reactor (SFR) fuel rods to transient accident conditions is an important safety concern. During transients the cladding strain caused by the stress due to pellet cladding mechanical interaction (PCMI) can lead to failure. Due to the fact that SFR fuel rods are commonly clad with strengthened material made of stainless steel (SS), cladding is usually treated as an elastic-perfectly-plastic material. However, viscoplastic behaviour can contribute to mechanical strain at high temperature (> 1000 K). (Author)

Sodium dodecyl benzene sulphonate mediated tautomerism of Eriochrome Black-T: Effect of charge transfer interaction

Science.gov (United States)

Ghosh, Sumit

2010-11-01

Interaction between anionic surfactant, sodium dodecyl benzene sulphonate, (SDBS) and an anionic dye Eriochrome Black-T, (EBT) has been investigated by visible spectroscopy, conductometry, dynamic light scattering and zeta potential measurements. Spectral changes of EBT observed on addition of SDBS indicate formation of quinone-hydrazone tautomer at pH 7.0, whereas in absence of SDBS this change appears at pH ˜ 9.45. However, at pH 7.0 this change in tautomerism is not observed in presence of sodium dodecyl sulphate (SDS). Experimental results indicate presence of charge transfer interaction between less stable quinone-hydrazone tautomer of EBT and SDBS molecules, which is confirmed using Benesi-Hildebrand and Scott equations.

Thermal and chemical interaction of hot liquid sodium with limestone concrete in argon atmosphere

International Nuclear Information System (INIS)

Fakir, Charan Parida; Sanjay, Kumar Das; Anil, Kumar Sharma; Ramesh, S.S.; Somayajulu, P.A.; Malarvizhi, B.; Kasinathan, N.; Rajan, M.

2007-01-01

bound water in the post-test concrete blocks was studied. Preliminary analysis of test results revealed that when hot sodium at 500degC was discharged on cold concrete block, it monotonically cooled with low degree of interaction. But supply of external heat to sodium pool has triggered considerable reaction with or without occurrence of ETT phase under the given test conditions. Critical analysis of published data on large scale tests has indicated that ETT phase has emerged without in situ heating of sodium pool on limestone concrete even at initial sodium temperature of 420degC. (author)

Benchmark for Neutronic Analysis of Sodium-cooled Fast Reactor Cores with Various Fuel Types and Core Sizes

International Nuclear Information System (INIS)

Stauff, N.E.; Kim, T.K.; Taiwo, T.A.; Buiron, L.; Rimpault, G.; Brun, E.; Lee, Y.K.; Pataki, I.; Kereszturi, A.; Tota, A.; Parisi, C.; Fridman, E.; Guilliard, N.; Kugo, T.; Sugino, K.; Uematsu, M.M.; Ponomarev, A.; Messaoudi, N.; Lin Tan, R.; Kozlowski, T.; Bernnat, W.; Blanchet, D.; Brun, E.; Buiron, L.; Fridman, E.; Guilliard, N.; Kereszturi, A.; Kim, T.K.; Kozlowski, T.; Kugo, T.; Lee, Y.K.; Lin Tan, R.; Messaoudi, N.; Parisi, C.; Pataki, I.; Ponomarev, A.; Rimpault, G.; Stauff, N.E.; Sugino, K.; Taiwo, T.A.; Tota, A.; Uematsu, M.M.; Monti, S.; Yamaji, A.; Nakahara, Y.; Gulliford, J.

2016-01-01

One of the foremost Generation IV International Forum (GIF) objectives is to design nuclear reactor cores that can passively avoid damage of the reactor when control rods fail to scram in response to postulated accident initiators (e.g. inadvertent reactivity insertion or loss of coolant flow). The analysis of such unprotected transients depends primarily on the physical properties of the fuel and the reactivity feedback coefficients of the core. Within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS), the Sodium Fast Reactor core Feed-back and Transient response (SFR-FT) Task Force was proposed to evaluate core performance characteristics of several Generation IV Sodium-cooled Fast Reactor (SFR) concepts. A set of four numerical benchmark cases was initially developed with different core sizes and fuel types in order to perform neutronic characterisation, evaluation of the feedback coefficients and transient calculations. Two 'large' SFR core designs were proposed by CEA: those generate 3 600 MW(th) and employ oxide and carbide fuel technologies. Two 'medium' SFR core designs proposed by ANL complete the set. These medium SFR cores generate 1 000 MW(th) and employ oxide and metallic fuel technologies. The present report summarises the results obtained by the WPRS for the neutronic characterisation benchmark exercise proposed. The benchmark definition is detailed in Chapter 2. Eleven institutions contributed to this benchmark: Argonne National Laboratory (ANL), Commissariat a l'energie atomique et aux energies alternatives (CEA of Cadarache), Commissariat a l'energie atomique et aux energies alternatives (CEA of Saclay), Centre for Energy Research (CER-EK), Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), Helmholtz Zentrum Dresden Rossendorf (HZDR), Institute of Nuclear Technology and Energy Systems (IKE), Japan Atomic Energy Agency (JAEA), Karlsruhe Institute of Technology (KIT

Molten fuel-coolant interaction behaviours of various fast reactor fuels (Paper No. HMT-45-87)

International Nuclear Information System (INIS)

Doshi, J.B.

1987-01-01

A parametric computational model of molten fuel-coolant interaction (MFCI) including a particle size distribution is developed and employed to analyse behaviours of various possible reactor fuels, such as oxide, carbide and metal in MFCI scenario. It is observed that while higher thermal conductivity and lower specific heat of carbide compared to oxide is responsible for higher peak pressure and work done per unit mass, the trend is not observed in the metal fuel. The reason for this is the lower operation temperature and latent heat of metallic fuel. (author). 9 refs., 1 fig

Molten fuel studies at Winfrith

International Nuclear Information System (INIS)

Edwards, A.J.; Knowles, J.B.; Tattersall, R.B.

1988-01-01

This report describes the experimental facilities available for molten fuel studies at Winfrith. These include a large facility capable of testing components at full LMFBR subassembly scale and also a high pressure facility for experiments at pressures up to 25 MPa, covering the whole range of temperatures and pressures of interest for the PWR. If the hypothetical accident conditions initiating the release of molten fuel do not produce an explosive transfer of thermal energy on contact of molten fuel with the reactor coolant, then an intermediate rate of heat transfer over several hundred milliseconds may occur. Theoretical work is described which is being carried out to predict the resulting pressurisation and the degree of mechanical loading on the reactor structure. Finally the current programme of molten fuel studies and recent progress are reviewed, and future plans, which are chiefly focussed on the study of thermal interactions between molten fuel and sodium coolant for the LMFBR are outlined. (author)

Experience on sodium removal from various components

Energy Technology Data Exchange (ETDEWEB)

Kamei, M; Kanbe, M; Yagisawa, H; Sasaki, S; Kataoka, H; Fukada, T; Ishii, Y; Saito, R; Mimoto, Y [O-arai Engineering Centre, PNC, Ibaraki-ken, Tokio (Japan)

1978-08-01

Since 1970, OEC (O-arai Engineering Center) has been Investigating the following methods for removal of sodium from the components of sodium plants: steam cleaning for the 50 MW Steam Generator, secondary proto-type pump of 'JOYO' and Dummy fuel assembly of 'JOYO', alcohol cleaning for Sector Model of Intermediate Heat Exchanger (IHX) of 'JOYO', a sector model of Sodium-to-Air cooler of 'JOYO' and a proto-type isolation valve of 'JOYO' and cleaning by vacuumization at high temperature for Regenerative Heat Exchanger. This report describes the outline of the Sodium Disposal Facility and experience of sodium removal processing on the 50 MW Steam Generator, the crevices of the experimental sub-assemblies, the Fuel Handling Machine of 'MONJU' and the Regenerative Heat Exchanger of the Sodium Flow Test Facility. Through these experiences it was noted that, (1) Removal of sodium from crevices such as in bolted joints are very difficult. (2) Consideration is needed in the removal process where material damage might occur from the generation of hydro-oxides. (3) Some detection device to tell the completion of sodium removal as well as the end of reaction is required. (4) Requalification rules should be clarified. Efforts in this direction have been made in the case of a 'JOYO' prototype pump by reinstalling it after sodium removal five times. (author)

Experience on sodium removal from various components

International Nuclear Information System (INIS)

Kamei, M.; Kanbe, M.; Yagisawa, H.; Sasaki, S.; Kataoka, H.; Fukada, T.; Ishii, Y.; Saito, R.; Mimoto, Y.

1978-01-01

Since 1970, OEC (O-arai Engineering Center) has been Investigating the following methods for removal of sodium from the components of sodium plants: steam cleaning for the 50 MW Steam Generator, secondary proto-type pump of 'JOYO' and Dummy fuel assembly of 'JOYO', alcohol cleaning for Sector Model of Intermediate Heat Exchanger (IHX) of 'JOYO', a sector model of Sodium-to-Air cooler of 'JOYO' and a proto-type isolation valve of 'JOYO' and cleaning by vacuumization at high temperature for Regenerative Heat Exchanger. This report describes the outline of the Sodium Disposal Facility and experience of sodium removal processing on the 50 MW Steam Generator, the crevices of the experimental sub-assemblies, the Fuel Handling Machine of 'MONJU' and the Regenerative Heat Exchanger of the Sodium Flow Test Facility. Through these experiences it was noted that, (1) Removal of sodium from crevices such as in bolted joints are very difficult. (2) Consideration is needed in the removal process where material damage might occur from the generation of hydro-oxides. (3) Some detection device to tell the completion of sodium removal as well as the end of reaction is required. (4) Requalification rules should be clarified. Efforts in this direction have been made in the case of a 'JOYO' prototype pump by reinstalling it after sodium removal five times. (author)

Experience on sodium removal from various components

International Nuclear Information System (INIS)

Kamei, M.; Kanbe, M.; Yagisawa, H.; Sasaki, S.; Kataoka, H.

1978-02-01

Since 1970, OEC (O-arai Engineering Center) has been investigating the following methods for removal of sodium from the components of sodium plants: steam cleaning for the 50 MW Steam Generator, secondary proto-type pump of ''JOYO'' and Dummy fuel assembly of ''JOYO'', alcohol cleaning for Sector Model of Intermediate Heat Exchanger (IHX) of ''JOYO'', a sector model of Sodium-to-Air cooler of ''JOYO'' and a proto-type Isolation valve of ''JOYO'' and cleaning by vacuumization at high temperature for Regenerative Heat Exchanger. This report describes the outline of the Sodium Disposal Facility and experience of sodium removal processing on the 50 MW Steam Generator, the crevices of the experimental subassemblies, the Fuel Handling Machine of ''MONJU'' and the Regenerative Heat Exchanger of the Sodium Flow Test Facility. Through these experiences it was noted that, (1) Removal of Sodium from crevices such as in bolted joints are very difficult. (2) Consideration is needed in the removal process where material damage might occur from the generation of hydro-oxides. (3) Some detection device to tell the completion of sodium removal as well as the end of reaction is required. (4) Requalification rules should be clarified. Efforts in this direction have been made in the case of a ''JOYO'' prototype pump by reinstalling it after sodium removal five times. (author)

Thermal expansion of the nuclear fuel-sodium reaction product Na3(U0.84(2),Na0.16(2))O4 - Structural mechanism and comparison with related sodium-metal ternary oxides

Science.gov (United States)

Illy, Marie-Claire; Smith, Anna L.; Wallez, Gilles; Raison, Philippe E.; Caciuffo, Roberto; Konings, Rudy J. M.

2017-07-01

Na3.16(2)UV,VI0.84(2)O4 is obtained from the reaction of sodium with uranium dioxide under oxygen potential conditions typical of a sodium-cooled fast nuclear reactor. In the event of a breach of the steel cladding, it would be the dominant reaction product forming at the rim of the mixed (U,Pu)O2 fuel pellets. High-temperature X-ray diffraction measurements show that a distortion of the uranium environment in Na3.16(2)UV,VI0.84(2)O4 results in a strongly anisotropic thermal expansion. A comparison with several related sodium metallates Nan-2Mn+On-1 - including Na3SbO4 and Na3TaO4, whose crystal structures are reported for the first time - has allowed us to assess the role played in the lattice expansion by the Mn+ cation radius and the Na/M ratio. On this basis, the thermomechanical behavior of the title compound is discussed, along with those of several related double oxides of sodium and actinide elements, surrogate elements, or fission products.

Automated and interactive fuel management tools: Past, present and future

International Nuclear Information System (INIS)

Cook, A.G.; Casadei, A.L.

1986-01-01

The past, present and future status of automated and interactive fuel management tools are reviewed. Issues such as who are the customers for these products and what are their needs are addressed. The nature of the fuel management problem is reviewed. The Westinghouse fuel management tools and methods are presented as an example of how the technology has evolved

Contribution to the study of the thermal interaction between uranium oxide and sodium

International Nuclear Information System (INIS)

Newman, W.H.

1982-01-01

A description is given of the experimental results of the fuel-coolant interactions carried out in the CORECT II device at the Grenoble Nuclear Study Centre. A description is then given of a theoretical model of interaction which comprises three points: first a study of the coolant, that is to say of its hydrodynamic and thermodynamic behaviour, then the study of the fuel, namely the phenomena of fragmentation and heat transfer between the fuel and the coolant, and last the treatment of heat leaks to the structures. A study follows on the effect of the various parameters on the theoretical model as well as the effects of the assumptions on the fragmentation, transfer and losses of heat. Last, the interpretations of a few experiments carried out with two models of fragmentation are described. A discussion of these interpretations enables some generalizations to be made on the nature of the thermal interaction [fr

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

Enthalpic interactions of N-glycylglycine with xylitol in aqueous sodium chloride and potassium chloride solutions at T = 298.15 K

International Nuclear Information System (INIS)

Liu Min; Wang Lili; Zhu Lanying; Li Hui; Sun Dezhi; Di Youying; Li Linwei

2010-01-01

The mixing enthalpies of N-glycylglycine with xylitol and their respective enthalpies of dilution in aqueous sodium chloride and potassium chloride solutions have been determined by using flow-mix isothermal microcalorimetry at the temperature of 298.15 K. These experimental results have been used to determine the heterotactic enthalpic interaction coefficients (h xy , h xxy , and h xyy ) according to the McMillan-Mayer theory. It has been found that the heterotactic enthalpic pairwise interaction coefficients h xy between N-glycylglycine and xylitol in aqueous sodium chloride and potassium chloride solutions are negative and become less negative with an increase in the molality of sodium chloride or potassium chloride. The results are discussed in terms of solute-solute and solute-solvent interactions.

Enthalpic interactions of N-glycylglycine with xylitol in aqueous sodium chloride and potassium chloride solutions at T = 298.15 K

Energy Technology Data Exchange (ETDEWEB)

Liu Min, E-mail: panpanliumin@163.co [College of Chemistry and Chemical Engineering, Liao Cheng University, Liaocheng, Shandong 252059 (China); Wang Lili [College of Chemistry and Chemical Engineering, Liao Cheng University, Liaocheng, Shandong 252059 (China); Zhu Lanying [College of Life Science and Bioengineering, Liao Cheng University, Liaocheng, Shandong 252059 (China); Li Hui; Sun Dezhi; Di Youying; Li Linwei [College of Chemistry and Chemical Engineering, Liao Cheng University, Liaocheng, Shandong 252059 (China)

2010-07-15

The mixing enthalpies of N-glycylglycine with xylitol and their respective enthalpies of dilution in aqueous sodium chloride and potassium chloride solutions have been determined by using flow-mix isothermal microcalorimetry at the temperature of 298.15 K. These experimental results have been used to determine the heterotactic enthalpic interaction coefficients (h{sub xy}, h{sub xxy}, and h{sub xyy}) according to the McMillan-Mayer theory. It has been found that the heterotactic enthalpic pairwise interaction coefficients h{sub xy} between N-glycylglycine and xylitol in aqueous sodium chloride and potassium chloride solutions are negative and become less negative with an increase in the molality of sodium chloride or potassium chloride. The results are discussed in terms of solute-solute and solute-solvent interactions.

Fundamental aspects of nuclear reactor fuel elements

Energy Technology Data Exchange (ETDEWEB)

Olander, D.R.

1976-01-01

The book presented is designed to function both as a text for first-year graduate courses in nuclear materials and as a reference for workers involved in the materials design and performance aspects of nuclear power plants. The contents are arranged under the following chapter headings: statistical thermodynamics, thermal properties of solids, crystal structures, cohesive energy of solids, chemical equilibrium, point defects in solids, diffusion in solids, dislocations and grain boundaries, equation of state of UO/sub 2/, fuel element thermal performance, fuel chemistry, behavior of solid fission products in oxide fuel elements, swelling due to fission gases, pore migration and fuel restructuring kinetics, fission gas release, mechanical properties of UO/sub 2/, radiation damage, radiation effects in metals, interaction of sodium and stainless steel, modeling of the structural behavior of fuel elements and assemblies. (DG)

Sodium leak on the fuel storage drum of Superphenix

International Nuclear Information System (INIS)

Acket, C.; Marcon, J.P.; Michoux, H.

1988-01-01

SUPERPHENIX the world's largest fast breeder prototype reached its nominal power 1200 MWe in December 1986. In March 1987 a sodium leakage was detected on the 'barillet'. This is a large double walled cylindrical sodium tank (14 m high, 9 m in diameter) made of ferritic steel and filled with 700 tonnes of sodium at a temperature of 200 0 C. Located close to the primary pool it is used in the refuelling process of the plant. The leakage of sodium through the main vessel was confined in the guard vessel. This paper presents the different stages of the operations undertaken: to guarantee and improve the safety until the complete drainage of sodium; to drain the vessels and localize the leakage; to characterise the defect and the presence or not of other similar or different defects; to define the next step between several solutions including the local repair and complete reconstruction. (author)

Development of sodium technology

International Nuclear Information System (INIS)

Hwang, Sung Tai; Nam, H. Y.; Choi, Y. D.

2000-05-01

The objective of present study is to produce the experimental data for development and verification of computer codes for development of LMR and to develop the preliminary technologies for the future large scale verification experiments. A MHD experimental test loop has been constructed for the quantitative analysis of the effect of magnetic field on the sodium flow and experiments are carried out for three EM pumps. The previous pressure drop correlations are evaluated using the experimental data obtained from the pressure drop experiment in a 19-pin fuel assembly with wire spacer. An dimensionless variable is proposed to describe the amplitude and frequency of the fluctuation of free surface using the experimental data obtained from free surface experimental apparatus and an empirical correlation is developed using this dimensionless variable. An experimental test loop is constructed to measure the flow characteristics in IHX shell side and the local pressure drop in fuel assembly, and to test the vibration behaviour of fuel pins due to flow induced vibration. The sodium two-phase flow measuring technique using the electromagnetic flowmeter is developed and the sodium differential pressure drop measuring technique using the method of direct contact of sodium and oil is established. The work on the analysis of sodium fire characteristics and produce data for vlidation of computer code is performed. Perfect reopen time of self plugged leak path was observed to be about 130 minutes after water leak initiation. Reopen shape of a specimen appeared to be double layer of circular type, and reopen size of this specimen surface was about 2mm diameter on sodium side. In small water leakage experiments, the following correlation equation about the reopen time between sodium temperature and initial leak rate was obtained, τ c = δ·g -0.83 ·10 (3570/T Na -3.34) , in 400-500 deg C of liquid sodium atmosphere. The characteristics of pressure propagation and gas flow, and

Development of sodium technology

Energy Technology Data Exchange (ETDEWEB)

Hwang, Sung Tai; Nam, H Y; Choi, Y D [and others

2000-05-01

The objective of present study is to produce the experimental data for development and verification of computer codes for development of LMR and to develop the preliminary technologies for the future large scale verification experiments. A MHD experimental test loop has been constructed for the quantitative analysis of the effect of magnetic field on the sodium flow and experiments are carried out for three EM pumps. The previous pressure drop correlations are evaluated using the experimental data obtained from the pressure drop experiment in a 19-pin fuel assembly with wire spacer. An dimensionless variable is proposed to describe the amplitude and frequency of the fluctuation of free surface using the experimental data obtained from free surface experimental apparatus and an empirical correlation is developed using this dimensionless variable. An experimental test loop is constructed to measure the flow characteristics in IHX shell side and the local pressure drop in fuel assembly, and to test the vibration behaviour of fuel pins due to flow induced vibration. The sodium two-phase flow measuring technique using the electromagnetic flowmeter is developed and the sodium differential pressure drop measuring technique using the method of direct contact of sodium and oil is established. The work on the analysis of sodium fire characteristics and produce data for vlidation of computer code is performed. Perfect reopen time of self plugged leak path was observed to be about 130 minutes after water leak initiation. Reopen shape of a specimen appeared to be double layer of circular type, and reopen size of this specimen surface was about 2mm diameter on sodium side. In small water leakage experiments, the following correlation equation about the reopen time between sodium temperature and initial leak rate was obtained, {tau}{sub c} = {delta}{center_dot}g{sup -0.83}{center_dot}10{sup (3570/T{sub Na}-3.34)}, in 400-500 deg C of liquid sodium atmosphere. The characteristics

Physical model and calculation code for fuel coolant interactions

International Nuclear Information System (INIS)

Goldammer, H.; Kottowski, H.

