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)

Influence of oxygen-metal ratio on mixed-oxide fuel performance

International Nuclear Information System (INIS)

Lawrence, L.A.; Leggett, R.D.

1979-04-01

The fuel oxygen-to-metal ratio (O/M) is recognized as an important consideration for performance of uranium--plutonium oxide fuels. An overview of the effects of differing O/M's on the irradiation performance of reference design mixed-oxide fuel in the areas of chemical and mechanical behavior, thermal performance, and fission gas behavior is presented. The pellet fuel has a nominal composition of 75 wt% UO 2 + 25 wt% PuO 2 at a pellet density of approx. 90% TD. for nominal conditions this results in a smeared density of approx. 85%. The cladding in all cases is 20% CW type 316 stainless steel with an outer diameter of 5.84 to 6.35 mm. O/M has been found to significantly influence fuel pin chemistry, mainly FCCI and fission product and fuel migration. It has little effect on thermal performance and overall mechanical behavior or fission gas release. The effects of O/M (ranging from 1.938 to 1.984) in the areas of fuel pin chemistry, to date, have not resulted in any reduction in fuel pin performance capability to goal burnups of approx. 8 atom% or more

Irradiation of mixed UO2-PuO2 oxide samples for fast neutron reactor fuel elements

International Nuclear Information System (INIS)

Mikailoff, H.; Mustelier, J.; Bloch, J.; Conte, M.; Hayet, L.; Lauthier, J.C.; Leclere, J.

1968-01-01

Thermal flux irradiation testings of small mixed oxide pellets UPuO 2 fuel elements were performed in support of the fuel reference design for the Phenix fast reactor. The effects of different parameters (stoichiometry, pellet density, pellet clad gap). on the behaviour of the oxide (temperature distribution, microstructural changes, fission gas release) were investigated in various irradiation conditions. In particular, the effect of fuel density decrease and power rate increase on thermal performances were determined on short term irradiations of porous fuels. (authors) [fr

Fabrication experience with mixed-oxide LWR fuels at the BELGONUCLEAIRE plant

International Nuclear Information System (INIS)

Vanhellemont, G.

1979-01-01

For nearly 20 years BELGONUCLEAIRE has been involved in a steadily growing effort to increase its production of mixed oxides. This programme has ranged from basic research and process development through a pilot-scale unit to today's mixed-oxide fuel fabrication plant at Dessel, which has been in operation for just over 5 years. The reference fabrication flow sheet includes UO 2 , PuO 2 and a scraped powder preparation, sintered ground pellets as well as rod fabrication and assembling. With regard to quality, attention is especially paid to the process monitoring and quality controls at the qualification step and during the routine production. Entirely different types of thermal UO 2 -PuO 2 fuel pellets, rods and assemblies have been manufactured for PWR and BWR operation. For these fabrications, some diagrams of the results with regard to the required technical specifications are presented. Special emphasis is placed on the occasional deviations of some finished products from the specifications and on the solutions applied to avoid such problems. Concerning the actual capacity of the mixed-oxide fuel fabrication plant, several limiting factors due to the nature of plutonium itself are discussed. Taking into account all these ambient limitations, a reference PWR mixed-oxide fuel output of nominally 18 t/a is obtained. The industrial feasibility of UO 2 -PuO 2 fuel fabrication has been thoroughly demonstrated by the present BELGONUCLEAIRE plant. The experience obtained has led to progressive improvements of the fabrication process and adaptation of the product controls in order to ensure the requested quality levels. (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

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)

Sodium-fuel interaction: dropping experiments and subassembly test

International Nuclear Information System (INIS)

Holtbecker, H.; Schins, H.; Jorzik, E.; Klein, K.

1978-01-01

Nine dropping tests, which bring together 2 to 4 kg of molten UO 2 with 150 l sodium, showed the incoherency and non-violence of these thermal interactions. The pressures can be described by sodium incipient boiling and bubble collapse; the UO 2 fragmentation by thermal stress and bubble collapse impact forces. The mildness of the interaction is principally due to the slowness and incoherency of UO 2 fragmentation. This means that parametric models which assume instantaneous mixing and fragmentation are of no use for the interpretation of dropping experiments. One parametric model, the Caldarola Fuel Coolant Interaction Variable Mass model, is being coupled to the two dimensional time dependent hydrodynamic REXCO-H code. In a first step the coupling is applicated to a monodimensional geometry. A subassembly test is proposed to validate the model. In this test rapid mixing between UO 2 and sodium has to be obtained. Dispersed molten UO 2 fuel is obtained by flashing injected sodium drops inside a UO 2 melt. This flashing is theoretically explained and modelled as a superheat limited explosion. The measured sodium drop dwell times of two experiments are compared to results obtained from the mentioned theory, which is the basis of the Press 2 Code

Fabrication of mixed oxide fuel using plutonium from dismantled weapons

International Nuclear Information System (INIS)

Blair, H.T.; Chidester, K.; Ramsey, K.B.

1996-01-01

A very brief summary is presented of experimental studies performed to support the use of plutonium from dismantled weapons in fabricating mixed oxide (MOX) fuel for commercial power reactors. Thermal treatment tests were performed on plutonium dioxide powder to determine if an effective dry gallium removal process could be devised. Fabrication tests were performed to determine the effects of various processing parameters on pellet quality. Thermal tests results showed that the final gallium content is highly dependent on the treatment temperature. Fabrication tests showed that the milling process, sintering parameters, and uranium feed did effect pellet properties. 1 ref., 1 tab

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

A method for determining an effective porosity correction factor for thermal conductivity in fast reactor uranium-plutonium oxide fuel pellets

International Nuclear Information System (INIS)

Inoue, Masaki; Abe, Kazuyuki; Sato, Isamu

2000-01-01

A reliable method has been developed for determining an effective porosity correction factor for calculating a realistic thermal conductivity for fast reactor uranium-plutonium (mixed) oxide fuel pellets. By using image analysis of the ceramographs of transverse sections of mixed-oxide fuel pellets, the fuel morphology could be classified into two basic types. One is a 'two-phase' type that consists of small pores dispersed in the fuel matrix. The other is a 'three-phase' type that has large pores in addition to the small pores dispersed in the fuel matrix. The pore sizes are divided into two categories, large and small, at the 30 μm area equivalent diameter. These classifications lead to an equation for calculating an effective porosity correction factor by accounting for the small and large pore volume fractions and coefficients. This new analytical method for determining the effective porosity correction factor for calculating the realistic thermal conductivity of mixed-oxide fuel was also experimentally confirmed for high-, medium- and low-density fuel pellets

Development of advanced mixed oxide fuels for plutonium management

International Nuclear Information System (INIS)

Eaton, S.; Beard, C.; Buksa, J.; Butt, D.; Chidester, K.; Havrilla, G.; Ramsey, K.

1997-01-01

A number of advanced Mixed Oxide (MOX) fuel forms are currently being investigated at Los Alamos National Laboratory that have the potential to be effective plutonium management tools. Evolutionary Mixed Oxide (EMOX) fuel is a slight perturbation on standard MOX fuel, but achieves greater plutonium destruction rates by employing a fractional nonfertile component. A pure nonfertile fuel is also being studied. Initial calculations show that the fuel can be utilized in existing light water reactors and tailored to address different plutonium management goals (i.e., stabilization or reduction of plutonium inventories residing in spent nuclear fuel). In parallel, experiments are being performed to determine the feasibility of fabrication of such fuels. Initial EMOX pellets have successfully been fabricated using weapons-grade plutonium. (author)

Development of advanced mixed oxide fuels for plutonium management

International Nuclear Information System (INIS)

Eaton, S.; Beard, C.; Buksa, J.; Butt, D.; Chidester, K.; Havrilla, G.; Ramsey, K.

1997-06-01

A number of advanced Mixed Oxide (MOX) fuel forms are currently being investigated at Los Alamos National Laboratory that have the potential to be effective plutonium management tools. Evolutionary Mixed Oxide (EMOX) fuel is a slight perturbation on standard MOX fuel, but achieves greater plutonium destruction rates by employing a fractional nonfertile component. A pure nonfertile fuel is also being studied. Initial calculations show that the fuel can be utilized in existing light water reactors and tailored to address different plutonium management goals (i.e., stabilization or reduction of plutonium inventories residing in spent nuclear fuel). In parallel, experiments are being performed to determine the feasibility of fabrication of such fuels. Initial EMOX pellets have successfully been fabricated using weapons-grade plutonium

Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet

International Nuclear Information System (INIS)

Amaya, Masaki; Nakamura, Jinichi; Nagase, Fumihisa; Fuketa, Toyoshi

2011-01-01

The thermal conductivity formula of fuel pellet which contains the effects of burnup and plutonium (Pu) addition was proposed based on the Klemens' theory and reported thermal conductivities of unirradiated (U, Pu) O 2 and irradiated UO 2 pellets. The thermal conductivity of high burnup MOX pellet was formulated by applying a summation rule between phonon scattering parameters which show the effects of plutonium addition and burnup. Temperature of high burnup MOX fuel was evaluated based on the thermal conductivity integral which was calculated from the above-mentioned thermal conductivity formula. Calculated fuel temperatures were plotted against the linear heat rates of the fuel rods, and were compared with the fuel temperatures measured in a test reactor. Since both values agreed well, it was confirmed that the proposed thermal conductivity formula of MOX pellets is adequate.

The thermal conductivity of mixed fuel U_xPu_1-xO₂: molecular dynamics simulations

Energy Technology Data Exchange (ETDEWEB)

Liu, Xiang-Yang [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Cooper, Michael William Donald [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanek, Christopher Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Andersson, Anders David Ragnar [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-10-16

Mixed oxides (MOX), in the context of nuclear fuels, are a mixture of the oxides of heavy actinide elements such as uranium, plutonium and thorium. The interest in the UO₂-PuO₂ system arises from the fact that these oxides are used both in fast breeder reactors (FBRs) as well as in pressurized water reactors (PWRs). The thermal conductivity of UO₂ fuel is an important material property that affects fuel performance since it is the key parameter determining the temperature distribution in the fuel, thus governing, e.g., dimensional changes due to thermal expansion, fission gas release rates, etc. For this reason it is important to understand the thermal conductivity of MOX fuel and how it differs from UO₂. Here, molecular dynamics (MD) simulations are carried out to determine quantitatively, the effect of mixing on the thermal conductivity of U_xPu_1-xO₂, as a function of PuO₂ concentrations, for a range of temperatures, 300 – 1500 K. The results will be used to develop enhanced continuum thermal conductivity models for MARMOT and BISON by INL. These models express the thermal conductivity as a function of microstructure state-variables, thus enabling thermal conductivity models with closer connection to the physical state of the fuel.

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.

Dissolution of mixed oxide fuel as a function of fabrication variables

International Nuclear Information System (INIS)

Lerch, R.E.

1979-08-01

Dissolution properties of mechanically blended mixed oxide fuel were very dependent on the six fuel fabrication variables studied. Fuel sintering temperature, source of PuO 2 and PuO 2 content of the fuel had major effects: (1) as the sintering temperature was increased from 1400 to 1700 0 C, pellet dissolution was more complete; (2) pellets made from burned metal derived PuO 2 were more completely dissolved than pellets made from calcined nitrate derived PuO 2 which in turn were more completely dissolved than pellets made from calcined nitrate derived PuO 2 ; (3) as the PuO 2 content decreased from 25 to 15 wt % PuO 2 , pellet dissolution was more complete. Preferential dissolution of uranium occurred in all the mechanically blended mixed oxide. Unirradiated mixed oxide fuel pellets made by the Sol Gel process were generally quite soluble in nitric acid. Unirradiated mixed oxide fuel pellets made by the coprecipitation process dissolved completely and rapidly in nitric acid. Fuel made by the coprecipitation process was more completely dissolved than fuel made by the Sol Gel process which, in turn, was more completely dissolved than fuel made by mechanically blending UO 2 and PuO 2 as shown below. Addition of uncomplexed fluoride to nitric acid during fuel dissolution generally rendered all fuel samples completely dissolvable. In boiling 12M nitric acid, 95 to 99% of the plutonium which was going to dissolve did so in the first hour. Irradiated mechanically blended mixed oxide fuel with known fuel fabrication conditions was also subjected to fuel dissolution tests. While irradiation was shown to increase completeness of plutonium dissolution, poor dissolubility due to adverse fabrication conditions (e.g., low sintering temperature) remained after irradiation

Kinetics study of thermal decomposition of sodium carboxylate salts

International Nuclear Information System (INIS)

Landoll, Michael P.; Holtzapple, Mark T.

2012-01-01

The MixAlco™ process ferments lignocellulosic biomass to carboxylate salts that are thermally decomposed into ketones, which are then chemically converted to a wide variety of chemicals and fuels. To perform these decompositions, suitable reaction models are necessary to properly design, scale, and optimize commercial reactors. For three salt types (sodium acetate, and two types of mixed sodium carboxylate salts), activation energy was determined using three isoconversional methods that employed TGA curves at different heating rates. For all three salt types, activation energy varied significantly with conversion. The average activation energy for sodium acetate was 226.65 kJ/mol, and the activation energies for the two mixed sodium carboxylate salts were 195.61, and 218.18 kJ/mol. In addition, three functions of conversion were employed to see which one best modeled the experimental data. The Sestak-Berggren model fits all three salt types best. -- Highlights: ► Sodium carboxylate salts from fermentation broth thermally decompose to ketones. ► Activation energy varies with conversion for all three salt types. ► Sestak-Berggren model provides best fit for all three salt types.

Transmutation of minor actinide using BWR fueled mixed oxide

International Nuclear Information System (INIS)

Susilo, Jati

2000-01-01

Nuclear spent fuel recycle has a strategic importance in the aspect of nuclear fuel economy and prevention of its spread-out. One among other application of recycle is to produce mixed oxide fuel (Mo) namely mixed Plutonium and uranium oxide. As for decreasing the burden of nuclear high level waste (HLW) treatment, transmutation of minor actinide (MA) that has very long half life will be carried out by conversion technique in nuclear reactor. The purpose of this study was to know influence of transition fuel cell regarding the percent weight of transmutation MA in the BWR fueled MOX. Calculation of cell BWR was used SRAC computer code, with assume that the reactor in equilibrium. The percent weight of transmutation MA to be optimum by increasing the discharge burn-up of nuclear fuel, raising ratio of moderator to fuel volume (Vm/Vf), and loading MA with percent weight about 3%-6% and also reducing amount of percent weight Pu in MOX fuel. For mixed fuel standard reactor, reactivity value were obtained between about -50pcm ∼ -230pcm for void coefficient and -1.8pcm ∼ -2.6pcm for fuel temperature coefficient

Plutonium spot of mixed oxide fuel, 2

International Nuclear Information System (INIS)

Suzuki, Yukio; Maruishi, Yoshihiro; Satoh, Masaichi; Aoki, Toshimasa; Muto, Tadashi

1974-01-01

In a fast reactor, the specification for the homogeneity of plutonium in plutonium-uranium mixed-oxide fuel is mainly dependent on the nuclear characteristics, whereas in a thermal reactor, on thermal characteristics. This homogeneity is measured by autoradiography as the plutonium spot size of the specimens which are arbitrarily chosen fuel pellets from a lot. Although this is a kind of random sampling, it is difficult to apply this method to conventional digital standards including JIS standards. So a special sampling inspection method was studied. First, it is assumed that the shape of plutonium spots is spherical, the size distribution is logarithmic normal, and the standard deviation is constant. Then, if standard deviation and mean spot size are given, the logarithmic normal distribution is decided unitarily, and further if the total weight of plutonium spots for a lot of pellets is known, the number of the spots (No) which does not conform to the specification can be obtained. Then, the fraction defective is defined as No devided by the number of pellets per lot. As to the lot with such fraction defective, the acceptance coefficient of the lot was obtained through calculation, in which the number of sampling, acceptable diameter limit observed and acceptable conditions were used as parameters. (Tai, I.)

Critical experiments with mixed oxide fuel

International Nuclear Information System (INIS)

Harris, D.R.

1997-01-01

This paper very briefly outlines technical considerations in performing critical experiments on weapons-grade plutonium mixed oxide fuel assemblies. The experiments proposed would use weapons-grade plutonium and Er 2 O 3 at various dissolved boron levels, and for specific fuel assemblies such as the ABBCE fuel assembly with five large water holes. Technical considerations described include the core, the measurements, safety, security, radiological matters, and licensing. It is concluded that the experiments are feasible at the Rensselaer Polytechnic Institute Reactor Critical Facility. 9 refs

High temperature transient deformation of mixed oxide fuels

International Nuclear Information System (INIS)

Slagle, O.D.

1986-01-01

The purpose of this paper is to present recent experimental results on fuel creep under transient conditions at high temperatures. The effect of temperature, stress, heating rate, density and grain size were considered. An empirical formulation is derived for the relationship between strain, stress, temperature and heating rate. This relationship provides a means for incorporating stress relief into the analysis of fuel-cladding interaction during an overpower transient. The effect of sample density and initial grain size is considered by varying the sample parameters. Previously derived steady-state creep relationships for the high temperature creep of mixed oxide fuel were combined with the time dependency of creep found for UO 2 to calculate a transient creep relationship for mixed oxide fuel. These calculated results were found to be in good agreement with the measured high temperature transient creep results

Mixed oxide thermal behaviour at BOL: COMETHE III-J models and impact on power-to-melt

International Nuclear Information System (INIS)

Vliet, J. van

1979-01-01

The mixed oxide thermal behaviour at beginning of life is very important because it can impose a limitation to the fuel pin peak power, and therefore to the reactor thermal output. The relevant physical processes leading to fuel restructuring are modelled in COMETHE III-J in a kinetic way. This ensures that the temperature and power history are properly taken into account. These models are described and their impact on the calculated power to melt early in life is analysed. (author)

Thermal ramp rate effects on mixed-oxide fuel swelling/gas release

International Nuclear Information System (INIS)

Hinman, C.A.; Randklev, E.H.

1979-01-01

Macroscopic swelling behavior of PNL-10 was compared to that of PNL-2 fuel and it was found that the swelling-threshold behavior is similar for similar thermal conditions. Transient fission gas release for the PNL-10 fuel is very similar to that observed for the PNL-2 fuel for similar thermal conditions

Introduction of mixed oxide fuel elements in the belgian cores

International Nuclear Information System (INIS)

Charlier, A.F.; Hollasky, N.A.

1994-01-01

The important amount of plutonium recovered from the reprocessing of spent fuel on the one hand, the national and international experience of the use of mixed oxide UO 2 -PuO 2 fuel in power reactors on the other hand, have led Belgian utilities to decide the introduction of Mixed-Oxide fuel in Doel unit 3 and Tihange unit 2 cores. The 'MOX' project has shown that it was possible without reducing safety or requiring modifications of the plant equipment. It has been approved by the Belgian 'Nuclear Safety Commission'. (authors). 1 tab., 2 figs

Modern methods of material accounting for mixed oxide fuel fabrication facility

International Nuclear Information System (INIS)

Eggers, R.F.; Pindak, J.L.; Brouns, R.J.; Williams, R.C.; Brite, D.W.; Kinnison, R.R.; Fager, J.E.

1981-01-01

The generic requirements loss detection, and response to alarms of a contemporary material control and accounting (MCandA) philosophy have been applied to a mixed oxide fuel fabrication plant to produce a detailed preliminary MCandA system design that is generally applicable to facilities of this type. This paper summarizes and discusses detailed results of the mixed oxide fuel fabrication plant study

Modern methods of material accounting for mixed-oxide fuel-fabrication facility

International Nuclear Information System (INIS)

Eggers, R.F.; Brouns, R.J.; Brite, D.W.; Pindak, J.L.

1981-07-01

The generic requirements loss detection, and response to alarms of a contemporary material control and accounting (MC and A) philosophy have been applied to a mixed-oxide fuel-fabrication plant to produce a detailed preliminary MC and A system design that is generally applicable to facilities of this type. This paper summarizes and discusses detailed results of the mixed-oxide fuel-fabrication plant study. Topics covered in this paper include: mixed-oxide fuel-fabrication process description, process disaggregation into MC and A system control units, quantitative results of analysis of control units for abrupt and recurring loss-detection capability, impact of short- and long-term holdup on loss-detection capability, response to alarms for abrupt loss, and response to alarms for recurring loss

Plutonium bearing oxide fuels for recycling in thermal reactors and fast breeder reactors

International Nuclear Information System (INIS)

Cunningham, G.W.

1977-01-01

Programs carried out in the past two decades have established the technical feasibility of using plutonium as a fuel material in both water-cooled power reactors and sodium-cooled fast breeder reactors. The problem facing the technical community is basically one of demonstrating plutonium fuel recycle under strict conditions of public safety, accountability, personnel exposure, waste management, transportation and diversion or theft which are still evolving. In this paper only technical and economic aspects of high volume production and the demonstration program required are discussed. This paper discusses the role of mixed oxide fuels in light water reactors and the objectives of the LMFBR required for continual growth of nuclear power during the next century. The results of studies showing the impact of using plutonium on uranium requirements, power costs, and the market share of nuclear power are presented. The influence of doubling time and the introduction date of LMFBRs on the benefits to be derived by its commercial use are discussed. Advanced fuel development programs scoped to meet future commerical LMFBR fuel requirements are described. Programs designed to provide the basic technology required for using plutonium fuels in a manner which will satisfy all requirements for public acceptance are described. Included are the high exposure plutonium fabrication development program centered around the High Performance Fuels Laboratory being built at the Hanford Engineering Development Laboratory and the program to confirm the technology required for the production of mixed oxide fuels for light water reactors which is being coordinated by Savannah River Laboratories

Investigation of coolant thermal mixing within 28-element CANDU fuel bundles using the ASSERT-PV thermal hydraulics code

International Nuclear Information System (INIS)

Lightston, M.F.; Rock, R.

1996-01-01

This paper presents the results of a study of the thermal mixing of single-phase coolant in 28-element CANDU fuel bundles under steady-state conditions. The study, which is based on simulations performed using the ASSERT-PV thermal hydraulic code, consists of two main parts. In the first part the various physical mechanisms that contribute to coolant mixing are identified and their impact is isolated via ASSERT-PV simulations. The second part is concerned with development of a preliminary model suitable for use in the fuel and fuel channel code FACTAR to predict the thermal mixing that occurs between flow annuli. (author)

Behavior of molybdenum in mixed-oxide fuel

International Nuclear Information System (INIS)

Giacchetti, G.; Sari, C.

1976-01-01

Metallic molybdenum, Mo--Ru--Rh--Pd alloys, barium, zirconium, and tungsten were added to uranium and uranium--plutonium oxides by coprecipitation and mechanical mixture techniques. This material was treated in a thermal gradient similar to that existing in fuel during irradiation to study the behavior of molybdenum in an oxide matrix as a function of the O/(U + Pu) ratio and some added elements. Result of ceramographic and microprobe analysis shows that when the overall O/(U + Pu) ratio is less than 2, molybdenum and Mo--Ru--Rh--Pd alloy inclusions are present in the uranium--plutonium oxide matrix. If the O/(U + Pu) ratio is greater than 2, molybdenum oxidizes to MoO 2 , which is gaseous at a temperature approximately 1000 0 C. Molybdenum oxide vapor reacts with barium oxide and forms a compound that exists as a liquid phase in the columnar grain region. Molybdenum oxide also reacts with tungsten oxide (tungsten is often present as an impurity in the fuel) and forms a compound that contains approximately 40 wt percent of actinide metals. The apparent solubility of molybdenum in uranium and uranium--plutonium oxides, determined by electron microprobe, was found to be less than 250 ppM both for hypo- and hyperstoichiometric fuels

Gas cooled fast breeder reactors using mixed carbide fuel

International Nuclear Information System (INIS)

Kypreos, S.

1976-09-01

The fast reactors being developed at the present time use mixed oxide fuel, stainless-steel cladding and liquid sodium as coolant (LMFBR). Theoretical and experimental designing work has also been done in the field of gas-cooled fast breeder reactors. The more advanced carbide fuel offers greater potential for developing fuel systems with doubling times in the range of ten years. The thermohydraulic and physics performance of a GCFR utilising this fuel is assessed. One question to be answered is whether helium is an efficient coolant to be coupled with the carbide fuel while preserving its superior neutronic performance. Also, an assessment of the fuel cycle cost in comparison to oxide fuel is presented. (Auth.)

Mixed oxide fuel pellet and manufacturing method thereof

International Nuclear Information System (INIS)

Yuda, Ryoichi; Ito, Ken-ichi; Masuda, Hiroshi.

1993-01-01

In a method of manufacturing nuclear fuel pellets which comprises compression molding a mixed oxide powder containing UO 2 and PuO 2 followed by sintering, a sintering agent having a composition comprising about 40 to 80 wt% of SiO 2 and the balance of Al 2 O 3 is mixed to a mixed oxide at a ratio of about 40ppm to about 0.5 wt% based on the total amount of the mixed oxide and the sintering agent, to prepare a mixture. The mixture is molded into a compression product and then sintered at a weakly acidic atmosphere at a temperature of about 1500degC to 1800degC. With such procedures, the sintering agent forms an eutectic product of a single liquid phase, PuO 2 is dispersed over the entire region of the pellet by way of the liquid phase, formation of a solid solution phase is promoted to annihilate a free PuO 2 phase. Further, growth of crystal grains is promoted. Accordingly, since the MOX fuel pellets prepared according to the present invention have a uniform solid solution state, and no free PuO 2 phase remains, increase of FP gas emission due to local nuclear fission of Pu can be avoided. (T.M.)

Passive measurements of mixed-oxide fuel for nuclear nonproliferation

International Nuclear Information System (INIS)

Dolan, Jennifer L.; Flaska, Marek; Pozzi, Sara A.; Chichester, David L.