1976-01-01

A physical model is proposed to describe fuel coolant interactions in shock-tube geometry. According to the experimental results, an interaction model which divides each cycle into three phases is proposed. The first phase is the fuel-coolant-contact, the second one is the ejection and recently of the coolant, and the third phase is the impact and fragmentation. Physical background of these phases are illustrated in the first part of this paper. Mathematical expressions of the model are exposed in the second part. A principal feature of the computational method is the consistent application of the fourier-equation throughout the whole interaction process. The results of some calculations, performed for different conditions are compiled in attached figures. (Aoki, K.)

The study of interaction of lanthanum-, cerium- and neodymium chlorides with sodium borohydride in pyridine- and tetrahydrofuran medium

International Nuclear Information System (INIS)

Mirsaidov, U.; Rotenberg, T.G.; Dymova, T.N.

1976-01-01

Bis-tetrahydrofurans of lanthanum and neodymium borohydrides and bis-pyridinates of lanthanum, cerium and neodymium borohydrides were obtained by interacting sodium borohydride with lanthanum-, cerium and neodymium chlorides in pyridine and tetrahydrofuran media. All operations involving reagent combination, sampling and phase separation are performed in inert atmosphere using argonvacuum equipment. The reaction in pyridine was virtually instantaneous and accompanied by flocculanet precipitation. The interaction of lanthanum chloride and neodymium chloride with sodium borohydride in tetrahydrofuran (THF) was a slow (23-30 hr) heterophase process. The interaction rate was affected by size reduction of the intial substances, temperature, reagent proportion and mixing rate. The reaction time was twice reduced with boiling tetrahydrofuran

NaBH4 (sodium borohydride) hydrogen generator with a volume-exchange fuel tank for small unmanned aerial vehicles powered by a PEM (proton exchange membrane) fuel cell

International Nuclear Information System (INIS)

Kim, Taegyu

2014-01-01

A proton exchange membrane fuel cell system integrated with a NaBH 4 (sodium borohydride) hydrogen generator was developed for small UAVs (unmanned aerial vehicles). The hydrogen generator was composed of a catalytic reactor, liquid pump and volume-exchange fuel tank, where the fuel and spent fuel exchange the volume within a single fuel tank. Co–B catalyst supported on a porous ceramic material was used to generate hydrogen from the NaBH 4 solution. Considering the power consumption according to the mission profile of a UAV, the power output of the fuel cell and auxiliary battery was distributed passively as an electrical load. A blended wing-body was selected considering the fuel efficiency and carrying capability of fuel cell components. First, the fuel cell stack and hydrogen generator were evaluated under the operating conditions, and integrated into the airframe. The ground test of the complete fuel cell UAV was performed under a range of load conditions. Finally, the fuel cell powered flight test was made for 1 h. The volume-exchange fuel tank minimized the fuel sloshing and the change in center of gravity due to fuel consumption during the flight, so that much stable operation of the fuel cell system was validated at different flight modes. - Highlights: • PEMFC system with a NaBH 4 hydrogen source was developed for small UAVs. • Volume-exchange fuel tank was used to reduce the size of the fuel cell system. • Passive power management was used for a stable power output during the flight. • BWB UAV was selected by taking the fuel cell integration into consideration. • Stable operation of the fuel cell system was verified from the flight test

Analysis of large scale UO2 Na interactions performed in Europe

International Nuclear Information System (INIS)

Berthoud, G.; Jacobs, H.; Knowles, B.

1994-01-01

Analysis of the European out of pile Fuel Sodium Interaction Experiments involving kilogram masses of molten oxide is reported i.e. CORECT 2 (CEA), SUS and MFTF-B (AEA), THINA (KfK). Then common conclusions are drawn. (author)

Pellet-clad interaction observations in boiling water reactor fuel elements

International Nuclear Information System (INIS)

Sahoo, K.C.; Bahl, J.K.; Sivaramakrishnan, K.S.; Roy, P.R.

1981-01-01

Under a programme to assess the performance of fuel elements of Tarapur Atomic Power Station, post-irradiation examination has been carried out on 18 fuel elements in the first phase. Pellet-clad mechanical interaction behaviour in 14 elements with varying burnup and irradiation history has been studied using eddy current testing technique. The data has been analysed to evaluate the role of pellet-clad mechanical interaction in PCI/SCC failure in power reactor operating conditions. (author)

Status of sodium removal and component decontamination technology in the SNR programme

Energy Technology Data Exchange (ETDEWEB)

Haubold, W [INTERATOM GmbH, Bergisch Gladbach (Germany); Smit, C Ch [MT-TNO Dept. 50-MW Sodium Component Test Facility, Hengelo (Netherlands); Stade, K Ch [Kernkraftwerk-Betriebsgesellschaft mbH, Eggenstein-Leopoldshafen (Germany)

1978-08-01

This paper summarizes the experience with sodium removal and component decontamination processes within the framework of the SNR project since the IAEA Specialists Meeting on 'Decontamination of Plant Components from Sodium and Radioactivity' at Dounreay, April 9-12, 1973. The moist nitrogen process has been successfully applied to remove sodium from all 66 fuel elements of the KNK I core. Progress has been obtained in removing sodium from fuel elements and large components by vacuum distillation. Areas where future development is required are identified. (author)

Status of sodium removal and component decontamination technology in the SNR programme

International Nuclear Information System (INIS)

Haubold, W.; Smit, C.Ch.; Stade, K.Ch.

1978-01-01

This paper summarizes the experience with sodium removal and component decontamination processes within the framework of the SNR project since the IAEA Specialists Meeting on 'Decontamination of Plant Components from Sodium and Radioactivity' at Dounreay, April 9-12, 1973. The moist nitrogen process has been successfully applied to remove sodium from all 66 fuel elements of the KNK I core. Progress has been obtained in removing sodium from fuel elements and large components by vacuum distillation. Areas where future development is required are identified. (author)

Parabrachial and hypothalamic interaction in sodium appetite

Science.gov (United States)

Dayawansa, S.; Peckins, S.; Ruch, S.

2011-01-01

Rats with bilateral lesions of the lateral hypothalamus (LH) fail to exhibit sodium appetite. Lesions of the parabrachial nuclei (PBN) also block salt appetite. The PBN projection to the LH is largely ipsilateral. If these deficits are functionally dependent, damaging the PBN on one side and the LH on the other should also block Na appetite. First, bilateral ibotenic acid lesions of the LH were needed because the electrolytic damage used previously destroyed both cells and axons. The ibotenic LH lesions produced substantial weight loss and eliminated Na appetite. Controls with ipsilateral PBN and LH lesions gained weight and displayed robust sodium appetite. The rats with asymmetric PBN-LH lesions also gained weight, but after sodium depletion consistently failed to increase intake of 0.5 M NaCl. These results dissociate loss of sodium appetite from the classic weight loss after LH damage and prove that Na appetite requires communication between neurons in the LH and the PBN. PMID:21270347

Laboratory-scale sodium-carbonate aggregate concrete interactions

International Nuclear Information System (INIS)

Westrich, H.R.; Stockman, H.W.; Suo-Anttila, A.

1983-09-01

A series of laboratory-scale experiments was made at 600 0 C to identify the important heat-producing chemical reactions between sodium and carbonate aggregate concretes. Reactions between sodium and carbonate aggregate were found to be responsible for the bulk of heat production in sodium-concrete tests. Exothermic reactions were initiated at 580+-30 0 C for limestone and dolostone aggregates as well as for hydrated limestone concrete, and at 540+-10 0 C for dehydrated limestone concrete, but were ill-defined for dolostone concrete. Major reaction products included CaO, MgO, Na 2 CO 3 , Na 2 O, NaOH, and elemental carbon. Sodium hydroxide, which forms when water is released from cement phases, causes slow erosion of the concrete with little heat production. The time-temperature profiles of these experiments have been modeled with a simplified version of the SLAM computer code, which has allowed derivation of chemical reaction rate coefficients

Comparison of KANEXT and SERPENT for fuel depletion calculations of a sodium fast reactor

International Nuclear Information System (INIS)

Lopez-Solis, R.C.; Francois, J.L.; Becker, M.; Sanchez-Espinoza, V.H.

2014-01-01

As most of Generation-IV systems are in development, efficient and reliable computational tools are needed to obtain accurate results in reasonably computer time. In this study, KANEXT code system is presented and validated against the well-known Monte Carlo SERPENT code, for fuel depletion calculations of a sodium fast reactor (SFR). The KArlsruhe Neutronic EXtended Tool (KANEXT) is a modular code system for deterministic reactor calculations, consisting of one kernel and several modules. Results obtained with KANEXT for the SFR core are in good agreement with the ones of SERPENT, e.g. the neutron multiplication factor and the isotopes evolution with burn-up. (author)

Fuel-cladding chemical interaction correlation for mixed-oxide fuel pins

International Nuclear Information System (INIS)

Lawrence, L.A.

1986-10-01

A revised wastage correlation was developed for FCCI with fabrication and operating parameters. The expansion of the data base to 305 data sets provided sufficient data to employ normal statistical techniques for calculation of confidence levels without unduly penalizing predictions. The correlation based on 316 SS cladding also adequately accounts for limited measured depths of interaction for fuel pins with D9 and HTq cladding

Passive safety features of low sodium void worth metal fueled cores in a bottom supported reactor vessel

International Nuclear Information System (INIS)

Chang, Y.I.; Marchaterre, J.F.; Wade, D.C.; Wigeland, R.A.; Kumaoka, Yoshio; Suzuki, Masao; Endo, Hiroshi; Nakagawa, Hiroshi

1991-01-01

A study has been performed on the passive safety features of low-sodium-void-worth metallic-fueled reactors with emphasis on using a bottom-supported reactor vessel design. The reactor core designs included self-sufficient types as well as actinide burners. The analyses covered the reactor response to the unprotected, i.e. unscrammed, transient overpower accident and the loss-of-flow accident. Results are given demonstrating the safety margins that were attained. 4 refs., 4 figs., 2 tabs

Nuclear fuel handling apparatus

International Nuclear Information System (INIS)

Andrea, C.; Dupen, C.F.G.; Noyes, R.C.

1977-01-01

A fuel handling machine for a liquid metal cooled nuclear reactor in which a retractable handling tube and gripper are lowered into the reactor to withdraw a spent fuel assembly into the handling tube. The handling tube containing the fuel assembly immersed in liquid sodium is then withdrawn completely from the reactor into the outer barrel of the handling machine. The machine is then used to transport the spent fuel assembly directly to a remotely located decay tank. The fuel handling machine includes a decay heat removal system which continuously removes heat from the interior of the handling tube and which is capable of operating at its full cooling capacity at all times. The handling tube is supported in the machine from an articulated joint which enables it to readily align itself with the correct position in the core. An emergency sodium supply is carried directly by the machine to provide make up in the event of a loss of sodium from the handling tube during transport to the decay tank. 5 claims, 32 drawing figures

Role of fuel bubble phenomenology in assessment of LMFBR source term

International Nuclear Information System (INIS)

Cho, D.H.; Condiff, D.W.; Chan, S.H.

1985-01-01

Phenomenological aspects of a fuel vapor bubble formed in the sodium pool in a hypothetical severe accident are considered. The potential for fuel bubble collapse in the sodium pool is analyzed. It appears that for a wide range of hypothetical LMFBR accidents involving core vaporization, the fuel vapor bubble would likely be quenched and collapse prior to migration to the cover gas region. Such rapid quenching is due mainly to radiative heat transfer from the fuel bubble, coupled with the inherent capability of the sodium pool (large subcooling and high thermal conductivity) to dissipate thermal energy. Major uncertainty in the analysis concerns fuel vapor condensation phenomena at the sodium interface and its effect on the sodium surface radiation absorptivity. This is discussed in detail

Studies of Lanthanide Transport in Metallic Fuel

Energy Technology Data Exchange (ETDEWEB)

Zhang, Jinsuo; Taylor, Christopher

2018-04-02

Metallic nuclear fuels were tested in fast reactor programs and performed well. However, metallic fuels have shown the phenomenon of FCCI that are due to deleterious reactions between lanthanide fission products and cladding material. As the burnup is increased, lanthanide fission products that contact with the cladding could react with cladding constituents such as iron and chrome. These reactions produce higher-melting intermetallic compounds and low-melting alloys, and weaken the mechanical integrity. The lanthanide interaction with clad in metallic fuels is recognized as a long-term, high-burnup cause of the clad failures. Therefore, one of the key concerns of using metallic fuels is the redistribution of lanthanide fission products and migration to the fuel surface. It is believed that lanthanide migration is in part due to the thermal gradient between the center and the fuel-cladding interface, but also largely in part due to the low solubility of lanthanides within the uranium-based metal fuel. PIE of EBR-II fuels shows that lanthanides precipitate directly and do not dissolve to an appreciable extent in the fuel matrix. Based on the PIE data from EBR-II, a recent study recommended a so-called “liquid-like” transport mechanism for lanthanides and certain other species. The liquid-like transport model readily accounts for redistribution of Ln, noble metal fission products, and cladding components in the fuel matrix. According to the novel mechanism, fission products can transport as solutes in liquid metals, such as liquid cesium or liquid cesium–sodium, and on pore surfaces and fracture surfaces for metals near their melting temperatures. Transport in such solutions is expected to be much more rapid than solid-state diffusion. The mechanism could explain the Ln migration to the fuel slug peripheral surface and their deposition with a sludge-like form. Lanthanides have high solubility in liquid cesium but have low solubility in liquid sodium. As a

A Development of Ethanol/Percarbonate Membraneless Fuel Cell

Directory of Open Access Journals (Sweden)

M. Priya

2014-01-01

Full Text Available The electrocatalytic oxidation of ethanol on membraneless sodium percarbonate fuel cell using platinum electrodes in alkaline-acidic media is investigated. In this cell, ethanol is used as the fuel and sodium percarbonate is used as an oxidant for the first time in an alkaline-acidic media. Sodium percarbonate generates hydrogen peroxide in aqueous medium. At room temperature, the laminar-flow-based microfluidic membraneless fuel cell can reach a maximum power density of 18.96 mW cm−2 with a fuel mixture flow rate of 0.3 mL min−2. The developed fuel cell features no proton exchange membrane. The simple planar structured membraneless ethanol fuel cell presents with high design flexibility and enables easy integration of the microscale fuel cell into actual microfluidic systems and portable power applications.

Technical committee meeting on fuel and cladding interaction. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1977-04-01

Experiments and experiences concerning fuel-cladding interaction in thermal and fast neutron flux burnup are dealt with. A number of results from in-pile and out-of pile experiments with different fuel pins with cladding made of different stainless steels showed the importance of corrosion process, dependent on the burnup, core temperature, metal-oxide ratio, and other steady state parameters in the core of fast reactors (most frequently LMFBRs). This is of importance for fuel pins design and fabrication. Mixed oxide fuel is treated in many cases.

Technical committee meeting on fuel and cladding interaction. Summary report

International Nuclear Information System (INIS)

1977-04-01

Experiments and experiences concerning fuel-cladding interaction in thermal and fast neutron flux burnup are dealt with. A number of results from in-pile and out-of pile experiments with different fuel pins with cladding made of different stainless steels showed the importance of corrosion process, dependent on the burnup, core temperature, metal-oxide ratio, and other steady state parameters in the core of fast reactors (most frequently LMFBRs). This is of importance for fuel pins design and fabrication. Mixed oxide fuel is treated in many cases

Solution interactions of diclofenac sodium and meclofenamic acid sodium with hydroxypropyl methylcellulose (HPMC).

Science.gov (United States)

Pygall, Samuel R; Griffiths, Peter C; Wolf, Bettina; Timmins, Peter; Melia, Colin D

2011-02-28

Many pharmaceutical agents require formulation in order to facilitate their efficacious delivery. However, the interaction between the active species and the formulation additives has the potential to significantly influence the pharmocokinetics of the active. In this study, the solution interactions between hydroxypropyl methylcellulose (HPMC) with two non-steroidal anti-inflammatories - the sodium salts of diclofenac and meclofenamate - were investigated using tensiometric, rheological, NMR, neutron scattering and turbidimetric techniques. The two drugs behaved very differently-meclofenamate addition to HPMC solutions led to substantial increases in viscosity, a depression of the gel point and a marked reduction in the self-diffusion coefficient of the drug, whereas diclofenac did not induce these changes. Collectively, these observations are evidence of meclofenamate forming self-assembled aggregates on the HPMC, a phenomenon not observed with diclofenac Na. Any process that leads to aggregation on a nonionic polymer will not be strongly favoured when the aggregating species is charged. Thus, it is hypothesised that the distinction between the two drugs arises as a consequence of the tautomerism present in meclofenamate that builds electron density on the carbonyl group that is further stabilised by hydrogen bonding to the HPMC. This mechanism is absent in the diclofenac case and thus no interaction is observed. These studies propose for the first time a molecular basis for the observed often-unexpected, concentration-dependant changes in HPMC solution properties when co-formulated with different NSAIDs, and underline the importance of characterising such fundamental interactions that have the potential to influence drug release in solid HPMC-based dosage forms. Copyright © 2010 Elsevier B.V. All rights reserved.

Modelling of pellet-cladding interaction for PWRs reactors fuel rods

International Nuclear Information System (INIS)

Esteves, A.M.

1991-01-01

The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyzes the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. Linear and non-linear material behaviors are allowed. Elastic, plastic and creep behaviors are considered for the cladding materials. The modelling is applied to Angra-II fuel rod design. The results are analyzed and compared. (author)

Interactive color graphics system for BWR fuel management

International Nuclear Information System (INIS)

Reese, A.P.

1986-01-01

An interactive color graphics system has been developed by the General Electric Company for fuel management engineers. The system consists of a Hewlett-Packard color graphics workstation in communication with a host mainframe. The system aids in such tasks as fuel cycle optimization, refueling bundle shuffle and control blade sequence design. Since being installed in 1983 turn-around time for a typical cycle reload and control blade pattern design has been reduced by a factor of four

Fuel Performance Modeling of U-Mo Dispersion Fuel: The thermal conductivity of the interaction layers of the irradiated U-Mo dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

Mistarhi, Qusai M.; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

2016-05-15

U-Mo/Al dispersion fuel performed well at a low burn-up. However, higher burn-up and higher fission rate irradiation testing showed enhanced fuel meat swelling which was caused by high interaction layer growth and pore formation. The performance of the dispersion type fuel in the irradiation and un-irradiation environment is very important. During the fabrication of the dispersion type fuel an Interaction Layer (IL) is formed due to the inter-diffusion between the U-Mo fuel particles and the Al matrix which is an intermetallic compound (U,Mo)Alx. During irradiation, the IL becomes amorphous causing a further decrease in the thermal conductivity and an increase in the centerline temperature of the fuel meat. Several analytical models and numerical methods were developed to study the performance of the unirradiated U-Mo/Al dispersion fuel. Two analytical models were developed to study the performance of the irradiated U-Mo/Al dispersion fuel. In these models, the thermal conductivity of the IL was assumed to be constant. The properties of the irradiated U-Mo dispersion fuel have been investigated recently by Huber et al. The objective of this study is to develop a correlation for IL thermal conductivity during irradiation as a function of the temperature and fission density from the experimentally measured thermal conductivity of the irradiated U-Mo/Al dispersion fuel. The thermal conductivity of IL during irradiation was calculated from the experimentally measured data and a correlation was developed from the thermal conductivity of IL as a function of T and fission density.

Upper limits to americium concentration in large sized sodium-cooled fast reactors loaded with metallic fuel

International Nuclear Information System (INIS)

Zhang, Youpeng; Wallenius, Janne

2014-01-01

Highlights: • The americium transmutation capability of Integral Fast Reactor was investigated. • The impact from americium introduction was parameterized by applying SERPENT Monte Carlo calculations. • Higher americium content in metallic fuel leads to a power penalty, preserving consistent safety margins. - Abstract: Transient analysis of a large sized sodium-cooled reactor loaded with metallic fuel modified by different fractions of americium have been performed. Unprotected loss-of-offsite power, unprotected loss-of-flow and unprotected transient-over-power accidents were simulated with the SAS4A/SASSYS code based on the geometrical model of an IFR with power rating of 2500 MW th , using safety parameters obtained with the SERPENT Monte Carlo code. The Ti-modified austenitic D9 steel, having higher creep rupture strength, was considered as the cladding and structural material apart from the ferritic/martensitic HT9 steel. For the reference case of U–12Pu–1Am–10Zr fuel at EOEC, the margin to fuel melt during a design basis condition UTOP is about 50 K for a maximum linear rating of 30 kW/m. In order to maintain a margin of 50 K to fuel failure, the linear power rating has to be reduced by ∼3% and 6% for 2 wt.% and 3 wt.% Am introduction into the fuel respectively. Hence, an Am concentration of 2–3 wt.% in the fuel would lead to a power penalty of 3–6%, permitting a consumption rate of 3.0–5.1 kg Am/TW h th . This consumption rate is significantly higher than the one previously obtained for oxide fuelled SFRs

A Review of Fragmentation Models Relative to Molten UO2 Breakup when Quenched in Sodium Coolant

International Nuclear Information System (INIS)

Cronenberg, A.W.; Grolmes, M.A.

1976-01-01

An important aspect of the fuel-coolant interaction problem relative to liquid metal fast breeder reactor (LMFBR) safety analysis is the fragmentation of molten oxide fuel during contact with liquid sodium coolant. A proper description of the kinetics of such an event requires an understanding of the breakup process and an estimate of the size and dispersion of such finely divided fuel in coolant. In recent years, considerable interest has centered on the problem of determining the nature of such fragmentation. In this paper, both analytic and experimental studies pertaining to such breakup are reviewed in light of recent developments in the understanding of heat transfer and solidification phenomena during quenching of UO 2 in sodium. A more extensive review of this subject can be found in Ref. 1. In conclusion: As discussed, a number of models have been proposed in an attempt to understand the nature of the UO 2 fragmentation process. The four principle mechanisms considered likely to cause such fragmentation (impact forces, boiling, violent gas release, and shell solidification) have been developed to the point where comparative analysis is possible. In addition, recent developments in the understanding of the physics of oxide fuel behavior in sodium coolant (boiling regime criteria, vapor nucleation theories, and prediction of solidification kinetics enable us to asses whether or not the various model assumptions are realistic. In view of this knowledge the following conclusions are made. For the case of hydrodynamic influence on fragmentation, it can be said that although the disruptive forces of impact and viscous drag may contribute to breakup, their effects are not controlling with respect to high temperature materials, including UO 2 -sodium. With respect to the vapor bubble growth and collapse mechanism it was shown that for sodium quenching, where coolant contact may, be expected (as opposed to water), the thermodynamic work potential of the bubble is

Calculation of DND-signals in case of fuel pin failures in KNK II with the computer code FICTION III

International Nuclear Information System (INIS)

Schmuck, I.

1990-11-01

In KNK II two delayed neutron detectors are installed for quick detection of fuel subassembly cladding failures. They record the release of the precursors of the emitters of delayed neutrons into the sodium. The computer code FICTION III calculates the expected delayed neutron signals for certain fuel pin failures, where the user has to set the boundary conditions interactively. In view of FICTION II the advancement of FICTION III consists of the following items: application of the data sets of 105 isotopes, distinction of thermal and fast neutron induced fission, partitioning of the sodium flow into two circuits, consideration of the specific fission rates in 10 fuel pin sections, elaboration of the user's interaction possibilities for input/ output. The capability of FICTION III is shown by means of two applications (UNi-test pin on position 100 and the third KNK fuel subassembly cladding failure). Object of further evaluations will be among other things the analysis of increased delayed neutron signals in regard to the fault location and dimension

Coupling analysis of deformation and thermal-hydraulics in a FBR fuel pin bundle using BAMBOO and ASFRE-IV Codes

International Nuclear Information System (INIS)

Ito, Masahiro; Imai, Yasutomo; Uwaba, Tomoyuki; Ohshima, Hiroyuki

2004-03-01

The bundle-duct interaction may occur in sodium cooled wire-wrapped FBR fuel subassemblies in high burn-up conditions. JNC has been developing a bundle deformation analysis code BAMBOO (Behavior Analysis code for Mechanical interaction of fuel Bundle under On-power Operation), a thermal hydraulics analysis code ASFRE-IV (Analysis of Sodium Flow in Reactor Elements - ver. IV) and their coupling method as a simulation system for the evaluation on the integrity of deformed FBR fuel pin bundles. In this study, the simulation system was applied to a coupling analysis of deformation and thermal-hydraulics in the fuel pin-bundle under a steady-state condition just after startup for the purpose of the verification of the simulation system. The iterative calculations of deformation and thermal-hydraulics employed in the coupling analysis provided numerically unstable solutions. From the result, it was found that improvement of the coupling algorithm of BAMBOO and ASFRE-IV is necessary to reduce numerical fluctuations and to obtain better convergence by introducing such computational technique as the optimized under-relaxation method. (author)

CERDEC Fuel Cell Team: Military Transitions for Soldier Fuel Cells

Science.gov (United States)

2008-10-27

Fuel Cell (DMFC) (PEO Soldier) Samsung: 20W DMFC (CRADA) General Atomics & Jadoo: 50W Ammonia Borane Fueled PEMFC Current Fuel Cell Team Efforts...Continued Ardica: 20W Wearable PEMFC operating on Chemical Hydrides Spectrum Brands w/ Rayovac: Hydrogen Generators and Alkaline Fuel Cells for AA...100W Ammonia Borane fueled PEMFC Ultralife: 150W sodium borohydride fueled PEMFC Protonex: 250W RMFC and Power Manager (ARO) NanoDynamics: 250W SOFC

Novel materials for fuel cells operating on liquid fuels

Directory of Open Access Journals (Sweden)

César A. C. Sequeira

2017-05-01

Full Text Available Towards commercialization of fuel cell products in the coming years, the fuel cell systems are being redefined by means of lowering costs of basic elements, such as electrolytes and membranes, electrode and catalyst materials, as well as of increasing power density and long-term stability. Among different kinds of fuel cells, low-temperature polymer electrolyte membrane fuel cells (PEMFCs are of major importance, but their problems related to hydrogen storage and distribution are forcing the development of liquid fuels such as methanol, ethanol, sodium borohydride and ammonia. In respect to hydrogen, methanol is cheaper, easier to handle, transport and store, and has a high theoretical energy density. The second most studied liquid fuel is ethanol, but it is necessary to note that the highest theoretically energy conversion efficiency should be reached in a cell operating on sodium borohydride alkaline solution. It is clear that proper solutions need to be developed, by using novel catalysts, namely nanostructured single phase and composite materials, oxidant enrichment technologies and catalytic activity increasing. In this paper these main directions will be considered.

Breached fuel location in FFTF by delayed neutron monitor triangulation

International Nuclear Information System (INIS)

Bunch, W.L.; Tang, E.L.

1985-10-01

The Fast Flux Test Facility (FFTF) features a three-loop, sodium-cooled 400 MWt mixed oxide fueled reactor designed for the irradiation testing of fuels and materials for use in liquid metal cooled fast reactors. To establish the ultimate capability of a particular fuel design and thereby generate information that will lead to improvements, many of the fuel irradiations are continued until a loss of cladding integrity (failure) occurs. When the cladding fails, fission gas escapes from the fuel pin and enters the reactor cover gas system. If the cladding failure permits the primary sodium to come in contact with the fuel, recoil fission products can enter the sodium. The presence of recoil fission products in the sodium can be detected by monitoring for the presence of delayed neutrons in the coolant. It is the present philosophy to not operate FFTF when a failure has occurred that permits fission fragments to enter the sodium. Thus, it is important that the identity and location of the fuel assembly that contains the failed cladding be established in order that it might be removed from the core. This report discusses method of location of fuel element when cladding is breached

The mechanism of interaction of polymethacrylic acid with sodium dodecylbenzenesulfonate in aqueous solutions

Science.gov (United States)

Sachko, A. V.; Zakordonskii, V. P.; Voloshinovskii, A. S.; Golod, T. Yu.

2009-07-01

A complex of physicochemical methods (light scattering, potentiometry, conductometry, viscometry, tensiometry, and fluorescence spectroscopy) were used to show the possibility of formation of intermolecular associates/complexes in systems with likely charged components. The driving forces of such interactions were analyzed and a possible scheme of complex formation between polymethacrylic acid and sodium dodecylbenzenesulfonate was suggested.

FAILED FUEL DISPOSITION STUDY

International Nuclear Information System (INIS)

THIELGES, J.R.

2004-01-01

In May 2004 alpha contamination was found on the lid of the pre-filter housing in the Sodium Removal Ion Exchange System during routine filter change. Subsequent investigation determined that the alpha contamination likely came from a fuel pin(s) contained in an Ident-69 (ID-69) type pin storage container serial number 9 (ID-69-9) that was washed in the Sodium Removal System (SRS) in January 2004. Because all evidence indicated that the wash water interacted with the fuel, this ID49 is designated as containing a failed fuel pin with gross cladding defect and was set aside in the Interim Examination and Maintenance (IEM) Cell until it could be determined how to proceed for long term dry storage of the fuel pin container. This ID49 contained fuel pins from the driver fuel assembly (DFA) 16392, which was identified as a Delayed Neutron Monitor (DNM) leaker assembly. However, this DFA was disassembled and the fuel pin that was thought to be the failed pin was encapsulated and was not located in this ID49 container. This failed fuel disposition study discusses two alternatives that could be used to address long term storage for the contents of ID-69-9. The first alternative evaluated utilizes the current method of identifying and storing DNM leaker fuel pin(s) in tubes and thus, verifying that the alpha contamination found in the SRS came from a failed pin in this pin container. This approach will require unloading selected fuel pins from the ID-69, visually examining and possibly weighing suspect fuel pins to identify the failed pin(s), inserting the failed pin(s) in storage tubes, and reloading the fuel pins into ID49 containers. Safety analysis must be performed to revise the 200 Area Interim Storage Area (ISA) Final Safety Analysis Report (FSAR) (Reference 1) for this fuel configuration. The second alternative considered is to store the failed fuel as-is in the ID-69. This was evaluated to determine if this approach would comply with storage requirements. This

FAILED FUEL DISPOSITION STUDY

Energy Technology Data Exchange (ETDEWEB)

THIELGES, J.R.

2004-12-20

In May 2004 alpha contamination was found on the lid of the pre-filter housing in the Sodium Removal Ion Exchange System during routine filter change. Subsequent investigation determined that the alpha contamination likely came from a fuel pin(s) contained in an Ident-69 (ID-69) type pin storage container serial number 9 (ID-69-9) that was washed in the Sodium Removal System (SRS) in January 2004. Because all evidence indicated that the wash water interacted with the fuel, this ID49 is designated as containing a failed fuel pin with gross cladding defect and was set aside in the Interim Examination and Maintenance (IEM) Cell until it could be determined how to proceed for long term dry storage of the fuel pin container. This ID49 contained fuel pins from the driver fuel assembly (DFA) 16392, which was identified as a Delayed Neutron Monitor (DNM) leaker assembly. However, this DFA was disassembled and the fuel pin that was thought to be the failed pin was encapsulated and was not located in this ID49 container. This failed fuel disposition study discusses two alternatives that could be used to address long term storage for the contents of ID-69-9. The first alternative evaluated utilizes the current method of identifying and storing DNM leaker fuel pin(s) in tubes and thus, verifying that the alpha contamination found in the SRS came from a failed pin in this pin container. This approach will require unloading selected fuel pins from the ID-69, visually examining and possibly weighing suspect fuel pins to identify the failed pin(s), inserting the failed pin(s) in storage tubes, and reloading the fuel pins into ID49 containers. Safety analysis must be performed to revise the 200 Area Interim Storage Area (ISA) Final Safety Analysis Report (FSAR) (Reference 1) for this fuel configuration. The second alternative considered is to store the failed fuel as-is in the ID-69. This was evaluated to determine if this approach would comply with storage requirements. This

Integrated Fuel-Coolant Interaction (IFCI 7.0) Code User's Manual

International Nuclear Information System (INIS)

Young, Michael F.

1999-01-01

The integrated fuel-coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, three-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a description of IFCI 7.0. The user's manual describes the hydrodynamic method and physical models used in IFCI 7.0. Appendix A is an input manual provided for the creation of working decks

Work plan: transient release from LMFBR fuel

International Nuclear Information System (INIS)

Kress, T.S.; Parker, G.W.; Fontana, M.H.

1975-09-01

The proposed LMFBR Transient Release Program at ORNL is designed to investigate, by means of ex-reactor experiments and analytical modeling, the release and transport of fuel, fission products, and transuranic elements from fast reactor cores in the event of certain hypothetical accidents. It is desired to experimentally produce energy depositions that are characteristic of severe hypothetical reactor transients by the application of direct electrical current to mixed-oxide fuels under sodium. The experimental program includes tests with and without sodium, investigations of alternative methods of generating fuel and sodium aerosols, the use of UO 2 as a fuel simulant, additions of tracers as fission product simulants, effects of radiation, and under-water and under-sodium efforts to study the behavior of the vapor bubble itself. Analytical modeling will accompany all phases of the program, and the data will be correlated with models developed. 21 references. (auth)

Specialists meeting on sodium removal and decontamination. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1978-08-01

This report covers experiences on sodium removal techniques developed or gained in a number of countries running sodium cooled reactors. This experience has created confidence that complex components can be cleaned of sodium, maintenance or repair operations carried out, and the components successfully re-used. Cleaning of sodium wetted components or fuel assemblies was achieved by applying different techniques including vacuum distillation, using different alcohols or evaporation processes.

Specialists meeting on sodium removal and decontamination. Summary report

International Nuclear Information System (INIS)

1978-08-01

This report covers experiences on sodium removal techniques developed or gained in a number of countries running sodium cooled reactors. This experience has created confidence that complex components can be cleaned of sodium, maintenance or repair operations carried out, and the components successfully re-used. Cleaning of sodium wetted components or fuel assemblies was achieved by applying different techniques including vacuum distillation, using different alcohols or evaporation processes

Brazed thermocouple pass-through for sodium service in a liquid-metal-cooled fast breeder reactor

International Nuclear Information System (INIS)

Walker, D.E.

1975-10-01

Sensors installed in special fuel elements for the EBR-II reactor had 30-ft-long leads that would pass from the sodium environment through a sealed bulkhead. A hydrogen-atmosphere, induction-heated brazing furnace was constructed to simultaneously braze 20-26 separate sensor leads at one time. The brazed seals were leak-tight, and the sheath wall has less than 10 percent interaction with the braze alloy

Soil-structure interaction in fuel handling building

International Nuclear Information System (INIS)

Elaidi, B.M.; Eissa, M.A.

1998-01-01

This paper presents an accurate three-dimensional seismic soil-structure interaction analysis for large structures. The method is applied to the fuel building in nuclear power plants. The analysis is performed numerically in the frequency domain and the responses are obtained by inverse Fourier transformation. The size of the structure matrices is reduced by transforming the equation of motion to the modal coordinate system. The soil is simulated as a layered media on top of viscoelastic half space. Soil impedance matrices are calculated from the principles of continuum mechanics and account for soil stiffness and energy dissipation. Effects of embedment on the field equations is incorporated through the scattering matrices or by simply scaling the soil impedance. Finite element methods are used to discretize the concrete foundation for the generation of the soil interaction matrices. Decoupling of the sloshing water in the spent fuel pools and the free-standing spent fuel racks is simulated. The input seismic motions are defined by three artificial time history accelerations. These input motions are generated to match the ground design basis response spectra and the target power spectral density function. The methods described in this paper can handle arbitrary foundation layouts, allows for large structural models, and accurately represents the soil impedance. Time history acceleration responses were subsequently used to generate floor response spectra at applicable damping values. (orig.)

Safety characteristics of mid-sized MOX fueled liquid metal reactor core of high converter type in the initiating phase of unprotected loss of flow accident. Effect of low specific fuel power density on ULOF behavior brought by employment of large diameter fuel pins

International Nuclear Information System (INIS)

Ishida, Masayoshi; Kawada, Kenichi; Niwa, Hajime

2003-07-01

-voided ro partially voided fuel subassemblies caused a high ramp rate insertion of sodium void reactivity due to fuel-coolant interactions, resulting in super-prompt critical power burst. Taking into account the ULOF analysis result of the UPL120 core, two alternative concepts of mid-sized core of high converter type were proposed in JFY2002, increasing the average specific fuel power density to 43 kW/kg and reducing the core height to 1 m and 0.8 m. These two cores, abbreviated here as Fsm 100 and Fsm 80, have core void reactivity worth of 5.7$ and 4.9$, respectively. The ULOF behaviors in initiating phase of both cores resulted in a mild power burst with net reactivity less than 1 $ under nominal design and best estimate analysis conditions, and even under conditions with a conservative 2σ uncertainty on Doppler constant, or on sodium void reactivity worth, the cores showed a mild power burst without exceeding prompt-criticality, though the conservative conditions resulted in a severer power burst. Hence the initiating phase of these alternative cores develops into a transition phase of ULOF scenario. Another parametric study showed that if the specific power density of these alternative cores was to be reduced by 20% from its 43 kW/kg to 34 kW/kg, the ULOF power burst became super-prompt critical. This study has shown that in the core design of high converter type with a large diameter (>10 mm) fuel pins, the specific power density of fuel is an important design parameter which affects the safety characteristics of the core in ULOF events, and the above mentioned ULOF results should be taken into account in the continuing core design study. (author)

Temperature and Burnup Correlated FCCI in U-10Zr Metallic Fuel

Energy Technology Data Exchange (ETDEWEB)

William J. Carmack

2012-05-01

Metallic fuels are proposed for use in advanced sodium cooled fast reactors. The experience basis for metallic fuels is extensive and includes development and qualification of fuels for the Experimental Breeder Reactor I, the Experimental Breeder Reactor II, FERMI-I, and the Fast Flux Test Facility (FFTF) reactors. Metallic fuels provide a number of advantages over other fuel types in terms of fabricability, performance, recyclability, and safety. Key to the performance of all nuclear fuel systems is the resistance to “breach” and subsequent release of fission products and fuel constituents to the primary coolant system of the nuclear power plant. In metallic fuel, the experience is that significant fuel-cladding chemical (FCCI) interaction occurs and becomes prevalent at high power-high temperature operation and ultimately leads to fuel pin breach and failure. Empirical relationships for metallic fuel pin failure have been developed from a large body of in-pile and out of pile research, development, and experimentation. It has been found that significant in-pile acceleration of the FCCI rate is experienced over similar condition out-of-pile experiments. The study of FCCI in metallic fuels has led to the quantification of in-pile failure rates to establish an empirical time and temperature dependent failure limit for fuel elements. Up until now the understanding of FCCI layer formation has been limited to data generated in EBR-II experiments. This dissertation provides new FCCI data extracted from the MFF-series of metallic fuel irradiations performed in the FFTF. These fuel assemblies contain valuable information on the formation of FCCI in metallic fuels at a variety of temperature and burnup conditions and in fuel with axial fuel height three times longer than EBR-II experiments. The longer fuel column in the FFTF and the fuel pins examined have significantly different flux, power, temperature, and FCCI profiles than that found in similar tests conducted in

Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

International Nuclear Information System (INIS)

Roake, W.E.; Adamson, M.G.; Hilbert, R.F.; Langer, S.

1977-01-01

Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to ∼60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

Energy Technology Data Exchange (ETDEWEB)

Roake, W E [Westinghouse-Hanford Co., Richland, WA (United States); Adamson, M G [General Electric Company, Vallecitos Nuclear Center, Pleasanton, CA (United States); Hilbert, R F; Langer, S

1977-04-01

Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to {approx}60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

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

Acceleration Test Method for Failure Prediction of the End Cap Contact Region of Sodium Cooled Fast Reactor Fuel Rod

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyung-Kyu; Lee, Young-Ho; Lee, Hyun-Seung; Lee, Kang-Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2017-05-15

This paper reports the results of an acceleration test to predict the contact-induced failure that could occur at the cylinder-to-hole joint for the fuel rod of a sodium-cooled fast reactor (SFR). To incorporate the fuel life of the SFR currently under development at KAERI (around 35,000 h), the acceleration test method of reliability engineering was adopted in this work. A finite element method was used to evaluate the flow-induced vibration frequency and amplitude for the test parameter values. Five specimens were tested. The failure criterion during the life of the SFR fuel was applied. The S-N curve of the HT-9, the material of concern, was used to obtain the acceleration factor. As a result, a test time of 16.5 h was obtained for each specimen. It was concluded that the B{sub 0.004} life would be guaranteed for the SFR fuel rods with 99% confidence if no failure was observed at any of the contact surfaces of the five specimens.

Metal-fuel modeling for inherently safe reactor designs

International Nuclear Information System (INIS)

Miles, K.J. Jr.