2013-01-01

We present new results on passive measurements and simulations of mixed-oxide fuel-pin assemblies. Potential tools for mixed-oxide fuel pin characterization are discussed for future nuclear-nonproliferation applications. Four EJ-309 liquid scintillation detectors coupled with an accurate pulse timing and digital, offline and optimized pulse-shape discrimination method were used. Measurement analysis included pulse-height distributions to distinguish between purely fission neutron sources and alpha-n plus fission neutrons sources. Time-dependent cross-correlation functions were analyzed to measure the fission neutron contribution to the measured sample's neutron source. The use of Monte Carlo particle transport code MCNPX-PoliMi is discussed in conjunction with the measurements

Thermal conductivity model of vibro-packed fuel

International Nuclear Information System (INIS)

Yeon Soo, Kim

2001-01-01

In an effort to dispose of excess weapons grade plutonium accumulated in the cold war era in the United States and the Russian Federation, one method currently under investigation is the conversion of the plutonium into mixed oxide (MOX) reactor fuel for LWRs and fast reactors in the Russian Federation. A fuel option already partly developed at the Research Institute of Atomic Reactors (RIAR) in Dimitrovgrad is that of vibro-packed MOX. Fuel rod fabrication using powder vibro-packing is attractive because it includes neither a process too complex to operate in glove boxes (or remotely), nor a waste-producing step necessary for the conventional pellet rod fabrication. However, because of its loose bonding between fuel particles at the beginning of life, vibro-packed MOX fuel has a somewhat less effective thermal conductivity than fully sintered pellet fuel, and undergoes more restructuring. Helium would also likely be pressurized in vibro-packed MOX fuel rods for LWRs to enhance initial fuel thermal conductivity. The combination of these two factors complicates development of an accurate thermal conductivity model. But clearly in order to predict fuel thermomechanical responses during irradiation of vibro-packed MOX fuel, fuel thermal conductivity must be known. The Vibropac fuel of interest in this study refers the fuel that is compacted with irregular fragments of mixed oxide fuel. In this paper, the thermal-conductivity models in the literature that dealt with relatively similar situations to the present case are examined. Then, the best model is selected based on accuracy of prediction and applicability. Then, the selected model is expanded to fit the various situations of interest. (author)

Sodium nitrate combustion limit tests

International Nuclear Information System (INIS)

Beitel, G.A.

1976-04-01

Sodium nitrate is a powerful solid oxidant. Energetically, it is capable of exothermically oxidizing almost any organic material. Rate-controlling variables such as temperature, concentration of oxidant, concentration of fuel, thermal conductivity, moisture content, size, and pressure severely limit the possibility of a self-supported exothermic reaction (combustion). The tests reported in this document were conducted on one-gram samples at atmospheric pressure. Below 380 0 C, NaNO 3 was stable and did not support combustion. At moisture concentrations above 22 wt percent, exothermic reactions did not propagate in even the most energetic and reactive compositions. Fresh resin and paraffin were too volatile to enable a NaNO 2 -supported combustion process to propagate. Concentrations of NaNO 3 above 95 wt percent or below 35 wt percent did not react with enough energy release to support combustion. The influence of sample size and confining pressure, both important factors, was not investigated in this study

Description of a reference mixed oxide fuel fabrication plant (MOFFP)

International Nuclear Information System (INIS)

1978-01-01

In order to evaluate the environment impact, due to the Mixed Oxide Fuel Fabrication Plants, work has been initiated to describe the general design and operating conditions of a reference Mixed Oxide Fuel Fabrication Plant (MOFFP) for the 1990 time frame. The various reference data and basic assumptions for the reference MOFFP plant have been defined after discussion with experts. The data reported in this document are only made available to allow an evaluation of the environmental impact due to a reference MOFFP plant. These data have therefore not to be used as recommandation, standards, regulatory guides or requirements

Safety aspects of LWR fuel reprocessing and mixed oxide fuel fabrication plants

International Nuclear Information System (INIS)

Fischer, M.; Leichsenring, C.H.; Herrmann, G.W.; Schueller, W.; Hagenberg, W.; Stoll, W.

1977-01-01

The paper is focused on the safety and the control of the consequences of credible accidents in LWR fuel reprocessing plants and in mixed oxide fuel fabrication plants. Each of these plants serve for many power reactor (about 50.000 Mwel) thus the contribution to the overall risk of nuclear energy is correspondingly low. Because of basic functional differences between reprocessing plants, fuel fabrication plants and nuclear power reactors, the structure and safety systems of these plants are different in many respects. The most important differences that influence safety systems are: (1) Both fuel reprocessing and fabrication plants do not have the high system pressure that is associated with power reactors. (2) A considerable amount of the radioactivity of the fuel, which is in the form of short-lived radionuclides has decayed. Therefore, fuel reprocessing plants and mixed oxide fuel fabrication plants are designed with multiple confinement barriers for control of radioactive materials, but do not require the high-pressure containment systems that are used in LWR plants. The consequences of accidents which may lead to the dispersion of radioactive materials such as chemical explosions, nuclear excursions, fires and failure of cooling systems are considered. A reasonable high reliability of the multiple confinement approach can be assured by design. In fuel reprocessing plants, forced cooling is necessary only in systems where fission products are accumulated. However, the control of radioactive materials can be maintained during normal operation and during the above mentioned accidents, if the dissolver off-gas and vessel off-gas treatment systems provide for effective removal of radioactive iodine, radioactive particulates, nitrogen oxides, tritium and krypton 85. In addition, the following incidents in the dissolver off-gas system itself must be controlled: failures of iodine filters, hydrogen explosion in O 2 - and NOsub(x)-reduction component, decomposition of

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

Sodium Borohydride/Hydrogen Peroxide Fuel Cells For Space Application

Science.gov (United States)

Valdez, T. I.; Deelo, M. E.; Narayanan, S. R.

2006-01-01

This viewgraph presentation examines Sodium Borohydride and Hydrogen Peroxide Fuel Cells as they are applied to space applications. The topics include: 1) Motivation; 2) The Sodium Borohydride Fuel Cell; 3) Sodium Borohydride Fuel Cell Test Stands; 4) Fuel Cell Comparisons; 5) MEA Performance; 6) Anode Polarization; and 7) Electrode Analysis. The benefits of hydrogen peroxide as an oxidant and benefits of sodium borohydride as a fuel are also addressed.

Fission gas behavior in mixed-oxide fuel during transient overpower

International Nuclear Information System (INIS)

Randklev, E.H.; Treibs, H.A.; Mastel, B.; Baldwin, D.L.

1979-01-01

Fission gas behavior can be important in determining fuel pin and core performance during a reactor transient. The results are presented of examinations characterizing the changes in microstructural distribution and retention of fission gas in fuel for a series of transient overpower (50 cents/s) tested mixed-oxide fuel pins and their steady state siblings

Using Coupled Mesoscale Experiments and Simulations to Investigate High Burn-Up Oxide Fuel Thermal Conductivity

Science.gov (United States)

Teague, Melissa C.; Fromm, Bradley S.; Tonks, Michael R.; Field, David P.

2014-12-01

Nuclear energy is a mature technology with a small carbon footprint. However, work is needed to make current reactor technology more accident tolerant and to allow reactor fuel to be burned in a reactor for longer periods of time. Optimizing the reactor fuel performance is essentially a materials science problem. The current understanding of fuel microstructure have been limited by the difficulty in studying the structure and chemistry of irradiated fuel samples at the mesoscale. Here, we take advantage of recent advances in experimental capabilities to characterize the microstructure in 3D of irradiated mixed oxide (MOX) fuel taken from two radial positions in the fuel pellet. We also reconstruct these microstructures using Idaho National Laboratory's MARMOT code and calculate the impact of microstructure heterogeneities on the effective thermal conductivity using mesoscale heat conduction simulations. The thermal conductivities of both samples are higher than the bulk MOX thermal conductivity because of the formation of metallic precipitates and because we do not currently consider phonon scattering due to defects smaller than the experimental resolution. We also used the results to investigate the accuracy of simple thermal conductivity approximations and equations to convert 2D thermal conductivities to 3D. It was found that these approximations struggle to predict the complex thermal transport interactions between metal precipitates and voids.

Mixed fuel strategy for carbon deposition mitigation in solid oxide fuel cells at intermediate temperatures.

Science.gov (United States)

Su, Chao; Chen, Yubo; Wang, Wei; Ran, Ran; Shao, Zongping; Diniz da Costa, João C; Liu, Shaomin

2014-06-17

In this study, we propose and experimentally verified that methane and formic acid mixed fuel can be employed to sustain solid oxide fuel cells (SOFCs) to deliver high power outputs at intermediate temperatures and simultaneously reduce the coke formation over the anode catalyst. In this SOFC system, methane itself was one part of the fuel, but it also played as the carrier gas to deliver the formic acid to reach the anode chamber. On the other hand, the products from the thermal decomposition of formic acid helped to reduce the carbon deposition from methane cracking. In order to clarify the reaction pathways for carbon formation and elimination occurring in the anode chamber during the SOFC operation, O2-TPO and SEM analysis were carried out together with the theoretical calculation. Electrochemical tests demonstrated that stable and high power output at an intermediate temperature range was well-maintained with a peak power density of 1061 mW cm(-2) at 750 °C. With the synergic functions provided by the mixed fuel, the SOFC was running for 3 days without any sign of cell performance decay. In sharp contrast, fuelled by pure methane and tested at similar conditions, the SOFC immediately failed after running for only 30 min due to significant carbon deposition. This work opens a new way for SOFC to conquer the annoying problem of carbon deposition just by properly selecting the fuel components to realize their synergic effects.

Results of the irradiation of mixed UO2 - PuO2 oxide fuel elements

International Nuclear Information System (INIS)

Mikailoff, H.; Mustelier, J.P.; Bloch, J.; Ezran, L.; Hayet, L.

1966-01-01

In order to study the behaviour of fuel elements used for the first charge of the reactor Rapsodie, a first batch of eleven needles was irradiated in the reactor EL3 and then examined. These needles (having a shape very similar lo that of the actual needles to be used) were made up of a stack of sintered mixed-oxide pellets: UO 2 containing about 10 per cent of PuO 2 . The density was 85 to 97 per cent of the theoretical, value. The diametral gap between the oxide and the stainless steel can was between 0,06 and 0,27 mm. The specific powers varied from 1230 to 2700 W/cm 3 and the can temperature was between 450 and 630 C. The maximum burn-up attained was 22000 MW days/tonne. Examination of the needles (metrology, radiography and γ-spectrography) revealed certain macroscopic changes, and the evolution of the fuel was shown by micrographic studies. These observations were used, together with flux measurements results, to calculate the temperature distribution inside the fuel. The volume of the fission gas produced was measured in some of the samples; the results are interpreted taking into account the temperature distribution in the oxide and the burn-up attained. Finally a study was made both of the behaviour of a fuel element whose central part was molten during irradiation, and of the effect of sodium which had penetrated into some of the samples following can rupture. (author) [fr

Thermal-spectrum recriticality energetics

International Nuclear Information System (INIS)

Schwinkendorf, K.N.

1993-12-01

Large computer codes have been created in the past to predict the energy release in hypothetical core disruptive accidents (CDA), postulated to occur in liquid metal reactors (LMR). These codes, such as SIMMER, are highly specific to LMR designs. More recent attention has focused on thermal-spectrum criticality accidents, such as for fuel storage basins and waste tanks containing fissile material. This paper resents results from recent one-dimensional kinetics simulations, performed for a recriticality accident in a thermal spectrum. Reactivity insertion rates generally are smaller than in LMR CDAs, and the energetics generally are more benign. Parametric variation of input was performed, including reactivity insertion and initial temperature

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)

Equipment to weld fuel rods of mixed oxides

International Nuclear Information System (INIS)

Aparicio, G.; Orlando, O.S.; Olano, V.R.; Toubes, B.; Munoz, C.A.

1987-01-01

Two welding outfits system T1G were designed and constructed to weld fuel rods with mixed oxides pellets (uranium and plutonium). One of them is connected to a glove box where the loading of sheaths takes place. The sheaths are driven to the welder through a removable plug pusher in the welding chamber. This equipment was designed to perform welding tests changing the parameters (gas composition and pressure, welding current, electrode position, etc.). The components of the welder, such as plug holder, chamber closure and peripheral accessories, were designed and constructed taking into account the working pressures in the machine, which is placed in a controlled area and connected to a glove box, where special safety conditions are necessary. The equipment to weld fuel bars is complemented by another machine, located in cold area, of the type presently used in the fuel elements factory. This equipment has been designed to perform some welding operations in sheaths and mixed oxide rods of the type Atucha I and II. Both machines have a programmed power supply of wide range and a vacuum, and pressurizing system that allows the change of parameters. Both systems have special features of handling and operation. (Author)

Process for recovery of plutonium from fabrication residues of mixed fuels consisting of uranium oxide and plutonium oxide

International Nuclear Information System (INIS)

Heremanns, R.H.; Vandersteene, J.J.

1983-01-01

The invention concerns a process for recovery of plutonium from fabrication residues of mixed fuels consisting of uranium oxide and plutonium oxide in the form of PuO 2 . Mixed fuels consisting of uranium oxide and plutonium oxide are being used more and more. The plants which prepare these mixed fuels have around 5% of the total mass of fuels as fabrication residue, either as waste or scrap. In view of the high cost of plutonium, it has been attempted to recover this plutonium from the fabrication residues by a process having a purchase price lower than the price of plutonium. The problem is essentially to separate the plutonium, the uranium and the impurities. The residues are fluorinated, the UF 6 and PuF 6 obtained are separated by selective absorption of the PuF 6 on NaF at a temperature of at least 400 0 C, the complex obtained by this absorption is dissolved in nitric acid solution, the plutonium is precipitated in the form of plutonium oxalate by adding oxalic acid, and the precipitated plutonium oxalate is calcined

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.

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

Thermal stability and oxidizing properties of mixed alkaline earth-alkali molten carbonates: A focus on the lithium-sodium carbonate eutectic system with magnesium additions

International Nuclear Information System (INIS)

Frangini, Stefano; Scaccia, Silvera

2013-01-01

Highlights: • TG/DSC analysis was conducted on magnesium-containing eutectic Li/Na eutectic carbonates. • Magnesium influence on the oxygen solubility properties of carbonate was also experimentally determined at 600 °C and 650 °C. • A reproducible partial decarbonation process in premelting region caused formation of magnesium oxycarbonate-like phases. • The acidobase buffering action of magnesium oxycarbonate species could explain the high basic/oxidizing properties of such carbonate melts. • A general correlation between thermal instability in premelting region and basic/oxidizing melt properties was established. - Abstract: A comparative study on thermal behavior and oxygen solubility properties of eutectic 52/48 lithium/sodium carbonate salt containing minor additions of magnesium up to 10 mol% has been made in order to determine whether a general correlation between these two properties can be found or not. Consecutive TG/DSC heating/cooling thermal cycles carried out under alternating CO 2 and N 2 gas flows allowed to assign thermal events observed in the premelting region to a partial decarbonation process of the magnesium-alkali mixed carbonates. The observed decarbonation process at 460 °C is believed to come from initial stage of thermal decomposition of magnesium carbonate resulting in the metastable formation of magnesium oxycarbonate-like phases MgO·2MgCO 3 , in a similar manner as previously reported for lanthanum. Reversible formation and decomposition of the magnesium carbonate phase has been observed under a CO 2 gas atmosphere. The intensity of the decomposition process shows a maximum for a 3 mol% MgO addition that gives also the highest oxygen solubility, suggesting therefore that instability thermal analysis in the premelting region can be considered as providing an effective measure of the basicity/oxidizing properties of alkali carbonate melts with magnesium or, in more general terms, with cations that are strong modifiers of

Uranium dioxide and beryllium oxide enhanced thermal conductivity nuclear fuel development

International Nuclear Information System (INIS)

Andrade, Antonio Santos; Ferreira, Ricardo Alberto Neto

2007-01-01

The uranium dioxide is the most used substance as nuclear reactor fuel for presenting many advantages such as: high stability even when it is in contact with water in high temperatures, high fusion point, and high capacity to retain fission products. The conventional fuel is made with ceramic sintered pellets of uranium dioxide stacked inside fuel rods, and presents disadvantages because its low thermal conductivity causes large and dangerous temperature gradients. Besides, the thermal conductivity decreases further as the fuel burns, what limits a pellet operational lifetime. This research developed a new kind of fuel pellets fabricated with uranium dioxide kernels and beryllium oxide filling the empty spaces between them. This fuel has a great advantage because of its higher thermal conductivity in relation to the conventional fuel. Pellets of this kind were produced, and had their thermophysical properties measured by the flash laser method, to compare with the thermal conductivity of the conventional uranium dioxide nuclear fuel. (author) (author)

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

Final Report: Fiscal Year 1997 demonstration of omnivorous non-thermal mixed waste treatment: Direct chemical oxidation of organic solids and liquids using peroxydisulfate

International Nuclear Information System (INIS)

Cooper, J.F.; Ballazs G.B.

1998-01-01

Direct Chemical Oxidation (DCO) is a non-thermal, ambient pressure, aqueous-based technology for the oxidative destruction of the organic components of hazardous or mixed waste streams. The process has been developed for applications in waste treatment, chemical demilitarization and decontamination at LLNL since 1992. The process uses solutions of the peroxydisulfate ion (typically sodium or ammonium salts) to completely mineralize the organics to carbon dioxide and water. The expended oxidant may be electrolytically regenerated to minimize secondary waste. The paper briefly describes: free radical and secondary oxidant formation; electrochemical regeneration; offgas stream; and throughput

Final Report: Fiscal Year 1997 demonstration of omnivorous non-thermal mixed waste treatment: Direct chemical oxidation of organic solids and liquids using peroxydisulfate

Energy Technology Data Exchange (ETDEWEB)

Cooper, J.F.

1998-01-01

Direct Chemical Oxidation (DCO) is a non-thermal, ambient pressure, aqueous-based technology for the oxidative destruction of the organic components of hazardous or mixed waste streams. The process has been developed for applications in waste treatment, chemical demilitarization and decontamination at LLNL since 1992. The process uses solutions of the peroxydisulfate ion (typically sodium or ammonium salts) to completely mineralize the organics to carbon dioxide and water. The expended oxidant may be electrolytically regenerated to minimize secondary waste. The paper briefly describes: free radical and secondary oxidant formation; electrochemical regeneration; offgas stream; and throughput.

Ruthenium release from thermally overheated nitric acid solution containing ruthenium nitrosyl nitrate and sodium nitrate to solidify

Energy Technology Data Exchange (ETDEWEB)

Sawada, Kayo; Ueda, Yasuyuki; Enokida, Youichi [Nuclear Chemical Engineering Laboratory, Nagoya University, Nagoya 4648603 (Japan)

2016-07-01

Radioactive ruthenium (Ru) is one of the dominant elemental species released into the environment from a fuel reprocessing plant in a hypothetical design accident due to its relatively higher fission yield and longer half-life. After the hypothetical accident assuming the loss of all electric power and cooling functions, high-level liquid waste (HLLW) may be overheated by the energetic decays of many fission products in it, and Ru may be oxidized to the volatile tetroxide, RuO{sub 4}, which is released through the off-gas pathway. At a reprocessing plant in Japan, alkaline solution from the solvent scrubbing stream is sometimes mixed with the HLLW followed by vitrification, which can be influenced by the addition of sodium nitrate to a simulated HLLW containing ruthenium nitrosyl nitrate that was experimentally evaluated on a small scale using the overheated nitric acid solution of 2 mol/dm{sup 3}, which was kept at 180 Celsius degrees in a glass evaporator placed in a thermostatic bath. The release fraction of Ru increased by approximately 30% by the addition of sodium nitrate. This may be partially explained by the existence of relatively highly concentrated nitrate ions in the liquid phase that oxidize the ruthenium species to RuO{sub 4} during the drying process. (authors)

Modeling of thermal expansion coefficient of perovskite oxide for solid oxide fuel cell cathode

Science.gov (United States)

Heydari, F.; Maghsoudipour, A.; Alizadeh, M.; Khakpour, Z.; Javaheri, M.

2015-09-01

Artificial intelligence models have the capacity to eliminate the need for expensive experimental investigation in various areas of manufacturing processes, including the material science. This study investigates the applicability of adaptive neuro-fuzzy inference system (ANFIS) approach for modeling the performance parameters of thermal expansion coefficient (TEC) of perovskite oxide for solid oxide fuel cell cathode. Oxides (Ln = La, Nd, Sm and M = Fe, Ni, Mn) have been prepared and characterized to study the influence of the different cations on TEC. Experimental results have shown TEC decreases favorably with substitution of Nd3+ and Mn3+ ions in the lattice. Structural parameters of compounds have been determined by X-ray diffraction, and field emission scanning electron microscopy has been used for the morphological study. Comparison results indicated that the ANFIS technique could be employed successfully in modeling thermal expansion coefficient of perovskite oxide for solid oxide fuel cell cathode, and considerable savings in terms of cost and time could be obtained by using ANFIS technique.

Fuel elements based on mixed oxides UO{sub 2} - PuO{sub 2}; Gorivni elementi na bazi mesanih oksida UO{sub 2} - PuO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Katanic-Popovic, J; Stevanovic, M [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

1978-07-01

Questions concerning utilization of plutonium as a fissionable material in fuel elements for nuclear power plants have been discussed. Characteristics and application of fuel elements with mixed UO{sub 2} - PuO{sub 2} fuel for thermal and fast breeder reactors have also been dealt with. In the presentation of technological processes for production of fuel elements based on mixed oxides specific characteristics are given with respect to the work with plutonium and relatively high production costs as compared to classical fuel elements based on sintered UO{sub 2}. (author)

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

Organic tank safety project: Preliminary results of energetics and thermal behavior studies of model organic nitrate and/or nitrite mixtures and a simulated organic waste

International Nuclear Information System (INIS)

Scheele, R.D.; Sell, R.L.; Sobolik, J.L.; Burger, L.L.

1995-08-01

As a result of years of production and recovery of nuclear defense materials and subsequent waste management at the Hanford Site, organic-bearing radioactive high-level wastes (HLW) are currently stored in large (up to 3. ML) single-shell storage tanks (SSTs). Because these wastes contain both fuels (organics) and the oxidants nitrate and nitrite, rapid energetic reactions at certain conditions could occur. In support of Westinghouse Hanford Company's (WHC) efforts to ensure continued safe storage of these organic- and oxidant-bearing wastes and to define the conditions necessary for reactions to occur, we measured the thermal sensitivities and thermochemical and thermokinetic properties of mixtures of selected organics and sodium nitrate and/or nitrite and a simulated Hanford organic-bearing waste using thermoanalytical technologies. These thermoanalytical technologies are used by chemical reactivity hazards evaluation organizations within the chemical industry to assess chemical reaction hazards

Organic tank safety project: Preliminary results of energetics and thermal behavior studies of model organic nitrate and/or nitrite mixtures and a simulated organic waste

Energy Technology Data Exchange (ETDEWEB)

Scheele, R.D.; Sell, R.L.; Sobolik, J.L.; Burger, L.L.

1995-08-01

As a result of years of production and recovery of nuclear defense materials and subsequent waste management at the Hanford Site, organic-bearing radioactive high-level wastes (HLW) are currently stored in large (up to 3. ML) single-shell storage tanks (SSTs). Because these wastes contain both fuels (organics) and the oxidants nitrate and nitrite, rapid energetic reactions at certain conditions could occur. In support of Westinghouse Hanford Company`s (WHC) efforts to ensure continued safe storage of these organic- and oxidant-bearing wastes and to define the conditions necessary for reactions to occur, we measured the thermal sensitivities and thermochemical and thermokinetic properties of mixtures of selected organics and sodium nitrate and/or nitrite and a simulated Hanford organic-bearing waste using thermoanalytical technologies. These thermoanalytical technologies are used by chemical reactivity hazards evaluation organizations within the chemical industry to assess chemical reaction hazards.

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

Postaccident heat removal: large-scale molten-fuel-sodium interaction experiments

International Nuclear Information System (INIS)

Johnson, T.R.; Pavlik, J.R.; Baker, L. Jr.

1975-02-01

Kilogram-scale interactions between molten UO 2 and sodium were performed in an unrestrained geometry to study the resulting energetics and fragmentation. The molten UO 2 was producted by the exothrmic reaction between uranium and MoO 3 powders. Under the conditions of the experiments completed to date, the short-rise-time pressure pulses created in the liquid phase had negligible work potential, and their magnitude did not increase with the amount of molten fuel. No significant gas-phase shock pressures were generated. The largest potential for mechanical work was the sodium vapor generated over a period of roughly 1 sec. About 20 percent of the heat was effective in generating vapor. The ex- perimental results show a marked tendency of molten UO 2 to form particulate after passage through only a few inches of sodium. Particle size distributions obtained under the conditions of the experiments were not significantly different from those obtained in prior small-scale tests and in TREAT tests. Also, the results indicate that the metallic component of the molten mixture formed larger particles than the oxide component. (U.S.)

Light water reactor mixed-oxide fuel irradiation experiment

International Nuclear Information System (INIS)

Hodge, S.A.; Cowell, B.S.; Chang, G.S.; Ryskamp, J.M.

1998-01-01

The United States Department of Energy Office of Fissile Materials Disposition is sponsoring and Oak Ridge National Laboratory (ORNL) is leading an irradiation experiment to test mixed uranium-plutonium oxide (MOX) fuel made from weapons-grade (WG) plutonium. In this multiyear program, sealed capsules containing MOX fuel pellets fabricated at Los Alamos National Laboratory (LANL) are being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The planned experiments will investigate the utilization of dry-processed plutonium, the effects of WG plutonium isotopics on MOX performance, and any material interactions of gallium with Zircaloy cladding

Experimental studies of thermal and chemical interactions between oxide and silicide nuclear fuels with water

Energy Technology Data Exchange (ETDEWEB)

farahani, A.A.; Corradini, M.L. [Univ. of Wisconsi, Madison, WI (United States)

1995-09-01

Given some transient power/cooling mismatch is a nuclear reactor and its inability to establish the necessary core cooling, energetic fuel-coolant interactions (FCI`s commonly called `vapor explosions`) could occur as a result of the core melting and coolant contact. Although a large number of studies have been done on energetic FCI`s, very few experiments have been performed with the actual fuel materials postulated to be produced in severe accidents. Because of the scarcity of well-characterized FCI data for uranium allows in noncommercial reactors (cermet and silicide fuels), we have conducted a series of experiments to provide a data base for the foregoing materials. An existing 1-D shock-tube facility was modified to handle depleted radioactive materials (U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al). Our objectives have been to determine the effects of the initial fuel composition and temperature and the driving pressure (triggering) on the explosion work output, dynamic pressures, transient temperatures, and the hydrogen production. Experimental results indicate limited energetics, mainly thermal interactions, for these fuel materials as compared to aluminum where more chemical reactions occur between the molten aluminum and water.

Importance of Sodium Fuel Interaction in Fast Reactor Safety Evaluation - CEA Point of View

International Nuclear Information System (INIS)

Tanguy, P.