1987-01-01

Current development of breeder reactor systems has led to the renewed interest in metal fuels. These fuels have properties that enhance the inherent safety of the system, such as high thermal conductivity, compatibility with liquid sodium, and low fuel/cladding mechanical interaction. While metal-fuel irradiation behavior is well understood, there are some areas where more information is needed to fully understand the various safety-related phenomena, such as fuel/cladding chemical interaction, eutectic melting and penetration, and axial relocation of molten fuel prior to cladding breach. Because many of these phenomena can cause changes in the reactivity state of the system, their effects on whole-core normal, anticipated, and hypothetical accident scenarios need to be studied. The metal-fuel behavior model DEFORM-5 is being developed to provide the necessary phenomenological basis for these studies. The first stage in the DEFORM-5 development has been completed. Presently, DEFORM-5 calculates the cladding strain, life fraction, and eutectic penetration thinning for Types D9, HT9, or 316 steels. This first stage of DEFORM-5 has been used to analyze the TREAT M2, M3, and M4 transients with irradiated Experimental Breeder Reactor-II driver fuel. The paper shows the DEFORM-5 and experimental results for failure times for the test pins. The results provide confidence and validation of the DEFORM-5 modeling of the cladding behavior

Modeling the influence of interaction layer formation on thermal conductivity of U–Mo dispersion fuel

International Nuclear Information System (INIS)

Burkes, Douglas E.; Casella, Andrew M.; Huber, Tanja K.

2015-01-01

Highlights: • Hsu equation provides best thermal conductivity estimate of U–Mo dispersion fuel. • Simple model considering interaction layer formation was coupled with Hsu equation. • Interaction layer thermal conductivity is not the most important attribute. • Effective thermal conductivity is mostly influenced by interaction layer formation. • Fuel particle distribution also influences the effective thermal conductivity. - Abstract: The Global Threat Reduction Initiative Program continues to develop existing and new test reactor fuels to achieve the maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Currently, the program is focused on assisting with the development and qualification of a fuel design that consists of a uranium–molybdenum (U–Mo) alloy dispersed in an aluminum matrix. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layer formation between the dispersed phase and matrix, porosity that forms during fabrication of the fuel plates or rods, and upon the concentration of the dispersed phase within the matrix. This paper develops and validates a simple model to study the influence of interaction layer formation, dispersed particle size, and volume fraction of dispersed phase in the matrix on the effective conductivity of the composite. The model shows excellent agreement with results previously presented in the literature. In particular, the thermal conductivity of the interaction layer does not appear to be as important in determining the effective conductivity of the composite, while formation of the interaction layer and subsequent consumption of the matrix reveals a rather significant effect. The effective thermal conductivity of the composite can be influenced by the dispersed particle distribution by minimizing interaction

IAEA sodium void reactivity benchmark calculations

International Nuclear Information System (INIS)

Hill, R.N.; Finck, P.J.

1992-01-01

In this paper, the IAEA-1 992 ''Benchmark Calculation of Sodium Void Reactivity Effect in Fast Reactor Core'' problem is evaluated. The proposed design is a large axially heterogeneous oxide-fueled fast reactor as described in Section 2; the core utilizes a sodium plenum above the core to enhance leakage effects. The calculation methods used in this benchmark evaluation are described in Section 3. In Section 4, the calculated core performance results for the benchmark reactor model are presented; and in Section 5, the influence of steel and interstitial sodium heterogeneity effects is estimated

Radio-contaminant behaviour in the cover-gas space and the containment building of a sodium-cooled fast reactor in accident conditions

International Nuclear Information System (INIS)

Mathe, Emmanuel

2014-01-01

In the context of the Generation IV initiative, the consequences of a severe-accident (SA) in a sodium-cooled fast reactor must be studied. A SFR (Sodium cooled Fast Reactor) severe accident involves the disruption of the core by super-criticality involving the destruction of a certain number of fuel assemblies. Subsequently the interaction between hot fuel and liquid sodium can lead to a vapor explosion which could create a breach in the primary system. Some contaminated liquid sodium would thus be ejected into the containment building. In this situation, the evaluation of potential releases to the environment (the source term) must forecast the quantity and the chemical speciation of the radio-contaminants likely to be released from the containment building. One critical risk of a SA is the production of contaminated aerosols in the containment building by spray ejection of primary-system sodium. Being pyrophoric, the sodium droplets react with oxygen first oxidizing then burning, with significant heat of combustion. As well as evaluating the consequences of a pressure rise inside the containment, the evolution of the sodium must be assessed since not only is it activated and contaminated but, in oxide form, very toxic. Ultimately, the aerosols are the main radiological risk acting as the vector for radionuclide transport to the environment in the event of a problem with the confinement. These aerosols could evolve and interact with the FP (Fissile Products) and these interactions could modify the physical and chemical nature of the PF. We model a large part of the events that occur during a SA inside a SFR from the sodium spray fire to the reaction between sodium aerosols and PF (iodine). At first, we develop a numerical model (NATRAC) that simulates the sodium spray fire, calculates the temperature and the pressure inside the containment as well as the mass of aerosols produced during this kind of fire. The simulation has been validated with different

Compatibility Behavior of the Ferritic-Martensitic Steel Cladding under the Liquid Sodium Environment

Energy Technology Data Exchange (ETDEWEB)

Kim, Jun Hwan; Baek, Jong Hyuk; Kim, Sung Ho; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-05-15

Fuel cladding is a component which confines uranium fuel to transport energy into the coolant as well as protect radioactive species from releasing outside. Sodium-cooled Fast Reactor (SFR) has been considered as one of the most probable next generation reactors in Korea because it can maximize uranium resource as well as reduce the amount of PWR spent fuel in conjunction with pyroprocessing. Sodium has been selected as the coolant of the SFR because of its superior fast neutron efficiency as well as thermal conductivity, which enables high power core design. However, it is reported that the fuel cladding materials like austenitic and ferritic stainless steel react sodium coolant so that the loss of the thickness, intergranular attack, and carburization or decarburization process may happen to induce the change of the mechanical property of the cladding. This study aimed to evaluate material property of the cladding material under the liquid sodium environment. Ferritic-martensitic steel (FMS) coupon and cladding tube were exposed at the flowing sodium then the microstructural and mechanical property were evaluated. mechanical property of the cladding was evaluated using the ring tension test

Transformation and fragmentation behavior of molten aluminum in sodium pool

International Nuclear Information System (INIS)

Nishimura, S.; Kinoshita, I.; Ueda, N.; Sugiyama, K. I.

2003-01-01

In order to investigate the possibility of fragmentation of the metallic alloy fuel on liquid phase formed by metallurgical reactions, which is important in evaluating the sequence of core disruptive accidents for metallic fuel fast reactors, a series of experiments was carried out using molten aluminum and sodium under the condition that the boiling of sodium on the surface of the melt does not occur. The melting point of aluminum (933K) is roughly equivalent to the liquefaction temperature between the U-Pu-Zr alloy fuel and the SUS cladding (about 923K). The thermal fragmentation of a molten aluminum with a solid crust in the sodium pool is caused by the transient pressurization within the melt confined by the solid crust even under the condition that the instantaneous contact interface temperature between the melt and the sodium is below the boiling point of sodium. This indicates the possibility that the metallic alloy fuel on liquid phase formed by metallurgical reactions can be fragmented without occurring the boiling of sodium on the surface of the melt. The transient pressurization within the melt is considered to be caused by following two mechanisms. i) the overheating of the coolant entrapped hydrodynamically inside the aluminum melt confined by solid crust ii) the progression of solid crust inward and the squeeze of inner liquid part of the aluminum melt confined by solid crust It is found that the degree of fragmentation defined by mass median diameter has the same tendency for different dropping modes (drop or jet) with different mass and ambient Weber number of the melt in the present experimental conditions

Sodium technology at EBR-II

International Nuclear Information System (INIS)

Holmes, J.T.; Smith, C.R.F.; Olson, W.H.

1976-01-01

Since the installation of purity monitoring systems in 1967, the control of the purity of the primary and secondary sodium and cover gas systems at the Experimental Breeder Reactor II (EBR-II) has been excellent. A rigorous monitoring program is being used to assure that operating limits for more than 25 chemical and radioactive impurities are not exceeded. The program involves the use of sophisticated sampling and analysis techniques and on-line monitors for both sodium and cover gas systems. Sodium purity control is accomplished by essentially continuous cold trapping of a small side stream of the total circulating sodium. The cold traps have been found to be very effective for the removal of the major chemical impurities (oxygen and hydrogen) and tritium but are almost ineffective for 131 I and 137 Cs that enter the sodium from fuel cladding breaks. Purging with pure argon maintains the cover gas purity

Study of mechanisms and kinetics of Sodium-CO2 interactions. Contribution to the evaluation of an energy conversion system with supercritical CO2 for sodium fast breeder reactors

International Nuclear Information System (INIS)

Gicquel, L.

2010-01-01

This PhD study consisted in studying reactive mechanisms and kinetics of sodium-CO 2 interactions, in the frame of the assessment of an energy conversion system with supercritical CO 2 for fast breeder reactors cooled by sodium. The approach was the following. First of all, the interactions between sodium and CO 2 have been brought to light by laboratory experiments associated with products analysis. They have enabled the establishment of a coherent mechanism, in agreement with literature data, and gave preliminary indications on the reaction kinetics. In order to estimate a more detailed reaction kinetics, we tried to approach the phenomenon that appears in the case of a leak in a sodium-CO 2 heat exchanger. Geometry of such heat exchangers is not fixed for the moment, even if the development of compact exchangers is foreseen. Then, free jets of CO 2 in liquid sodium have been modeled in order to obtain, by identification, kinetics parameters of the reaction. Those parameters, estimated with such a geometry, will remain valid with a much complex geometry, that will better represent the real exchanger. An experimental bench has been defined and built to realize those jets. The first laboratory experiments have concluded in the existence of different reactive mechanisms according to the temperature level. A threshold has been brought to light around 500 C. Below this one, reaction appears moderated, or even, slow, with a medium exothermicity, and appears after an induction period that depends on the temperature,and which duration could reach several hours. At contrary, above this threshold, it seems rapid and more exothermic. Below 500 C, sodium oxalate is produced, and then reacts with sodium in an exothermic way, following the reactions: CO 2 + Na →1/4 Na 2 C 2 O 4 + 1/4 CO + 1/4 Na 2 CO 3 (5) 4 Na + Na 2 C 2 O 4 → 3 Na 2 O + CO + C (6) Above 500 C, sodium carbonate is produced, and can then possibly react with sodium in an endothermic way, following the

Safety analyses for sodium-cooled fast reactors with pelletized and sphere-pac oxide fuels within the FP-7 European project PELGRIMM - 15386

International Nuclear Information System (INIS)

Maschek, W.; Andriolo, L.; Matzerath-Boccaccini, C.; Delage, F.; Parisi, C.; Del Nevo, A.; Abbate, G.; Schmitt, D.

2015-01-01

The European FP-7 project PELGRIMM addresses the development of Minor-Actinide (MA) bearing oxide fuel for Sodium-cooled Fast Reactors. Optionally, both MA homogeneous recycling and heterogeneous recycling is investigated with pellet and sphere-pac fuel. A first safety assessment of sphere-pac fuelled cores should be given in the Work Package 4 of the project. This assessment is in continuity with the former FP-7 CP-ESFR project. Within the CP-ESFR project the CONF2 core design has been developed characterized by a core with a large upper sodium plenum to reduce the coolant void worth. This optimized core has been chosen for the safety analyses in PELGRIMM. The task within the PELGRIMM project is thus a safety assessment of the CONF2 core loaded either with pellets or with sphere-pac fuel. The investigations started with the design of the CONF2 core with sphere-pac fuel and the determination of core safety parameters and burn-up behavior. The neutronic analyses have been performed with the MCNPX code. Variants of the CONF2 core contain up to 4% Am in the fuel. The results revealed an extended void worth (core + upper plenum) for an Am free core of 1 up to 3 dollars for the 4% Am core. Thermal-hydraulic design analyses have been performed by RELAP5-3D. The accident simulations should be performed by different codes, some of which focus on the initiation phase of the accident, as SAS4A, BELLA and the MAT5DYN code, whereas the SIMMER-III code will also deal with the later accident phases and a potential whole core melting. The codes had to be adapted to the specifics of the sphere-pac fuel, in particular to the thermal conductivity and gap conditions. Analyses showed that the safety assessment has to take into account two main phases. Starting up the core, the green fuel shows a reduced fuel thermal conductivity. After restructuring within a couple of hours, the thermal conductivity recovers and the fuel temperature decreases. The main objective of the safety analyses

Specialists' meeting on sodium fires and prevention. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1979-03-15

The purpose of the specialists' meeting was to summarize the IWGFR member countries' knowledge of sodium combustion and extinguishment technology, including prevention and detection of sodium fires and protective clothes and to review and discuss critical features of sodium fires contaminated with fission products and fuel, evolution and filtration of aerosols and to determine the critical gaps in our knowledge and what should be done to develop knowledge in this area. The technical parts of the meeting were divided into three major sections, as follows: sodium fires; prevention and extinguishing of sodium fires, and aerosols.

Specialists' meeting on sodium fires and prevention. Summary report

International Nuclear Information System (INIS)

1979-03-01

The purpose of the specialists' meeting was to summarize the IWGFR member countries' knowledge of sodium combustion and extinguishment technology, including prevention and detection of sodium fires and protective clothes and to review and discuss critical features of sodium fires contaminated with fission products and fuel, evolution and filtration of aerosols and to determine the critical gaps in our knowledge and what should be done to develop knowledge in this area. The technical parts of the meeting were divided into three major sections, as follows: sodium fires; prevention and extinguishing of sodium fires, and aerosols

Sodium cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Hokkyo, N; Inoue, K; Maeda, H

1968-11-21

In a sodium cooled fast neutron reactor, an ultrasonic generator is installed at a fuel assembly hold-down mechanism positioned above a blanket or fission gas reservoir located above the core. During operation of the reactor an ultrsonic wave of frequency 10/sup 3/ - 10/sup 4/ Hz is constantly transmitted to the core to resonantly inject the primary bubble with ultrasonic energy to thereby facilitate its growth. Hence, small bubbles grow gradually to prevent the sudden boiling of sodium if an accident occurs in the cooling system during operation of the reactor.

Hydrogen generation systems and methods utilizing sodium silicide and sodium silica gel materials

Energy Technology Data Exchange (ETDEWEB)

Wallace, Andrew P.; Melack, John M.; Lefenfeld, Michael

2017-12-19

Systems, devices, and methods combine thermally stable reactant materials and aqueous solutions to generate hydrogen and a non-toxic liquid by-product. The reactant materials can sodium silicide or sodium silica gel. The hydrogen generation devices are used in fuels cells and other industrial applications. One system combines cooling, pumping, water storage, and other devices to sense and control reactions between reactant materials and aqueous solutions to generate hydrogen. Springs and other pressurization mechanisms pressurize and deliver an aqueous solution to the reaction. A check valve and other pressure regulation mechanisms regulate the pressure of the aqueous solution delivered to the reactant fuel material in the reactor based upon characteristics of the pressurization mechanisms and can regulate the pressure of the delivered aqueous solution as a steady decay associated with the pressurization force. The pressure regulation mechanism can also prevent hydrogen gas from deflecting the pressure regulation mechanism.

Method of detecting a fuel element failure

International Nuclear Information System (INIS)

Cohen, P.

1975-01-01

A method is described for detecting a fuel element failure in a liquid-sodium-cooled fast breeder reactor consisting of equilibrating a sample of the coolant with a molten salt consisting of a mixture of barium iodide and strontium iodide (or other iodides) whereby a large fraction of any radioactive iodine present in the liquid sodium coolant exchanges with the iodine present in the salt; separating the molten salt and sodium; if necessary, equilibrating the molten salt with nonradioactive sodium and separating the molten salt and sodium; and monitoring the molten salt for the presence of iodine, the presence of iodine indicating that the cladding of a fuel element has failed. (U.S.)

Corrosion inhibition studies in support of the long term storage of AGR fuel

Energy Technology Data Exchange (ETDEWEB)

Standring, P [Sellafield Limited (United Kingdom)

2012-07-01

Thorp Receipt and Storage (at Sellafield, UK) is currently being investigated as a bridging solution for the storage of AGR fuel pending the out-come of a national review into spent fuel management. AGR spent fuel is known to be susceptible to corrosion through inter-granular attack. To avoid this, the chosen storage regime for AGR fuel is sodium hydroxide dosed pond water to pH 11.4; now 22 years of operating experience. The conversion of TR and S will require a phased transition. During this transition sodium hydroxide cannot be used due to materials compatibility issues. Alternative corrosion inhibitors have been investigated as an interim measure and sodium nitrate has been selected as a suitable candidate. The efficiency of sodium nitrate to inhibit propagating inter-granular attack of active AGR materials has yet to be established. In the longer term sodium hydroxide will be deployed along with a move to a closed loop pond water management system. Given that carbon dioxide is known to be absorbed by sodium hydroxide dosed water and can affect fuel integrity, in the case of Magnox fuel, there is a need to establish its impact on AGR fuel. The objectives are: To establish the impact of carbonate on AGR fuel corrosion; To establish the efficiency of sodium nitrate to inhibit propagating inter-granular attack of irradiated AGR materials.

Study on characteristics of U-Mo/Al-Si interaction layers of dispersion fuel plates

International Nuclear Information System (INIS)

Liu Lijian; Yin Changgeng; Chen Jiangang; Sun Changlong; Liu Yunming

2014-01-01

In this paper, we analyzed the characteristics of U-Mo/Al-Si interaction layers of dispersion fuel plates. The results show that the interaction layers (IL) are with irregular morphology and uneven thickness, and are mainly formed in the internal micro cracks of the dispersion fuel particles or at the interface between the particles and the substrates. The diffusion mechanism of U-Mo/Al-Si is the vacancy diffusion, Al and Si are migrating elements, and the diffusion reaction is that Al and Si diffuse to U-Mo alloy. Inside the interaction layers, the Al content keeps constant basically, but the Si content gradually increases with the substrate-fuel direction, and the maximum content of Si appears interaction layers near the U-Mo side. Adding about 5 wt% Si into Al matrix can restrain the diffusion reaction, and improve the performance of dispersion fuel plates finally. (authors)

An investigation of chitosan and sodium dodecyl sulfate interactions in acetic media

Directory of Open Access Journals (Sweden)

Petrović Lidija B.

2016-01-01

Full Text Available Polymer/surfactant association is a cooperative phenomenon where surfactant binds to the polymer in the form of aggregates, usually through electrostatic or hydrophobic forces. As already known, polyelectrolytes may interact with oppositely charged surfactants through electrostatic attraction that results in polymer/surfactant complex formation. This behavior could be desirable in wide range of application of polymer/surfactant mixtures, such as improving colloid stability, gelling, emulsification and microencapsulation. In the present study surface tension, turbidity, viscosity and electrophoretic mobility measurements were used to investigate interactions of cationic polyelectrolyte chitosan (Ch and oppositely charged anionic surfactant, sodium dodecyl sulfate (SDS, in buffered water. Obtained results show the presence of interactions that lead to Ch/SDS complexes formation at all investigated pH and for all investigated polymer concentrations. Mechanisms of interaction, as well as characteristics of formed Ch/SDS complexes, are highly dependent on their mass ratio in the mixtures, while pH has no significant influence. [Projekat Ministarstva nauke Republike Srbije, br. II46010

AGR fuel pin pellet-clad interaction failure limits and activity release fractions

International Nuclear Information System (INIS)

Hughes, H.; Hargreaves, R.

1985-01-01

The limiting conditions beyond which pellet-clad interaction can flail AGR fuel are described. They have been determined by many experiments involving post-irradiation examination and testing, loop experiments and cycling and up-rating of both individual fuel stringers and the whole WAGR core. The mechanisms causing this interaction are well understood and are quantitatively expressed in computer codes. Strain concentration effects over fuel cracks determine power cycling endurance while additional strain concentrations at clad ridges and from cross pin temperature gradients contribute to up-rating failures. An equation summarising tube burst test data so as to determine the ductility available at any transient is given. The hollow fuel and more ductile clad of the Civil AGR fuel pins leads to a much improved performance over the original fuel design. The Civil AGRs operate well within these limiting conditions and substantial increases beyond the design burn-up are confidently expected. The activity release on pin failure and its development during continued operation of failed fuel have also been investigated. A retention of radioiodine and caesium of 90-99% compared to the noble gases has been demonstrated. Measured fission gas releases into the free volume of Civil AGR fuel pins have been very low (< 0.1%)

Measurement of dynamic interaction between a vibrating fuel element and its support

Energy Technology Data Exchange (ETDEWEB)

Fisher, N.J.; Tromp, J.H.; Smith, B.A.W. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada). Chalk River Labs.

1996-12-01

Flow-induced vibration of CANDU{reg_sign} fuel can result in fretting damage of the fuel and its support. A WOrk-Rate Measuring Station (WORMS) was developed to measure the relative motion and contact forces between a vibrating fuel element and its support. The fixture consists of a small piece of support structure mounted on a micrometer stage. This arrangement permits position of the support relative to the fuel element to be controlled to within {+-} {micro}m. A piezoelectric triaxial load washer is positioned between the support and micrometer stage to measure contact forces, and a pair of miniature eddy-current displacement probes are mounted on the stage to measure fuel element-to-support relative motion. WORMS has been utilized to measure dynamic contact forces, relative displacements and work-rates between a vibrating fuel element and its support. For these tests, the fuel element was excited with broadband random force excitation to simulate flow-induced vibration due to axial flow. The relationship between fuel element-to-support gap or preload (i.e., interference or negative gap) and dynamic interaction (i.e., relative motion, contact forces and work-rates) was derived. These measurements confirmed numerical simulations of in-reactor interaction predicted earlier using the VIBIC code.