1976-01-01

The consequences of interactions between molten metal (aluminium-uranium alloy) and water have long been a subject of concern for those in charge of reactor safety, following accidents observed or induced in certain reactors (BORAX, SL1, SPERT 1 D). In such accidents, as in similar cases occurring in traditional industries (aluminium foundries, steel works, paper mills...) the contact between the hot liquid product and the coolant entails rapid vaporization of the latter with effects identical to that of an explosive. Although chemical reactions of water decomposition occur in some cases, the main phenomenon is the conversion of the thermal energy stored in the hot substance into mechanical energy. Despite the fact that a molten oxide fuel differs from an aluminium-uranium alloy, as does sodium from water, the consequences of possible contact between the molten mixed uranium and plutonium oxide and sodium must be carefully studied since such a contact may occur in accident conditions in sodium-cooled fast neutron reactors. The essential purpose of an evaluation of reactor safety in accident conditions is in fact to ensure the containment of dangerous products Consequently, any phenomenon likely to endanger containment barriers must be carefully examined. In conclusion: Whereas an accident within an assembly seems to show little likelihood of creating conditions seriously endangering fuel containment, the gravity of problems associated with an overall accident on the core is worthy of thorough and attentive study. In the case of an overall accident on the core of a fast reactor, the interaction between the molten fuel and the sodium is of consequence at two levels. The first is the retention of mechanical energy which may be considerable. The second is the recovery of fuel fragments in an overall cooled configuration but where local cooling problems may give rise to interaction. A greater effort is required in performing tests and mastering their results to

Effect of 17 x 17 fuel assembly geometry on interchannel thermal mixing

International Nuclear Information System (INIS)

Motley, F.E.; Wenzell, A.H.; Cadek, F.F.

1975-01-01

A test to determine the value of the thermal diffusion coefficient (TDC) in the 17 x 17 fuel assembly geometry was conducted. The test section was a 5 x 5 rod bundle with a radial power difference of 4.5 to 1. The rod OD and pitch are identical to the 17 x 17 fuel assembly, as is the mixing vane grid design. The value of thermal diffusion coefficient (TDC) was determined by matching the experimental exit enthalpy distribution to that predicted by the THINC computer code. The mean value of TDC for the 17 x 17 fuel assembly geometry is TDC = .059. 6 references

Mixed U/Pu oxide fabrication facility for gel-sphere-pac fuel

International Nuclear Information System (INIS)

1978-09-01

This paper describes a conceptual plant which uses the gel-sphere-pac process to fabricate mixed oxide (MOX) fuel and covers (1) fabrication of co-processed MOX fuel and (2) fabrication of co-processed spiked MOX fuel, using 60 Co. The report describes: the fuel fabrication process and plant layout, including scrap and waste processing; and maintenance safety and ventilation measures. A description of the conversion of U and Pu nitrate using a gel sphere process is given in Appendix A

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.

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

Chemical and Radiochemical Composition of Thermally Stabilized Plutonium Oxide from the Plutonium Finishing Plant Considered as Alternate Feedstock for the Mixed Oxide Fuel Fabrication Facility

International Nuclear Information System (INIS)

Tingey, Joel M.; Jones, Susan A.

2005-01-01

Eighteen plutonium oxide samples originating from the Plutonium Finishing Plant (PFP) on the Hanford Site were analyzed to provide additional data on the suitability of PFP thermally stabilized plutonium oxides and Rocky Flats oxides as alternate feedstock to the Mixed Oxide Fuel Fabrication Facility (MFFF). Radiochemical and chemical analyses were performed on fusions, acid leaches, and water leaches of these 18 samples. The results from these destructive analyses were compared with nondestructive analyses (NDA) performed at PFP and the acceptance criteria for the alternate feedstock. The plutonium oxide materials considered as alternate feedstock at Hanford originated from several different sources including Rocky Flats oxide, scrap from the Remote Mechanical C-Line (RMC) and the Plutonium Reclamation Facility (PRF), and materials from other plutonium conversion processes at Hanford. These materials were received at PFP as metals, oxides, and solutions. All of the material considered as alternate feedstock was converted to PuO2 and thermally stabilized by heating the PuO2 powder at 950 C in an oxidizing environment. The two samples from solutions were converted to PuO2 by precipitation with Mg(OH)2. The 18 plutonium oxide samples were grouped into four categories based on their origin. The Rocky Flats oxide was divided into two categories, low- and high-chloride Rocky Flats oxides. The other two categories were PRF/RMC scrap oxides, which included scrap from both process lines and oxides produced from solutions. The two solution samples came from samples that were being tested at Pacific Northwest National Laboratory because all of the plutonium oxide from solutions at PFP had already been processed and placed in 3013 containers. These samples originated at the PFP and are from plutonium nitrate product and double-pass filtrate solutions after they had been thermally stabilized. The other 16 samples originated from thermal stabilization batches before canning at

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)

Safety analysis of MOX fuels by fuel performance code

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-12-01

Performance of plutonium rick mixed oxide fuels specified for the Reduced-Moderation Water Reactor (RMWR) has been analysed by modified fuel performance code. Thermodynamic properties of these fuels up to 120 GWd/t burnup have not been measured and estimated using existing uranium fuel models. Fission product release, pressure rise inside fuel rods and mechanical loads of fuel cans due to internal pressure have been preliminarily assessed based on assumed axial power distribution history, which show the integrity of fuel performance. Detailed evaluation of fuel-cladding interactions due to thermal expansion or swelling of fuel pellets due to high burnup will be required for safety analysis of mixed oxide fuels. Thermal conductivity and swelling of plutonium rich mixed oxide fuels shall be taken into consideration. (T. Tanaka)

Standardization of thermal and epithermal INAA methods for simultaneous determination of U and Th in mixed oxide samples

International Nuclear Information System (INIS)

Acharya, R.; Pujari, P.K.; Chandra, Ruma

2010-01-01

Full text: Uranium and thorium are important fuel materials for nuclear power program. In recent years utilization of thoria based fuel has assumed significance due to higher energy requirements. Thorium based mixed oxide is the proposed fuel for Advanced Heavy Water Reactors (AHWR). In this respect, studies are carried out through preparation of natural U and Th mixed oxides by powder metallurgical route, wherein composition of U and Th is specific and requires strict control in terms their contents and homogeneity in the mixture. Stringent chemical quality control necessitates compositional characterization of the fuel material i.e. accurate and precise determination of U and Th. A suitable method which does not need any chemical dissolution and yields high precision results with minima sample handling is desirable. Instrumental neutron activation analysis (INAA) using reactor neutron is the technique of choice. In view of this, INAA methods namely thermal lNAA (TNAA) (utilizing whole reactor neutrons) and epithermal INAA (ENAA) (utilizing epicadmium neutrons) were standardized for the determination of U and Th in presence of each other in mixed oxide samples. In the present work pneumatic carrier facility (PCF) of Dhruva reactor and self-serve facility of CIRUS reactor were used for TNAA and ENAA respectively. Standards, synthetic samples and mixed oxide samples prepared in cellulose matrix, were irradiated for 1 minute at PCF of Dhruva reactor and for 1 hour at CIRUS reactor under cadmium cover (0.5 mm). Radioactive assay was carried out using 40% relative efficiency HPGe detector. Peak areas under the full energy peaks were evaluated by peak fit method using the PHAST software. Both activation and daughter products of U ( 239 U, 74.6 keV and 239 Np, 277 keV) and Th ( 233 Th, 86 keV and 233 Pa, 312 keV) were used for their concentration determination. The method was validated by analyzing synthetic mixed oxide samples (6-48%U-Th mixed oxide). The % deviations

Safety criteria related to microheterogeneities in LWR mixed oxide fuels

International Nuclear Information System (INIS)

Renard, A.; Mostin, N.

1978-01-01

The main safety aspets of PuO 2 microheterogeneities in the pellets of LWR mixed oxide fuels are reviewed. Points of interest are studied, especially the transient behaviour in accidental conditions and criteria are deduced for use in the specification and quality control of the fabricated product. (author)

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.

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

International Nuclear Information System (INIS)

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

1976-05-01

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

Safety problems related to microheterogeneities in physically mixed oxide fuels

International Nuclear Information System (INIS)

Renard, A.; Evrard, G.; Vanhellemont, G.

1976-01-01

The safety aspects of microheterogeneities in LMFBR mixed oxide fuel are reviewed from the point of view of the pin behaviour dynamic study, the fabrication and the quality control. The paper emphasizes some significant parameters in transient conditions, the prevention means in the fabrication process and the analysis methods for the control

Fuel-coolant interaction-phenomena under prompt burst conditions

International Nuclear Information System (INIS)

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

1979-01-01

The Prompt Burst Energetics (PBE) experiments conducted at Sandia Laboratories are a series of in-pile tests with fresh uranium oxide or uranium carbide fuel pins in stagnant sodium. Fuel-coolant-interactions in PBE-9S (oxide/sodium system) and PBE-SG2 (carbide/sodium) have been analyzed with the MURTI parametric FCI code. The purpose is to gain insight into possible FCI scenarios in the experiments and sensitivity of results to input parameters. Results are in approximate agreement for the second (triggered) event in PBE-9S (32 MPa peak) and the initial interaction in PBE-SG2

Melting temperature of uranium - plutonium mixed oxide fuel

Energy Technology Data Exchange (ETDEWEB)

Ishii, Tetsuya; Hirosawa, Takashi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1997-08-01

Fuel melting temperature is one of the major thermodynamical properties that is used for determining the design criteria on fuel temperature during irradiation in FBR. In general, it is necessary to evaluate the correlation of fuel melting temperature to confirm that the fuel temperature must be kept below the fuel melting temperature during irradiation at any conditions. The correlations of the melting temperature of uranium-plutonium mixed oxide (MOX) fuel, typical FBR fuel, used to be estimated and formulized based on the measured values reported in 1960`s and has been applied to the design. At present, some experiments have been accumulated with improved experimental techniques. And it reveals that the recent measured melting temperatures does not agree well to the data reported in 1960`s and that some of the 1960`s data should be modified by taking into account of the recent measurements. In this study, the experience of melting temperature up to now are summarized and evaluated in order to make the fuel pin design more reliable. The effect of plutonium content, oxygen to metal ratio and burnup on MOX fuel melting was examined based on the recent data under the UO{sub 2} - PuO{sub 2} - PuO{sub 1.61} ideal solution model, and then formulized. (J.P.N.)

Melting temperature of uranium - plutonium mixed oxide fuel

International Nuclear Information System (INIS)

Ishii, Tetsuya; Hirosawa, Takashi

1997-08-01

Fuel melting temperature is one of the major thermodynamical properties that is used for determining the design criteria on fuel temperature during irradiation in FBR. In general, it is necessary to evaluate the correlation of fuel melting temperature to confirm that the fuel temperature must be kept below the fuel melting temperature during irradiation at any conditions. The correlations of the melting temperature of uranium-plutonium mixed oxide (MOX) fuel, typical FBR fuel, used to be estimated and formulized based on the measured values reported in 1960's and has been applied to the design. At present, some experiments have been accumulated with improved experimental techniques. And it reveals that the recent measured melting temperatures does not agree well to the data reported in 1960's and that some of the 1960's data should be modified by taking into account of the recent measurements. In this study, the experience of melting temperature up to now are summarized and evaluated in order to make the fuel pin design more reliable. The effect of plutonium content, oxygen to metal ratio and burnup on MOX fuel melting was examined based on the recent data under the UO 2 - PuO 2 - PuO 1.61 ideal solution model, and then formulized. (J.P.N.)

Reliability of fast reactor mixed-oxide fuel during operational transients

International Nuclear Information System (INIS)

Boltax, A.; Neimark, L.A.; Tsai, Hanchung; Katsuragawa, M.; Shikakura, S.

1991-07-01

Results are presented from the cooperative DOE and PNC Phase 1 and 2 operational transient testing programs conducted in the EBR-2 reactor. The program includes second (D9 and PNC 316 cladding) and third (FSM, AST and ODS cladding) generation mixed-oxide fuel pins. The irradiation tests include duty cycle operation and extended overpower tests. the results demonstrate the capability of second generation fuel pins to survive a wide range of duty cycle and extended overpower events. 15 refs., 9 figs., 4 tabs

Thermal gradient effects on the oxidation of Zircaloy fuel cladding

International Nuclear Information System (INIS)

Klein, A.C.; Reyes, J.N. Jr.; Maguire, M.A.

1990-01-01

A Thermal Gradient Test Facility (TGTF) has been designed and constructed to measure the thermal gradient effect on pressurized water reactor (PWR) fuel rod cladding. The TGTF includes a heat flux simulator assembly capable of producing a wide range of PWR operating conditions including water flow velocities and temperatures, water chemistry conditions, cladding temperatures, and heat fluxes ranging to 160 W/cm 2 . It is fully instrumented including a large number of thermocouples both inside the water flow channel and inside the cladding. Two test programs are in progress. First, cladding specimens are pre-oxidized in air at 500 deg. C and in 400 deg. C steam for various lengths of time to develop a range of uniform oxide thicknesses from 1 to 60 micrometers. The pre-oxidized specimens are placed in the TGTF to characterize the oxide thermal conductivity under a variety of water flow and heat flux conditions. Second, to overcome the long exposure times required under typical PWR conditions a series of tests with the addition of high concentrations of lithium hydroxide to the water are being considered. Static autoclave tests have been conducted with lithium hydroxide concentrations ranging from 0 to 2 moles per liter at 300, 330, and 360 deg. C for up to 36 hours. Results for zircaloy-4 show a considerable increase in the weight gain for the exposed samples with oxidation rate enhancement factors as high as 70 times that of pure water. Operation of the TGTF with elevated lithium hydroxide levels will yield real-time information concerning the effects of a heat flux on the oxidation kinetics of zircaloy fuel rod cladding. (author). 5 refs, 5 figs, 2 tabs

Effects of solid fission products forming dissolved oxide (Nd) and metallic precipitate (Ru) on the thermal conductivity of uranium base oxide fuel

International Nuclear Information System (INIS)

Kim, Dong-Joo; Yang, Jae-Ho; Kim, Jong-Hun; Rhee, Young-Woo; Kang, Ki-Won; Kim, Keon-Sik; Song, Kun-Woo

2007-01-01

The effects of solid fission products on the thermal conductivity of uranium base oxide nuclear fuel were experimentally investigated. Neodymium (Nd) and ruthenium (Ru) were added to represent the physical states of solid fission products such as 'dissolved oxide' and 'metallic precipitate', respectively. Thermal conductivity was determined on the basis of the thermal diffusivity, density and specific heat values. The effects of the additives on the thermal conductivity were quantified in the form of the thermal resistivity equation - the reciprocal of the phonon conduction equation - which was determined from the measured data. It is concluded that the thermal conductivity of the irradiated nuclear fuel is affected by both the 'dissolved oxide' and the 'metallic precipitate', however, the effects are in the opposite direction and the 'dissolved oxide' influences the thermal conductivity more significantly than that of the 'metallic precipitate'

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

Thermal-hydraulics and neutronics studies on the FP7 CP-ESFR oxide and carbide cores

Energy Technology Data Exchange (ETDEWEB)

Ammirabile, L.; Tsige-Tamirat, H. [European Commission, JRC, Inst. for Energy, Petten (Netherlands)

2011-07-01

In the framework of the the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) two core designs that are currently being proposed for the 3600 MWth sodium-cooled reactor concept: one is based on oxide fuel and the other on carbide fuel. Using the European Safety Assessment Platform (ESAP), JRC-IE has conducted static calculation on neutronics (incl. reactivity coefficients) and thermal-hydraulic characteristics for both oxide and carbide reference cores. The quantities evaluated include: keff, coolant heat-up, void, and Doppler reactivity coefficients, axial and radial expansion reactivity coefficients, pin-by-pin calculated power profiles, average and peak channel temperatures. This paper presents the ESAP models applied in the study together with the relevant results for the oxide and carbide core. (author)

Thermal-hydraulics and neutronics studies on the FP7 CP-ESFR oxide and carbide cores

International Nuclear Information System (INIS)

Ammirabile, L.; Tsige-Tamirat, H.

2011-01-01

In the framework of the the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) two core designs that are currently being proposed for the 3600 MWth sodium-cooled reactor concept: one is based on oxide fuel and the other on carbide fuel. Using the European Safety Assessment Platform (ESAP), JRC-IE has conducted static calculation on neutronics (incl. reactivity coefficients) and thermal-hydraulic characteristics for both oxide and carbide reference cores. The quantities evaluated include: keff, coolant heat-up, void, and Doppler reactivity coefficients, axial and radial expansion reactivity coefficients, pin-by-pin calculated power profiles, average and peak channel temperatures. This paper presents the ESAP models applied in the study together with the relevant results for the oxide and carbide core. (author)

Fuel-coolant interaction-phenomena under prompt burst conditions. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

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

1979-01-01

The Prompt Burst Energetics (PBE) experiments conducted at Sandia Laboratories are a series of in-pile tests with fresh uranium oxide or uranium carbide fuel pins in stagnant sodium. Fuel-coolant-interactions in PBE-9S (oxide/sodium system) and PBE-SG2 (carbide/sodium) have been analyzed with the MURTI parametric FCI code. The purpose is to gain insight into possible FCI scenarios in the experiments and sensitivity of results to input parameters. Results are in approximate agreement for the second (triggered) event in PBE-9S (32 MPa peak) and the initial interaction in PBE-SG2 (190 MPa peak).

Sodium-based hydrides for thermal energy applications

Science.gov (United States)

Sheppard, D. A.; Humphries, T. D.; Buckley, C. E.

2016-04-01

Concentrating solar-thermal power (CSP) with thermal energy storage (TES) represents an attractive alternative to conventional fossil fuels for base-load power generation. Sodium alanate (NaAlH4) is a well-known sodium-based complex metal hydride but, more recently, high-temperature sodium-based complex metal hydrides have been considered for TES. This review considers the current state of the art for NaH, NaMgH3- x F x , Na-based transition metal hydrides, NaBH4 and Na3AlH6 for TES and heat pumping applications. These metal hydrides have a number of advantages over other classes of heat storage materials such as high thermal energy storage capacity, low volume, relatively low cost and a wide range of operating temperatures (100 °C to more than 650 °C). Potential safety issues associated with the use of high-temperature sodium-based hydrides are also addressed.

Thermal signature measurements for ammonium nitrate/fuel mixtures by laser heating

International Nuclear Information System (INIS)

Nazarian, Ashot; Presser, Cary

2016-01-01

Highlights: • LDTR is a useful diagnostic for characterizing AN/fuel mixture thermochemical behavior. • Each AN/fuel mixture thermal signature was different. • AN/fuel mixture signature features were defined by the individual constituents. • Baseline signatures changed after an experiment. - Abstract: Measurements were carried out to obtain thermal signatures of several ammonium nitrate/fuel (ANF) mixtures, using a laser-heating technique referred to as the laser-driven thermal reactor (LDTR). The mixtures were ammonium nitrate (AN)/kerosene, AN/ethylene glycol, AN/paraffin wax, AN/petroleum jelly, AN/confectioner's sugar, AN/cellulose (tissue paper), nitromethane/cellulose, nitrobenzene/cellulose, AN/cellulose/nitromethane, AN/cellulose/nitrobenzene. These mixtures were also compared with AN/nitromethane and AN/diesel fuel oil, obtained from an earlier investigation. Thermograms for the mixtures, as well as individual constituents, were compared to better understand how sample thermal signature changes with mixture composition. This is the first step in development of a thermal-signature database, to be used along with other signature databases, to improve identification of energetic substances of unknown composition. The results indicated that each individual thermal signature was associated unambiguously with a particular mixture composition. The signature features of a particular mixture were shaped by the individual constituent signatures. It was also uncovered that the baseline signature was modified after an experiment due to coating of unreacted residue on the substrate surface and a change in the reactor sphere oxide layer. Thus, care was required to pre-oxidize the sphere prior to an experiment. A minimum sample mass (which was dependent on composition) was required to detect the signature characteristics. Increased laser power served to magnify signal strength while preserving the signature features. For the mixtures examined, the thermal

Thermal diffusivity and conductivity of thorium- uranium mixed oxides

Science.gov (United States)

Saoudi, M.; Staicu, D.; Mouris, J.; Bergeron, A.; Hamilton, H.; Naji, M.; Freis, D.; Cologna, M.

2018-03-01

Thorium-uranium oxide pellets with high densities were prepared at the Canadian Nuclear Laboratories (CNL) by co-milling, pressing, and sintering at 2023 K, with UO2 mass contents of 0, 1.5, 3, 8, 13, 30, 60 and 100%. At the Joint Research Centre, Karlsruhe (JRC-Karlsruhe), thorium-uranium oxide pellets were prepared using the spark plasma sintering (SPS) technique with 79 and 93 wt. % UO2. The thermal diffusivity of (Th1-xUx)O2 (0 ≤ x ≤ 1) was measured at CNL and at JRC-Karlsruhe using the laser flash technique. ThO2 and (Th,U)O2 with 1.5, 3, 8 and 13 wt. % UO2 were found to be semi-transparent to the infrared wavelength of the laser and were coated with graphite for the thermal diffusivity measurements. This semi-transparency decreased with the addition of UO2 and was lost at about 30 wt. % of UO2 in ThO2. The thermal conductivity was deduced using the measured density and literature data for the specific heat capacity. The thermal conductivity for ThO2 is significantly higher than for UO2. The thermal conductivity of (Th,U)O2 decreases rapidly with increasing UO2 content, and for UO2 contents of 60% and higher, the conductivity of the thorium-uranium oxide fuel is close to UO2. As the mass difference between the Th and U atoms is small, the thermal conductivity decrease is attributed to the phonon scattering enhanced by lattice strain due to the introduction of uranium in ThO2 lattice. The new results were compared to the data available in the literature and were evaluated using the classical phonon transport model for oxide systems.

Development of ORIGEN libraries for mixed oxide (MOX) fuel assembly designs

International Nuclear Information System (INIS)

Mertyurek, Ugur; Gauld, Ian C.

2016-01-01

Highlights: • ORIGEN MOX library generation process is described. • SCALE burnup calculations are validated against measured MOX fuel samples from the MALIBU program. • ORIGEN MOX libraries are verified using the OECD Phase IV-B benchmark. • There is good agreement for calculated-to-measured isotopic distributions. - Abstract: ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup. The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. The nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.

Dissolution of mixed oxide spent fuel from FBR

International Nuclear Information System (INIS)

Sanyoshi, H.; Nishina, H.; Toyota, O.; Yamamoto, R.; Nemoto, S.; Okamoto, F.; Togashi, A.; Kawata, T.; Hayashi, S.

1991-01-01

At the Tokai Works of the Power Reactor and Nuclear Fuel Development Corporation (PNC), the Chemical Processing Facility (CPF) has been continuing operation since 1982 for laboratory scale hot experiments on reprocessing of FBR mixed oxide fuel. As a part of these experiments, dissolution experiments have been performed to define the key parameters affecting dissolution rates such as concentration of nitric acid, temperature and burnup and also to confirm the amount of insoluble residue. The dissolution rate of the irradiated fuel was determined to be in proportion to the 1.7 power of the nitric acid concentration. The activation energy determined from the experiments varied from 6 to 11 kcal/mol depending on the method of dissolution. The dissolution rate decreased as the fuel burnup increased in low nitric acid media below 5 mol/l. However, it was found that the effect of the burnup became negligible in a high concentration of nitric acid media. The amount of insoluble residue and its constituents were evaluated by changing the dissolution condition. (author)

Analysis of fuel sodium interaction in a fast breeder reactor

International Nuclear Information System (INIS)

Tezuka, M.; Suzuki, K.; Sasanuma, K.; Nagasima, K.; Kawaguchi, O.

A code ''SUGAR'' has been developed to evaluate molten Fuel Sodium Interaction (FSI) in a fast breeder reactor. This code computes thermohydrodynamic behavior by heat transfer from fuel to sodium and dynamic deformation of reactor structures simultaneously. It was applied to evaluate FSI in local fuel melting accident in a fuel assembly and in core disassembly accident for the 300MWe fast breeder reactor under development in Japan. The analytical methods of the SUGAR code are mainly shown in the following: 1) the thermal and dynamic model of FSI is mainly based on Cho-Wright's model; 2) the axial and radial expansions of surroundings of FSI region are calculated with one-dimensional and compressive hydrodynamics equation; 3) the structure response is calculated with one-dimensional and dynamic stress equation. Our studies show that mass of fuel interacted with sodium, ratio of fuel mass to sodium mass, fuel particle size, heat transfer coefficient from fuel to sodium, and structure's force have great effect on pressure amplitude and deformation of reactor structures

Synthesis and characterization of brannerite wasteforms for the immobilization of mixed oxide fuel residues

Energy Technology Data Exchange (ETDEWEB)

Bailey, D.J.; Stennett, M.C.; Hyatt, N.C. [Immobilisation Science Laboratory, Department of Materials Science and Engineering, University of Sheffield, Sheffield, S1 3JD (United Kingdom)

2016-07-01

A possible method for the reduction of civil Pu stockpiles is the reuse of Pu in mixed oxide fuel (MOX). During MOX fuel production, residues unsuitable for further recycle will be produced. Due to their high actinide content MOX residues require immobilization within a robust host matrix. Although it is possible to immobilize actinides in vitreous wasteforms; ceramic phases, such as brannerite (UTi{sub 2}O{sub 6}), are attractive due to their high waste loading capacity and relative insolubility. A range of uranium brannerite, formulated Gd{sub x}U{sub 1-x}Ti{sub 2}O{sub 6}, were prepared using a mixed oxide route. Charge compensation of divalent and trivalent cations was expected to occur via the oxidation of U{sup 4+} to higher valence states (U{sup 5+} or U{sup 6+}). Gd{sup 3+} was added to act as a neutron absorber in the final Pu bearing wasteform. X-ray powder diffraction of synthesised specimens found that phase distribution was strongly affected by processing atmosphere (air or Ar). In all cases prototypical brannerite was formed accompanied by different secondary phases dependent on processing atmosphere. Microstructural analysis (SEM) of the sintered samples confirmed the results of the X-ray powder diffraction. The preliminary results presented here indicate that brannerite is a promising host matrix for mixed oxide fuel residues. (authors)

Chemical Characterization and Reactivity of Fuel-Oxidizer Reaction Product

Science.gov (United States)

David, Dennis D.; Dee, Louis A.; Beeson, Harold D.

1997-01-01

Fuel-oxidizer reaction product (FORP), the product of incomplete reaction of monomethylhydrazine and nitrogen tetroxide propellants prepared under laboratory conditions and from firings of Shuttle Reaction Control System thrusters, has been characterized by chemical and thermal analysis. The composition of FORP is variable but falls within a limited range of compositions that depend on three factors: the fuel-oxidizer ratio at the time of formation; whether the composition of the post-formation atmosphere is reducing or oxidizing; and the reaction or post-reaction temperature. A typical composition contains methylhydrazinium nitrate, ammonium nitrate, methylammonium nitrate, and trace amounts of hydrazinium nitrate and 1,1-dimethylhydrazinium nitrate. Thermal decomposition reactions of the FORP compositions used in this study were unremarkable. Neither the various compositions of FORP, the pure major components of FORP, nor mixtures of FORP with propellant system corrosion products showed any unusual thermal activity when decomposed under laboratory conditions. Off-limit thruster operations were simulated by rapid mixing of liquid monomethylhydrazine and liquid nitrogen tetroxide in a confined space. These tests demonstrated that monomethylhydrazine, methylhydrazinium nitrate, ammonium nitrate, or Inconel corrosion products can induce a mixture of monomethylhydrazine and nitrogen tetroxide to produce component-damaging energies. Damaging events required FORP or metal salts to be present at the initial mixing of monomethylhydrazine and nitrogen tetroxide.