Study of pellet clad interaction defects in Dresden-3 fuel rods

International Nuclear Information System (INIS)

Pasupathi, V.; Perrin, J.S.

1979-01-01

During Cycle-3 operation of Dresden-3, fuel rod failures occurred following a transient power increase. Ten fuel rods from five of the leaking fuel assemblies were examined at Battelle's Columbus Laboratory and General Electric-Vallecitos Nuclear Center. Examinations consisted of nondestructive and destructive methods including metallography and scanning electron microscopy (SEM). Results showed the cause of fuel rod failure to be pellet clad interaction involving stress corrosion cracking. Results of SEM studies of the cladding crack surfaces and deposits on clad inner surfaces were in agreement with those reported by other investigators

Fuel-cladding interaction. Framatome CEA experiment on pencils preirradiated in nuclear power plants

International Nuclear Information System (INIS)

Atabek, Rosemarie; Vignesoult, Nicole

1979-01-01

The study of the fuel-cladding interaction is the subject of an important joint research programme between Framatome and the CEA. Tests are performed either on whole fuel rods, not exceeding two metres in length, from BR3 or the CAP (PRISCA experiment) or on fuel rods refabricated in hot cells from fuel rods of power reactors (FABRICE experiment). The first results reveal the two mechanical and chemical aspects of the interaction phenomenon: the permissible power surge of the fuel elements passes through a minimum for an integrated fast dose (E>1MeV) of around 1.5x10 21 n/cm 2 ; a study made with the electronic microprobe and the scanning microscope shows that the Te, I and Cs fission products are the corrosive agents of the cladding [fr

Fission product vapour - aerosol interactions in the containment: simulant fuel studies

International Nuclear Information System (INIS)

Beard, A.M.; Benson, C.G.; Bowsher, B.R.

1988-12-01

Experiments have been conducted in the Falcon facility to study the interaction of fission product vapours released from simulant fuel samples with control rod aerosols. The aerosols generated from both the control rod and fuel sample were chemically distinct and had different deposition characteristics. Extensive interaction was observed between the fission product vapours and the control rod aerosol. The two dominant mechanisms were condensation of the vapours onto the aerosol, and chemical reactions between the two components; sorption phenomena were believed to be only of secondary importance. The interaction of fission product vapours and reactor materials aerosols could have a major impact on the transport characteristics of the radioactive emission from a degrading core. (author)

A Mechanistic Source Term Calculation for a Metal Fuel Sodium Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Grabaskas, David; Bucknor, Matthew; Jerden, James

2017-06-26

A mechanistic source term (MST) calculation attempts to realistically assess the transport and release of radionuclides from a reactor system to the environment during a specific accident sequence. The U.S. Nuclear Regulatory Commission (NRC) has repeatedly stated its expectation that advanced reactor vendors will utilize an MST during the U.S. reactor licensing process. As part of a project to examine possible impediments to sodium fast reactor (SFR) licensing in the U.S., an analysis was conducted regarding the current capabilities to perform an MST for a metal fuel SFR. The purpose of the project was to identify and prioritize any gaps in current computational tools, and the associated database, for the accurate assessment of an MST. The results of the study demonstrate that an SFR MST is possible with current tools and data, but several gaps exist that may lead to possibly unacceptable levels of uncertainty, depending on the goals of the MST analysis.

Preliminary results on a contact between 4kg of molten UO2 and liquid sodium

International Nuclear Information System (INIS)

Amblard, M.

1976-01-01

The CORECT II Experiment consists in simulating the penetration of sodium into an assembly when the fuel is molten. It is a shock-tube type experiment with dimensions representative of a full scale assembly. Six tests were performed which have always resulted in fine fragmentation without any violent interaction. Grain size measurements were carried out. The following assumptions were made: the grains are formed in a very short time; the grains are formed from the liquid state; the grains are intimately blended with the sodium whose mass is one of the parameters. But computations from grain size data using these assumptions give results that have no bearing on the effects actually observed [fr

Sodium voiding analysis in Kalimer

International Nuclear Information System (INIS)

Chang, Won-Pyo; Jeong, Kwan-Seong; Hahn, Dohee

2001-01-01

A sodium boiling model has been developed for calculations of the void reactivity feedback as well as the fuel and cladding temperatures in the KALIMER core after onset of sodium boiling. The sodium boiling in liquid metal reactors using sodium as coolant should be modeled because of phenomenon difference observed from that in light water reactor systems. The developed model is a multiple -bubble slug ejection model. It allows a finite number of bubbles in a channel at any time. Voiding is assumed to result from formation of bubbles that fill the whole cross section of the coolant channel except for liquid film left on the cladding surface. The vapor pressure, currently, is assumed to be uniform within a bubble. The present study is focused on not only demonstration of the sodium voiding behavior predicted by the developed model, but also confirmation on qualitative acceptance for the model. In results, the model catches important phenomena for sodium boiling, while further effort should be made for the complete analysis. (author)

Modelling of molten fuel/concrete interactions

International Nuclear Information System (INIS)

Muir, J.F.; Benjamin, A.S.

1980-01-01

A computer program modelling the interaction between molten core materials and structural concrete (CORCON) is being developed to provide quantitative estimates of fuel-melt accident consequences suitable for risk assessment of light water reactors. The principal features of CORCON are reviewed. Models developed for the principal interaction phenomena, inter-component heat transfer, concrete erosion, and melt/gas chemical reactions, are described. Alternative models for the controlling phenomenon, heat transfer from the molten pool to the surrounding concrete, are presented. These models, formulated in conjunction with the development of CORCON, are characterized by the presence or absence of either a gas film or viscous layer of molten concrete at the melt/concrete interface. Predictions of heat transfer based on these models compare favorably with available experimental data

Sodium fast reactor safety and licensing research plan - Volume II

International Nuclear Information System (INIS)

Ludewig, H.; Powers, D.A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A.; Phillips, J.; Zeyen, R.; Clement, B.; Garner, Frank; Walters, Leon; Wright, Steve; Ott, Larry J.; Suo-Anttila, Ahti Jorma; Denning, Richard; Ohshima, Hiroyuki; Ohno, S.; Miyhara, S.; Yacout, Abdellatif; Farmer, M.; Wade, D.; Grandy, C.; Schmidt, R.; Cahalen, J.; Olivier, Tara Jean; Budnitz, R.; Tobita, Yoshiharu; Serre, Frederic; Natesan, Ken; Carbajo, Juan J.; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Thomas, Justin; Wei, Tom; Sofu, Tanju; Flanagan, George F.; Bari, R.; Porter D.

2012-01-01

Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

Sodium fast reactor safety and licensing research plan. Volume II.

Energy Technology Data Exchange (ETDEWEB)

Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.

2012-05-01

Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

Ex-vessel nuclear fuel transfer system

International Nuclear Information System (INIS)

1977-01-01

A system is described for transferring reactor fuel assemblies between a fuel storage area and a fuel transfer area while the fuel assemblies remain completely submerged in a continuous body of coolant. The invention relates particularly to sodium cooled fast breeder reactors. (UK)

Design and selection of materials for sodium cooled fast reactors

International Nuclear Information System (INIS)

Chetal, S.C.

2011-01-01

Sodium cooled fast reactors are currently in operation, under construction or under design by a number of countries. The design of sodium cooled fast reactor is covered by French RCC - MR code and ASME code NH. The codes cover rules as regards to materials, design and construction. These codes do not cover the effect of irradiation and environment. Elevated temperature design criteria in nuclear codes are much stringent in comparison to non nuclear codes. Sodium corrosion is not an issue in selection of materials provided oxygen impurity in sodium is controlled for which excellent reactor operating experience is available. Austenitic stainless steels have remained the choice for the permanent structures of primary sodium system. Stabilized austenitic stainless steel are rejected because of poor operating experience and non inclusion in the design codes. Route for improved creep behaviour lies in compositional modifications in 316 class steel. However, the weldability needs to be ensured. For cold leg component is non creep regime, SS 304 class steel is favoured from overall economics. Enhanced fuel burn up can be realized by the use of 9-12%Cr 1%Mo class steel for the wrapper of MOX fuel design, and cladding and wrapper for metal fuel reactors. Minor compositional modifications of 20% cold worked 15Cr-15Ni class austenitic stainless steel will be a strong candidate for the cladding of MOX fuel design in the short term. Long term objective for the cladding will be to develop oxide dispersion strengthened steel. 9%Cr 1%Mo class steel (Gr 91) is an ideal choice for integrated once through sodium heated steam generators. One needs to incorporate operating experience from reactors and thermal power stations, industrial capability and R and D feedback in preparing the technical specifications for procurement of wrought products and welding consumables to ensure reliable operation of the components and systems over the design life. The paper highlights the design approach

Radiochemical measurement of mass transport in sodium

International Nuclear Information System (INIS)

Cooper, M.H.; Chiang, S.H.

1976-01-01

Mass transport processes in the sodium coolant of Liquid Metal Fast Breeder Reactors (LMFBRs) are significant in determining rates of corrosion and deposition of radioactive nuclides from the fuel cladding, deposition and cold trapping of fission products from defect or failed fuel, carbon and nitrogen redistribution in the containment materials, and removal of impurities by cold trapping or hot trapping. Mass transport between rotating, concentric cylinders in molten sodium has been investigated using a unique radiochemical method. Long-lived (33 year) cesium-137, dissolved in the sodium, decays radioactively emitting a beta to barium-137m, which decays with a short half-life (2.6 minutes) emitting a gamma. Cesium is weakly adsorbed and remains in solution, while the barium is strongly adsorbed on the stainless steel surfaces. Hence, by measuring the barium-137m activity on movable stainless steel surfaces, one can calculate the mass transport to that surface. Mass transfer coefficients in sodium measured by this method are in agreement with published heat transfer correlations when the effect of the volumetric mass source is taken into account. Hence, heat transfer correlations can be confidently utilized by analogy in estimating mass transfer in liquid-metal systems

The improving of sodium cleaning device with atomized water

International Nuclear Information System (INIS)

Guo Huanfang; Zhang Xiufeng; Yu Chunli; Hong Shunzhang; Xing Chaoqing; Yuan Waimei

1999-04-01

The atomized water in N 2 under pressure can react with sodium on the sodium-soiled surfaces under control. This method has the following main advantages: The reaction proceeds at room temperature; high efficiency; fast reaction velocity; small quantity of water; the mechanical and/or corrosion damage by sodium reaction products can be avoided. The results of washing fuel element of CEFR (China Experimental Fast Reactor) were satisfactory

Nuclear fuel rod with burnable plate and pellet-clad interaction fix

International Nuclear Information System (INIS)

Boyle, R.F.

1987-01-01

This patent describes a nuclear fuel rod comprising a metallic tubular cladding containing nuclear fuel pellets, the pellets containing enriched uranium-235. The improvement described here comprises: ceramic wafers, each wafter comprising a sintered mixture of gadolinium oxide and uranium dioxide, the uranium oxide having no more uranium-235 than is present in natural uranium dioxide. Each of the wafers is axially disposed between a major portion of adjacent the nuclear fuel pellets, whereby the wafers freeze out volatile fission products produced by the nuclear fuel and prevent interaction of the fission products with the metallic tubing cladding

Fuel chemistry and pellet-clad interaction related to high burnup fuel. Proceedings of the technical committee

International Nuclear Information System (INIS)

2000-10-01

The purpose of the meeting was to review new developments in clad failures. Major findings regarding the causes of clad failures are presented in this publication, with the main topics being fuel chemistry and fission product behaviour, swelling and pellet-cladding mechanical interaction, cladding failure mechanism at high burnup, thermal properties and fuel behaviour in off-normal conditions. This publication contains 17 individual presentations delivered at the meeting; each of them was indexed separately

Conceptual core designs for a 1200 MWe sodium cooled fast reactor

International Nuclear Information System (INIS)

Joo, H. K.; Lee, K. B.; Yoo, J. W.; Kim, Y. I.

2008-01-01

The conceptual core design for a 1200 MWe sodium cooled fast reactor is being developed under the framework of the Gen-IV SFR development program. To this end, three core concepts have been tested during the development of a core concept: a core with an enrichment split fuel, a core with a single-enrichment fuel with a region-wise varying clad thickness, and a core with a single-enrichment fuel with non-fuel rods. In order to optimize a conceptual core configuration which satisfies the design targets, a sensitivity study of the core design parameters has been performed. Two core concepts, the core with an enrichment-split fuel and the core with a single-enrichment fuel with a region-wise varying clad thickness, have been proposed as the candidates of the conceptual core for a 1200 MWe sodium cooled fast reactor. The detailed core neutronic, fuel behavior, thermal, and safety analyses will be performed for the proposed candidate core concepts to finalize the core design concept. (authors)

Analyses of Interaction Phases of U Mo Dispersion Fuel by Synchrotron X ray Diffraction

Energy Technology Data Exchange (ETDEWEB)

Kim, Woo Jeong; Nam, Ji Min; Ryu, Ho Jin; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Herve, Palancher; Charollais, Francois [Saint Paul Lez Durance Cedex, Rhone (France); Bonnin, Anne; Honkimaeki, Veijo [Grenoble Cedex, Grenoble (France); Patrick Lemoined [Gif sur Yvette, Paris (France)

2012-10-15

Gamma phase U Mo alloys are one of the promising candidates to be used as advanced high uranium density fuel for high power research reactors due to their excellent irradiation performance. However, formation of interaction layers between the U Mo particles and Al matrix degrades the irradiation performance of U Mo dispersion fuel. One of the remedies to the interaction problem is a Si addition to the Al matrix. Recent irradiation tests have shown that the use of Al (2{approx}5wt%)Si matrices retarded the growth of interaction layers effectively during irradiation. Recently, KAERI has proposed silicide or nitride coated U Mo fuel for the minimization of the interaction layer growth. The silicide or nitride coatings are expected to act as interdiffusion barriers and their out of pile tests showed the improved diffusion barrier performances of the silicide and nitride layers. In order to characterize constituent phases in the coated layers on U Mo particles and the interaction layers of coated U Mo particle dispersed fuel, synchrotron X ray diffraction experiments have been performed at the ESRF (European Synchrotron Radiation Facility), France as a KAERI CEA cooperation program.

Methods of preventing fast breeder reactor shield plug from adhesion of sodium

International Nuclear Information System (INIS)

Hashiguchi, Koh; Hara, Johji; Nei, Hiromichi; Daiku, Motoichi; Wagatsuma, Kenji

1980-01-01

The shield plug, which is located at the upper part of a reactor vessel of a sodium-cooled fast breeder reactor, is composed of a rotating and a stationary plug. Fuel exchange is performed easily by the rotation of the rotating plug. The vapor or mist of sodium evaporated from liquid sodium deposits on the gap surfaces of the rotating and stationary plugs and is solidified or changed into a solid reactant. If such condition continues for a long period, harmful effects are exerted on the fuel exchange operation. In order to develop methods of preventing the sodium deposition, investigation was made on the phenomenon of sodium deposition. By the use of the testing equipment simulating the shield plug, deposition tests and specimen measurements were made for different gap width test section size and condition. On the basis of the effects of these parameters clarified by experiments, the effectiveness of three kinds of mechanism for preventing sodium deposition were investigated experimentally. In addition, by using a thermo-siphon analogical model, analysis was performed to deduce experimental equations for sodium deposition. (author)

Japanese studies on sodium fires, design and testing

International Nuclear Information System (INIS)

Mitsutsuka, N.; Yoshida, N.

1983-01-01

Considerations of sodium fires are very important for the design and licensing of LMFBRs. Continuing effort has been made in the study of sodium fires and their consequences since the beginning of the Japanese fast breeder reactor development program. Recent effort is mainly focussed on studies related to Monju, especially on the design and testing of primary cell liners against large sodium spills. Experimental and analytical studies on sodium fires, water release from concrete and sodium concrete reactions are conducted as a part of this study. Some extinguishing agents are also tested against sodium fires. In addition, considerable effort is being made in the development of detection systems for the small sodium leaks before a pipe rupture. This paper briefly summarizes the Japanese status of these sodium fire related activities conducted by Fast Breeder Reactor Development Project of the Power Reactor and Nuclear Fuel Development Corporation (PNC)

Demonstration of fuel resistant to pellet-cladding interaction. Phase I. Final report

International Nuclear Information System (INIS)

Rosenbaum, H.S.

1979-03-01

This program has as its ultimate objective the demonstration of an advanced fuel design that is resistant to the failure mechanism known as fuel pellet-cladding interaction (PCI). Two fuel concepts are being developed for possible demonstration within this program: (a) Cu-barrier fuel, and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to protect the Zircaloy cladding tube from the harmful effects of localized stress, and reactive fission products during reactor service. This is the final report for PHASE 1 of this program. Support tests have shown that the barrier fuel resists PCI far better than does the conventional Zircaloy-clad fuel. Power ramp tests thus far have shown good PCI resistance for Cu-barrier fuel at burnup > 12 MWd/kg-U and for Zr-liner fuel > 16 MWd/kg-U. The program calls for continued testing to still higher burnup levels in PHASE 2

Creep and creep rupture properties of cladding tube (type 316) in high temperature sodium

International Nuclear Information System (INIS)

Atsumo, H.

1977-01-01

The thin walled small sized seamless AISI 316 steel tubes, which are designated to be domestically used as the fuel cladding tube for sodium cooled fast breeder reactors in Japan, are irradiated in the following sodium of high temperature in the range of 370 deg. C to 700 deg. C, and receive gradually increased internal pressure caused by the fission produced gas generating from the nuclear fuel burn-up inside the cladding tube. Consequently, the creep behavior of fuel cladding tubes under a high temperature sodium environment is an important problem which must be determined and clarified together with their characteristic features under irradiation and in air. In relation to the creep performance of fuel cladding tubes made of AISI 316 steel and other comparable austenitic stainless steels, hardly any studies are found that are made systematically to examine the effect of sodium with sodium purity as parameter or any comparative studies with in-air data at various different temperatures. The present research work was aimed to obtain certain basic design data relating to in-sodium creep performance of the domestic made fuel cladding tubes for fast breeder reactors, and also to gain further date as considered necessary under several sodium conditions. That is, together with establishment of the technology for tensile creep test and internal pressure creep rupture test in flowing sodium of high temperature, a series of tests and studies were performed on the trial made cladding tubes of AISI Type-316 steel. In the first place, two kinds of purity conditions of sodium, close to the actual reactor-operating condition, (oxygen concentration of 10 ppm and 5 ppm respectively) were established, and then uniaxial tensile creep test and rupture test under various temperatures were performed and the resulting data were compared and evaluated against the in-air data. Then, secondly, an internal pressure creep rupture test was conducted under a single purity sodium environment

Oxygen Chemical Diffusion Coefficients of (Pu,Am)O2 Fuels

International Nuclear Information System (INIS)

Watanabe, M.; Kato, M.; Matsumoto, T.

2015-01-01

Minor actinide (MA)-bearing MOX fuels have been developed as candidate fuels which are used in fast neutron spectrum cores such as sodium-cooled fast reactor (SFR) cores and experimental accelerator driven system (ADS) cores. Americium (Am) which is one of the MA elements significantly affects basic properties. It is known that Am content causes oxygen potential to increase and that influences irradiation behaviour such as fuel-cladding chemical interaction (FCCI) and chemical state of fission products. However, the effects of Am content on changes of basic properties are not clear. In this work, the oxygen chemical diffusion coefficients were calculated from measured data and the relationship between oxygen diffusion and oxygen potential of (Pu,Am)O 2-x was discussed. (authors)

Specialists' meeting on sodium fires

International Nuclear Information System (INIS)

Kozlov, F.A.; Kuznetsova, R.I.

1989-01-01

The four sessions of the meeting covered the following topics: 1. general approach to fast reactor safety, standards of fire safety, maximum design basis accidents for sodium leaks and fires, status of sodium fires in different countries; 2. physical and chemical processes during combustion of sodium and its interaction with structural and technological materials and methods for structural protection; 3. methods of sodium fires extinguishing and measures for localizing aerosol combustion products, organization of fire fighting procedures, instruction and training of fire personnel; 4. elimination of the consequences of sodium fires

Specialists' meeting on sodium fires

Energy Technology Data Exchange (ETDEWEB)

Kozlov, F A; Kuznetsova, R I [eds.

1989-07-01

The four sessions of the meeting covered the following topics: 1. general approach to fast reactor safety, standards of fire safety, maximum design basis accidents for sodium leaks and fires, status of sodium fires in different countries; 2. physical and chemical processes during combustion of sodium and its interaction with structural and technological materials and methods for structural protection; 3. methods of sodium fires extinguishing and measures for localizing aerosol combustion products, organization of fire fighting procedures, instruction and training of fire personnel; 4. elimination of the consequences of sodium fires.