Behavior of mixed-oxide fuel elements during an overpower transient

International Nuclear Information System (INIS)

Tsai, H.; Shikakura, S.

1993-01-01

A slow-ramp (0.1%/s), extended overpower (∼90%) transient test was conducted in EBR-II on 19 mixed-oxide fuel elements with conservative, moderate, and aggressive designs. Claddings for the elements were Type 316, D9, or PNC-316 stainless steel. Before the transient, the elements were preirradiated under steady-state or steady-state plus duty-cycle (periodic 15% overpower transient) conditions to burnups of 2.5-9.7 at%. Cladding integrity during the transient test was maintained by all fuel elements except one, which had experienced substantial overtemperature in the earlier stedy-state irradiation. Extensive centerline fuel melting occurred in all test elements. Significantly, this melting did not cause any elements to breach, although it did have a strong effect on the other aspects of fuel element behavior. (orig.)

Emission computer tomography on a Dodewaard mixed oxide fuel pin

International Nuclear Information System (INIS)

Buurveld, H.A.; Dassel, G.

1993-12-01

A nondestructive technique as well as a destructive PIE technique have been used to verify the results obtained with a newly 8-e computer tomography (GECT) system. Multi isotope Scanning (MIS), electron probe micro analysis (EPMA) and GECT were used on a mixed oxide (MOX) fuel rod from the Dodewaard reactor with an average burnup of 24 MWd/kg fuel. GECT shows migration of Cs to the periphery of fuel pellets and to radial cracks and pores in the fuel, whereas MIS shows Cs migration to pellet interfaces. The EPMA technique appeared not to be useful to show migration of Cs but, it shows the distribution of fission products from Pu. EPMA clearly shows the distribution of fission products from Pu, but did not reveal the Cs-migration. (orig./HP)

Measurements of Flow Mixing at Subchannels in a Wire-Wrapped 37-Rod Bundle for a Sodium Cooled Fast Reactor

International Nuclear Information System (INIS)

Kim, Hyungmo; Bae, Hwang; Chang, Seok-Kyu; Choi, Sun Rock; Lee, Dong Won; Ko, Yung Joo; Choi, Hae Seob; Euh, Dong-Jin; Lee, Hyeong-Yeon

2014-01-01

For a safety analysis in a core thermal design of a sodium-cooled fast reactor (SFR), flow mixing characteristics at subchannels in a wire-wrapped rod bundle are very important. Wrapped wires make a cross flow in a around the fuel rod) of the fuel rod, and this effect lets flow be mixed. Experimental results of flow mixing can be meaningful for verification and validation of thermal mixing correlation in a reactor core thermo-hydraulic design code. A wire mesh sensing technique can be useful method for measuring of flow mixing characteristics. A wire mesh sensor has been traditionally used to measure the void fraction of a two-phase flow field, i.e. gas and liquid. However, it has been recently reported that the wire mesh sensor can be used successfully to recognize the flow field in liquid phase by injecting a tracing liquid with a different level of electric conductivity. This can be powerfully adapted to recognize flow mixing characteristics by wrapped wires in SFR core thermal design. In this work, we conducted the flow mixing experiments using a custom designed wire mesh sensor. To verify and validate computer codes for the SFR core thermal design, mixing experiments were conducted at a hexagonally arrayed 37-pin wire-wrapped fuel rod bundle test section. The well-designed wire mesh sensor was used to measure flow mixing characteristics. The developed post-processing method has its own merits, and flow mixing results were reasonable. In addition, by uncertainty analysis, the system errors and the random error were estimated in experiments. Therefore, the present results and methods can be used for design code verification and validation

Cesium relocation in mixed-oxide fuel pins resulting from increased temperature reirradiation

International Nuclear Information System (INIS)

Lawrence, L.A.; Woodley, R.E.; Weber, E.T.

1976-06-01

Mixed-oxide fuel pins from EBR-II test subassemblies PNL-3 and PNL-4 were reirradiated in the GETR to study effects of increased fuel and cladding temperatures on chemical and thermomechanical behavior. Radial and axial distributions of cesium were obtained using postirradiation nondestructive precision gamma-scanning techniques. Data presented relate to the dependence of cesium distribution and transport processes on temperature gradients which were altered after substantial steady-state operation

Reactors as a Source of Antineutrinos: Effects of Fuel Loading and Burnup for Mixed-Oxide Fuels

Science.gov (United States)

Bernstein, Adam; Bowden, Nathaniel S.; Erickson, Anna S.

2018-01-01

In a conventional light-water reactor loaded with a range of uranium and plutonium-based fuel mixtures, the variation in antineutrino production over the cycle reflects both the initial core fissile inventory and its evolution. Under an assumption of constant thermal power, we calculate the rate at which antineutrinos are emitted from variously fueled cores, and the evolution of that rate as measured by a representative ton-scale antineutrino detector. We find that antineutrino flux decreases with burnup for low-enriched uranium cores, increases for full mixed-oxide (MOX) cores, and does not appreciably change for cores with a MOX fraction of approximately 75%. Accounting for uncertainties in the fission yields in the emitted antineutrino spectra and the detector response function, we show that the difference in corewide MOX fractions at least as small as 8% can be distinguished using a hypothesis test. The test compares the evolution of the antineutrino rate relative to an initial value over part or all of the cycle. The use of relative rates reduces the sensitivity of the test to an independent thermal power measurement, making the result more robust against possible countermeasures. This rate-only approach also offers the potential advantage of reducing the cost and complexity of the antineutrino detectors used to verify the diversion, compared to methods that depend on the use of the antineutrino spectrum. A possible application is the verification of the disposition of surplus plutonium in nuclear reactors.

Determination of melting point of mixed-oxide fuel irradiated in a fast breeder reactor

International Nuclear Information System (INIS)

Tachibana, Toshimichi

1985-01-01

The melting point of fuel is important to set its in-reactor maximum temperature in fuel design. The fuel melting point measuring methods are broadly the filament method and the capsule sealing method. The only instance of measuring the melting point of irradiated mixed oxide (U, Pu)O 2 fuel by the filament method is by GE in the United States. The capsule sealing method, while the excellent means, is difficult in weld sealing the irradiated fuel in a capsule within the cell. In the fast reactor development program, the remotely operated melting point measuring apparatus in capsule sealing the mixed (U, Pu)O 2 fuel irradiated in the experimental FBR Joyo was set in the cell and the melting point was measured, for the first time in the world. (Mori, K.)

Thermal stress analysis of sulfur deactivated solid oxide fuel cells

Science.gov (United States)

Zeng, Shumao; Parbey, Joseph; Yu, Guangsen; Xu, Min; Li, Tingshuai; Andersson, Martin

2018-03-01

Hydrogen sulfide in fuels can deactivate catalyst for solid oxide fuel cells, which has become one of the most critical challenges to stability. The reactions between sulfur and catalyst will cause phase changes, leading to increase in cell polarization and mechanical mismatch. A three-dimensional computational fluid dynamics (CFD) approach based on the finite element method (FEM) is thus used to investigate the polarization, temperature and thermal stress in a sulfur deactivated SOFC by coupling equations for gas-phase species, heat, momentum, ion and electron transport. The results indicate that sulfur in fuels can strongly affect the cell polarization and thermal stresses, which shows a sharp decrease in the vicinity of electrolyte when 10% nickel in the functional layer is poisoned, but they remain almost unchanged even when the poisoned Ni content was increased to 90%. This investigation is helpful to deeply understand the sulfur poisoning effects and also benefit the material design and optimization of electrode structure to enhance cell performance and lifetimes in various hydrocarbon fuels containing impurities.

Thermal expansion studies on uranium-neodymium mixed oxide solid solutions

International Nuclear Information System (INIS)

Panneerselvam, G.; Venkata Krishnan, R.; Antony, M.P.; Nagarajan, K.

2012-01-01

Uranium-Neodymium mixed oxides solid solutions (U 1-y Nd y ) O 2 (y=0.2-0.95) were prepared by combustion synthesis using citric acid as fuel. Structural characterization and computation of lattice parameter was carried out from room temperature X-ray diffraction measurements. Single-phase fluorite structure was observed up to y=0.80. For solid solutions with y>0.80 additional Nd 2 O 3 lines were visible

Calculation of parameters for inspection planning and evaluation: mixed-oxide fuel fabrication facilities

International Nuclear Information System (INIS)

Reardon, P.T.; Mullen, M.F.

1982-08-01

As part of Task C.35 (Calculation of Parameters for Inspection Planning and Evaluation) of the US Program of Technical Assistance to IAEA Safeguards, Pacific Northwest Laboratory has performed some quantitative analyses of IAEA inspection activities for mixed-oxide fuel fabrication facilities. There were four distinct efforts involved in this task. These were as follows: show the effect on a material balance verification of using two variables measurement methods in some strata; perform additional calculations for the reference facility described in STR-89; modify the INSPECT computer programs to be used as an after-inspection analysis tool, as well as a preinspection planning tool; provide written comments and explantations of text and graphs of the first draft of STR-89, Safeguards Considerations for Mixed-Oxide Fuel Element Fabrication Facilities, by W. Bahm, T. Shea, and D. Tolchenkov, System Studies Section, IAEA

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

Behavior of mixed-oxide fuel elements during the TOPI-1E transient overpower test

International Nuclear Information System (INIS)

Tsai, H.; Neimark, L.A.; Yamamoto, K.; Hirai, K.; Shikakura, S.

1993-12-01

A slow-ramp, extended overpower transient test was conducted on a group of nineteen preirradiated mixed-oxide fuel elements in EBR-II. During the transient two of the test elements with high-density fuel and tempered martensitic cladding (PNC-FMS) breached at an overpower of ∼75%. Fuel elements with austenitic claddings (D9, PNC316, and PNC150), many with aggressive design features and high burnups, survived the overpower transient and incurred little or no cladding strain. Fuel elements with annual fuel or heterogeneous fuel columns also behaved well

Chemical states of fission products in irradiated uranium-plutonium mixed oxide fuel

International Nuclear Information System (INIS)

Kurosaki, Ken; Uno, Masayoshi; Yamanaka, Shinsuke

1999-01-01

The chemical states of fission products (FPs) in irradiated uranium-plutonium mixed oxide (MOX) fuel for the light water reactor (LWR) were estimated by thermodynamic equilibrium calculations on system of fuel and FPs by using ChemSage program. A stoichiometric MOX containing 6.1 wt. percent PuO 2 was taken as a loading fuel. The variation of chemical states of FPs was calculated as a function of oxygen potential. Some pieces of information obtained by the calculation were compared with the results of the post-irradiation examination (PIE) of UO 2 fuel. It was confirmed that the multicomponent and multiphase thermodynamic equilibrium calculation between fuel and FPs system was an effective tool for understanding the behavior of FPs in fuel. (author)

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.

Thermal imaging of solid oxide fuel cell anode processes

Energy Technology Data Exchange (ETDEWEB)

Pomfret, Michael B.; Kidwell, David A.; Owrutsky, Jeffrey C. [Chemistry Division, U.S. Naval Research Laboratory, Washington, DC 20375 (United States); Steinhurst, Daniel A. [Nova Research Inc., Alexandria, VA 22308 (United States)

2010-01-01

A Si-charge-coupled device (CCD), camera-based, near-infrared imaging system is demonstrated on Ni/yttria-stabilized zirconia (YSZ) fragments and the anodes of working solid oxide fuel cells (SOFCs). NiO reduction to Ni by H{sub 2} and carbon deposition lead to the fragment cooling by 5 {+-} 2 C and 16 {+-} 1 C, respectively. When air is flowed over the fragments, the temperature rises 24 {+-} 1 C as carbon and Ni are oxidized. In an operational SOFC, the decrease in temperature with carbon deposition is only 4.0 {+-} 0.1 C as the process is moderated by the presence of oxides and water. Electrochemical oxidation of carbon deposits results in a {delta}T of +2.2 {+-} 0.2 C, demonstrating that electrochemical oxidation is less vigorous than atmospheric oxidation. While the high temperatures of SOFCs are challenging in many respects, they facilitate thermal imaging because their emission overlaps the spectral response of inexpensive Si-CCD cameras. Using Si-CCD cameras has advantages in terms of cost, resolution, and convenience compared to mid-infrared thermal cameras. High spatial ({proportional_to}0.1 mm) and temperature ({proportional_to}0.1 C) resolutions are achieved in this system. This approach provides a convenient and effective analytical technique for investigating the effects of anode chemistry in operating SOFCs. (author)

Thermal imaging of solid oxide fuel cell anode processes

Science.gov (United States)

Pomfret, Michael B.; Steinhurst, Daniel A.; Kidwell, David A.; Owrutsky, Jeffrey C.

A Si-charge-coupled device (CCD), camera-based, near-infrared imaging system is demonstrated on Ni/yttria-stabilized zirconia (YSZ) fragments and the anodes of working solid oxide fuel cells (SOFCs). NiO reduction to Ni by H 2 and carbon deposition lead to the fragment cooling by 5 ± 2 °C and 16 ± 1 °C, respectively. When air is flowed over the fragments, the temperature rises 24 ± 1 °C as carbon and Ni are oxidized. In an operational SOFC, the decrease in temperature with carbon deposition is only 4.0 ± 0.1 °C as the process is moderated by the presence of oxides and water. Electrochemical oxidation of carbon deposits results in a Δ T of +2.2 ± 0.2 °C, demonstrating that electrochemical oxidation is less vigorous than atmospheric oxidation. While the high temperatures of SOFCs are challenging in many respects, they facilitate thermal imaging because their emission overlaps the spectral response of inexpensive Si-CCD cameras. Using Si-CCD cameras has advantages in terms of cost, resolution, and convenience compared to mid-infrared thermal cameras. High spatial (∼0.1 mm) and temperature (∼0.1 °C) resolutions are achieved in this system. This approach provides a convenient and effective analytical technique for investigating the effects of anode chemistry in operating SOFCs.

Experimental evaluation of different mixing promoter for nuclear fuel element by means of a new thermal tracing technique

International Nuclear Information System (INIS)

Silin, Nicolas; Juanico, Luis; Delmastro, Dario

2004-01-01

In this work a new experimental method is used to experimentally evaluate the performance of different appendages promoting the turbulent mixing between the coupled subchannels of nuclear fuel elements.The method used will be introduced in another presentation and consists in the generation and measurement of small thermal traces in the refrigerating water flow between the fuel rods.Because it is suitable for heterogeneous and compact subchannels (as Argentinean fuels) with high water flows in simple and affordable tests at atmospheric pressure, this new method is specially well suited for the design of fuel elements, while it offers advantages over other methods of mixing measurement.The experiments carried out on a small test section proved that the buttons brazed to the fuel rods (similar to the 'turbulence promoters' of the Canflex fuel) had an excellent thermohydraulic performance as compared to different mixing vane designs studied.The thermal traces method developed has shown its potential as a thermohydraulic design tool for the development of advanced nuclear fuels, that eventually incorporate mixing promoter elements. In the case of CARA, and as it includes spacer grids, it could be possible to use them to incorporate these elements without the need of brazing them to the rods (as is the case in Canflex), and therefore without penalizing its integrity [es

Simulated physical inventory verification exercise at a mixed-oxide fuel fabrication facility

International Nuclear Information System (INIS)

Reilly, D.; Augustson, R.

1985-01-01

A physical inventory verification (PIV) was simulated at a mixed-oxide fuel fabrication facility. Safeguards inspectors from the International Atomic Energy Agency (IAEA) conducted the PIV exercise to test inspection procedures under ''realistic but relaxed'' conditions. Nondestructive assay instrumentation was used to verify the plutonium content of samples covering the range of material types from input powders to final fuel assemblies. This paper describes the activities included in the exercise and discusses the results obtained. 5 refs., 1 fig., 6 tabs

Behaviour of defective CANDU fuel: fuel oxidation kinetic and thermodynamic modelling

International Nuclear Information System (INIS)

Higgs, J.

2005-01-01

The thermal performance of operating CANDU fuel under defect conditions is affected by the ingress of heavy water into the fuel element. A mechanistic model has been developed to predict the extent of fuel oxidation in defective fuel and its affect on fuel thermal performance. A thermodynamic treatment of such oxidized fuel has been performed as a basis for the boundary conditions in the kinetic model. Both the kinetic and thermodynamic models have been benchmarked against recent experimental work. (author)

Studies on the dissolution of mixed oxide spent fuel from FBR

International Nuclear Information System (INIS)

Nemoto, Shin-ichi; Shibata, Atsuhiro; Shioura, Takao; Okamoto, Fumitoshi; Tanaka, Yasumasa

1995-01-01

At the Chemical Processing Facility(CPF) in the Tokai Works of the Power Reactor and Nuclear Fuel Development Corporation(PNC), since 1982 Laboratory scale hot experiments have been carried out on the development of reprocessing technology for FBR mixed oxide fuel. The spent fuel pins which have been used in out experiments were irradiated in Experimental Fast Reactor 'Joyo' Phenix (France) and DFR(UK). Burn-up of the fuel pins were 4,400-100,000 MWd/t. This paper Summarizes a dissolution study that have been performed to define the Key parameters affecting dissolution rate such as concentration of nitric acid, burn-up, and temperature. And this paper also discusses about the character of releasing 85 Kr in chopping and dissolution process, and about the amount of insoluble residue. (author)

On the use of thermal NF3 as the fluorination and oxidation agent in treatment of used nuclear fuels

Science.gov (United States)

Scheele, Randall; McNamara, Bruce; Casella, Andrew M.; Kozelisky, Anne

2012-05-01

This paper presents results of our investigation on the use of nitrogen trifluoride as a fluorination or fluorination/oxidation agent for separating valuable constituents from used nuclear fuels by exploiting the different volatilities of the constituent fission product and actinide fluorides. Our thermodynamic calculations show that nitrogen trifluoride has the potential to produce volatile fission product and actinide fluorides from oxides and metals that can form volatile fluorides. Simultaneous thermogravimetric and differential thermal analyses show that the oxides of lanthanum, cerium, rhodium, and plutonium are fluorinated but do not form volatile fluorides when treated with nitrogen trifluoride at temperatures up to 550 °C. However, depending on temperature, volatile fluorides or oxyfluorides can form from nitrogen trifluoride treatment of the oxides of niobium, molybdenum, ruthenium, tellurium, uranium, and neptunium. Thermoanalytical studies demonstrate near-quantitative separation of uranium from plutonium in a mixed 80% uranium and 20% plutonium oxide. Our studies of neat oxides and metals suggest that the reactivity of nitrogen trifluoride may be adjusted by temperature to selectively separate the major volatile fuel constituent uranium from minor volatile constituents, such as Mo, Tc, Ru and from the non-volatile fuel constituents based on differences in their reaction temperatures and kinetic behaviors. This reactivity is novel with respect to that reported for other fluorinating reagents F2, BrF5, ClF3.

EBSD and TEM Characterization of High Burn-up Mixed Oxide Fuel

International Nuclear Information System (INIS)

Teague, Melissa C; Gorman, Brian P.; Miller, Brandon D; King, Jeffrey

2014-01-01

Understanding and studying the irradiation behavior of high burn-up oxide fuel is critical to licensing of future fast breeder reactors. Advancements in experimental techniques and equipment are allowing for new insights into previously irradiated samples. In this work dual column focused ion beam (FIB)/scanning electron microscope (SEM) was utilized to prepared transmission electron microscope samples from mixed oxide fuel with a burn-up of 6.7% FIMA. Utilizing the FIB/SEM for preparation resulted in samples with a dose rate of <0.5 mRem/h compared to approximately 1.1 R/h for a traditionally prepared TEM sample. The TEM analysis showed that the sample taken from the cooler rim region of the fuel pellet had approximately 2.5x higher dislocation density than that of the sample taken from the mid-radius due to the lower irradiation temperature of the rim. The dual column FIB/SEM was additionally used to prepared and serially slice approximately 25 um cubes. High quality electron back scatter diffraction (EBSD) were collected from the face at each step, showing, for the first time, the ability to obtain EBSD data from high activity irradiated fuel

Energetic analysis of a syngas-fueled chemical-looping combustion combined cycle with integration of carbon dioxide sequestration

International Nuclear Information System (INIS)

Jiménez Álvaro, Ángel; Paniagua, Ignacio López; Fernández, Celina González; Carlier, Rafael Nieto; Martín, Javier Rodríguez

2014-01-01

Chemical-looping combustion for power generation has significant advantages over conventional combustion. Mainly, it allows an integration of CO 2 capture in the power plant without energy penalty; secondly, a less exergy destruction in the combustion chemical transformation is achieved, leading to a greater overall thermal efficiency. Most efforts have been devoted to systems based on methane as a fuel, although other systems for alternative fuels have can be proposed. This paper focus on the study of the energetic performance of this concept of combustion in a gas turbine combined cycle when synthesis gas is used as fuel. After optimization of some thermodynamic parameters of the cycle, the power plant performance is evaluated under diverse working conditions and compared to a conventional gas turbine system. Energy savings related with CO 2 capture and storage have been quantified. The overall efficiency increase is found to be significant, reaching values of around 5% (even more in some cases). In order to analyze the influence of syngas composition on the results, different H 2 -content fuels are considered. In a context of real urgency to reduce green house gas emissions, this work is intended to contribute to the conceptual development of highly efficient alternative power generation systems. - Highlights: • Analysis of the energetic performance of a CLC (chemical-looping combustion) gas turbine system is done. • Syngas as fuel and iron oxides as oxygen carrier are considered. • Different H 2 -content syngas are under study. • Energy savings accounting CO 2 sequestration and storage are quantified. • A significant increase on thermal efficiency of about 5–6% is found

Measurements of Flow Mixing at Subchannels in a Wire-Wrapped 61-Rod Bundle for a Sodium Cooled Fast Reactor

International Nuclear Information System (INIS)

Lee, Dong Won; Kim, Hyungmo; Ko, Yung Joo; Choi, Hae Seob; Euh, Dong-Jin; Jeong, Ji-Young; Lee, Hyeong-Yeon

2015-01-01

For a safety analysis in a core thermal design of a sodium-cooled fast reactor (SFR), flow mixing characteristics at subchannels in a wire-wrapped rod bundle are crucial factor for the design code verification and validation. Wrapped wires make a cross flow in a circumference of the fuel rod, and this effect lets flow be mixed. Therefore the sub-channel analysis method is commonly used for thermal hydraulic analysis of a SFR, a wire wrapped sub-channel type. To measure flow mixing characteristics, a wire mesh sensing technique can be useful method. A wire mesh sensor has been traditionally used to measure the void fraction of a two-phase flow field, i.e. gas and liquid. However, the recent reports that the wire mesh sensor can be used successfully to recognize the flow field in liquid phase by injecting a tracing liquid with a different level of electric conductivity. The subchannel flow characteristics analysis method is commonly used for the thermal hydraulic analysis of a SFR, a wire wrapped subchannel type. In this study, mixing experiments were conducted successfully at a hexagonally arrayed 61-pin wire-wrapped fuel rod bundle test section. Wire mesh sensor was used to measure flow mixing characteristics. The developed post-processing method has its own merits, and flow mixing results were reasonable

Development of an engineered safeguards system concept for a mixed-oxide fuel fabrication facility

International Nuclear Information System (INIS)

Chapman, L.D.; de Montmollin, J.M.; Deveney, J.E.; Fienning, W.C.; Hickman, J.W.; Watkins, L.D.; Winblad, A.E.

1976-08-01

An initial concept of an Engineered Safeguards System for a representative commercial mixed-oxide fuel fabrication facility is presented. Computer simulation techniques for evaluation and further development of the concept are described. An outline of future activity is included

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)

Proceedings of the third specialist meeting on sodium/fuel interaction in fast reactors

International Nuclear Information System (INIS)

1976-01-01

This specialist meeting, sponsored by the OECD-NEA and organized by the Power Reactor and Nuclear Fuel Development Corporation, was attended by 56 delegates from 6 countries and the CEC (Commission of the European Communities). The purpose of the meeting was to bring together and discuss in depth the Fuel-Sodium Interaction, a phenomenon of major importance in the assessment of the Hypothetical Core Disruptive Accident in the Liquid Metal Fast Breeder Reactor. The meeting was essentially a follow-up of an earlier meeting held at Ispra in December 1973. In all, 29 papers were presented, covering the following topics: 1. Current perspective on sodium-fuel interaction in LMFBR safety; 2. Basic experimental and theoretical studies including other materials; 3. In-pile and out-of-pile experimental studies on sodium-fuel interaction; 4. Theoretical models for the interpretation of experiments and for application to reactor situations. The meeting is considered useful in narrowing down the chain of events necessary to get energetic interaction, large work potential, but many points are being clarified on the gap between the basic vapor explosions and the real fuel sodium interactions in the HCDA scenario of LMFBR. Finally another meeting of the same nature as this one has been recommended

Thermal-grating contributions to degenerate four-wave mixing in nitric oxide

International Nuclear Information System (INIS)

Danehy, P.M.; Paul, P.H.; Farrow, R.L.

1995-01-01

We report investigations of degenerate four-wave mixing (DFWM) line intensities in the A 2 Σ + left-arrow X 2 Π electronic transitions of nitric oxide. Contributions from population gratings (spatially varying perturbations in the level populations of absorbing species) and thermal gratings (spatially varying perturbations in the overall density) were distinguished and compared by several experimental and analytical techniques. For small quantities of nitric oxide in a strongly quenching buffer gas (carbon dioxide), we found that thermal-grating contributions dominated at room temperature for gas pressures of ∼0.5 atm and higher. In a nearly nonquenching buffer (nitrogen) the population-grating mechanism dominated at pressures of ∼1.0 atm and lower. At higher temperatures in an atmospheric-pressure methane/air flame, population gratings of nitric oxide also dominated. We propose a simple model for the ratio of thermal- to population-grating scattering intensities that varies as P 4 T -4.4 . Preliminary investigations of the temperature dependence and detailed studies of the pressure dependence are in agreement with this model. Measurements of the temporal evolution and the peak intensity of isolated thermal-grating signals are in detailed agreement with calculations based on a linearized hydrodynamic model [J. Opt. Soc. Am. B 12, 384 (1995)]. copyright 1995 Optical Society of America

Autoradiographic measurement of Pu distribution in mixed-oxide nuclear fuel

International Nuclear Information System (INIS)

Green, D.R.; Rasmussen, D.E.; Gray, W.H.