Interactions of Genes and Sodium Intake on the Development of Hypertension: A Cohort-Based Case-Control Study

Directory of Open Access Journals (Sweden)

Mei-Ling Chen

2018-05-01

Full Text Available There have been few studies investigating interactions of G-protein beta3 subunit (GNB3 C825T (rs5443 and dietary sodium intake on the risk of hypertension, i.e., BP salt sensitivity. The study aims to evaluate joint effects of GNB3 polymorphisms and sodium consumption on the development of hypertension. A cohort-based case-control study was conducted in 2014. There are 233 participants with newly diagnosed hypertension in the case group and 699 participants in the gender-matched control group. The primary outcome is the development of hypertension over a 10-year period. The determinants of hypertension were three genotypes of SNP in GNB3 (TT; CT; and CC and two dietary salt categories on the basis of the level of sodium consumption representing high (>4800 mg/day and low-sodium (<2400 mg/day diets. The development of hypertension increased with participants carrying TT genotype and high-sodium diets comparing with those carrying TC or CC genotype with low-sodium diets (adjusted OR 3.23, 95% CI 1.52–6.83 (Rothman synergy index = 3.79. The study suggests that GNB3 C825T polymorphism may influence the response of the renin-angiotensin system to high-sodium diet. It implies that GNB3 can be served as an easy, inexpensive, and early genetic marker of salt sensitivity to blood pressure. Salt-sensitive individuals should pay more attention to salt intake to reduce cardiovascular morbidity or mortality.

Fuel-Coolant Interactions: Visualization and Mixing Measurements

International Nuclear Information System (INIS)

Loewen, Eric P.; Bonazza, Riccardo; Corradini, Michael L.; Johannesen, Robert E.

2002-01-01

Dynamic X-ray imaging of fuel-coolant interactions (FCI), including quantitative measurement of fuel-coolant volume fractions and length scales, has been accomplished with a novel imaging system at the Nuclear Safety Research Center at the University of Wisconsin, Madison. The imaging system consists of visible-light high-speed digital video, low-energy X-ray digital imaging, and high-energy X-ray digital imaging subsystems. The data provide information concerning the melt jet velocity, melt jet configuration, melt volume fractions, void fractions, and spatial and temporal quantification of premixing length scales for a model fuel-coolant system of molten lead poured into a water pool (fuel temperatures 500 to 1000 K; jet diameters 10 to 30 mm; coolant temperatures 20 to 90 deg. C). Overall results indicate that the FCI has three general regions of behavior, with the high fuel-coolant temperature region similar to what might be expected under severe accident conditions. It was observed that the melt jet leading edge has the highest void fraction and readily fragments into discrete masses, which then subsequently subdivide into smaller masses of length scales <10 mm. The intact jet penetrates <3 to 5 jet length/jet diameter before this breakup occurs into discrete masses, which continue to subdivide. Hydrodynamic instabilities can be visually identified at the leading edge and along the jet column with an interfacial region that consists of melt, vapor, and water. This interface region was observed to grow in size as the water pool temperature was increased, indicating mixing enhancement by boiling processes

UO2 - Zr chemical interaction of PHWR fuel pins under high temperature

International Nuclear Information System (INIS)

Majumdar, P.; Mukhopadhyay, D.; Gupta, S.K.

2001-01-01

At high temperature Zircaloy clad interacts with the UO 2 fuel as well as with the steam to produce oxide layer of a-Zr(O) and ZrO 2 . This layer formation significantly reduces the structural strength of the clad. A computer code SFDCPA/MOD1 has been developed to simulate the interaction and predict the oxide layer thickness for any accidental transient condition. It is well validated with published experimental data on the isothermal and transient temperature condition. The program is applied to Indian Pressurized Heavy Water Reactor (PHWR) fuel pin under certain severe transient condition where it experiences temperature above 1000 C. The study gives an idea of the un-oxidized thickness of Zircaloy, which is an important criterion for fuel integrity. (author)

Power output of microbial fuel cell emphasizing interaction of anodic binder with bacteria

Science.gov (United States)

Li, Hongying; Liao, Bo; Xiong, Juan; Zhou, Xingwang; Zhi, Huozhen; Liu, Xiang; Li, Xiaoping; Li, Weishan

2018-03-01

Electrochemically active biofilm is necessary for the electron transfer between bacteria and anodic electrode in microbial fuel cells and selecting the type of anodic electrode material that favours formation of electrochemically active biofilm is crucial for the microbial fuel cell operation. We report a new finding that the interaction of anodic binder with bacteria plays more important role than its hydrophilicity for forming an electrochemically active biofilm, which is emphasized by applying poly(bisphenol A-co-epichorohydrin) as an anodic binder of the microbial fuel cell based on carbon nanotubes as anodic electrode and Escherichia coli as bacterium. The physical characterizations and electrochemical measurements demonstrate that poly(bisphenol A-co-epichorohydrin) exhibits a strong interaction with bacteria and thus provides the microbial fuel cell with excellent power density output. The MFC using poly(bisphenol A-co-epichorohydrin) reaches a maximum power density output of 3.8 W m-2. This value is larger than that of the MFCs using polytetrafluoroethylene that has poorer hydrophilicity, or polyvinyl alcohol that has better hydrophilicity but exhibits weaker interaction with bacteria than poly(bisphenol A-co-epichorohydrin).

Study of sodium film-boiling heat transfer from a high-temperature sphere

International Nuclear Information System (INIS)

Le-Belguet, A.

2013-01-01

During a severe accident in a sodium-cooled fast reactor, molten fuel may come into contact with the surrounding liquid sodium, resulting in a so-called Fuel-Coolant Interaction. This work aims at providing a better understanding and knowledge of the associated heat transfer, likely to be in the film-boiling regime and required to study the risks related to a vapor explosion. Scarce literature has been found on sodium film boiling, both from an experimental and a theoretical point of view. Only one experiment has been conducted to investigate sodium pool film-boiling heat transfer. In our analysis of the experiment, two film-boiling regimes have been identified: a stable film boiling regime, without liquid-solid contact, and an unstable film-boiling regime, with contacts. Besides, the only theoretical model dedicated to sodium film boiling has shown some weaknesses. First, a scaling analysis of the problem has been proposed for free and forced convection, considering the two extreme cases of saturated and highly subcooled liquid. This simplified approach, which shows a good agreement with the experimental data, provides the dimensionless numbers which should be used to build correlations. A theoretical model has been developed to describe sodium film-boiling heat transfer from a hot sphere in free and forced convection, whatever the liquid subcooling. It is based on a two-phase laminar boundary layer integral method and includes the inertial and convective terms in the vapor momentum and energy equations, usually neglected. The radiation has been taken into account in the interfacial energy balance and contributes directly to produce vapor. This model enables to predict the heat lost from a hot body within an acceptable error compared to the tests results especially when the experimental uncertainties are considered. The heat partition between liquid heating and vaporization, essential to study the vapor explosion phenomenon, is also estimated. The influence of

Effect of sodium sulfite, sodium dodecyl sulfate, and urea on the molecular interactions and properties of whey protein isolate-based films

Science.gov (United States)

Schmid, Markus; Prinz, Tobias K.; Stäbler, Andreas; Sängerlaub, Sven

2016-12-01

Whey protein coatings and cast films are promising for use as food packaging materials. Ongoing research is endeavoring to reduce their permeability. The intention of this study was to evaluate the effect of the reactive additives sodium sulfite, sodium dodecyl sulfate (SDS), and urea on the oxygen barrier, water vapor barrier, and protein solubility of whey protein cast films. The concentration of the reactive additives was 1 to 20 wt.-%. Dried whey protein cast films were used as substrate materials. The water vapor transmission rate, the oxygen permeability, and the protein solubility were measured. Effective diffusion coefficients and effective sorption coefficients were calculated from the results of the water vapor sorption experiments. The presence of sodium sulfite resulted in an increased number of hydrophobic interactions and hydrogen bonds and a slightly decreased number of disulfide bonds. The oxygen permeability decreased from 68 to 46 cm³ (STP / standard temperature and pressure) 100 µm (m² d bar)-1 for 1 wt.-% SDS in the whey protein cast film. The water vapor transmission rate decreased from 165 to 44 g 100 µm (m² d)-1 measured at 50 to 0 % r. h. for 20 wt.-% SDS in the whey protein cast film. The reduction in the water vapor transmission rate correlated with the lower effective diffusion coefficient.

FFTF sodium and cover gas characterization and purification

International Nuclear Information System (INIS)

McCown, J.J.; Bloom, G.R.; Meadows, G.E.; Mettler, G.W.

1980-02-01

The FFTF Primary and Secondary Heat Transport System (HTS) sodium is purified with cold traps which have packed crystallizers and external economizers. The Primary HTS cold trap is NaK cooled and the Secondary HTS cold traps are air cooled. The FFTF cold traps have maintained high purity in the sodium since sodium fill. Plant operational procedures during fill and initial sodium heatup to 800 0 F were controlled to assure low release rates of impurities to the sodium. The FFTF sodium systems are monitored by plugging temperature indicators and by several sampling methods. During reactor fill and non-fueled operations at 400 to 800 0 F, impurity changes in the sodium were followed by continuous plugging indicator coverage, by exposing wires and foils to measure carbon, hydrogen and oxygen, and by bulk sample analysis of all other trace constituents. The sampling and analysis methods and data are presented, impurity excursions in the cover gas and sodium are described, and impurity trends are discussed

SPECTROSCOPIC STUDY OF INTERACTION OF SODIUM DOLUTEGRAVIR WITH HUMAN SERUM ALBUMIN

Directory of Open Access Journals (Sweden)

A. V. Yegorova

2017-11-01

Full Text Available Drug-protein binding has become an important research field in life sciences, chemistry and clinical medicine. Under physiological conditions, in vitro interaction between the antiviral drug 2 Sodium (4R, 12aS-9-{[(2,4-difluorophenylmethyl]carbamoyl}-4-methyl-6,8-dioxo3,4,6,8, 12,12a-hexahydro-2H-pyrido[1’,2’:4,5]pyrazino[2, 1-b][1,3]oxazin-7 –olate (dolutegravir sodium, DN and human serum albumin (HSA was investigated at excitation wavelength 280 nm and at different temperatures (298 K and 313 K by fluorescence emission spectroscopy. The emission of HSA was characterized by a broad emission band at 346 nm. The results of the experiment showed that DN quench the intrinsic fluorescence of the protein as a result of static interaction in the HSA -DN system, which is confirmed by shifts in the difference UV spectra of the HSA -DN and the reduction of the binding constant for the HSA -DN system with increasing temperature. The constant (KA =9,82· 103 L·mol-1 at 298 K and the number of binding sites of the HSA –DN system are established. The negative values of enthalpy change (ΔHº and entropy change (ΔSº can be attributed in part to van der Waals forces and in part to the formation of hydrogen bonds. A value of 2,14 nm for the average distance r between DN (acceptor and tryptophan residues of HSA (donor was derived from the fluorescence resonance energy transfer. The overlap of the absorbance spectrum of DN with the fluorescence emission spectrum of HAS has been shown. Since, the pharmaceutical firms need standardized screens for protein binding in the first step of new drug design, this kind of study of interaction between HSA with DN would be useful in pharmaceutical industry and clinical medicine.

The importance of input interactions in the uncertainty and sensitivity analysis of nuclear fuel behavior

Energy Technology Data Exchange (ETDEWEB)

Ikonen, T., E-mail: timo.ikonen@vtt.fi; Tulkki, V.

2014-08-15

Highlights: • Uncertainty and sensitivity analysis of modeled nuclear fuel behavior is performed. • Burnup dependency of the uncertainties and sensitivities is characterized. • Input interactions significantly increase output uncertainties for irradiated fuel. • Identification of uncertainty sources is greatly improved with higher order methods. • Results stress the importance of using methods that take interactions into account. - Abstract: The propagation of uncertainties in a PWR fuel rod under steady-state irradiation is analyzed by computational means. A hypothetical steady-state scenario of the Three Mile Island 1 reactor fuel rod is modeled with the fuel performance FRAPCON, using realistic input uncertainties for the fabrication and model parameters, boundary conditions and material properties. The uncertainty and sensitivity analysis is performed by extensive Monte Carlo sampling of the inputs’ probability distribution and by applying correlation coefficient and Sobol’ variance decomposition analyses. The latter includes evaluation of the second order and total effect sensitivity indices, allowing the study of interactions between input variables. The results show that the interactions play a large role in the propagation of uncertainties, and first order methods such as the correlation coefficient analyses are in general insufficient for sensitivity analysis of the fuel rod. Significant improvement over the first order methods can be achieved by using higher order methods. The results also show that both the magnitude of the uncertainties and their propagation depends not only on the output in question, but also on burnup. The latter is due to onset of new phenomena (such as the fission gas release) and the gradual closure of the pellet-cladding gap with increasing burnup. Increasing burnup also affects the importance of input interactions. Interaction effects are typically highest in the moderate burnup (of the order of 10–40 MWd

A data processing program for transient sodium boiling and fuel failure propagation tests, (2)

International Nuclear Information System (INIS)

Hasebe, Takeshi; Isozaki, Tadashi; Satoh, Akihiro; Yamaguchi, Katsuhisa; Haga, Kazuo.

1983-01-01

Transient Sodium Boiling Tests and Fuel Failure Propagation Tests are being conducted with the out-of-pile test facility, SIENA, in the Core Safety Section of O-arai Engineering Center. The experimental data are recorded using a digital data acquisition system controlled by a HP-1000E computer. The SICILIAN (Speedy Illustration Code for Inspection Line Anomaly) code was developed to obtain quick graphic outputs of data recorded in the magnetic tapes. The program is written in BASIC and Assembler languages and uses a data processing system composed of a desktop computer HP 9845B, a magnetic tape system, a magnetic disc and an eightcolor plotter. The SICILIAN code enables us to get graphic outputs soon after a run. These outputs are very helpful to inspect anomaly in the instrument circuit and to check the experimental conditions of coming runs. (author)

In-reactor performance of methods to control fuel-cladding chemical interaction

International Nuclear Information System (INIS)

Weber, E.T.; Gibby, R.L.; Wilson, C.N.; Lawrence, L.A.; Adamson, M.G.

1979-01-01

Inner surface corrosion of austenitic stainless steel cladding by oxygen and reactive fission product elements requires a 50 μm wastage allowance in current FBR reference oxide fuel pin design. Elimination or reduction of this wastage allowance could result in better reactor efficiency and economics through improvements in fuel pin performance and reliability. Reduction in cladding thickness and replacement of equivalent volume with fuel result in improved breeding capability. Of the factors affecting fuel-cladding chemical interaction (FCCI), oxygen activity within the fuel pin can be most readily controlled and/or manipulated without degrading fuel pin performance or significantly increasing fuel fabrication costs. There are two major approaches to control oxygen activity within an oxide fuel pin: (1) control of total oxygen inventory and chemical activity (Δ anti GO 2 ) by use of low oxygen-to-metal ratio (O/M) fuel; and (2) incorporation of a material within the fuel pin to provide in-situ control of oxygen activity (Δ anti GO 2 ) and fixation of excess oxygen prior to, or in preference to reaction with the cladding. The paper describes irradiation tests which were conducted in EBR-II and GETR incorporating oxygen buffer/getter materials and very low O/M fuel to control oxygen activity in sealed fuel pins

Investigation on the interaction of catalase with sodium lauryl sulfonate and the underlying mechanisms.

Science.gov (United States)

Wang, Jing; Jia, Rui; Wang, Jiaxi; Sun, Zhiqiang; Wu, Zitao; Liu, Rutao; Zong, Wansong

2018-02-01

As a classic type of anionic surfactants, sodium lauryl sulfonate (SLS) might change the structure and function of antioxidant enzyme catalase (CAT) through their direct interactions. However, the underlying molecular mechanism is still unknown. This study investigated the direct interaction of SLS with CAT molecule and the underlying mechanisms using multi-spectroscopic methods, isothermal titration calorimetry, and molecular docking studies. No obvious effects were observed on CAT structure and activity under low SLS concentration exposure. The particle size of CAT molecule decreased and CAT activity was slightly inhibited under high SLS concentration exposure. SLS prefers to bind to the interface of CAT mainly via van der Waals' forces and hydrogen bonds. Subsequently, SLS interacts with the amino acid residues around the heme groups of CAT via hydrophobic interactions and might inhibit CAT activity. © 2017 Wiley Periodicals, Inc.

UO2/sodium thermal interaction experiments at AWRE foulness since January 1972

International Nuclear Information System (INIS)

Darby, K.I.; Pottinger, R.C.; Turner, R.G.; Rees, N.J.M.

With an interaction rig described in an earlier report a number of performance tests were carried out. The unwanted gas content of the pyrotechnic charge has been identified as adsorbed water vapor. Calorimetric firings have confirmed the total theoretical energy release and chemical analyses of the debris showed that the pyrotechnic reaction had burnt to completion: experiments to correlate rate of charge burning with the pressure rise are proceeding. Charges with UO 2 diluent instead of the previously used Al 2 O 3 have been fired under sodium and water producing identical effects. One tries to separate debris from sodium by distillation. Pyrotechnic charges diluted with aluminium or silver were fired in a viewing cell, the design of which is described. A very high gas blanket pressure probably due to vapor pressure effects from the charge were observed. One frictional plate has been used with a negligible effect on measured gas blanket pressure. As a means to limit the first bubble excursion same firings were made with an increased gas blanket pressure

Integrated Fuel-Coolant Interaction (IFCI 7.0) Code User's Manual

Energy Technology Data Exchange (ETDEWEB)

Young, Michael F.

1999-05-01

The integrated fuel-coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, three-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a description of IFCI 7.0. The user's manual describes the hydrodynamic method and physical models used in IFCI 7.0. Appendix A is an input manual provided for the creation of working decks.

Improvement the value of sodium void reactivity effect of the fast neutron reactor by the instrumentality of the Monte Carlo code

OpenAIRE

P.A. Maslov; V.I. Matveev; I.V. Malysheva

2015-01-01

To fulfill safety of fast sodium reactors in a beyond design-basis accident (BDBA) like unprotected loss of flow accident (ULOF), the sodium void reactivity effect (SVRE) should be close to zero. Its value depends on the fuel burnup – the higher burnup the higher value of SVRE. We analyze limitation of the fuel burnup in the core of a large sodium reactor imposed by SVRE. The model of a large sodium-cooled reactor core is chosen for analysis. Two fuel types are considered – MOX and nitride...

The combustion of sodium

International Nuclear Information System (INIS)

Newman, R.N.

1978-01-01

The burning rates of sodium in the form of vapour jets, droplets, sprays and unconfined and confined pools have been reviewed. Attention has been paid to assessing the value of models in the various combustion modes. Additional models have been constructed for the descriptions of laminar and turbulent vapour jets, stationary droplets, forced convection over ambient pool fires together with correlations for peak pressures in confined pool environments. Where appropriate experiments with sodium have not been conducted, the likely behaviour is predicted by comparison with the burning of other fuels, particularly in the field of large free ambient fires. Some areas where further knowledge is required are highlighted. (author)

Interactions of alfalfa hay and sodium propionate on dairy calf performance and rumen development.

Science.gov (United States)

Beiranvand, H; Ghorbani, G R; Khorvash, M; Nabipour, A; Dehghan-Banadaky, M; Homayouni, A; Kargar, S

2014-01-01

The objective of this experiment was to investigate the effects of different levels of alfalfa hay (AH) and sodium propionate (Pro) added to starter diets of Holstein calves on growth performance, rumen fermentation characteristics, and rumen development. Forty-two male Holstein calves (40±2kg of birth weight) were used in a complete randomized design with a 3×2 factorial arrangement of treatments. Dietary treatments were as follows: (1) control = concentrate only; (2) Pro = concentrate with 5% sodium propionate [dry matter (DM) basis]; (3) 5% AH = concentrate + 5% alfalfa hay (DM basis); (4) 5% AH + Pro = concentrate + 5% alfalfa hay + 5% sodium propionate (DM basis); (5) 10% AH = concentrate + 10% alfalfa hay (DM basis); and (6) 10% AH + Pro = concentrate + 10% alfalfa hay + 5% sodium propionate (DM basis). All calves were housed in individual pens bedded with sawdust until 10wk of age. They were given ad libitum access to water and starter throughout the experiment and were fed 2L of milk twice daily. Dry matter intake was recorded daily and body weight weekly. Calves from the control, 10% AH, and 10% AH + Pro treatments were euthanized after wk 10, and rumen wall samples were collected. Feeding of forage was found to increase overall dry matter intake, average daily gain, and final weight; supplementing sodium propionate had no effect on these parameters. Calves consuming forage had lower feed efficiency than those on the Pro diet. Rumen fluid in calves consuming forage had higher pH and greater concentrations of total volatile fatty acids and molar acetate. Morphometric parameters of the rumen wall substantiated the effect of AH supplementation, as plaque formation decreased macroscopically. Overall, the interaction between forage and sodium propionate did not affect calf performance parameters measured at the end of the experiment. Furthermore, inclusion of AH in starter diets positively enhanced the growth performance of male Holstein calves and influenced

Fuel clad chemical interactions in fast reactor MOX fuels

Energy Technology Data Exchange (ETDEWEB)

Viswanathan, R., E-mail: rvis@igcar.gov.in

2014-01-15

Clad corrosion being one of the factors limiting the life of a mixed-oxide fast reactor fuel element pin at high burn-up, some aspects known about the key elements (oxygen, cesium, tellurium, iodine) in the clad-attack are discussed and many Fuel–Clad-Chemical-Interaction (FCCI) models available in the literature are also discussed. Based on its relatively superior predictive ability, the HEDL (Hanford Engineering Development Laboratory) relation is recommended: d/μm = ({0.507 ⋅ [B/(at.% fission)] ⋅ (T/K-705) ⋅ [(O/M)_i-1.935]} + 20.5) for (O/M){sub i} ⩽ 1.98. A new model is proposed for (O/M){sub i} ⩾ 1.98: d/μm = [B/(at.% fission)] ⋅ (T/K-800){sup 0.5} ⋅ [(O/M){sub i}-1.94] ⋅ [P/(W cm{sup −1})]{sup 0.5}. Here, d is the maximum depth of clad attack, B is the burn-up, T is the clad inner surface temperature, (O/M){sub i} is the initial oxygen-to-(uranium + plutonium) ratio, and P is the linear power rating. For fuels with [n(Pu)/n(M = U + Pu)] > 0.25, multiplication factors f are recommended to consider the potential increase in the depth of clad-attack.

EBR-II spent fuel treatment demonstration project

International Nuclear Information System (INIS)

Benedict, R.W.; Henslee, S.P.

1997-01-01

For approximately 10 years, Argonne National Laboratory was developed a fast reactor fuel cycle based on dry processing. When the US fast reactor program was canceled in 1994, the fuel processing technology, called the electrometallurgical technique, was adapted for treating unstable spent nuclear fuel for disposal. While this technique, which involves electrorefining fuel in a molten salt bath, is being developed for several different fuel categories, its initial application is for sodium-bonded metallic spent fuel. In June 1996, the Department of Energy (DOE) approved a radiation demonstration program in which 100 spent driver assemblies and 25 spent blanket assemblies from the Experimental Breeder Reactor-II (EBR-II) will be treated over a three-year period. This demonstrated will provide data that address issues in the National Research Council's evaluation of the technology. The planned operations will neutralize the reactive component (elemental sodium) in the fuel and produce a low enriched uranium product, a ceramic waste and a metal waste. The fission products and transuranium elements, which accumulate in the electrorefining salt, will be stabilized in the glass-bonded ceramic waste form. The stainless steel cladding hulls, noble metal fission products, and insoluble residues from the process will be stabilized in a stainless steel/zirconium alloy. Upon completion of a successful demonstration and additional environmental evaluation, the current plans are to process the remainder of the DOE sodium bonded fuel

Analysis of advanced sodium-cooled fast reactor core designs with improved safety characteristics

Energy Technology Data Exchange (ETDEWEB)

Sun, K.

2012-09-15

Currently, the large majority of nuclear power plants are operated with thermal-neutron spectra and need regular fuel loading of enriched uranium. According to the identified conventional uranium resources and their current consumption rate, only about 100 years’ nuclear fuel supply is foreseen. A reactor operated with a fast-neutron spectrum, on the other hand, can induce self-sustaining, or even breeding, conditions for its inventory of fissile material, which effectively allow it, after the initial loading, to be refueled using simply natural or depleted uranium. This implies a much more efficient use of uranium resources. Moreover, minor actinides become fissionable in a fast-neutron spectrum, enabling full closure of the fuel cycle and leading to a minimization of long-lived radioactive wastes. The sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declared goals. In comparison to other Generation IV systems, there is considerable design experience related to the SFR, and also more than 300 reactor years of practical operation. As a fast-neutron-spectrum system, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. Although the SFR has many advantageous characteristics, it has one dominating neutronics drawback: there is generally a positive reactivity effect when sodium coolant is removed from the core. This so-called sodium void effect becomes even stronger in the equilibrium closed fuel cycle. The goal of the present doctoral research is to improve the safety characteristics of advanced SFR core designs, in particular, from the viewpoint of the positive sodium void reactivity effect. In this context, particular importance has been given to the dynamic core behavior under a hypothetical unprotected loss-of-flow (ULOF) accident scenario, in which sodium boiling occurs. The proposed

Analysis of advanced sodium-cooled fast reactor core designs with improved safety characteristics

International Nuclear Information System (INIS)

Sun, K.

2012-09-01

Currently, the large majority of nuclear power plants are operated with thermal-neutron spectra and need regular fuel loading of enriched uranium. According to the identified conventional uranium resources and their current consumption rate, only about 100 years’ nuclear fuel supply is foreseen. A reactor operated with a fast-neutron spectrum, on the other hand, can induce self-sustaining, or even breeding, conditions for its inventory of fissile material, which effectively allow it, after the initial loading, to be refueled using simply natural or depleted uranium. This implies a much more efficient use of uranium resources. Moreover, minor actinides become fissionable in a fast-neutron spectrum, enabling full closure of the fuel cycle and leading to a minimization of long-lived radioactive wastes. The sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declared goals. In comparison to other Generation IV systems, there is considerable design experience related to the SFR, and also more than 300 reactor years of practical operation. As a fast-neutron-spectrum system, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. Although the SFR has many advantageous characteristics, it has one dominating neutronics drawback: there is generally a positive reactivity effect when sodium coolant is removed from the core. This so-called sodium void effect becomes even stronger in the equilibrium closed fuel cycle. The goal of the present doctoral research is to improve the safety characteristics of advanced SFR core designs, in particular, from the viewpoint of the positive sodium void reactivity effect. In this context, particular importance has been given to the dynamic core behavior under a hypothetical unprotected loss-of-flow (ULOF) accident scenario, in which sodium boiling occurs. The proposed

Technical committee meeting on material-coolant interactions and material movement and relocation in liquid metal fast reactors

International Nuclear Information System (INIS)

1994-01-01

The Technical Committee Meeting on Material-Coolant Interactions and Material Movement and Relocation in Liquid Metal Fast Reactors was sponsored by the International Working Group on Fast Reactors (IWGFR), International Atomic Energy Agency (IAEA) and hosted by PNC, on behalf of the Japanese government. A broad range of technical subjects was discussed in the TCM, covering entire aspects of material motion and interactions relevant to the safety of LMFRs. Recent achievement and current status in research and development in this area were presented including European out-of-pile test of molten material movement and relocation; molten material-sodium interaction; molten fuel-coolant interaction; core disruptive accidents; sodium boiling; post accident material relocation, heat removal and relevant experiments already performed or planned

Technical committee meeting on material-coolant interactions and material movement and relocation in liquid metal fast reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-07-01

The Technical Committee Meeting on Material-Coolant Interactions and Material Movement and Relocation in Liquid Metal Fast Reactors was sponsored by the International Working Group on Fast Reactors (IWGFR), International Atomic Energy Agency (IAEA) and hosted by PNC, on behalf of the Japanese government. A broad range of technical subjects was discussed in the TCM, covering entire aspects of material motion and interactions relevant to the safety of LMFRs. Recent achievement and current status in research and development in this area were presented including European out-of-pile test of molten material movement and relocation; molten material-sodium interaction; molten fuel-coolant interaction; core disruptive accidents; sodium boiling; post accident material relocation, heat removal and relevant experiments already performed or planned.

Effect of interactions between vehicles and pedestrians on fuel consumption and emissions

Science.gov (United States)

Li, Xiang; Sun, Jian-Qiao

2014-12-01

This paper presents a study of variations of fuel consumption and emissions of vehicles due to random street crossings of pedestrians. The pedestrian and vehicle movement models as well as the interaction model between the two entities are presented. Extensive numerical simulations of single and multiple cars are carried out to investigate the traffic flow rate, vehicle average speed, fuel consumption, CO, HC and NOx emissions. Generally more noncompliant road-crossings of pedestrians lead to higher level of fuel consumptions and emissions of vehicles, and the traffic situation can be improved by imposing higher vehicle speed limit to some extent. Different traffic characteristics in low and high vehicle density regions are studied. The traffic flow is more influenced by crossing pedestrians in the low vehicle density region, while in the high vehicle density region, the interactions among vehicles dominate. The main contribution of this paper lies in the qualitative analysis of the impact of the interactions between pedestrians and vehicles on the traffic, its energy economy and emissions.

Interaction between chlorhexidine digluconate and sodium monofluorophosphate in vitro

International Nuclear Information System (INIS)

Barkvoll, P.; Roella, G.; Bellagamba, S.

1988-01-01

The aim of the present study was to investigate the compatibility of chlorhexidine digluconate and sodium monofluorophosphate since these agents are potential ingredients in future products in preventive dentistry. Varying combinations of chlorhexidine digluconate and sodium monofluorophosphate in water, covering the possible ranges of clinically relevant concentrations of both compounds, were made, incubated for 24 h and observed for precipitation of insoluble salts. The mixtures were analyzed for presence of free chlorhexidine and monofluorophosphate after incubation. The results showed that chlorhexidine digluconate and sodium monofluorophosphate are not compatible in clinically relevant concentrations. A chlorhexidinemonofluorophosphate salt of low solubility in water is presumably formed. (author)

Thermodynamic solution properties of pefloxacin mesylate and its interactions with organized assemblies of anionic surfactant, sodium dodecyl sulphate

International Nuclear Information System (INIS)

Usman, Muhammad; Rashid, Muhammad Abid; Mansha, Asim; Siddiq, Mohammad

2013-01-01

Graphical abstract: - Highlights: • Free energy of adsorption is more negative than free energy of micellization. • Micellization becomes more spontaneous at high temperature. • There is strong interaction between PFM and SDS. - Abstract: This manuscript reports the physicochemical behavior of antibiotic amphiphilic drug pefloxacin mesylate (PFM) and its interaction with anionic surfactant, sodium dodecyl sulfate (SDS). The data of surface tension and electrical conductivity are helpful to detect the CMC as well as to calculate surface parameters, i.e. surface pressure, π, surface excess concentration, Γ, area per molecule of drug and standard Gibbs free energy of adsorption, ΔG ads and thermodynamic parameters like standard free energy of micellization, ΔG m , standard enthalpy of micellization, ΔH m and standard entropy of micellization, ΔS m . The interaction of this drug with anionic surfactant, sodium dodecyl sulfate (SDS) was studied by electrical conductivity and UV/visible spectroscopy. This enabled us to compute the values of partition coefficient (K x ), free energy of partition, ΔG p , binding constant, K b , free energy of binding, ΔG b , number of drug molecules per micelle, n, and thermodynamic parameters of drug–surfactant interaction

Design comparisons of TRU burner cores with similar sodium void worth

International Nuclear Information System (INIS)

Sang Ji, Kim; Young Il, Kim; Young Jin, Kim; Nam Zin, Cho

2001-01-01

This study summarizes the neutronic performance and fuel cycle behavior of five geometrically-different transuranic (TRU) burner cores with similar low sodium void reactivity. The conceptual cores encompass core geometries for annular, two-region homogeneous, dual pin type, pan-shaped and H-shaped cores. They have been designed with the same assembly specifications and managed to have similar end-of-cycle sodium void reactivities and beginning-of-cycle peak power densities through the changes in the core size and configuration. The requirement of low sodium void reactivity is shown to lead each design concept to characteristic neutronics performance and fuel cycle behavior. The H-/pan-shaped cores allow the core compaction as well as higher rate of TRU burning. (author)

Operational experience of the fuel cleaning facility of Joyo

International Nuclear Information System (INIS)

Mukaibo, R.; Matsuno, Y.; Sato, I.; Yoneda, Y.; Ito, H.

1978-01-01

Spent fuel assemblies in 'Joyo', after they are taken out of the core, are taken to the Fuel Cleaning Facility in the reactor service building and sodium removal is done. The cleaning process is done by cooling the assembly with argon gas, steam charging and rinsing by demineralized water. Deposited sodium was 50 ∼ 60 g per assembly. The sodium and steam reaction takes about 15 minutes to end and the total time the fuel is placed in the pot is about an hour. The total number of assemblies cleaned in the facility was 95 as of November 1977. In this report the operational experience together with discussions of future improvements are given. (author)

Operational experience of the fuel cleaning facility of Joyo

Energy Technology Data Exchange (ETDEWEB)

Mukaibo, R; Matsuno, Y; Sato, I; Yoneda, Y; Ito, H [O-arai Engineering Centre, PNC, Ibaraki-ken, Tokio (Japan)

1978-08-01

Spent fuel assemblies in 'Joyo', after they are taken out of the core, are taken to the Fuel Cleaning Facility in the reactor service building and sodium removal is done. The cleaning process is done by cooling the assembly with argon gas, steam charging and rinsing by demineralized water. Deposited sodium was 50 {approx} 60 g per assembly. The sodium and steam reaction takes about 15 minutes to end and the total time the fuel is placed in the pot is about an hour. The total number of assemblies cleaned in the facility was 95 as of November 1977. In this report the operational experience together with discussions of future improvements are given. (author)

Irradiation performance of U-Pu-Zr metal fuels for liquid-metal-cooled reactors

International Nuclear Information System (INIS)

Tsai, H.; Cohen, A.B.; Billone, M.C.; Neimark, L.A.

1994-10-01

This report discusses a fuel system utilizing metallic U-Pu-Zr alloys which has been developed for advanced liquid metal-cooled reactors (LMRs). Result's from extensive irradiation testing conducted in EBR-II show a design having the following key features can achieve both high reliability and high burnup capability: a cast nominally U-20wt %Pu-10wt %Zr slug with the diameter sized to yield a fuel smear density of ∼75% theoretical density, low-swelling tempered martensitic stainless steel cladding, sodium bond filling the initial fuel/cladding gap, and an as-built plenum/fuel volume ratio of ∼1.5. The robust performance capability of this design stems primarily from the negligible loading on the cladding from either fuel/cladding mechanical interaction or fission-gas pressure during the irradiation. The effects of these individual design parameters, e.g., fuel smear density, zirconium content in fuel, plenum volume, and cladding types, on fuel element performance were investigated in a systematic irradiation experiment in EBR-II. The results show that, at the discharge burnup of ∼11 at. %, variations on zirconium content or plenum volume in the ranges tested have no substantial effects on performance. Fuel smear density, on the other hand, has pronounced but countervailing effects: increased density results in greater cladding strain, but lesser cladding wastage from fuel/cladding chemical interaction

Study on MAs transmutation of accelerator-driven system sodium-cooled fast reactor loaded with metallic fuel

International Nuclear Information System (INIS)

Han Song; Yang Yongwei

2007-01-01

Through the analysis of the effect of heavy metal actinides on the effective multiplication constant (k eff ) of the core in accelerator-driven system (ADS) sodium-cooled fast reactor loaded with metallic fuel, we gave the method for determining fuel components. the characteristics of minor actinides (MAs) transmutation was analyzed in detail. 3D burn-up code COUPLE, which couples MCNP4c3 and ORIGEN2, was applied to the neutron simulation and burn up calculation. The results of optimized scheme shows that adjusting the proportion of 239 Pu and maintaining the value during the burn-up cycle is an efficient method of designing k eff and keeping stable during the burn-up cycle. Spallation neutrons lead to the neutron spectrum harder at inner core than that at outer core. It is in favor of improving MA's fission cross sections and the capture-to-fission ratio. The total MAs transmutation support ratio 8.3 achieves excellent transmutation effect. For higher flux at inner core leads to obvious differences on transmutation efficiency,only disposing MAs at inner core is in favor of decreasing the loading mass and improving MAs transmutation effect. (authors)

Core performance of equilibrium fast reactors for different coolant materials and fuel types

International Nuclear Information System (INIS)

Mizutani, Akihiko; Sekimoto, Hiroshi

1998-01-01

Parametric studies with several coolant and fuel materials in the equilibrium state are performed for fast reactors in which natural uranium is fed and all of the actinides are confined. Sodium, sodium-potassium, lead, lead-bismuth and helium coolant materials, and oxide, nitride and metal fuels are employed to compare the neutronic characteristics in the equilibrium state. As to the criticality performance, sodium-potassium shows the best performance among the liquid metal coolants and the metallic fuel indicates the best performance

Mechatronics of fuel handling mechanism for fast experimental reactor 'Joyo'

Energy Technology Data Exchange (ETDEWEB)

Fujiwara, Akikazu (Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center)

1984-01-01

The outline of the fast experimental reactor ''Joyo'' is introduced, and the fuel handling mechanism peculiar to fast reactors is described. The objectives of the construction of Joyo are to obtain the techniques for the design, construction, manufacture, installation, operation and maintenance of sodium-cooled fast reactors independently, and to use it as an irradiation facility for the development of fuel and materials for fast breeder reactors. At present, the reactor is operated at 100 MW maximum thermal output for the second objective. Since liquid sodium is used as the coolant, the atmosphere of the fuel handling course changes such as liquid sodium at 250 deg C, argon gas at 200 deg C and water, in addition, the spent fuel taken out has the decay heat of 2.1 kW at maximum. The fuel handling works in the reactor and fuel transfer works, and the fuel handling mechanism of a fuel exchanger and that of a cask car for fuel handling are described. Relay sequence control system is used for the fuel handling mechanism of Joyo.

Interaction of cadmium and indium nitrate mixture with sodium tungstate in aqueous solution

Energy Technology Data Exchange (ETDEWEB)

Belousova, E E; Krivobok, V I; Gruba, A I [Donetskij Gosudarstvennyj Univ. (Ukrainian SSR)

1982-01-01

The interaction of the mixture of cadmium and indium nitrates with sodium tungstate in aqueous solution is studied using the methods of ''residual concentrations'', pH potentiometry and conductometry. Independent of the ratio of components in the initial solution a mixture of coprecipitated normal tungstates of cadmium and indium is formed in the system. Heat treatment of the precipitates at 800 deg C for 50 hrs with subsequent hardening results in the formation of solid solutions on the basis of normal cadmium and indium tungstates.

Influence of hydroxypropylmethyl cellulose-sodium laurylsulfate interaction on rheological properties of the solution

Directory of Open Access Journals (Sweden)

Šaletić Jelena V.

2004-01-01

Full Text Available Interactions between the polymers and surfactants in solution have widely been investigated because of their scientific and technological importance. These interactions can be utilized to modify the physicochemical properties of system in many food products, pharmaceutical formulations, personal care products, paints, pesticides, etc. Interaction between nonionic polymer - hydroxypropylmethyl cellulose (HPMC and anionic surfactant - sodium laurylsulfate (SDS in solution has been investigated in this paper by rheological measurements. Rheological measurements are performed by rotational viscometer at 20°C and changes of rheological characteristics of HPMC solutions (0.5-1.5% with increasing SDS concentrations (0-4.0% were determined. The results of these investigations showed that viscosity of the solution is dependant on HPMC-SDS interaction. At particular SDS concentration viscosity increases, reach maximum and after that decreases until reach constant value. From the viscosity changes the characteristic concentrations of SDS, critical aggregation concentration (cac and polymer saturation point (psp, were determined. These concentrations are in linear relationships with HPMC concentrations. Rheological properties of the solution are strong influenced by HPMC-SDS interaction and exhibits more or less pronounced pseudoplastic behavior, which changes to Newtonian one after the psp has been reached.

Analysis of Accident Scenarios for the Development of Probabilistic Safety Assessment Model for the Metallic Fuel Sodium-Cooled Fast Reactor

International Nuclear Information System (INIS)

Kim, Tae Woon; Park, S. Y.; Yang, J. E.; Kwon, Y. M.; Jeong, H. Y.; Suk, S. D.; Lee, Y. B.

2009-03-01

The safety analysis reports which were reported during the development of sodium cooled fast reactors in the foreign countries are reviewed for the establishment of Probabilistic Safety Analysis models for the domestic SFR which are under development. There are lots of differences in the safety characteristics between the mixed oxide (MOX) fuel SFR and metallic fuel SFR. Metallic fuel SFR is under development in Korea while MOX fuel SFR is under development in France, Japan, India and China. Therefore the status on the development of fast reactors in the foreign countries are reviewed at first and then the safety characteristics between the MOX fuel SFR and the metallic fuel SFR are reviewed. The core damage can be defined as coolant voiding, fuel melting, cladding damage. The melting points of metallic fuel and the MOX fuel is about 1000 .deg. C and 2300 .deg. C, respectively. The high energy stored in the MOX fuel have higher potential to voiding of coolant compared to the possibility in the metallic fuel. The metallic fuel has also inherent reactivity feedback characteristic that the metallic fuel SFR can be shutdown safely in the events of transient overpower, loss of flow, and loss of heat sink without scram. The metallic fuel has, however, lower melting point due to the eutectic formation between the uranium in metallic fuel and the ferrite in metallic cladding. It is needed to identify the core damage accident scenarios to develop Level-1 PSA model. SSC-K computer code is used to identify the conditions in which the core damage can occur in the KALIMER-600 SFR. The accident cases which are analyzed are the triple failure accidents such as unprotected transient over power events, loss of flow events, and loss of heat sink events with impaired safety systems or functions. Through the analysis of the triple failure accidents for the KALIMER-600 SFR, it is found that the PSA model developed for the PRISM reactor design can be applied to KALIMER-600. However

Under sodium reliability tests on core components and in-core instrumentation

International Nuclear Information System (INIS)

Ruppert, E.; Stehle, H.; Vinzens, K.