1976-09-01

The autoradiographic method described was developed for rapid, economical determination of the Pu distribution and microhomogeneity in mixed oxide fuel. High Pu concentration regions of any size down to 13 microns in diameter can be reproducibly resolved using this method. The new method uses computerized scanning and analysis, and includes automatic self-calibration to virtually elimate variations resulting from photographic film and processing. The speed of this new method allows analysis of enough data to ensure statistical reliability of occurrence frequencies, even for sparse populations of Pu-rich regions with diameters greater than 60 microns. Determination of these occurrence frequencies is an important factor in controlling fuel quality to ensure safe, efficient operation in a Liquid Metal Fast Breeder Reactor

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

Thermal analysis of a mix up sodium tank and its ebb pipeline for SS-050 circuit during a thermal shock

International Nuclear Information System (INIS)

Jesus Miranda, C.A. de; Gebrim, A.N.

1988-12-01

In this work a thermo-hydraulic model was developed in order to obtain the sodium temperature time history between the mixup tank (TM) and the drain tank of the SS-050 sodium test loop. Results are presented relative to a thermal shock whith initial and final sodium inlet temperature of 600 0 C and 400 0 C respectively, with a thermal gradient of-200 0 C/s. This sodium loop will be briefly installed in the IEN/RJ area. From the sodium temperature time-history during the thermal shock transient the temperature field for the walls of the TM bottom and outlet nozzle is obtained. (author) [pt

Effect of a time varying power level in EBR-II on mixed-oxide fuel burnup

International Nuclear Information System (INIS)

Stone, I.Z.; Jost, J.W.; Baker, R.B.

1979-01-01

A refined prediction of burnup of mixed-oxide fuel in EBR-2 is compared with measured data. The calculation utilizes a time-varying power factor and results in a general improvement to previous calculations

Some alternatives to the mixed oxide fuel cycle

International Nuclear Information System (INIS)

Deonigi, D.E.; Eschbach, E.A.; Goldsmith, S.; Pankaskie, P.J.; Rohrmann, C.A.; Widrig, R.D.

1977-02-01

While on initial examination each of the six fuel cycle concepts (tandem cycle, extended burnup, fuel rejuvenation, coprocessing, partial reprocessing, and thorium) described in the report may have some potential for improving safeguards, none of the six appears to have any other major or compelling advantages over the mixed oxide (MOX) fuel cycle. Compared to the MOX cycle, all but coprocessing appear to have major disadvantages, including severe cost penalties. Three of the concepts-tandem, extended burnup, and rejuvenation--share the basic problems of the throwaway cycle (GESMO Alternative 6): without reprocessing, high-level waste volumes and costs are substantially increased, and overall uranium utilization decreases for three reasons. First, the parasitic fission products left in the fuel absorb neutrons in later irradiation steps reducing the overall neutronic efficiencies of these cycles. Second, discarded fuel still has sufficient fissile values to warrant recycle. Third, perhaps most important, the plutonium needed for breeder start-up will not be available; without the breeder, uranium utilization would drop by about a factor of sixty. Two of the concepts--coprocessing and partial reprocessing--involve variations of the basic MOX fuel cycle's chemical reprocessing step to make plutonium diversion potentially more difficult. These concepts could be used with the MOX fuel cycle or in conjunction with the tandem, extended burnup and rejuvenation concepts to eliminate some of the problems with those cycles. But in so doing, the basic impetus for those cycles--elimination of reprocessing for safeguards purposes--no longer exists. Of all the concepts considered, only coprocessing--and particularly the ''master blend'' version--appears to have sufficient promise to warrant a more detailed study. The master blend concept could possibly make plutonium diversion more difficult with minimal impact on the reprocessing and MOX fuel fabrication operations

Performance of fast reactor mixed-oxide fuels pins during extended overpower transients

International Nuclear Information System (INIS)

Tsai, H.; Neimark, L.A.; Asaga, T.; Shikakura, S.

1991-02-01

The Operational Reliability Testing (ORT) program, a collaborative effort between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corp. (PNC) of Japan, was initiated in 1982 to investigate the behavior of mixed-oxide fuel pin under various slow-ramp transient and duty-cycle conditions. In the first phase of the program, a series of four extended overpower transient tests, with severity sufficient to challenge the pin cladding integrity, was conducted. The objectives of the designated TOPI-1A through -1D tests were to establish the cladding breaching threshold and mechanisms, and investigate the thermal and mechanical effects of the transient on pin behavior. The tests were conducted in EBR-2, a normally steady-state reactor. The modes of transient operation in EBR-2 were described in a previous paper. Two ramp rates, 0.1%/s and 10%/s, were selected to provide a comparison of ramp-rate effects on fuel behavior. The test pins chosen for the series covered a range of design and pre-test irradiation parameters. In the first test (1A), all pins maintained their cladding integrity during the 0.1%/s ramp to 60% peak overpower. Fuel pins with aggressive designs, i.e., high fuel- smear density and/or thin cladding, were, therefore, included in the follow-up 1B and 1C tests to enhance the likelihood of achieving cladding breaching. In the meantime, a higher pin overpower capability, to greater than 100%, was established by increasing the reactor power limit from 62.5 to 75 MWt. In this paper, the significant results of the 1B and 1C tests are presented. 4 refs., 5 figs., 1 tab

Advanced fuels safety comparisons

International Nuclear Information System (INIS)

Grolmes, M.A.

1977-01-01

The safety considerations of advanced fuels are described relative to the present understanding of the safety of oxide fueled Liquid Metal Fast Breeder Reactors (LMFBR). Safety considerations important for the successful implementation of advanced fueled reactors must early on focus on the accident energetics issues of fuel coolant interactions and recriticality associated with core disruptive accidents. It is in these areas where the thermal physical property differences of the advanced fuel have the greatest significance

An assessment of the energetic flows in a commercial PEM fuel-cell system

International Nuclear Information System (INIS)

Jovan, Vladimir; Perne, Matija; Petrovcic, Janko

2010-01-01

Some primary issues have not yet been fully investigated on the way towards the commercialization of fuel-cell-based systems (FCS), e.g., their actual efficiency, reliability, safety, degradation, maintainability, etc. This article deals with an estimation of the real energetic flows and the corresponding electrical efficiency of a commercial proton-exchange-membrane fuel-cell hydrogen-fed generator set (PEMFCS). The fuel-cell power system considered here is planned to be the source of both electrical and thermal energy in a mobile dwelling container unit with in-built fuel-cell-based cogeneration system, and for the design of a cogeneration unit the actual amount of disposable energy from the PEMFC unit should be estimated. The assessment of the actual energetic flows, the disposable energy and the consequent electrical efficiency of the case-study PEMFCS is carried out using commercial technical data for the PEMFCS.

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)

Design of an engineered safeguards system for a mixed-oxide fuel fabrication facility

International Nuclear Information System (INIS)

Winblad, A.E.; McKnight, R.P.; Fienning, W.C.; Fenchel, B.R.

1977-06-01

Several Engineered Safeguards System concepts and designs are described that provide increased protection against a wide spectrum of adversary threats. An adversary sequence diagram that outlines all possible adversary paths through the safeguards elements in a mixed-oxide fuel fabrication facility is shown. An example of a critical adversary path is given

A recommendation for the thermal conductivity of oxide fuels

International Nuclear Information System (INIS)

Kang, K. H.; Ryu, H. J.; Song, K. C.; Yang, M. S.; Na, S. H.; Lee, Y. W.; Moon, H. S.; Kim, H. S.

2004-01-01

The thermal conductivity of nuclear fuel is one of the most important properties because it affects the fuel operating temperature. Therefore, it influences almost all the important processes occurred in nuclear fuel during irradiation, such as gas release, swelling and grain growth. The model of the thermal conductivity of nuclear fuel should be used in the codes to evaluate the performance of it analytically and be required in the nuclear fuel research and development. The thermal conductivity, k, of UO 2 depends on the deviation from stoichiometry, x, the burnup, b, and the fractional porosity, p, as well as the temperature, T: k = k(x, b, p, T), (1) Changes in thermal conductivity occur during irradiation because of fission-gas bubble formation, pores, cracks, fission product build-up and possible changes in the oxygen to uranium ratio (O/U). The dependence on temperature and porosity has been well studied and incorporated in computer codes used for the in-pile fuel behavior analysis. There are several studies on the effect of impurity on the thermal conductivity of UO 2 . In this paper, the variables affected on the thermal conductivity were studied. The available data of the thermal conductivity of UO 2 , UO 2+x , (U, Pu)O 2 , (U, Pu)O 2 and simulated fuel for irradiation fuel were reviewed and analyzed. The best models were recommended

Characterization of aerosols from industrial fabrication of mixed-oxide nuclear reactor fuels

International Nuclear Information System (INIS)

Hoover, M.D.; Newton, G.J.

1997-01-01

Recycling plutonium into mixed-oxide (MOX) fuel for nuclear reactors is being given serious consideration as a safe and environmentally sound method of managing plutonium from weapons programs. Planning for the proper design and safe operation of the MOX fuel fabrication facilities can take advantage of studies done in the 1970s, when recycling of plutonium from nuclear fuel was under serious consideration. At that time, it was recognized that the recycle of plutonium and uranium in irradiated fuel could provide a significant energy source and that the use of 239 Pu in light water reactor fuel would reduce the requirements for enriched 235 U as a reactor fuel. It was also recognized that the fabrication of uranium and plutonium reactor fuels would not be risk-free. Despite engineered safety precautions such as the handling of uranium and plutonium in glove-box enclosures, accidental releases of radioactive aerosols from normal containment might occur. Workers might then be exposed to the released materials by inhalation

Neutronic feasibility of PWR core with mixed oxide fuels in the Republic of Korea

International Nuclear Information System (INIS)

Kim, Y.J.; Joo, H.K.; Jung, H.G.; Sohn, D.S.

1997-01-01

Neutronic feasibility of a PWR core with mixed oxide (MOX) fuels has been investigated as part of the feasibility study for recycling spent fuels in Korea. A typical 3-loop PWR with 900 MWe capacity is selected as reference plant to develop equilibrium core designs with low-leakage fuel management scheme, while incorporating various MOX loading. The fuel management analyses and limited safety analyses show that, safely stated, MOX recycling with 1/3 reload fraction can be accommodated for both annual and 18 month fuel cycle schemes in Korean PWRs, without major design modifications on the reactor systems. (author). 12 refs, 4 figs, 3 tabs

Effect of burnup on the response of stainless steel-clad mixed-oxide fuels to simulated thermal transients

International Nuclear Information System (INIS)

Fenske, G.R.; Badyopadhyay, G.

1981-01-01

Direct electrical heating experiments were performed on irradiated fuel to study the fuel and cladding response as a function of burnup during a slow thermal transient. The results indicated that the nature and extent of the fuel and cladding behavior depended on the quantity of fission gas retained in the fuel. Fission-gas-driven fuel ejection occurred as the molten cladding flowed down the stack exposing bare, radially unrestrained fuel. The fuel dispersion occurred in the absence of molten fuel and the amount of fuel ejected increased with increasing burnup. 31 refs

Coolant-fuel interaction in Sodium-cooled Fast Reactors: Structural investigations of The Na-An-O (An = U, Np, Pu) systems

International Nuclear Information System (INIS)

Smith, A.L.; Raison, P.E.; Bykov, D.M.; Konings, R.J.; Caciuffo, R.; Cheetham, A.K.

2014-01-01

Nuclear energy has the potential to provide Europe with a secure and sustainable electricity supply at a competitive price and to make a significant contribution to the reduction of greenhouse gases emissions. The interest for Sodium-cooled-Fast-spectrum Reactors (SFRs), when compared to Pressurized Water Reactors (PWRs), lies in their more efficient management of plutonium and other actinides as well as their ability to use almost all of the energy in the natural uranium versus 1% utilized in thermal spectrum systems. The high fuel efficiency of fast reactors could greatly dampen concerns about fuel supply. But these reactors have also several drawbacks when compared to PWRs (i.e sodium fire, Na reaction with O2 and H2O, interaction of sodium with oxide fuels). Their development at an industrial scale needs therefore an exhaustive safety assessment that comprises both experimental work and development of sophisticated modelling tools able to describe the reactor behaviour in normal or incidental conditions

Study of Advanced Reactor Mixed Oxide Fuel Production of (U,Th)O2

International Nuclear Information System (INIS)

Busron-Masduki; Damunir; Pristi-Hartati; R-Sukarsono; Bangun-Wasito

2000-01-01

The high price and starting scarcity of reserved of oil drive the people to drill the alternative nuclear energy. Accelerator-driven Transmutation Waste (ATW) is a prospective technology to solve the problem of used fuel waste, to reduce the anxiety of long term disposal waste, to increase the public acceptance of nuclear energy enter into the third millennium. The future of large nuclear energy appears in many-branched industry will depend on the capability to generate relatively low priced fuel on the basis of commercial nuclear energy. Utilization of uranium-233 -thorium cycle insures long-term fuel supply, makes the nuclear energy production more flexible and enables the self-provision regime to be realized in future. Flowsheet of mixed oxide fuel production for advanced reactor of (U,Th)O 2 is a combination of existing manufacturing equipment and quality assurance program from commercial LWR and HTR. The front-end of flowsheet using sol-gel process. The external sol-gel process is chosen due to simple equipment can anticipate refabrication of U-233 which always contains a few hundred ppm of U-232 and its gamma-emitting daughters, besides yielding smaller waste. The decision to choose external sol-gel process encourages to develop External Gelation Thorium (EGT). In order to get higher density and relatively low compaction pressures (i.e. for advanced LWR) adopted flowsheet EGT is developed to be Sol-Gel Microsphere Pelletization (SGMP). Using the optimal parameters, SGMP become established flowsheet for producing mixed oxide fuel of (U,Th)O 2 for advanced reactor. (author)

Fuel cracking in relation to fuel oxidation in support of an out-reactor instrumented defected fuel experiment

Energy Technology Data Exchange (ETDEWEB)

Quastel, A.; Thiriet, C. [Atomic Energy of Canada Limited, Chalk River, ON (Canada); Lewis, B., E-mail: brent.lewis@uoit.ca [Univ. of Ontario Inst. of Tech., Oshawa, ON (Canada); Corcoran, E., E-mail: emily.corcoran@rmc.ca [Royal Military College of Canada, Kingston, ON (Canada)

2014-07-01

An experimental program funded by the CANDU Owners Group (COG) is studying an out-reactor instrumented defected fuel experiment in Stern Laboratories (Hamilton, Ontario) with guidance from Atomic Energy of Canada Limited (AECL). The objective of this test is to provide experimental data for validation of a mechanistic fuel oxidation model. In this experiment a defected fuel element with UO{sub 2} pellets will be internally heated with an electrical heater element, causing the fuel to crack. By defecting the sheath in-situ the fuel will be exposed to light water coolant near normal reactor operating conditions (pressure 10 MPa and temperature 265-310{sup o}C) causing fuel oxidation, especially near the hotter regions of the fuel in the cracks. The fuel thermal conductivity will change, resulting in a change in the temperature distribution of the fuel element. This paper provides 2D r-θ plane strain solid mechanics models to simulate fuel thermal expansion, where conditions for fuel crack propagation are investigated with the thermal J integral to predict fuel crack stress intensity factors. Finally since fuel crack geometry can affect fuel oxidation this paper shows that the solid mechanics model with pre-set radial cracks can be coupled to a 2D r-θ fuel oxidation model. (author)

Thermal conductivity degradation analyses of LWR MOX fuel by the quasi-two phase material model

International Nuclear Information System (INIS)

Kosaka, Yuji; Kurematsu, Shigeru; Kitagawa, Takaaki; Suzuki, Akihiro; Terai, Takayuki

2012-01-01

The temperature measurements of mixed oxide (MOX) and UO 2 fuels during irradiation suggested that the thermal conductivity degradation rate of the MOX fuel with burnup should be slower than that of the UO 2 fuel. In order to explain the difference of the degradation rates, the quasi-two phase material model is proposed to assess the thermal conductivity degradation of the MIMAS MOX fuel, which takes into account the Pu agglomerate distributions in the MOX fuel matrix as fabricated. As a result, the quasi-two phase model calculation shows the gradual increase of the difference with burnup and may expect more than 10% higher thermal conductivity values around 75 GWd/t. While these results are not fully suitable for thermal conductivity degradation models implemented by some industrial fuel manufacturers, they are consistent with the results from the irradiation tests and indicate that the inhomogeneity of Pu content in the MOX fuel can be one of the major reasons for the moderation of the thermal conductivity degradation of the MOX fuel. (author)

Nondestructive characterization of mixed oxide pellets in welded nuclear fuel pins by neutron radiography and gamma-autoradiography

International Nuclear Information System (INIS)

Panakkal, J.P.; Ghosh, J.K.; Roy, P.R.

1989-01-01

Nondestructive evaluation of nuclear fuel pellets after the welding of fuel pins plays a vital role in assuring a safe and reliable operation of reactors. Some of the important characteristics to be monitored in low plutonium enriched mixed oxide fuel pellets are plutonium enrichment, size of plutonium dioxide agglomerates, incorrect loading and geometric shape. Experiments were carried out at Bhabha Atomic Research Centre, Bombay on experimental fuel pins containing mixed oxide pellets of different geometry (solid and annular), of different plutonium enrichment (0-6 w% of plutonium dioxide) and containing PuO 2 agglomerates of size 125-2000 microns to evaluate these characteristics nondestructively. Neutron radiography of these fuel pins was carried out using a swimming pool type reactor 'APSARA'. Results of quantitative evaluation of the neutron radiographs and a simple model correlating neutron interaction probability and the optical density are presented. Gamma autoradiography of these fuel pins showed that these parameters could be evaluated with a few limitations. This paper presents the experimental details, quantitative analysis of the radiographs by microdensitometry and merits and demerits of neutron radiography and gamma autoradiography for nondestructive charcterisation of nuclear fuel pellets. (orig.)

Fission product phases in irradiated carbide fuels

International Nuclear Information System (INIS)

Ewart, F.T.; Sharpe, B.M.; Taylor, R.G.

1975-09-01

Oxide fuels have been widely adopted as 'first charge' fuels for demonstration fast reactors. However, because of the improved breeding characteristics, carbides are being investigated in a number of laboratories as possible advanced fuels. Irradiation experiments on uranium and mixed uranium-plutonium carbides have been widely reported but the instances where segregate phases have been found and subjected to electron probe analysis are relatively few. Several observations of such segregate phases have now been made over a period of time and these are collected together in this document. Some seven fuel pins have been examined. Two of the irradiations were in thermal materials testing reactors (MTR); the remainder were experimental assemblies of carbide gas bonded oxycarbide and sodium bonded oxycarbide in the Dounreay Fast Reactor (DFR). All fuel pins completed their irradiation without failure. (author)

Status of plutonium recycle from mixed oxide fuel fabrication wastes (U,Pu)O2 facility activities

International Nuclear Information System (INIS)

Quesada, Calixto A.; Adelfang, Pablo; Greiner, G.; Orlando, Oscar S.; Mathot, Sergio R.

1999-01-01

Within the specific subject of mixed oxides corresponding to the Fuel Cycle activities performed at CNEA, the recovery of plutonium from wastes originated during tests and pre-fabrication stages is performed. (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)

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

Role of thermal analysis in uranium oxide fuel fabrication process

International Nuclear Information System (INIS)

Balaji Rao, Y.; Yadav, R.B.

2006-01-01

The present paper discusses the application of thermal analysis, particularly, differential thermal analysis (Dta) at various stages of fuel fabrication process. The useful role of Dta in knowing the decomposition pattern and calcination temperature of Adu along with de-nitration temperature is explained. The decomposition pattern depends upon the type of drying process adopted for wet ADU cake (ADU C). Also, the paper highlights the utility of DTA in determining the APS and SSA of UO 2+x and U 3 O 8 powders as an alternate technique. Further, the temperature difference (ΔT max ) between the two exothermic peaks obtained in UO 2+x powder oxidation is related to sintered density of UO 2 pellets. (author)

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

Thermal Signature Measurements for Ammonium Nitrate/Fuel Mixtures by Laser Heating.

Science.gov (United States)

Nazarian, Ashot; Presser, Cary

2016-01-10

Measurements were carried out to obtain thermal signatures of several ammonium nitrate/fuel (ANF) mixtures, using a laser-heating technique referred to as the laser-driven thermal reactor (LDTR). The mixtures were ammonium nitrate (AN)/kerosene, AN/ethylene glycol, AN/paraffin wax, AN/petroleum jelly, AN/confectioner's sugar, AN/cellulose (tissue paper), nitromethane/cellulose, nitrobenzene/cellulose, AN/cellulose/nitromethane, AN/cellulose/nitrobenzene. These mixtures were also compared with AN/nitromethane and AN/diesel fuel oil, obtained from an earlier investigation. Thermograms for the mixtures, as well as individual constituents, were compared to better understand how the sample thermal signature changes with mixture composition. This is the first step in development of a thermal-signature database, to be used along with other signature databases, to improve identification of energetic substances of unknown composition. The results indicated that each individual thermal signature was associated unambiguously with a particular mixture composition. The signature features of a particular mixture were shaped by the individual constituent signatures. It was also uncovered that the baseline signature was modified after an experiment due to coating of unreacted residue on the substrate surface and a change in the reactor sphere oxide layer. Thus, care was required to pre-oxidize the sphere prior to an experiment. A minimum sample mass (which was dependent on composition) was required to detect the signature characteristics. Increased laser power served to magnify signal strength while preserving the signature features. For the mixtures examined, the thermal response of each ANF mixture was found to be different, which was based on the mixture composition and the thermal behavior of each mixture constituent.

Coordinated safeguards for materials management in a mixed-oxide fuel facility

International Nuclear Information System (INIS)

Shipley, J.P.; Cobb, D.D.; Dietz, R.J.; Evans, M.L.; Schelonka, E.P.; Smith, D.B.; Walton, R.B.

1977-02-01

A coordinated safeguards system is described for safeguarding strategic quantities of special nuclear materials in mixed-oxide recycle fuel fabrication facilities. The safeguards system is compatible with industrial process requirements and combines maximum effectiveness consistent with modest cost and minimal process interference. It is based on unit process accounting using a combination of conventional and state-of-the-art NDA measurement techniques. The effectiveness of the system against single and multiple thefts is evaluated using computer modeling and simulation techniques

Coordinated safeguards for materials management in a mixed-oxide fuel facility

Energy Technology Data Exchange (ETDEWEB)

Shipley, J.P.; Cobb, D.D.; Dietz, R.J.; Evans, M.L.; Schelonka, E.P.; Smith, D.B.; Walton, R.B.

1977-02-01

A coordinated safeguards system is described for safeguarding strategic quantities of special nuclear materials in mixed-oxide recycle fuel fabrication facilities. The safeguards system is compatible with industrial process requirements and combines maximum effectiveness consistent with modest cost and minimal process interference. It is based on unit process accounting using a combination of conventional and state-of-the-art NDA measurement techniques. The effectiveness of the system against single and multiple thefts is evaluated using computer modeling and simulation techniques.

Measurement of PuO2 spot in the mixed-oxide fuels

International Nuclear Information System (INIS)

Yokosuka, Yoshifumi; Kaneda, Kenichiro; Ishikawa, Eiji; Nakajima, Katsuaki; Kashima, Sadamitsu

1974-01-01

The homogeneity of mixed-oxide fuel is evaluated with the maximum diameter of PuO 2 spots. PuO 2 spot distribution is influenced by fuel production process and raw materials used. Analytical measurement of the PuO 2 spot distribution was carried out. α-autoradiograph was taken with nitrocellulose film as a detector. The film was exposed for 15 and 30 minutes to the sample pellets, and etched in 10% NaOH for 2 minutes at 60 0 C. Measurement of PuO 2 spots was performed manually with a projector. PuO 2 spot distribution was measured. There is no difference of spot distribution owing to the location in a pellet. However some discrepancy exists in spot number and maximum diameter between surface and inner part. This phenomenon is explained by considering that the diffusion on surface is more effective than in inner part in sintering process. Three pellets were taken out from a lot at random, and α-autoradiograph was taken longitudinally with a pellet and transversely with the others. PuO 2 spot number counted from three autoradiographs was about 150/cm 2 , and the standard deviation was about 1.5. No difference was observed among them. Concerning the influence of mixing method on PuO 2 spot formation, three methods were examined; ball mill mixing (with and without lining) and Kneader mixing. Kneader mixing gave the worst results in maximum spot diameter and spot distribution range. (Tai, I.)

Role of Oxides and Porosity on High-Temperature Oxidation of Liquid-Fueled HVOF Thermal-Sprayed Ni50Cr Coatings

Science.gov (United States)

Song, B.; Bai, M.; Voisey, K. T.; Hussain, T.

2017-02-01

High chromium content in Ni50Cr thermally sprayed coatings can generate a dense and protective scale at the surface of coating. Thus, the Ni50Cr coating is widely used in high-temperature oxidation and corrosion applications. A commercially available gas atomized Ni50Cr powder was sprayed onto a power plant steel (ASME P92) using a liquid-fueled high velocity oxy-fuel thermal spray with three processing parameters in this study. Microstructure of as-sprayed coatings was examined using oxygen content analysis, mercury intrusion porosimetry, scanning electron microscope (SEM), energy-dispersive x-ray spectroscopy (EDX) and x-ray diffraction (XRD). Short-term air oxidation tests (4 h) of freestanding coatings (without boiler steel substrate) in a thermogravimetric analyzer at 700 °C were performed to obtain the kinetics of oxidation of the as-sprayed coating. Long-term air oxidation tests (100 h) of the coated substrates were performed at same temperature to obtain the oxidation products for further characterization in detail using SEM/EDX and XRD. In all samples, oxides of various morphologies developed on top of the Ni50Cr coatings. Cr2O3 was the main oxidation product on the surface of all three coatings. The coating with medium porosity and medium oxygen content has the best high-temperature oxidation performance in this study.

The underwater coincidence counter for plutonium measurements in mixed-oxide fuel assemblies manual

International Nuclear Information System (INIS)

Eccleston, G.W.; Menlove, H.O.; Abhold, M.; Baker, M.; Pecos, J.

1999-01-01

This manual describes the Underwater Coincidence Counter (UWCC) that has been designed for the measurement of plutonium in mixed-oxide (MOX) fuel assemblies prior to irradiation. The UWCC uses high-efficiency 3 He neutron detectors to measure the spontaneous-fission and induced-fission rates in the fuel assembly. Measurements can be made on MOX fuel assemblies in air or underwater. The neutron counting rate is analyzed for singles, doubles, and triples time correlations to determine the 240 Pu effective mass per unit length of the fuel assembly. The system can verify the plutonium loading per unit length to a precision of less than 1% in a measurement time of 2 to 3 minutes. System design, components, performance tests, and operational characteristics are described in this manual

Efficient solar-driven synthesis, carbon capture, and desalinization, STEP: solar thermal electrochemical production of fuels, metals, bleach

Energy Technology Data Exchange (ETDEWEB)

Licht, S. [Department of Chemistry, George Washington University, Washington, DC (United States)

2011-12-15

STEP (solar thermal electrochemical production) theory is derived and experimentally verified for the electrosynthesis of energetic molecules at solar energy efficiency greater than any photovoltaic conversion efficiency. In STEP the efficient formation of metals, fuels, chlorine, and carbon capture is driven by solar thermal heated endothermic electrolyses of concentrated reactants occuring at a voltage below that of the room temperature energy stored in the products. One example is CO{sub 2}, which is reduced to either fuels or storable carbon at a solar efficiency of over 50% due to a synergy of efficient solar thermal absorption and electrochemical conversion at high temperature and reactant concentration. CO{sub 2}-free production of iron by STEP, from iron ore, occurs via Fe(III) in molten carbonate. Water is efficiently split to hydrogen by molten hydroxide electrolysis, and chlorine, sodium, and magnesium from molten chlorides. A pathway is provided for the STEP decrease of atmospheric carbon dioxide levels to pre-industrial age levels in 10 years. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Evolution of thermal stress and failure probability during reduction and re-oxidation of solid oxide fuel cell

Science.gov (United States)

Wang, Yu; Jiang, Wenchun; Luo, Yun; Zhang, Yucai; Tu, Shan-Tung

2017-12-01

The reduction and re-oxidation of anode have significant effects on the integrity of the solid oxide fuel cell (SOFC) sealed by the glass-ceramic (GC). The mechanical failure is mainly controlled by the stress distribution. Therefore, a three dimensional model of SOFC is established to investigate the stress evolution during the reduction and re-oxidation by finite element method (FEM) in this paper, and the failure probability is calculated using the Weibull method. The results demonstrate that the reduction of anode can decrease the thermal stresses and reduce the failure probability due to the volumetric contraction and porosity increasing. The re-oxidation can result in a remarkable increase of the thermal stresses, and the failure probabilities of anode, cathode, electrolyte and GC all increase to 1, which is mainly due to the large linear strain rather than the porosity decreasing. The cathode and electrolyte fail as soon as the linear strains are about 0.03% and 0.07%. Therefore, the re-oxidation should be controlled to ensure the integrity, and a lower re-oxidation temperature can decrease the stress and failure probability.

Energetic materials and methods of tailoring electrostatic discharge sensitivity of energetic materials

Energy Technology Data Exchange (ETDEWEB)

Daniels, Michael A.; Heaps, Ronald J.; Wallace, Ronald S.; Pantoya, Michelle L.; Collins, Eric S.

2016-11-01

An energetic material comprising an elemental fuel, an oxidizer or other element, and a carbon nanofiller or carbon fiber rods, where the carbon nanofiller or carbon fiber rods are substantially homogeneously dispersed in the energetic material. Methods of tailoring the electrostatic discharge sensitivity of an energetic material are also disclosed.

Fuel Thermal Expansion (FTHEXP)

International Nuclear Information System (INIS)

Reymann, G.A.

1978-07-01

A model is presented which deals with dimensional changes in LWR fuel pellets caused by changes in temperature. It is capable of dealing with any combination of UO 2 and PuO 2 in solid, liquid or mixed phase states, and includes expansion due to the solid-liquid phase change. The function FTHEXP models fuel thermal expansion as a function of temperature, fraction of PuO 2 , and the fraction of fuel which is molten

Fuel-steel mixing and radial mesh effects in power excursion simulations

International Nuclear Information System (INIS)

Chen, X.-N.; Rineiski, A.; Gabrielli, F.; Andriolo, L.; Vezzoni, B.; Li, R.; Maschek, W.; Kiefhaber, E.

2016-01-01

Highlights: • Fuel-steel mixing and radial mesh effects are significant on power excursion. • The earliest power peak is reduced and retarded by these two effects. • Unprotected loss of coolant transients in ESFR core are calculated. - Abstract: This paper deals with SIMMER-III once-through simulations of the earliest power excursion initiated by an unprotected loss of flow (ULOF) in the Working Horse design of the European Sodium Cooled Fast Reactor (ESFR). Since the sodium void effect is strictly positive in this core and dominant in the transient, a power excursion is initiated by sodium boiling in the ULOF case. Two major effects, namely (1) reactivity effects due to fuel-steel mixing after melting and (2) the radial mesh size, which were not considered originally in SIMMER simulations for ESFR, are studied. The first effect concerns the reactivity difference between the heterogeneous fuel/clad/wrapper configuration and the homogeneous mixture of steel and fuel. The full core homogenization (due to melting) effect is −2 $, though a smaller effect takes place in case of partial core melting. The second effect is due to the SIMMER sub-assembly (SA) coarse mesh treatment, where a simultaneous sodium boiling onset in all SAs belonging to one ring leads to an overestimated reactivity ramp. For investigating the influence of fuel/steel mixing effects, a lumped “homogenization” reactivity feedback has been introduced, being proportional to the molten steel mass. For improving the coarse mesh treatment, we employ finer radial meshes to take the subchannel effects into account, where the side and interior channels have different coolant velocities and temperatures. The simulation results show that these two effects have significant impacts on the earliest power excursion after the sodium boiling.

Stabilized super-thermite colloids: A new generation of advanced highly energetic materials

Science.gov (United States)

Elbasuney, Sherif; Gaber Zaky, M.; Radwan, Mostafa; Mostafa, Sherif F.

2017-10-01

One of the great impetus of nanotechnology on energetic materials is the achievement of nanothermites (metal-oxide/metal) which are characterized by massive heat output. Yet, full exploitation of super-thermites in highly energetic systems has not been achieved. This manuscript reports on the sustainable fabrication of colloidal Fe2O3 and CuO nanoparticles for thermite applications. TEM micrographs demonstrated mono-dispersed Fe2O3 and CuO with an average particle size of 3 and 15 nm respectively. XRD diffractograms demonstrated highly crystalline materials. SEM micrographs demonstrated a great tendency of the developed oxides to aggregate over drying process. The effective integration and dispersion of mono-dispersed colloidal thermite particles into energetic systems are vital for enhanced performance. Aluminum is of interest as highly energetic metal fuel. In this paper, synthesized Fe2O3 and CuO nanoparticles were re-dispersed in isopropyl alcohol (IPA) with aluminum nanoparticles using ultrasonic prope homogenizer. The colloidal thermite peraticles can be intgegrated into highly energetic system for subsequent nanocomposite development. Thanks to stabilization of colloidal CuO nanoparticles in IPA which could offer intimate mixing between oxidizer and metal fuel. The stabilization mechanism of CuO in IPA was correlated to steric stabilization with solvent molecules. This approach eliminated nanoparticle drying and the re-dispersion of dry aggregates into energetic materials. This manuscript shaded the light on the real development of colloidal thermite mixtures and their integration into highly energetic systems.

Transient redistribution of intragranular fission gas in irradiated mixed oxide

International Nuclear Information System (INIS)

Hinman, C.A.; Randklev, E.H.

1981-01-01

Safety analyses for an LMFBR require a knowledge of the fuel and fission gas behavior under transient conditions. Analyses of microstructural data derived from transiently heated, irradiated, mixed oxide fuel specimens have allowed the calculation of the degree of nonequilibrium of intragranular bubbles formed during the transient. It is hypothesized that the observed over-pressurization of the intragranular bubbles mechanically loads the fuel within the grain, leading to a stress gradient derived force upon near-grain-surface bubbles, driving them preferentially to the grain boundaries. Using existing models for forced diffusion it can be estimated that the stress derived forces on bubbles are within the same magnitude, and possibly greater, than the forces derived from the thermal gradient

The use of the average plutonium-content for criticality evaluation of boiling water reactor mixed oxide-fuel transport and storage packages

International Nuclear Information System (INIS)

Mattera, C.

2003-01-01

Currently in France, criticality studies in transport configurations for Boiling Water Reactor Mixed Oxide fuel assemblies are based on conservative hypothesis assuming that all rods (Mixed Oxide (Uranium and Plutonium), Uranium Oxide, Uranium and (Gadolinium Oxide rods) are Mixed Oxide rods with the same Plutonium-content, corresponding to the maximum value. In that way, the real heterogeneous mapping of the assembly is masked and covered by an homogenous Plutonium-content assembly, enriched at the maximum value. As this calculation hypothesis is extremely conservative, Cogema Logistics (formerly Transnucleaire) has studied a new calculation method based on the use of the average Plutonium-content in the criticality studies. The use of the average Plutonium-content instead of the real Plutonium-content profiles provides a highest reactivity value that makes it globally conservative. This method can be applied for all Boiling Water Reactor Mixed Oxide complete fuel assemblies of type 8 x 8, 9 x 9 and 10 x 10 which Plutonium-content in mass weight does not exceed 15%; it provides advantages which are discussed in the paper. (author)

Autoxidation of jet fuels: Implications for modeling and thermal stability

Energy Technology Data Exchange (ETDEWEB)

Heneghan, S.P. [Univ. of Dayton Research Institute, OH (United States); Chin, L.P. [Systems Research Laboratories, Inc., Dayton, OH (United States)

1995-05-01

The study and modeling of jet fuel thermal deposition is dependent on an understanding of and ability to model the oxidation chemistry. Global modeling of jet fuel oxidation is complicated by several facts. First, liquid jet fuels are hard to heat rapidly and fuels may begin to oxidize during the heat-up phase. Non-isothermal conditions can be accounted for but the evaluation of temperature versus time is difficult. Second, the jet fuels are a mixture of many compounds that may oxidize at different rates. Third, jet fuel oxidation may be autoaccelerating through the decomposition of the oxidation products. Attempts to model the deposition of jet fuels in two different flowing systems showed the inadequacy of a simple two-parameter global Arrhenius oxidation rate constant. Discarding previous assumptions about the form of the global rate constants results in a four parameter model (which accounts for autoacceleration). This paper discusses the source of the rate constant form and the meaning of each parameter. One of these parameters is associated with the pre-exponential of the autoxidation chain length. This value is expected to vary inversely to thermal stability. We calculate the parameters for two different fuels and discuss the implication to thermal and oxidative stability of the fuels. Finally, we discuss the effect of non-Arrhenius behavior on current modeling of deposition efforts.

Kinetics of sodium borohydride direct oxidation and oxygen reduction in sodium hydroxide electrolyte

International Nuclear Information System (INIS)

Chatenet, Marian; Micoud, Fabrice; Roche, Ivan; Chainet, Eric

2006-01-01

The direct oxidation of sodium borohydride in concentrated sodium hydroxide medium has been studied by cyclic and linear voltammetry, chronoamperometry and chronopotentiometry for silver and gold electrocatalysts, either bulk and polycrystalline or nanodispersed over high area carbon blacks. Gold and silver yield rather complete utilisation of the reducer: around 7.5 electrons are delivered on these materials, versus 4 at the most for platinum as a result of the BH 4 - non-negligible hydrolysis taking place on this latter material. The kinetic parameters for the direct borohydride oxidation are better for gold than for silver. A strong influence of the ratio of sodium hydroxide versus sodium borohydride is found: whereas the theoretical stoichiometry does forecast that eight hydroxide ions are needed for each borohydride ion, our experimental results prove that a larger excess hydroxide ion is necessary in quasi-steady state conditions. When the above-mentioned ratio is unity (1 M NaOH and 1 M NaBH 4 ), the tetrahydroborate ions direct oxidation is limited by the hydroxide concentration, and their hydrolysis is no longer negligible. The hydrolysis products are probably BH 3 OH - ions, for which gold displays a rather good oxidation activity. Additionally, silver, which is a weak BH 4 - oxidation electrocatalyst, exhibits the best activity of all the studied materials towards the BH 3 OH - direct oxidation. Finally, carbon-supported gold nanoparticles seem promising as anode material to be used in direct borohydride fuel cells

Fe-catalyzed thermal conversion of sodium lignosulfonate to graphene

Science.gov (United States)

Sung Phil Mun; Zhiyong Cai; Jilei Zhang

2013-01-01

Sodium lignosulfonate (LS) from sulfite pulping processing was used as a carbon source to synthesize graphene. LS was mixed with Fe nanoparticles (FeNPs) as a catalyst and thermally treated at 1000 °C for 1 h. The Raman spectrum and X-ray diffraction pattern suggested that graphene sheets were formed in LS thermally treated with FeNPs (Fe-HTLS). Scanning...

Fuel and core design study of the sodium-cooled fast reactors. Studies on metallic fuel cores in the JFY2002

International Nuclear Information System (INIS)

Sugino, Kazuteru; Mizuno, Tomoyasu

2003-06-01

Based on the results obtained in the former feasibility study, the metallic fueled core of ordinary-type, that is, 2-region homogeneous core, has been established aiming at the improvement in the core performance, and subsequent comparison has been performed with the mixed oxide fueled core. Further, the attractive concept of the metallic fueled core of high outlet temperature has been constructed which has good nuclear features as a metallic fueled core and has identical outlet temperature to mixed oxide fuelled core. Following items have been found as a result of the investigation on the ordinary-type core. The metallic fueled core whose maximum fast neutron fluence (En>0.1MeV) is set identical (5x10 23 n/cm 2 ) to the mixed oxide fueled cores with core discharge burnup 150GWd/t has sufficient core performances as a metallic fueled core, e.g. higher breeding ratio and longer operation period compared with mixed oxide fueled cores, but the core discharge burnup is limited up to 100GWd/t. However effective discharge burnup including the contribution of the blanket region is comparative to mixed oxide cores under the same breeding ratio condition. In order to enlarge the core discharge burnup to 150GWd/t keeping the core performance identical to above mentioned core's, the irradiation deformation of structural material should be reduced to that of mixed oxide fueled cores. Further the maximum fast neutron fluence reaches to 7-8x10 23 n/cm 2 (En>0.1MeV). The investigations on the core of high outlet temperature have clarified following items. Even in the change of core regions by pin-diameter form 3-region to 2-region and in the limited maximum fuel pin diameter 8.5 mm, realization of the identical outlet/inlet temperatures to the mixed oxide cores (550/395degC) is feasible under the criteria of the maximum temperature 650degC at the inner surface of the cladding. The constructed core accommodates the targets of breeding ratio from about 1.0 to 1.2 only by adjusting

Neutronic and Logistic Proposal for Transmutation of Plutonium from Spent Nuclear Fuel as Mixed-Oxide Fuel in Existing Light Water Reactors

International Nuclear Information System (INIS)

Trellue, Holly R.

2004-01-01

The use of light water reactors (LWRs) for the destruction of plutonium and other actinides [especially those in spent nuclear fuel (SNF)] is being examined worldwide. One possibility for transmutation of this material is the use of mixed-oxide (MOX) fuel, which is a combination of uranium and plutonium oxides. MOX fuel is used in nuclear reactors worldwide, so a large experience base for its use already exists. However, to limit implementation of SNF transmutation to only a fraction of the LWRs in the United States with a reasonable number of license extensions, full cores of MOX fuel probably are required. This paper addresses the logistics associated with using LWRs for this mission and the design issues required for full cores of MOX fuel. Given limited design modifications, this paper shows that neutronic safety conditions can be met for full cores of MOX fuel with up to 8.3 wt% of plutonium

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

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

Portable calorimeter system for nondestructive assay of mixed-oxide fuels

International Nuclear Information System (INIS)

Roche, C.T.; Perry, R.B.; Lewis, R.N.; Jung, E.A.; Haumann, J.R.

1978-04-01

Calorimetric assay provides a precise, nondestructive method to determine sample Pu content based on the heat emitted by decaying radionuclides. This measurement, in combination with a gamma-spectrometer analysis of sample isotopic content, yields the total sample Pu mass. The technique is applicable to sealed containers and is essentially independent of sample matrix configuration and elemental composition. Conventional calorimeter designs employ large water-bath heat sinks and lack the portability needed by inspection personnel. The ANL air-chamber isothermal calorimeters are low-thermal-capacitance devices which eliminate the need for large constant-temperature heat sinks. These instruments are designed to use a feedback system that applies power to maintain the sample chamber at a constant electrical resistance and, therefore, at a constant temperature. The applied-power difference between a Pu-containing sample and a blank determines the radioactive-decay power. The operating characteristics of a calorimeter designed for assaying mixed-oxide powders, fuel pellets, and Pu-containing solutions are discussed. This device consists of the calorimeter, sample preheatr, and a microprocessor-controlled data-acquisition system. The small-sample device weighs 18 kg and has a measurement cycle of 20 min, with a precision of 0.1% at 10 mW. A 100-min gamma-ray measurement gives the specific power with a precision of better than 1% for samples containing 1 to 2 g of plutonium

Model for Fuel-Sodium Interaction - Application to the JEF Experiments

International Nuclear Information System (INIS)

Breton, J.P.; Antonakas, D.

1976-01-01

A model of sodium-fuel interaction, referred to as TRACONABUEE, has been developed. The fuel particles are assumed to be introduces in the interacting zone within a finite mixing time, according to a given function (not necessarily linear). The equations for heat transfer inside fuel particles are those of Cho and Wright (transient conduction for phase A and quasi-steady state heat transfer for phase B). During phase B several options for heat transfer from fuel to sodium can be assumed (no transfer, transfer proportional to the volume fraction of liquid sodium, given duration of transfer, etc... ) Two versions are available: a spherical one (EPISCOPOS) and an axial one (TEXAS). For application to the JEF experiments a model of heat losses along the cold column had to be introduced into TEXAS. It was found that the phenomenon is essentially governed by the heat losses. The velocity of the cold sodium in the column presents marked maxima and minima. The agreement with experiment is satisfactory. In conclusion: Due to their simple well-defined geometry, the JEF experiments can be profitably interpreted. They are inadequate for the determination of the interacting sodium mass. On the other hand they allow to fit a simple, parametric, two-phase heat transfer model, suitable for this type of experiments. Finally they show the great importance of the heat losses when the mass of molten fuel is small. These- latter alone explain the phenomenon

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

Trends for Methane Oxidation at Solid Oxide Fuel Cell Conditions

DEFF Research Database (Denmark)

Kleis, Jesper; Jones, Glenn; Abild-Pedersen, Frank

2009-01-01

First-principles calculations are used to predict a plausible reaction pathway for the methane oxidation reaction. In turn, this pathway is used to obtain trends in methane oxidation activity at solid oxide fuel cell (SOFC) anode materials. Reaction energetics and barriers for the elementary...... the Ni surfaces to other metals of interest. This allows the reactivity over the different metals to be understood in terms of two reactivity descriptors, namely, the carbon and oxygen adsorption energies. By combining a simple free-energy analysis with microkinetic modeling, activity landscapes of anode...

GAP pre-polymer, as an energetic binder and high performance additive for propellants and explosives: A review

Directory of Open Access Journals (Sweden)

Mehmet S. Eroglu

2017-08-01

Full Text Available In preparation of energetic composite formulations, functionally terminated pre-polymers have been used as binder. After physically mixing the pre-polymers with oxidizing components, metallic fuel, burning rate modifier and other minor ingredients, they are cured with a suitable curing agent to provide physical and chemical stability. These pre-polymers could be functionalized with carboxyl, epoxide or hydroxyl groups at varying average chain functionalities. For carboxyl-terminated pre-polymers, an epoxy functional curing agents could be used. If the pre-polymer possesses hydroxyl groups, isocyanate functional curing agents are the most suitable curing agents in terms of easy and efficient processing. Glycidyl azide polymer (GAP is one of the well-known low-molecular weight energetic liquid pre-polymer, which was developed to use as energetic binder, high performance additive and gas generator for high performance smokeless composite propellant and explosive formulations. Linear or branched GAP can be synthesized by nucleophilic substitution reaction of corresponding poly(epichlorohydrin (PECH with sodium azide through replacement of chloromethyl groups of PECH with pendant energetic azido-methyl groups on the polyether main chain. Positive heat of formation (+957 kJ/kg enables exothermic and rapid decomposition of GAP producing fuel rich gases. Its polyether main chain provides GAP with relatively low glass transition temperature (Tg= - 48 oC and presence of hydroxyl functional groups allows it to have easy processing in curing with isocyanate curing agents to form covalently crosslinked polyurethane structure. These outstanding properties of GAP enable it to be used as energetic polymeric binder and high performance additive in preparation of energetic materials and low vulnerable explosives.

Increase of thermal conductivity of uranium dioxide nuclear fuel pellets with beryllium oxide addition

International Nuclear Information System (INIS)

Camarano, D.M.; Mansur, F.A.; Santos, A.M.M. dos; Ferraz, W.B.

2016-01-01

The UO_2 fuel is one of the most used nuclear fuel in thermal reactors and has many advantages such as high melting point, chemical compatibility with cladding, etc. However, its thermal conductivity is relatively low, which leads to a premature degradation of the fuel pellets due to a high radial temperature gradient during reactor operation. An alternative to avoid this problem is to increase the thermal conductivity of the fuel pellets, by adding beryllium oxide (BeO). Pellets of UO_2 and UO_2-BeO were obtained from a homogenized mixture of powders of UO_2 and BeO, containing 2% and 3% by weight of BeO and sintering at 1750 °C for 3 h under H_2 atmosphere after uniaxial pressing at 400 MPa. The pellet densities were obtained by xylol penetration-immersion method and the thermal diffusivity, specific heat and thermal conductivity were determined according to ASTM E-1461 at room temperature (25 deg C) and 100 deg C. The thermal diffusivity measurements were carried out employing the laser flash method. The thermal conductivity obtained at 25 deg C showed an increase with the addition of 2% and 3% of BeO corresponding to 19% and 28%, respectively. As for the measurements carried out at 100 deg C, there was an increase in the thermal conductivity for the same BeO contents of 20% and 31%. These values as a percentage of increased conductivity were obtained in relation to the UO_2 pellets. (author)

Uranium and plutonium distribution in unirradiated mixed oxide fuel from industrial fabrication

International Nuclear Information System (INIS)

Hanus, D.; Kleykamp, H.

1982-01-01

Different process variants developed in the last few years by the firm ALKEM to manufacture FBR and LWR mixed oxide fuel are given. The uranium and plutonium distribution is determined on the pellets manufactured with the help of the electron beam microprobe. The stepwise improvement of the uranium-plutonium homogeneity in the short-term developed granulate variants and in the long-term developed new processes are illustrated starting with early standard processes for FBR fuel. An almost uniform uranium-plutonium distribution could be achieved for the long-term developed new processes (OKOM, AuPuC). The uranium-plutonium homogeneity are quantified in the pellets manufactured according to the considered process variants with a newly defined quality number. (orig.)

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

Energy Technology Data Exchange (ETDEWEB)

Roh, Gyu Hong; Choi, Hang Bok

2005-09-15

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

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

International Nuclear Information System (INIS)

Roh, Gyu Hong; Choi, Hang Bok

2005-09-01

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

Mosaic-shaped cathode for highly durable solid oxide fuel cell under thermal stress

Science.gov (United States)

Joo, Jong Hoon; Jeong, Jaewon; Kim, Se Young; Yoo, Chung-Yul; Jung, Doh Won; Park, Hee Jung; Kwak, Chan; Yu, Ji Haeng

2014-02-01

In this study, we propose a novel "mosaic structure" for a SOFC (solid oxide fuel cell) cathode with high thermal expansion to improve the stability against thermal stress. Self-organizing mosaic-shaped cathode has been successfully achieved by controlling the amount of binder in the dip-coating solution. The anode-supported cell with mosaic-shaped cathode shows itself to be highly durable performance for rapid thermal cycles, however, the performance of the cell with a non-mosaic cathode exhibits severe deterioration originated from the delamination at the cathode/electrolyte interface after 7 thermal cycles. The thermal stability of an SOFC cathode can be evidently improved by controlling the surface morphology. In view of the importance of the thermal expansion properties of the cathode, the effects of cathode morphology on the thermal stress stability are discussed.

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

Development of DIPRES feed for the fabrication of mixed-oxide fuels for fast breeder reactors

International Nuclear Information System (INIS)

Griffin, C.W.; Rasmussen, D.E.; Lloyd, M.H.

1983-01-01

The DIrect PREss Spheroidized feed process combines the conversion of uranium-plutonium solutions into spheres by internal gelation with conventional pellet fabrication techniques. In this manner, gel spheres could replace conventional powders as the feed material for pellet fabrication of nuclear fuels. Objective of the DIPRES feed program is to develop and qualify a process to produce mixed-oxide fuel pellets from gel spheres for fast breeder reactors. This process development includes both conversion and fabrication activities

Review of behavior of mixed-oxide fuel elements in extended overpower transient tests in EBR-II

International Nuclear Information System (INIS)

Tsai, H.; Neimark, L.A.

1994-10-01

From a series of five tests conducted in EBR-II, a substantial data base has been established on the performance of mixed-oxide fuel elements in a liquid-metal-cooled reactor under slow-ramp transient overpower conditions. Each test contained 19 preirradiated fuel elements with varying design and prior operating histories. Elements with aggressive design features, such as high fuel smear density and/or thin cladding, were included to accentuate transient effects. The ramp rates were either 0.1 or 10% ΔP/P/s and the overpowers ranged between ∼60 and 100% of the elements' prior power ratings. Six elements breached during the tests, all with aggressive design parameters. The other elements, including all those with moderate design features for the reference or advanced long-life drivers for PNC's prototype fast reactor Monju, maintained their cladding integrity during the tests. Posttest examination results indicated that fuel/cladding mechanical interaction (FCMI) was the most significant mechanism causing the cladding strain and breach. In contrast, pressure loading from the fission gas in the element plenum was less important, even in high-burnup elements. During an overpower transient, FCMI arises from fuel/cladding differential thermal expansion, transient fuel swelling, and, significantly, the gas pressure in the sealed central cavity of elements with substantial centerline fuel melting. Fuel performance data from these tests, including cladding breaching margin and transient cladding strain, are correlatable with fuel-element design and operating parameters. These correlations are being incorporated into fuel-element behavior codes. At the two tested ramp rates, fuel element behavior appears to be insensitive to transient ramp rate and there appears to be no particular vulnerability to slow ramp transients as previously perceived

Thermal hydraulic analysis of Pb-Bi cooled HYPER fuel assemblies using SLTHEN code

International Nuclear Information System (INIS)

Tak, Nam Il; Song, Tae Y.; Park, Won S.; Kim, Chang Hyun

2002-12-01

In the present work, the existing SLTHEN code, which had been originally developed for subchannel analysis of sodium cooled fast reactors, was modified and applied to the Pb-Bi cooled HYPER core which consists of 237 fuel assemblies (TRU assemblies). In the analysis of single fuel assembly having chopped cosine power profile, the validation and the assessment of usefulness of the modified SLTHEN were focused. In the quantitative comparison, the results of the modified SLTHEN agreed well with those of analytical calculations and of MATRA. For the qualitative approaches, the sensitivity calculations for intra-assembly gap flow and turbulent mixing parameter were used. The sensitivity analysis results showed that the modified SLTHEN can provide reasonable simulations of subchannel thermal hydraulics. In particular, turbulent mixing parameter which is known as the most uncertain parameter in subchannel analyses did not affect largely the maximum cladding temperature. Therefore, it can be said that the results of single assembly show the usefulness of the modified SLTHEN code for thermal hydraulic analysis and design of HYPER under the conceptual design stage. In order to assess intra-assembly heat transfer, subchannel analyses were implemented for two types of 7 assemblies; 1) artificial 7 fuel assemblies to maximize intra-assembly heat transfer, 2) central 7 fuel assemblies in the HYPER reference core. The results showed that the modified SLTHEN can reasonably simulate intra-heat transfer and the amount of intra-assembly heat transfer is not so large in HYPER conditions. Particularly, intra-heat transfer did not affect the maximum coolant and the maximum cladding temperatures which are major parameters in conceptual core designs. The capability of full core thermal hydraulic analysis was confirmed by the analysis of 45 fuel assemblies in 1/6 HYPER core at the first cycle. The SLTHEN predicted that the reference design parameters are acceptable in terms of thermal

Green colorants based on energetic azole borates.