1977-01-01

A sodium test facility for fast breeder core components (AKB), built by INTERATOM at Bensberg, has been operating since 1971 to test fuel dummies and blanket elements as well as absorber elements under simulated normal and extreme reactor conditions. Individual full-scale fuel or blanket elements and arrays of seven elements, modelling a section of the SNR-300 reactor core, have been tested under a wide range of sodium mass flow and isothermal test conditions up to 925K as well as under cyclic changed temperature transients. Besides endurance testing of the core components a special sodium and high-temperature instrumentation is provided to investigate thermohydraulic and vibrational behaviour of the test objects. During all test periods the main subassembly characteristics could be reproduced and the reliability of the instrumentation could be proven. (orig.) [de

CANDU fuel : design/manufacturing interaction

International Nuclear Information System (INIS)

Graham, N.A.

1999-01-01

The design of CANDU fuel has been the product of intense cooperation among fuel designers and fuel manufacturers. The developments of some of the novel processes in fuel manufacture are outlined. These include the brazed-split-spacer design, the resistance welded endcap and CANLUB coatings. (author)

Fuel swelling and interaction layer formation in the SELENIUM Si and ZrN coated U(Mo) dispersion fuel plates irradiated at high power in BR2

Energy Technology Data Exchange (ETDEWEB)

Leenaers, A., E-mail: aleenaer@sckcen.be [Nuclear Materials Science Institute, SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Van den Berghe, S.; Koonen, E.; Kuzminov, V. [Nuclear Materials Science Institute, SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Detavernier, C. [Department of Solid State Sciences, Ghent University, Krijgslaan 281/S1, 9000 Ghent (Belgium)

2015-03-15

In the framework of the SELENIUM project two full size flat fuel plates were produced with respectively Si and ZrN coated U(Mo) particles and irradiated in the BR2 reactor at SCK• CEN. Non-destructive analysis of the plates showed that the fuel swelling profiles of both SELENIUM plates were very similar to each other and none of the plates showed signs of pillowing or excessive swelling at the end of irradiation at the highest power position (local maximum 70% {sup 235}U). The microstructural analysis showed that the Si coated fuel has less interaction phase formation at low burn-up but at the highest burn-ups, defects start to develop on the IL–matrix interface. The ZrN coated fuel, shows a virtual absence of reaction between the U(Mo) and the Al, up to high fission densities after which the interaction layer formation starts and defects develop in the matrix near the U(Mo) particles. It was found and is confirmed by the SELENIUM (Surface Engineering of Low ENrIched Uranium–Molybdenum) experiment that there are two phenomena at play that need to be controlled: the formation of an interaction layer and swelling of the fuel. As the interaction layer formation occurs at the U(Mo)–matrix interface, applying a diffusion barrier (coating) at that interface should prevent the interaction between U(Mo) and the matrix. The U(Mo) swelling, observed to proceed at an accelerating rate with respect to fission density accumulation, is governed by linear solid state swelling and fission gas bubble swelling due to recrystallization of the fuel. The examination of the SELENIUM fuel plates clearly show that for the U(Mo) dispersion fuel to be qualified, the swelling rate at high burn-up needs to be reduced.

Fuel swelling and interaction layer formation in the SELENIUM Si and ZrN coated U(Mo) dispersion fuel plates irradiated at high power in BR2

Science.gov (United States)

Leenaers, A.; Van den Berghe, S.; Koonen, E.; Kuzminov, V.; Detavernier, C.

2015-03-01

In the framework of the SELENIUM project two full size flat fuel plates were produced with respectively Si and ZrN coated U(Mo) particles and irradiated in the BR2 reactor at SCK•CEN. Non-destructive analysis of the plates showed that the fuel swelling profiles of both SELENIUM plates were very similar to each other and none of the plates showed signs of pillowing or excessive swelling at the end of irradiation at the highest power position (local maximum 70% 235U). The microstructural analysis showed that the Si coated fuel has less interaction phase formation at low burn-up but at the highest burn-ups, defects start to develop on the IL-matrix interface. The ZrN coated fuel, shows a virtual absence of reaction between the U(Mo) and the Al, up to high fission densities after which the interaction layer formation starts and defects develop in the matrix near the U(Mo) particles. It was found and is confirmed by the SELENIUM (Surface Engineering of Low ENrIched Uranium-Molybdenum) experiment that there are two phenomena at play that need to be controlled: the formation of an interaction layer and swelling of the fuel. As the interaction layer formation occurs at the U(Mo)-matrix interface, applying a diffusion barrier (coating) at that interface should prevent the interaction between U(Mo) and the matrix. The U(Mo) swelling, observed to proceed at an accelerating rate with respect to fission density accumulation, is governed by linear solid state swelling and fission gas bubble swelling due to recrystallization of the fuel. The examination of the SELENIUM fuel plates clearly show that for the U(Mo) dispersion fuel to be qualified, the swelling rate at high burn-up needs to be reduced.

Fuel burn analysis of a sodium fast reactor with KANEXT and Serpent; Analisis de quemado de combustible de un reactor rapido de sodio con KANEXT y SERPENT

Energy Technology Data Exchange (ETDEWEB)

Lopez S, R. C.; Francois L, J. L., E-mail: rcarlos.lope@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

2015-09-15

The fast reactors cooled by sodium are one of the options considered in the Generation IV. Since most of the reactors of Fourth Generation are still in development stage, is necessary to have efficient and reliable computational tools, this in order to obtain accurate results in reasonable computational times. In this paper is introduced and describes the deterministic code KANEXT (KArlsruhe Neutronic EXtended Tool) and is compared against a Monte Carlo code of more diffusion: Serpent. KANEXT, being a modular code requires the interaction of different modules to perform a job, this interaction of modules is described in this article. The parameters to be compared are the results of the neutron multiplication effective factor and the evolution of isotopes during the burning. The mentioned comparison is carried out for a fast reactor cooled by sodium of relatively small size compared to commercial size reactors. In this paper the particularities of the reactor are described, important for the analysis such as geometry, enrichments, reflector, etc. The considerations in the implementation in both codes are also described, as are simplifications, length of the burning steps, possible solutions of the Bateman equations for the burning fuel in Serpent and the solution options for transport (P3) and diffusion (P1) in KANEXT. The results show good correspondence between Serpent and KANEXT, which give confidence to continue using KANEXT as the main tool. Respect to computation time, time saving is evident with the use of deterministic codes instead of Monte Carlo codes, in this particular case, the time savings using KANEXT is about 98.5% of the time used by Serpent. (Author)

Studying the processes of sodium-water interaction in the BOR-60 reactor micromodule steam generator

International Nuclear Information System (INIS)

Tsykanov, V.A.; Antipin, G.K.; Borisov, V.V.

1981-01-01

Main results of experimental studies of emergency regimes of micromodule steam generator (MSG) at small and big leaks of water into sodium, realized using the 30 MW MSG, operating in the BOR-o0 reactor, are considered. The aims of the study are as follows: the modelling of macroleak in ''Nadja'' steam generator for the BN-350 reactor; testing the conceptions of alarm signalling and MSG protection; testing under real conditions of new perspective systems of leak detection; gaining the experimence and development of the ways to eliminate the consequences of accident caused by big water leak into sodium; accumulation of knowledge on restoration of MSG operating ability after accident; experimental test of calculational techniques for big leak accidents to use them in future for calculational studies of similar situations at other reactors equipped with sodium-water steam generators; refinement of characteristics of hydrodynamic and thermal effects interaction zone for big leak in real circuit during the plant operation. A series of experiments with the imitation of water leak into sodium by means of argon and steam supply through injection devices, located before the steam superheater module of one of the sections and between evaporator module of the same section, is conducted. The range of steam flow rate is 0.02-0.45 g/s. Duration of steam supply is 100-400 s. A conclusion is made that the results obtained can be used for steam generator of the BN-350 reactor [ru

Potentiometric and spectrophotometric titration study of interaction of tungstovanadophosphoric heteropolyacids with sodium hydroxide

International Nuclear Information System (INIS)

Borkoyakov, S.A.; Fisun, L.A.

1988-01-01

The methods of potentiometric and spectrophotometric titration are used to study H s+n PW 12-n V n O 40 (n=1,2) (P-W-V HPA) decomposition by sodium hydroxide. It is shown that at the first stage of heteropolyanion interaction with alkali (pH) > 4 P-W-V HPA structural reconstruction takes place. It is accompanied by the formation of complexes with a higher content of vanadium atoms stable at pH 4-7/ P-W-V HPA decomposition to initial salts occurs at pH > 8

FRACAS: a subcode for the analysis of fuel pellet-cladding mechanical interaction

International Nuclear Information System (INIS)

Bohn, M.P.

1977-04-01

This report describes FRACAS (Fuel Rod and Cladding Analysis Subcode), a computer code which performs the mechanical analysis in the FRAP fuel rod codes. At each loadstep, FRACAS obtains a complete elastic-plastic-creep solution for the stresses, strains, and displacements in the fuel rod cladding. The cladding is modeled as a thin cylindrical shell with prescribed temperature, pressures, and radial displacement of the inside surface. The displacement of the fuel pellets is assumed to be due to thermal gradients only. Three different regimes of pellet-cladding mechanical interaction are considered: (a) open gap, (b) closed gap, and (c) trapped stack. Both transient and steady state creep calculations are performed. The capabilities of the code are illustrated by an example problem, and comparisons are made with data obtained from two experimental fuel rods

Large scale sodium interactions. Part 3. Chemical phenomena with limestone concrete

International Nuclear Information System (INIS)

Sallach, R.A.

1977-01-01

The description of the chemical processes and reaction products resulting from the exposure of concrete to molten sodium metal is important for a thorough, realistic assessment of the safety of CRBR-type reactors. Concretes are in general complex heterogenous substances whose ingredients can be derived from many sources. Consequently a wide variety of reaction processes and products might be anticipated. Initial attention has focused on a concrete in which both the aggregate and sandy components are derived from limestone. Presented are the chemical observations and experimental data from tests in which molten sodium metal at approximately 500 0 C is dropped into cold limestone concrete crucibles. Thermocouples immersed in the sodium pool indicate that the reaction proceeds in two stages. In the first stage which lasts 5 to 8 minutes, the temperature of the reacting mass hovers around 500 0 C. This stage is followed by a second stage of longer duration--greater than 100 minutes--where the temperature is 700 to 800 0 C. The main reaction product is a hard, fused, black slag which contains about 3/4 of the sodium in the initial charge. A secondary product is sodium oxide aerosol which accounts for the remaining 1/4 of the charge. It is significant that no free sodium metal is found in the slag; all sodium has completely reacted

Fragmentation of a single molten copper and silver droplets penetrating a sodium pool with solid crust

International Nuclear Information System (INIS)

Wataru Itagaki; Ken-ichiro Sugiyama; Satoshi Nishimura; Izumi Kinoshita

2005-01-01

As a basic study of molten fuel-coolant interaction in liquid metal fast cooled reactors, we carried out a series of experiments for the fragmentation of molten copper droplet penetrating sodium pool at instantaneous contact interface temperatures below its freezing point. A single molten copper droplet with 5g in weight and with superheating varied from 0 degree C to 131 degree C was dropped into a sodium pool in a wide range of ambient Weber numbers 24 to 228. In addition to the experiment of molten copper droplet, molten silver droplet with 5gs in weight and with superheating varied from 3 degree C to 174 degree C was dropped into the sodium pool at an ambient Weber number of about 80. From the observation of the cross section of solidified silver droplet without fragmentation, it was clearly confirmed that sodium micro jet is driven into the inside from the upper surface of molten droplet keeping liquid phase, which is clear evidence for the thermal fragmentation mechanism proposed in the previous paper. Large scattering in the values of dimensionless mass median diameter observed in the present experimental study is recognized to be dependent on whether latent heat instantaneously released due to the injection of sodium micro jet can be effectively utilized for fragmentation. (authors)

Evaluation of molten lead mixing in sodium coolant by diffusion for application to PAHR

International Nuclear Information System (INIS)

Chawla, T.C.; Pedersen, D.R.; Leaf, G.; Minkowycz, W.J.

1983-01-01

In post-accident heat removal (PAHR) applications the use of a lead slab is being considered for protecting a porous bed of steel shots in ex-vessel cavity from direct impingement of molten steel or fuel upon vessel failure following a hypothetical core dissembly accident in an LMFBR. The porous bed is provided to increase coolability of the fuel debris by the sodium coolant. The objectives of the present study are (1) to determine melting rates of lead slabs of various thicknesses in contact with sodium coolant and (2) to evaluate the extent of penetration and mixing rates of molten lead into sodium coolant by molecular diffusion alone

Quality evaluation and in vitro interaction studies between levofloxacin 250mg and diclofenac sodium 50mg tablets.

Science.gov (United States)

Fayyaz, Muhammad; Yousuf, Rabia Ismail; Shoaib, Muhammad Harris; Ali, Tariq; Nasiri, Iqbal; Ashraf, Nida

2015-01-01

Fluoroquinolones are broad-spectrum antibiotics, work against Gram-positive and Gram-negative bacteria and are a clinically proven option for many resistant infections. Among fluoroquinolones Levofloxacin works best against acute sinusitis, inflammation of the lower airways, acute exacerbation of chronic bronchitis, community acquired pneumonia, complicated urinary tract infection including Pyelonephritis, chronic bacterial prostatitis and skin and soft tissue infection. Levofloxacin is a frequently prescribed antibacterial agent with Diclofenac Sodium for pain management in infectious conditions. The objective of the present work is to evaluate the level of interaction between Levofloxacin and Diclofenac Sodium. In this work market available brands of both drugs were also evaluated for quality.The physiochemical parameters like weight variation, thickness variation, and mechanical strength were determined. Similarly the percentage drug release and content uniformity test were also analyzed; the tested quality attributes were found within the recommended pharmacopeia ranges except brand L(6) that had high drug content 124.629±3.614 while brand L(4) and L(5) were not found similar in pH 1.2. When subjected to model dependent analysis Levofloxacin showed compliance with (first order, Higuchi, Hixson Crowell and Weibull) at pH (1.2, 4.5 and 6.8). However Diclofenac Sodium showed adherence with (first order, Hixson Crowell and Weibull) at pH (1.2, 4.5 and 6.8) but following Higuchi at pH 1.2 and 4.5 only. The interaction studies were also performed spectrophotometrically and simultaneous equation was used to estimate the percentage availability of both the drugs at pH 4.5, 6.8, FaSSGF and FaSSIF. The studies showed that the percent availability of Levofloxacin was increased significantly in FaSSIF i.e. 129.173±0.323 at 45 minutes in the presence of Diclofenac Sodium.

HOT CELL SYSTEM FOR DETERMINING FISSION GAS RETENTION IN METALLIC FUELS

Energy Technology Data Exchange (ETDEWEB)

Sell, D. A.; Baily, C. E.; Malewitz, T. J.; Medvedev, P. G.; Porter, D. L.; Hilton, B. A.

2016-09-01

A system has been developed to perform measurements on irradiated, sodium bonded-metallic fuel elements to determine the amount of fission gas retained in the fuel material after release of the gas to the element plenum. During irradiation of metallic fuel elements, most of the fission gas developed is released from the fuel and captured in the gas plenums of the fuel elements. A significant amount of fission gas, however, remains captured in closed porosities which develop in the fuel during irradiation. Additionally, some gas is trapped in open porosity but sealed off from the plenum by frozen bond sodium after the element has cooled in the hot cell. The Retained fission Gas (RFG) system has been designed, tested and implemented to capture and measure the quantity of retained fission gas in characterized cut pieces of sodium bonded metallic fuel. Fuel pieces are loaded into the apparatus along with a prescribed amount of iron powder, which is used to create a relatively low melting, eutectic composition as the iron diffuses into the fuel. The apparatus is sealed, evacuated, and then heated to temperatures in excess of the eutectic melting point. Retained fission gas release is monitored by pressure transducers during the heating phase, thus monitoring for release of fission gas as first the bond sodium melts and then the fuel. A separate hot cell system is used to sample the gas in the apparatus and also characterize the volume of the apparatus thus permitting the calculation of the total fission gas release from the fuel element samples along with analysis of the gas composition.

Corrosion behavior of sodium-exposed stainless steels in chloride-containing aqueous solutions

International Nuclear Information System (INIS)

Wu, P.C.S.; Grundy, B.R.; Miller, R.L.

1979-01-01

The corrosion behavior of sodium-exposed stainless steels in chloride-containing aqueous solutions was investigated. Results showed that sodium-corroded Type 316 stainless steel (prototypic Liquid Metal Fast Breeder Reactor (LMFBR) fuel cladding) maintains its integrity after five months exposure in these solutions at 82 0 C and with chloride content up to 500 ppM. In contrast, sensitized and sodium mass transfer deposit-containing Type 304 stainless steel failed in the high chloride solution (500 ppM) within ten days at the same temperature. The failure was initiated by pitting and subsequently accelerated by intergranular attack. The results also show that high pH tends to reduce the susceptibility to failure while procedures commonly used for sodium removal have no significant effect on the water corrosion behavior of the test material. Based on the current results, it is concluded that water shortage is feasible for spent fuels in a LMFBR reprocessing plant

The study of flow and proton exchange interactions in the cylindrical solid oxide fuel cell

International Nuclear Information System (INIS)

Saievar-Iranizad, E.; Malekifar, A.

2002-01-01

The solid oxide fuel cell operates at high temperature of about 1000 deg C. In this temperature, some known materials such as Ni, ... which is abundant in the nature, can be used as a catalyst in the electrodes. The electrolytes of such cell solid oxide fuel cell can be made through non-porous solid ceramics such as Zircon's (ZrO 2 ). It can be stabilized using a doped Yttrium oxide. The importance of Yttria-stabilised Zirconia at high temperature belongs to the transport of oxygen ions through the electrolyte. Oxygen using in the hot cathode side causes a considerable reduction in the concentration of oxygen molecules. The oxygen ions exchange through the electrolyte relates to the molecular oxygen concentration gradient between the anode and cathode. Applying fuels such as hydrogen or natural gas in the anode and its chemical reaction with oxygen ions transfer from cathode through the electrolyte, produce electricity, water and heat. To study the ion exchange and its interaction into solid oxide fuel cell, a mathematical model had been considered in this article. This model simulates and illustrates the interaction, diffusion and oxygen ions exchange into fuel cell. The electrical power of fuel cell due to the ion exchange can be obtained using a simulation method. The ion exchange simulation, diffusion of molecules, their interactions and system development through the mathematical model has been discussed in this paper

Fuel-coolant interactions: preliminary experiments on the effect of gases dissolved in the 'coolant'

International Nuclear Information System (INIS)

Asher, R.C.; Davies, D.; Jones, P.G.

1976-12-01

A simple apparatus has been used to study fuel-coolant interactions under reasonably well controlled conditions. Preliminary experiments have used water as the 'coolant' and molten tin at 800 0 C as the 'fuel' and have investigated how the violence of the interaction is affected by dissolving gases (oxygen, nitrogen, carbon dioxide and nitrous oxide) in the water. It was found that saturating the water with carbon dioxide or nitrous oxide completely suppresses the violent interaction. Experiments in which the concentrations of these gases were varied showed that a certain critical concentration was needed; below this concentration the dissolved gas has no significant effect but above it the suppression is

Fuel Cycle Concept with Advanced METMET and Composite Fuel in LWRs

International Nuclear Information System (INIS)

Savchenko, A.; Skupov, M.; Vatulin, A.; Glushenkov, A.; Kulakov, G.; Lipkina, K.

2014-01-01

The basic factor that limits the serviceability of fuel elements developing in the framework of RERTR Program (transition from HEU to LEU fuel of research reactors) is interaction between U10Mo fuel and aluminium matrix . Interaction results in extra swelling of fuels, disappearance of a heat conducting matrix, a temperature rise in the fuel centre, penetration porosity, etc. Several methods exist to prevent fuel-matrix interaction. In terms of simplifying fuel element fabrication technology and reducing interaction, doping of fuel is the most optimal version

Blockages in LMFBR fuel assemblies: a review