Science.gov (United States)

Glück, Johann; Klapötke, Thomas M; Rusan, Magdalena; Stierstorfer, Jörg

2014-11-24

The investigation of green-burning boron-based compounds as colorants in pyrotechnic formulations as alternative for barium nitrate, which is a hazard to health and to the environment, is reported. Metal-free and nitrogen-rich dihydrobis(5-aminotetrazolyl)borate salts and dihydrobis(1,3,4-triazolyl)borate salts have been synthesized and characterized by NMR spectroscopy, elemental analysis, mass spectrometry, and vibrational spectroscopy. Their thermal and energetic properties have been determined as well. Several pyrotechnic compositions using selected azolyl borate salts as green colorants were investigated. Formulations with ammonium dinitramide and ammonium nitrate as oxidizers and boron and magnesium as fuels were tested. The burn time, dominant wavelength, spectral purity, luminous intensity, and luminous efficiency as well as the thermal and energetic properties of these compositions were measured. © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

76 FR 65544 - Standard Format and Content of License Applications for Mixed Oxide Fuel Fabrication Facilities

Science.gov (United States)

2011-10-21

... NUCLEAR REGULATORY COMMISSION [NRC-2009-0323] Standard Format and Content of License Applications... revision to regulatory guide (RG) 3.39, ``Standard Format and Content of License Applications for Mixed Oxide Fuel Fabrication Facilities.'' This guide endorses the standard format and content for license...

Energetics and Defect Interactions of Complex Oxides for Energy Applications

Science.gov (United States)

Solomon, Jonathan Michael

The goal of this dissertation is to employ computational methods to gain greater insights into the energetics and defect interactions of complex oxides that are relevant for today's energy challenges. To achieve this goal, the development of novel computational methodologies are required to handle complex systems, including systems containing nearly 650 ions and systems with tens of thousands of possible atomic configurations. The systems that are investigated in this dissertation are aliovalently doped lanthanum orthophosphate (LaPO4) due to its potential application as a proton conducting electrolyte for intermediate temperature fuel cells, and aliovalently doped uranium dioxide (UO2) due to its importance in nuclear fuel performance and disposal. First we undertake density-functional-theory (DFT) calculations on the relative energetics of pyrophosphate defects and protons in LaPO4, including their binding with divalent dopant cations. In particular, for supercell calculations with 1.85 mol% Sr doping, we investigate the dopant-binding energies for pyrophosphate defects to be 0.37 eV, which is comparable to the value of 0.34 eV calculated for proton-dopant binding energies in the same system. These results establish that dopant-defect interactions further stabilize proton incorporation, with the hydration enthalpies when the dopants are nearest and furthest from the protons and pyrophosphate defects being -1.66 eV and -1.37 eV, respectively. Even though our calculations show that dopant binding enhances the enthalpic favorability of proton incorporation, they also suggest that such binding is likely to substantially lower the kinetic rate of hydrolysis of pyrophosphate defects. We then shift our focus to solid solutions of fluorite-structured UO 2 with trivalent rare earth fission product cations (M3+=Y, La) using a combination of ionic pair potential and DFT based methods. Calculated enthalpies of formation with respect to constituent oxides show higher

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

The 3D thermal-hydraulic numerical simulation for the fuel zone outlet of China experimental fast reactor

International Nuclear Information System (INIS)

Xue Xiuli; Yang Hongyi; Yang Fuchang

2008-01-01

Detailed 3D thermal-hydraulic numerical analyses to the fuel zone outlet are actualized with the STAR-CD CFD code. The performance of sodium mixing is studied and detailed velocity and temperature distribution are obtained in this region which will offer foundations and references to study the rationality of temperature monitoring-spot arrangement and to assess the effect of temperature fluctuations to control rod guide tubes in this region, and so on. (authors)

Synthesis, characterization and thermal expansion studies on thorium-praseodymium mixed oxide solid solutions

International Nuclear Information System (INIS)

Panneerselvam, G.; Antony, M.P.; Srinivasan, T.G.; Vasudeva Rao, P.R.

2010-01-01

Full text: Thorium-praseodymium mixed oxide solid solutions containing 15, 25, 40 and 55 mole percent of praseodymia were synthesized by mixing the solutions of thorium nitrate in water and praseodymium oxide (Pr 6 O 11 ) in conc. HNO 3 . Subsequently, their hydroxides were co-precipitated by the addition of aqueous ammonia. Further the precipitate was dried at 50 deg C, calcined at 600 deg C for 4 hours and sintered at 1200 deg C for 6 h in air. X-ray diffraction measurements were performed for phase identification and lattice parameter derivation. Single-phase fluorite structure was observed for all the compositions. Bulk and theoretical densities of solid solutions were also determined by immersion and X-ray techniques. Thermal expansion coefficients and percentage linear thermal expansion of the solid solutions were determined using high temperature X-ray diffraction technique in the temperature range 300 to 1700 K for the first time. The room temperature lattice constants estimated for above compositions are 0.5578, 0.5565, 0.5545 and 0.5526 nm, respectively. The mean linear thermal expansion coefficients for the solid solutions are 15.48 x 10 -6 K -1 , 18.35 x 10 -6 K -1 , 22.65 x 10 -6 K -1 and 26.95 x 10 -6 K -1 , respectively. The percentage linear thermal expansions in this temperature range are 1.68, 1.89, 2.21 and 2.51 respectively. It is seen that the solid solutions are stable up to 1700 K. It is also seen that the effect and nature of the dopant are the important parameters influencing the thermal expansion of the ThO 2 . The lattice parameter of the solid solutions exhibited a decreasing trend with respect to praseodymia addition. The percentage linear thermal expansion of the solid solutions increases steadily with increasing temperature

Mixing enhancement in a scramjet combustor using fuel jet injection swirl

Science.gov (United States)

Flesberg, Sonja M.

The scramjet engine has proven to be a viable means of powering a hypersonic vehicle, especially after successful flights of the X-51 WaveRider and various Hy-SHOT test vehicles. The major challenge associated with operating a scramjet engine is the short residence time of the fuel and oxidizer in the combustor. The fuel and oxidizer have only milliseconds to mix, ignite and combust in the combustion chamber. Combustion cannot occur until the fuel and oxidizer are mixed on a molecular level. Therefore the improvement of mixing is of utmost interest since this can increase combustion efficiency. This study investigated mixing enhancement of fuel and oxidizer within the combustion chamber of a scramjet by introducing swirl to the fuel jet. The investigation was accomplished with numerical simulations using STAR-CCM+ computational fluid dynamic software. The geometry of the University of Virginia Supersonic Combustion Facility was used to model the isolator, combustor and nozzle of a scramjet engine for simulation purposes. Experimental data from previous research at the facility was used to verify the simulation model before investigating the effect of fuel jet swirl on mixing. The model used coaxial fuel jet with a swirling annular jet. Single coaxial fuel jet and dual coaxial fuel jet configurations were simulated for the investigation. The coaxial fuel jets were modelled with a swirling annular jet and non-swirling core jet. Numerical analysis showed that fuel jet swirl not only increased mixing and entrainment of the fuel with the oxidizer but the mixing occurred further upstream than without fuel jet swirl. The burning efficiency was calculated for the all the configurations. An increase in burning efficiency indicated an increase in the mixing of H2 with O2. In the case of the single fuel jet models, the maximum burning efficiency increase due to fuel injection jet swirl was 23.3%. The research also investigated the possibility that interaction between two

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

Water experiments on thermal striping in reactor vessel of advanced sodium-cooled fast reactor. Influence of flow collector of backup CR guide tube

International Nuclear Information System (INIS)

Kobayashi, Jun; Ezure, Toshiki; Tanaka, Masaaki; Kamide, Hideki

2016-01-01

Design study of an advanced large-scale sodium-cooled fast reactor (SFR) has been conducted in JAEA. In the region between the bottom of the Upper Internal Structure (UIS) and the core outlet, the hot sodium from the fuel subassembly mixes with the cold sodium from the neighbor control rod (CR) channel. Therefore, temperature fluctuation due to mixing fluids at different temperatures may cause high cycle thermal fatigue at the bottom of the UIS. In the advanced design, installation of a flow guide structure named Flow-Collector (FC) to the backup control rod (BCR) guide tube is considered to enhance reliable operation of self-actuated shutdown system (SASS) and to ensure reactor shutdown operation. Previously, water experiments without the FC model had been examined in JAEA to investigate effective countermeasures to the significant temperature fluctuation generation at the bottom of the UIS. Since the FC may affect the thermal mixing behavior at the bottom of the UIS, influence of the FC on characteristics of the temperature fluctuation around the BCR channels was investigated using a water experimental facility with structure model of the FC. Through the experiment, small influence of the FC on the temperature fluctuation distribution at the bottom of the UIS was indicated. (author)

Establishment of cooperation basis of joint research on the mixed waste molten salt oxidation technology

International Nuclear Information System (INIS)

Yang, Hee Chul; Cho, Y. J.; Kim, J. H.; Yoo, J. H.; Yun, H. C.; Lee, D. G.

2005-08-01

Molten salt oxidation, MSO for short, is a robust technology that can effectively treat mixed waste (radioactive waste including hazardous metals or organics). It can safely and economically treat the difficult wastes such as not-easily destroyable toxic organic waste, medical waste, chemical warfare and energetic materials such as propellant and explosives, all of which are not easily treated by an incinerator or other currently existing thermal treatment system. Therefore, molten salt oxidation technology should be developed and utilized to treat a lot of niche waste stored in the nuclear and environmental industries. So, if we put the MSO technology to practical use by Korea-Vietnam joint research, we can reduce R and D fund for MSO technology by ourselves and we can expect an export of the outcome of nuclear R and D in Korea. For Establishment of cooperation basis of joint research concerning molten salt oxidation technology between KOREA and VIETNAM, in this research, We invited two Vietnamese researchers and we introduced our experimental scale molten salt oxidation system in order to let them understand molten salt oxidation technology. We also visited Viet man and we consulted about molten salt oxidation process. We held seminar on the mixed waste molten salt oxidation technology, discussed on the joint research on the mixed waste molten salt oxidation technology and finally we wrote MOU for joint research

Establishment of cooperation basis of joint research on the mixed waste molten salt oxidation technology

Energy Technology Data Exchange (ETDEWEB)

Yang, Hee Chul; Cho, Y. J.; Kim, J. H.; Yoo, J. H.; Yun, H. C.; Lee, D. G

2005-08-01

Molten salt oxidation, MSO for short, is a robust technology that can effectively treat mixed waste (radioactive waste including hazardous metals or organics). It can safely and economically treat the difficult wastes such as not-easily destroyable toxic organic waste, medical waste, chemical warfare and energetic materials such as propellant and explosives, all of which are not easily treated by an incinerator or other currently existing thermal treatment system. Therefore, molten salt oxidation technology should be developed and utilized to treat a lot of niche waste stored in the nuclear and environmental industries. So, if we put the MSO technology to practical use by Korea-Vietnam joint research, we can reduce R and D fund for MSO technology by ourselves and we can expect an export of the outcome of nuclear R and D in Korea. For Establishment of cooperation basis of joint research concerning molten salt oxidation technology between KOREA and VIETNAM, in this research, We invited two Vietnamese researchers and we introduced our experimental scale molten salt oxidation system in order to let them understand molten salt oxidation technology. We also visited Viet man and we consulted about molten salt oxidation process. We held seminar on the mixed waste molten salt oxidation technology, discussed on the joint research on the mixed waste molten salt oxidation technology and finally we wrote MOU for joint research.

Method of manufacturing mixed stock powders for nuclear fuel elements

International Nuclear Information System (INIS)

Hirayama, Satoshi.

1980-01-01

Purpose: To alleviate the limit of the present reactor operating conditions by uniformly mixing an additive to the main content as an uranium dioxide or mixture of the uranium dioxide with plutonium dioxide. Method: A mixed stock powder is obtained by adding an additive of at least two of aluminium oxide, beryllium oxide, calcium oxide, magnesium oxide, silicon oxide, sodium oxide, potassium oxide, phosphorus oxide, titanium oxide and iron oxide to suspension having ammonia water as dispersion medium to start the deposition of precipitation at a step of precipitating ammonium diuranate or plutionium hydroxide of a main content of uranium dioxide or mixture of uranium dioxide and plutonium dioxide and deposited precipitate is calcinated and reduced. (Yoshihara, H.)

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

Experimental and thermodynamic evaluation of the melting behavior of irradiated oxide fuels

International Nuclear Information System (INIS)

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

1985-01-01

Onset of melting is an important performance limit for irradiated UO 2 and UO 2 -based nuclear reactor fuels. Melting (solidus) temperatures are reasonably well known for starting fuel materials such as UO 2 and (U,PU)O 2 , however the influence of burnup on oxide fuel melting behavior continues to represent an area of considerable uncertainty. In this paper we report the results of a variety of melting temperature measurements on pseudo-binary fuel-fissia mixtures such as UO 2 -PUO 2 , UO 2 -CeO 2 , UO 2 -BaO, UO 2 -SrO, UO 2 -BaZrO 3 and UO 2 -SrZrO 3 . These measurements were performed using the thermal arrest technique on tungsten-encapsulated specimens. Several low melting eutectics, the existence of which had previously been inferred from post-irradiation examinations of high burnup mixed oxide fuels, were characterized in the course of the investigation. Also, an assessment of melting temperature changes in irradiated oxide fuels due to the production and incorporation of soluble oxidic fission products was performed by application of solution theory to the available pseudo-binary phase diagram data. The results of this assessment suggest that depression of oxide fuel solidus temperatures by dissolved fission products is substantially less than that indicated by earlier experimental studies. (orig.)

Off gas processing device for degreasing furnace for uranium/plutonium mixed oxide fuel

International Nuclear Information System (INIS)

Ueda, Masaya; Akasaka, Takayuki; Noura, Takeshi.

1996-01-01

A low melting ingredient capturing-cooling trap connected to a degreasing sintering furnace by way of sealed pipelines, a burning/decomposing device for decomposing high melting ingredient gases discharged from the cooling trap by burning them and a gas sucking means for forming the flow of off gases are contained in a glovebox, the inside pressure of which is kept negative. Since the degreasing sintering furnace for uranium/plutonium mixed oxide fuels is disposed outside of the glovebox, operation can be performed safely without greatly increasing the scale of the device, and the back flow of gases is prevented easily by keeping the pressure in the inside of the glovebox negative. Further, a heater is disposed at the midway of the sealed pipelines from the degreasing sintering furnace to the cooling trap, the temperature is kept high to prevent deposition of low melting ingredients to prevent clogging of the sealed pipelines. Further, a portion of the pipelines is made extensible in the axial direction to eliminate thermal stresses caused by temperature change thereby enabling to extend the life of the sealed pipelines. (N.H.)

Pathfinder irradiation of advanced fuel (Th/U mixed oxide) in a power reactor

International Nuclear Information System (INIS)

Brant Pinheiro, R.

1993-01-01

Within the joint Brazilian-German cooperative R and D Program on Thorium Utilization in Pressurized Water Reactors carried out from 1979 to 1988 by Nuclebras/CDTN, KFA-Juelich, Siemens/KWU and NUKEM, a pathfinder irradiation of Th/U mixed oxide fuel in the Angra 1 nuclear power reactor was planned. The objectives of this irradiation testing, the irradiation strategy, the work performed and the status achieved at the end of the joint Program are presented. (author)

Synthesis and Characterization of Oxide Feedstock Powders for the Fuel Cycle R and D Program

International Nuclear Information System (INIS)

Voit, Stewart L.; Vedder, Raymond James; Johnson, Jared A.

2010-01-01

Nuclear fuel feedstock properties, such as physical, chemical, and isotopic characteristics, have a significant impact on the fuel fabrication process and, by extension, the in-reactor fuel performance. This has been demonstrated through studies with UO 2 spanning greater than 50 years. The Fuel Cycle R and D Program with The Department of Energy Office of Nuclear Energy has initiated an effort to develop a better understanding of the relationships between oxide feedstock, fresh fuel properties, and in-reactor fuel performance for advanced mixed oxide compositions. Powder conditioning studies to enable the use of less than ideal powders for ceramic fuel pellet processing are ongoing at Los Alamos National Laboratory (LANL) and an understanding of methods to increase the green density and homogeneity of pressed pellets has been gained for certain powders. Furthermore, Oak Ridge National Laboratory (ORNL) is developing methods for the co-conversion of mixed oxides along with techniques to analyze the degree of mixing. Experience with the fabrication of fuel pellets using co-synthesized multi-constituent materials is limited. In instances where atomically mixed solid solutions of two or more species are needed, traditional ceramic processing methods have been employed. Solution-based processes may be considered viable synthesis options, including co-precipitation (AUPuC), direct precipitation, direct-conversion (Modified Direct Denitration or MDD) and internal/external gelation (sol-gel). Each of these techniques has various advantages and disadvantages. The Fiscal Year 2010 feedstock development work at ORNL focused on the synthesis and characterization of one batch of UO x and one batch of U 80 Ce 20 O x . Oxide material synthesized at ORNL is being shipped to LANL for fuel fabrication process development studies. The feedstock preparation was performed using the MDD process which utilizes a rotary kiln to continuously thermally denitrate double salts of ammonium

Thermal conductivity thermal diffusivity of UO{sub 2}-BeO nuclear fuel pellets

Energy Technology Data Exchange (ETDEWEB)

Mansur, Fábio A.; Camarano, Denise M.; Santos, Ana M. M.; Ferraz, Wilmar B.; Silva, Mayra A.; Ferreira, Ricardo A.N., E-mail: fam@cdtn.br, E-mail: dmc@cdtn.br, E-mail: amms@cdtn.br, E-mail: ferrazw@cdtn.br, E-mail: mayra.silva@cdtn.br, E-mail: ricardoanf@yahoo.com.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2017-07-01

The temperature distribution in nuclear fuel pellets is of vital importance for the performance of the reactor, as it affects the heat transfer, the mechanical behavior and the release of fission gas during irradiation, reducing safety margins in possible accident scenarios. One of the main limitation for the current uranium dioxide nuclear fuel (UO{sub 2}) is its low thermal conductivity, responsible for the higher temperature of the pellet center and, consequently, for a higher radial temperature gradient. Thus, the addition of another material to increase the UO{sub 2} fuel thermal conductivity has been considered. Among the additives that are being investigated, beryllium oxide (BeO) has been chosen due to its high thermal conductivity, with potential to optimize power generation in pressurized light water reactors (PWR). In this work, UO{sub 2}-BeO pellets were obtained by the physical mixing of the powders with additions of 2wt% and 3wt% of BeO. The thermal diffusivity and conductivity of the pellets were determined from room temperature up to 500 °C. The results were normalized to 95% of the theoretical density (TD) of the pellets and varied according to the BeO content. The range of the values of thermal diffusivity and conductivity were 1.22 mm{sup 2}∙s{sup -1} to 3.69 mm{sup 2}∙s{sup -1} and 3.80 W∙m{sup -}'1∙K{sup -1} to 9.36 W∙m{sup -1}∙K{sup -1}, respectively. (author)

Assessment of accident energetics in LMFBR core-disruptive accidents

International Nuclear Information System (INIS)

Fauske, H.K.

1977-01-01

An assessment of accident energetics in LMFBR core-disruptive accidents is given with emphasis on the generic issues of energetic recriticality and energetic fuel-coolant interaction events. Application of a few general behavior principles to the oxide-fueled system suggests that such events are highly unlikely following a postulated core meltdown event

Fuel removal from plasma-facing components by oxidation-based techniques. An overview of surface conditions after oxidation

International Nuclear Information System (INIS)

Rubel, M.J.; De Temmerman, G.; Sergienko, G.; Sundelin, P.; Emmoth, B.; Philipps, V.

2007-01-01

Oxygen-assisted fuel removal is reported for laboratory-prepared a-C:D films and for layers obtained by boronisation in a tokamak and then exposed to a helium-oxygen glow discharge in TEXTOR. Oxidation of thick mixed-material co-deposits under laboratory conditions is also presented. The essential results are following: (i) laboratory-prepared amorphous deuterated carbon (a-C:D) layers are decomposed efficiently by the He-O 2 glow: D and C contents are decreased by a factor of 45-220 and 25-60, respectively; (ii) the same treatment of the boronised films leads to the release of D but no removal of carbon is observed; (iii) the thermal oxidation (at 300 deg. C in air under laboratory conditions) of co-deposits on PFC and probes exposed to the SOL reduces the D content by a factor of 4-5 after 2 h, whereas nearly complete fuel removal (98%) occurs after 10 h at 300 deg. C. The study shows that the fuel removal efficiency is dependent on the overall composition of the mixed layer. It is high from pure a-C:D films but distinctly less efficient from real co-deposits

Strategies for Lowering Solid Oxide Fuel Cells Operating Temperature

Directory of Open Access Journals (Sweden)

Albert Tarancón

2009-11-01

Full Text Available Lowering the operating temperature of solid oxide fuel cells (SOFCs to the intermediate range (500–700 ºC has become one of the main SOFC research goals. High operating temperatures put numerous requirements on materials selection and on secondary units, limiting the commercial development of SOFCs. The present review first focuses on the main effects of reducing the operating temperature in terms of materials stability, thermo-mechanical mismatch, thermal management and efficiency. After a brief survey of the state-of-the-art materials for SOFCs, attention is focused on emerging oxide-ionic conductors with high conductivity in the intermediate range of temperatures with an introductory section on materials technology for reducing the electrolyte thickness. Finally, recent advances in cathode materials based on layered mixed ionic-electronic conductors are highlighted because the decreasing temperature converts the cathode into the major source of electrical losses for the whole SOFC system. It is concluded that the introduction of alternative materials that would enable solid oxide fuel cells to operate in the intermediate range of temperatures would have a major impact on the commercialization of fuel cell technology.

Prediction of the pressure-time history due to fuel-sodium interaction in a subassembly

International Nuclear Information System (INIS)

Jacobs, H.

1975-01-01

A local cooling disturbance may lead to complete voiding of a subassembly and melt down of the fuel pins. Thus molten fuel may be accumulated and mixed with liquid sodium returning accidentally into the subassembly. The resulting fuel-sodium interaction (FSI) produces a pressure load on the surrounding core structures. It is necessary to prove that the corresponding core deformation neither initiates a nuclear excursion nor renders the shut down system inoperable. This requires the knowledge of the initiating FSI pressure time history. In this paper a theoretical pressure time history is presented which differs completely from all calculations known so far. (Auth.)

Revision of the second basic plans of power reactor development in Power Reactor and Nuclear Fuel Development Corporation

International Nuclear Information System (INIS)

1978-01-01

Revision of the second basic plans concerning power reactor development in PNC (Power Reactor and Nuclear Fuel Development Corporation) is presented. (1) Fast breeder reactors: As for the experimental fast breeder reactor, after reaching the criticality, the power is raised to 50 MW thermal output within fiscal 1978. The prototype fast breeder reactor is intended for the electric output of 200 MW -- 300 MW, using mixed plutonium/uranium oxide fuel. Along the above lines, research and development will be carried out on reactor physics, sodium technology, machinery and parts, nuclear fuel, etc. (2) Advanced thermal reactor: The prototype advanced thermal reactor, with initial fuel primarily of slightly enriched uranium and heavy water moderation and boiling water cooling, of 165 MW electric output, is brought to its normal operation by the end of fiscal 1978. Along the above lines, research and development will be carried out on reactor physics, machinery and parts, nuclear fuel, etc. (Mori, K

A thermal hydraulic analysis in PWR reactors with UO2 or (U-Th)O2 fuel rods employing a simplified code

International Nuclear Information System (INIS)

Santos, Thiago A. dos; Maiorino, José R.; Stefanni, Giovanni L. de

2017-01-01

In order to project a nuclear reactor, the neutronic calculus must be validated, so that its thermal limits and safety parameters are respected. Considering this issue, this research aims to evaluate the APTh-100 reactor thermal limits. This PWR is a project developed in Universidade Federal do ABC (UFABC) using fuel composed of Uranium and Thorium oxide mixed (U,Th)O 2 . For this purpose, a simplified, although conservative, code was developed in a MATLAB environment named STC-MOX-Th 'Simplified Thermal-hydraulics Code-Mixed Oxide Thorium'. This code provides axial and radial temperature distribution, as well as DNBR distribution over the hottest channel of the reactor core. Moreover, it brings other hydraulic quantities, such as pressure drop over the fuel rod, considering any fuel proportion of (U,Th)O 2 .The software uses basic laws of conservation of mass, momentum and energy, it also calculates the thermal conduction equation, considering the thermal conductive coefficient as a temperature function. In order to solve this equation, the finite elements method was used. Furthermore, the proportion of 36% of UO 2 was used to evaluate the temperature over the fuel rod and DNBR minimum in three burn conditions: beginning, middle and ending. The program has proven to be efficient in every condition and the results evidenced that the APTh-1000 reactor, in an initial analysis, has its thermal limits within the recommended security parameters. (author)

The electrical and thermal properties of sodium sulfate mixed with lithium sulfate, yttrium sulfate, and silicon dioxide

International Nuclear Information System (INIS)

Imanaka, N.; Yamaguchi, Y.; Adachi, G.; Shiokawa, J.

1986-01-01

Sodium sulfate mixed with lithium sulfate, yttrium sulfate, and silicon dioxide was prepared. The thermal and electrical properties of its phases were investigated. The Na 2 SO 4 -Li 2 SO 4 -Y 2 (SO 4 ) 3 SiO 2 samples are similar to the Na 2 SO 4 -I phase (a high temperature phase), which is appreciably effective for Na + ionic conduction. Phase transformation was considerably suppressed by mixing. Electromotive force (EMF) was measured, using Na 2 SO 4 -Li 2 SO 4 -Y 2 (SO 4 ) 3 -SiO 2 as a solid electrolyte, by constructing an SO 2 gas concentration cell. The measured EMF's at 823 and 773 K were in fairly good accordance with the calculated EMF's for inlet SO 2 gas concentration between 30 ppm and 1%, and 500 ppm and 0.5% respectively

Development of uranium dioxide fuel pellets with addition of beryllium oxide for increasing of thermal conductivity

International Nuclear Information System (INIS)

Queiroz, Carolinne Mol; Ferreira, Ricardo Alberto Neto

2011-01-01

The CDTN - Centro de Desenvolvimento de Tecnologia Nuclear presents a project named 'Beryllium Project' viewing to increasing the thermal conductivity of UO 2 fuel pellets, increasing the lifetime of those pellets in the reactor, generating a greater economy. This increase of conductivity is obtained by means of Be O addition to the UO 2 fuel pellets, which is very used for the production of nuclear energy. The UO 2 pellets however present a thermal conductivity relatively low, generating a high temperature gradient between the center and his side surface. The addition of beryllium oxide, with higher thermal conductivity gives pellets which will present lower temperature gradient and, consequently, more durability and better utilization of energy potential of the pellet in the reactor. (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

Evaluation of thermal physical properties for fast reactor fuels. Melting point and thermal conductivities

International Nuclear Information System (INIS)

Kato, Masato; Morimoto, Kyoichi; Komeno, Akira; Nakamichi, Shinya; Kashimura, Motoaki; Abe, Tomoyuki; Uno, Hiroki; Ogasawara, Masahiro; Tamura, Tetsuya; Sugata, Hirotada; Sunaoshi, Takeo; Shibata, Kazuya

2006-10-01

Japan Atomic Energy Agency has developed a fast breeder reactor (FBR), and plutonium and uranium mixed oxide (MOX) having low density and 20-30%Pu content has used as a fuel of the FBR, Monju. In plutonium, Americium has been accumulated during long-term storage, and Am content will be increasing up to 2-3% in the MOX. It is essential to evaluate the influence of Am content on physical properties of MOX on the development of FBR in the future. In this study melting points and thermal conductivities which are important data on the fuel design were measured systematically in wide range of composition, and the effects of Am accumulated were evaluated. The solidus temperatures of MOX were measured as a function of Pu content, oxygen to metal ratio (O/M) and Am content using thermal arrest technique. The sample was sealed in a tungsten capsule in vacuum for measuring solidus temperature. In the measurements of MOX with Pu content of more than 30%, a rhenium inner capsule was used to prevent the reaction between MOX and tungsten. In the results, it was confirmed that the melting points of MOX decrease with as an increase of Pu content and increase slightly with a decrease of O/M ratio. The effect of Am content on the fuel design was negligible small in the range of Am content up to 3%. Thermal conductivities of MOX were evaluated from thermal diffusivity measured by laser flash method and heat capacity calculated by Neumann- Kopp's law. The thermal conductivity of MOX decreased slightly in the temperature of less than 1173K with increasing Am content. The effect of Am accumulated in long-term storage fuel was evaluated from melting points and thermal conductivities measured in this study. It is concluded that the increase of Am in the fuel barely affect the fuel design in the range of less than 3%Am content. (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

COMTA - a computer code for fuel mechanical and thermal analysis

International Nuclear Information System (INIS)

Basu, S.; Sawhney, S.S.; Anand, A.K.; Anantharaman, K.; Mehta, S.K.

1979-01-01

COMTA is a generalized computer code for integrity analysis of the free standing fuel cladding, with natural UO 2 or mixed oxide fuel pellets. Thermal and Mechanical analysis is done simultaneously for any power history of the fuel pin. For analysis, the fuel cladding is assumed to be axisymmetric and is subjected to axisymmetric load due to contact pressure, gas pressure, coolant pressure and thermal loads. Axial variation of load is neglected and creep and plasticity are assumed to occur at constant volume. The pellet is assumed to be made of concentric annuli. The fission gas release integral is dependent on the temperature and the power produced in each annulus. To calculate the temperature distribution in the fuel pin, the variation of bulk coolant temperature is given as an input to the code. Gap conductance is calculated at every time step, considering fuel densification, fuel relocation and gap closure, filler gas dilution by released fission gas, gap closure by expansion and irradiation swelling. Overall gap conductance is contributed by heat transfer due to the three modes; conduction convection and radiation as per modified Ross and Stoute model. Equilibrium equations, compatibility equations, stress strain relationships (including thermal strains and permanent strains due to creep and plasticity) are used to obtain triaxial stresses and strains. Thermal strain is assumed to be zero at hot zero power conditions. The boundary conditions are obtained for radial stresses at outside and inside surfaces by making these equal to coolant pressure and internal pressure respectively. A multi-mechanism creep model which accounts for thermal and irradiation creep is used to calculate the overall creep rate. Effective plastic strain is a function of effective stress and material constants. (orig.)

Thermal performance of a meso-scale liquid-fuel combustor

International Nuclear Information System (INIS)

Vijayan, V.; Gupta, A.K.

2011-01-01

Research highlights: → Demonstrated successful combustion of liquid fuel-air mixtures in a novel meso-scale combustor. → Flame quenching was eliminated using heat recirculation in a swiss roll type combustor that also extended the flammability limits. → Liquid fuel was rapidly vaporized with the use of hot narrow channel walls that eliminated the need of a fuel atomizer. → Maximum power density of the combustor was estimated to be about 8.5 GW/m3 and heat load in the range of 50-280W. → Overall efficiency of the combustor was estimated in the range of 12 to 20%. - Abstract: Combustion in small scale devices poses significant challenges due to the quenching of reactions from wall heat losses as well as the significantly reduced time available for mixing and combustion. In the case of liquid fuels there are additional challenges related to atomization, vaporization and mixing with the oxidant in the very short time-scale liquid-fuel combustor. The liquid fuel employed here is methanol with air as the oxidizer. The combustor was designed based on the heat recirculating concept wherein the incoming reactants are preheated by the combustion products through heat exchange occurring via combustor walls. The combustor was fabricated from Zirconium phosphate, a ceramic with very low thermal conductivity (0.8 W m -1 K -1 ). The combustor had rectangular shaped double spiral geometry with combustion chamber in the center of the spiral formed by inlet and exhaust channels. Methanol and air were introduced immediately upstream at inlet of the combustor. The preheated walls of the inlet channel also act as a pre-vaporizer for liquid fuel which vaporizes the liquid fuel and then mixes with air prior to the fuel-air mixture reaching the combustion chamber. Rapid pre-vaporization of the liquid fuel by the hot narrow channel walls eliminated the necessity for a fuel atomizer. Self-sustained combustion of methanol-air was achieved in a chamber volume as small as 32.6 mm 3

Status of steady-state irradiation testing of mixed-carbide fuel designs

International Nuclear Information System (INIS)

Harry, G.R.

1983-01-01

The steady-state irradiation program of mixed-carbide fuels has demonstrated clearly the ability of carbide fuel pins to attain peak burnup greater than 12 at.% and peak fluences of 1.4 x 10 23 n/cm 2 (E > 0.1 MeV). Helium-bonded fuel pins in 316SS cladding have achieved peak burnups of 20.7 at.% (192 MWd/kg), and no breaches have occurred in pins of this design. Sodium-bonded fuel pins in 316SS cladding have achieved peak burnups of 15.8 at.% (146 MWd/kg). Breaches have occurred in helium-bonded fuel pins in PE-16 cladding (approx. 5 at.% burnup) and in D21 cladding (approx. 4 at.% burnup). Sodium-bonded fuel pins achieved burnups over 11 at.% in PE-16 cladding and over 6 at.% in D9 and D21 cladding

Formulation and Testing of Paraffin-Based Solid Fuels Containing Energetic Additives for Hybrid Rockets

Science.gov (United States)

Larson, Daniel B.; Boyer, Eric; Wachs,Trevor; Kuo, Kenneth K.; Story, George

2012-01-01

Many approaches have been considered in an effort to improve the regression rate of solid fuels for hybrid rocket applications. One promising method is to use a fuel with a fast burning rate such as paraffin wax; however, additional performance increases to the fuel regression rate are necessary to make the fuel a viable candidate to replace current launch propulsion systems. The addition of energetic and/or nano-sized particles is one way to increase mass-burning rates of the solid fuels and increase the overall performance of the hybrid rocket motor.1,2 Several paraffin-based fuel grains with various energetic additives (e.g., lithium aluminum hydride (LiAlH4) have been cast in an attempt to improve regression rates. There are two major advantages to introducing LiAlH4 additive into the solid fuel matrix: 1) the increased characteristic velocity, 2) decreased dependency of Isp on oxidizer-to-fuel ratio. The testing and characterization of these solid-fuel grains have shown that continued work is necessary to eliminate unburned/unreacted fuel in downstream sections of the test apparatus.3 Changes to the fuel matrix include higher melting point wax and smaller energetic additive particles. The reduction in particle size through various methods can result in more homogeneous grain structure. The higher melting point wax can serve to reduce the melt-layer thickness, allowing the LiAlH4 particles to react closer to the burning surface, thus increasing the heat feedback rate and fuel regression rate. In addition to the formulation of LiAlH4 and paraffin wax solid-fuel grains, liquid additives of triethylaluminum and diisobutylaluminum hydride will be included in this study. Another promising fuel formulation consideration is to incorporate a small percentage of RDX as an additive to paraffin. A novel casting technique will be used by dissolving RDX in a solvent to crystallize the energetic additive. After dissolving the RDX in a solvent chosen for its compatibility

In-pile measurement of the thermal conductivity of irradiated metallic fuel

International Nuclear Information System (INIS)

Bauer, T.H.; Holland, J.W.

1995-01-01

Transient test data and posttest measurements from recent in-pile overpower transient experiments are used for an in situ determination of metallic fuel thermal conductivity. For test pins that undergo melting but remain intact, a technique is described that relates fuel thermal conductivity to peak pin power during the transient and a posttest measured melt radius. Conductivity estimates and their uncertainty are made for a database of four irradiated Integral Fast Reactor-type metal fuel pins of relatively low burnup (<3 at.%). In the assessment of results, averages and trends of measured fuel thermal conductivity are correlated to local burnup. Emphasis is placed on the changes of conductivity that take place with burnup-induced swelling and sodium logging. Measurements are used to validate simple empirically based analytical models that describe thermal conductivity of porous media and that are recommended for general thermal analyses of irradiated metallic fuel

Thermal hydraulic model validation for HOR mixed core fuel management

International Nuclear Information System (INIS)

Gibcus, H.P.M.; Vries, J.W. de; Leege, P.F.A. de

1997-01-01

A thermal-hydraulic core management model has been developed for the Hoger Onderwijsreactor (HOR), a 2 MW pool-type university research reactor. The model was adopted for safety analysis purposes in the framework of HEU/LEU core conversion studies. It is applied in the thermal-hydraulic computer code SHORT (Steady-state HOR Thermal-hydraulics) which is presently in use in designing core configurations and for in-core fuel management. An elaborate measurement program was performed for establishing the core hydraulic characteristics for a variety of conditions. The hydraulic data were obtained with a dummy fuel element with special equipment allowing a.o. direct measurement of the true core flow rate. Using these data the thermal-hydraulic model was validated experimentally. The model, experimental tests, and model validation are discussed. (author)

Study of thermal sensitivity and thermal explosion violence of energetic materials in the LLNL ODTX system

International Nuclear Information System (INIS)

Hsu, P C; Hust, G; Zhang, M X; Lorenz, T K; Reynolds, J G; Fried, L; Springer, H K; Maienschein, J L

2014-01-01

Incidents caused by fire and combat operations can heat energetic materials that may lead to thermal explosion and result in structural damage and casualty. Some explosives may thermally explode at fairly low temperatures (< 100 °C) and the violence from thermal explosion may cause significant damage. Thus it is important to understand the response of energetic materials to thermal insults. The One Dimensional Time to Explosion (ODTX) system at the Lawrence Livermore National Laboratory has been used for decades to measure times to explosion, threshold thermal explosion temperature, and determine kinetic parameters of energetic materials. Samples of different configurations (pressed part, powder, paste, and liquid) can be tested in the system. The ODTX testing can also provide useful data for assessing the thermal explosion violence of energetic materials. Recent ODTX experimental data are reported in the paper.

Chemical, mass spectrometric, and spectrochemical analysis of nuclear-grade mixed oxides [(U,Pu)O2

International Nuclear Information System (INIS)

Anon.

1981-01-01

Mixed oxide, a mixture of uranium and plutonium oxides, is used as a nuclear-reactor fuel in the form of pellets. The plutonium content may be up to 10 wt %, and the diluent uranium may be of any U-235 enrichment. In order to be suitable for use as a nuclear fuel, the material must meet certain criteria for combined uranium and plutonium content, effective fissile content, and impurity content. Analytical procedures used to determine if mixed oxides comply with specifications are: uranium by controlled-potential coulometry; plutonium by controlled-potential coulometry; plutonium by amperometric titration with iron (II); nitrogen by distillation spectrophotometry using Nessler reagent; carbon (total) by direct combustion-thermal-conductivity; total chlorine and fluorine by pyrohydrolysis; sulfur by distillation-spectrophotometry; moisture by the coulometric, electrolytic moisture analyzer; isotopic composition by mass spectrometry; rare earths by copper spark spectroscopy; trace impurities by carrier distillation spectroscopy; impurities by spark-source mass spectrography; total gas in reactor-grade mixed dioxide pellets; tungsten by dithiol-spectrophotometry; rare earth elements by spectroscopy; plutonium-238 isotopic abundance by alpha spectrometry; uranium and plutonium isotopic analysis by mass spectrometry; oxygen-to-metal atom ratio by gravimetry

Numerical investigation on turbulence mixing characteristics under thermal striping flows. Investigations on fluid temperature fluctuation phenomena in air and sodium

Energy Technology Data Exchange (ETDEWEB)

Murakami, Satoshi [Customer System Co. Ltd., Tokai, Ibaraki (Japan); Muramatsu, Toshiharu

1999-05-01

A three-dimensional thermal striping analysis was carried out using a direct numerical simulation code DINUS-3, for a coaxial jet configuration using air and sodium as a working fluid, within the framework of the EJCC thermo-hydraulic division. From the analysis, the following results have been obtained: (1) Calculated potential core length in air and sodium turbulence flows agreed with a theoretical value (5d - 7d ; d : diameter of jet nozzle) in the two-dimensional free jet theory. (2) Hydraulic characteristics in sodium flows as the potential core length can be estimated by the use of that of air flow characteristics. (3) Shorter thermally potential core length defined by spatial temperature distribution was evaluated in sodium flows, compared with that in air flows. This is due to the higher thermal conductivity of sodium. (4) Thermal characteristics in sodium flows as the thermally potential core length can not be evaluated, based on that air thermal characteristics. (author)

A new thermodynamic model of energetic molten fuel-coolant interactions

International Nuclear Information System (INIS)

Hall, A.N.

1987-01-01

A new thermodynamic model of energetic molten fuel-coolant interactions is presented, in which the response of fluid around the interaction zone is treated explicitly. By assuming that this fluid is compressed reversibly and adiabatically, a qualified lower limit to the efficiency of conversion of thermal energy to mechanical work is obtained. A detailed comparison of the model predictions with the results of the SUW series of experiments at AEE Winfrith is made. The predicted efficiencies are found to be in close agreement with those determined experimentally. Model predictions for a system of infinite volume are also presented. (author)

Fabrication and control of fuels made of mixed carbides (U, Pu)C

International Nuclear Information System (INIS)

Lorenzelli, R.; Delaroche, P.

1980-01-01

Fabrication of this type of advanced fuel is described. The fuel is prepared by reduction of oxides with carbon and natural sintering. Density, thermal stability and thermal conductibility are more particularly studied [fr

Effects of ashes in solid fuels on fuel particle charging during combustion in an air stream

Energy Technology Data Exchange (ETDEWEB)

Zakharov, A.G.; Fialkov, B.S.; Mel' nichuk, A.Yu.; Khvan, L.A.

1982-09-01

Black coal from the Karaganda basin is mixed with sodium chloride and graphite. Coal characteristics are given in a table (density, ashes, content of silica, aluminium oxides, iron oxides, calcium oxides, potassium oxides and magnesium oxides). Effects of ash fluctuations on electric potential of fuel particles during combustion are analyzed. Analyses show that with increasing ash content electric potential of fuel particles decreases and reaches the minimum when ash content ranges from 70 to 80 %. Particles with electric potential are generated during chemical processes between carbon and oxygen when coal is burned in an air stream. (5 refs.) (In Russian)

Uranium plutonium oxide fuels

International Nuclear Information System (INIS)

Cox, C.M.; Leggett, R.D.; Weber, E.T.

1981-01-01

Uranium plutonium oxide is the principal fuel material for liquid metal fast breeder reactors (LMFBR's) throughout the world. Development of this material has been a reasonably straightforward evolution from the UO 2 used routinely in the light water reactor (LWR's); but, because of the lower neutron capture cross sections and much lower coolant pressures in the sodium cooled LMFBR's, the fuel is operated to much higher discharge exposures than that of a LWR. A typical LMFBR fuel assembly is shown. Depending on the required power output and the configuration of the reactor, some 70 to 400 such fuel assemblies are clustered to form the core. There is a wide variation in cross section and length of the assemblies where the increasing size reflects a chronological increase in plant size and power output as well as considerations of decreasing the net fuel cycle cost. Design and performance characteristics are described

Beryllium Project: developing in CDTN of uranium dioxide fuel pellets with addition of beryllium oxide to increase the thermal conductivity

International Nuclear Information System (INIS)

Ferreira, Ricardo Alberto Neto; Camarano, Denise das Merces; Miranda, Odair; Grossi, Pablo Andrade; Andrade, Antonio Santos; Queiroz, Carolinne Mol; Gonzaga, Mariana de Carvalho Leal

2013-01-01

Although the nuclear fuel currently based on pellets of uranium dioxide be very safe and stable, the biggest problem is that this material is not a good conductor of heat. This results in an elevated temperature gradient between the center and its lateral surface, which leads to a premature degradation of the fuel, which restricts the performance of the reactor, being necessary to change the fuel before its full utilization. An increase of only 5 to 10 percent in its thermal conductivity, would be a significant increase. An increase of 50 percent would be a great improvement. A project entitled 'Beryllium Project' was developed in CDTN - Centro de Desenvolvimento da Tecnologia Nuclear, which aimed to develop fuel pellets made from a mixture of uranium dioxide microspheres and beryllium oxide powder to obtain a better heat conductor phase, filling the voids between the microspheres to increase the thermal conductivity of the pellet. Increases in the thermal conductivity in the range of 8.6% to 125%, depending on the level of addition employed in the range of 1% to 14% by weight of beryllium oxide, were obtained. This type of fuel promises to be safer than current fuels, improving the performance of the reactor, in addition to last longer, resulting in great savings. (author)

Thermal analysis of thermo-gravimetric measurements of spent nuclear fuel oxidation rates

International Nuclear Information System (INIS)

Cramer, E.R.

1997-01-01

A detailed thermal analysis was completed of the sample temperatures in the Thermo-Gravimetric Analysis (TGA) system used to measure irradiated N Reactor fuel oxidation rates. Sample temperatures during the oxidation process did not show the increase which was postulated as a result of the exothermic reactions. The analysis shows the axial conduction of heat in the sample holder effectively removes the added heat and only a very small, i.e., <10 C, increase in temperature is calculated. A room temperature evaporation test with water showed the sample thermocouple sensitivity to be more than adequate to account for a temperature change of approximately 5 C. Therefore, measured temperatures in the TGA are within approximately 10 C of the actual sample temperatures and no adjustments to reported data to account for the heat input from the oxidation process are necessary

Prompt Burst Energetics (PBE) experiment analyses using the SIMMER-II computer code

International Nuclear Information System (INIS)

Tomkins, J.L.; Hitchcock, J.T.; Young, M.F.

1979-01-01

Two of the Prompt Burst Energetics (PBE) in-pile experiments conducted at Sandia Laboratories PBE-5S and PBE-SG2, have been investigated with SIMMER-II. These two tests utilize fresh uranium oxide and fresh uranium carbide pins, respectively, in stagnant sodium. The purpose of the analysis is to investigate the applicability of SIMMER-II to this type of experiment. Qualitative agreement with measured data is seen for PBE-5S. PBE-SG2 results agree somewhat less well but demonstrate SIMMER-II's potential for describing fuel-coolant-interactions with further model development

The spectrographic analysis of plutonium oxide or mixed plutonium oxide/uranium oxide fuel pellets by the dried residue technique

International Nuclear Information System (INIS)

Jarbo, G.J.; Faught, P.; Hildebrandt, B.

1980-05-01

An emission spectrographic method for the quantitative determination of metallic impurities in plutonium oxide and mixed plutonium oxide/uranium oxide is described. The fuel is dissolved in nitric acid and the plutonium and/or uranium extracted with tributyl phosphate. A small aliquot of the aqueous residue is dried on a 'mini' pyrolitic graphite plate and excited by high voltage AC spark in an oxygen atmosphere. Spectra are recorded in a region which has been specially selected to record simultaneously lines of boron and cadmium in the 2nd order and all the other elements of interest in the 1st order. Indium is used as an internal standard. The excitation of very small quantities of the uraniumm/plutonium free residue by high voltage spark, together with three separate levels of containment reduce the hazards to personnel and the environment to a minimum with limited effect on sensitivity and accuracy of the results. (auth)

Experience with oxide fuel for advanced reactors

International Nuclear Information System (INIS)

Leggett, R.D.

1984-01-01

This paper focuses on the use and potential of oxide fuel systems for the LMFBR. The flawless performance of mixed oxide (UO 2 -PuO 2 ) fuel in FFTF to 100,000 MWd/MTM is reviewed and means for achieving 200,000 MWd/MTM are presented. This includes using non-swelling alloys for cladding and ducts to overcome the limitations caused by swelling of the current alloys. Examples are provided of the inherently safe characteristics of oxide fuel including a large negative Doppler coefficient, its dispersive nature under hypothetical accident scenarios, and the low energy molten fuel-coolant interaction. Developments in fuel fabrication and reprocessing that stress safety and reduced personnel exposure are presented. Lastly, the flexibility to design for maximum fuel supply (high breeding gain) or minimum fuel cost (long lifetime) is shown

Experience with oxide fuel for advanced reactors

International Nuclear Information System (INIS)

Leggett, R.D.

1984-04-01

This paper focuses on the use and potential of oxide fuel system for the LMFBR. The flawless performance of mixed oxide (UO 2 -PuO 2 ) fuel in FFTF to 100,000 MWd/MTM is reviewed and means for achieving 200,000 MWd/MTM are presented. This includes using non-swelling alloys for cladding and ducts to overcome the limitations caused by swelling of the current alloys. Exampled are provided of the inherently safe characteristics of oxide fuel including a large negative Doppler coefficient, its dispersive nature under hypothetical accident scenarios, and the low energy molten fuel-coolant interaction. Developments in fuel fabrication and reprocessing that stress safety and reduced personnel exposure are presented. Lastly, the flexibility to design for maximum fuel supply (high breeding gain) or minimum fuel cost (long lifetime) is shown

Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

Energy Technology Data Exchange (ETDEWEB)

Cowell, B.S.; Fisher, S.E.

1999-02-01

The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

International Nuclear Information System (INIS)

Cowell, B.S.; Fisher, S.E.

1999-01-01

The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option

General features of conceptual design for the pilot plant to manufacture fuel rods from mixed oxides

International Nuclear Information System (INIS)

Quesada, C.A.; Adelfang, P.; Esteban, A.; Aparicio, G.; Friedenthal, M.; Orlando, O.S.

1987-01-01

This paper conceptually describes: 1) the processes in the manufacturing lines; 2) the distribution of quality controls and glove boxes in manufacturing lines; 3) the Control and Radiological Safety Room; 4) the Dressing Room; 5) the requirements of the ventilation system. The plant will be located in the first floor of the Radiochemical Processes Laboratory building, occupying a surface of 600 m 2 . The necessary equipment for the following manufacturing lines will be provided: a) conversion from Pu(NO3)4 to PuO 2 (through Pu(III)oxalate); b) manufacture of homogeneous of mixed oxides of U and Pu; c) manufacture of (U,Pu)O 2 pellets; d) manufacture of fuel rods of mixed uranium and plutonium oxides. (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

Modeling Thermal Ignition of Energetic Materials

National Research Council Canada - National Science Library

Gerri, Norman J; Berning, Ellen

2004-01-01

This report documents an attempt to computationally simulate the mechanics and thermal regimes created when a threat perforates an armor envelope and comes in contact with stowed energetic material...

Laminar simulation of intersubchannel mixing in a triangular nuclear fuel bundle geometry

International Nuclear Information System (INIS)

Zaretsky, A.; Lightstone, M.F.; Tullis, S.

2015-01-01

Highlights: • Quasi-periodic flow was observed through rod-to-wall gaps. • Triangular subchannel flows were fundamentally irregular. • Cross-gap flow was influenced both by local and adjacent cross-gap intensity. • Phase-linking between gaps induced cross-plane peripheral circulation through rod–wall gaps. • Cross-gap flow structure was dependent on subchannel geometry. - Abstract: Predicting temperature distributions in fuel rod bundles is an important component of nuclear reactor safety analysis. Intersubchannel mixing acts to homogenize coolant temperatures thus reducing the likelihood of localized regions of high fuel temperature. Previous research has shown that intersubchannel mixing in nuclear fuel rod bundles is enhanced by a large-scale quasi-periodic energetic fluid motion, which transports fluid on the cross-plane between the narrow gaps connecting subchannels. This phenomenon has also been observed in laminar flows. Unsteady laminar flow simulations were performed in a simplified bundle of three rods with a pipe. Three similar geometries of varying gap width were examined, and a thermal trace was implemented on the first geometry. Thermal mixing was driven by the advection of energy between subchannels by the cross-plane flow. Flow through the rod-to-wall gaps in the wall subchannels alternated with a dominant frequency, particularly when rod-to-wall gaps were smaller than rod-to-rod gaps. Significant phase-linking between rod-to-wall gaps was also observed such that a peripheral circulation occurred through each gap simultaneously. Cross-plane flow through the rod-to-rod gaps in the triangular subchannel was irregular in each case. This was due to the fundamental irregularity of the triangular subchannel geometry. Vortices were continually broken up by cross-plane flow from other gaps due to the odd number of fluid pathways within the central subchannel. Cross-plane flow in subchannel geometries is highly interconnected between gaps. The

A thermal hydraulic analysis in PWR reactors with UO{sub 2} or (U-Th)O{sub 2} fuel rods employing a simplified code

Energy Technology Data Exchange (ETDEWEB)

Santos, Thiago A. dos; Maiorino, José R., E-mail: thiago.santos@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil); Stefanni, Giovanni L. de, E-mail: giovanni.stefanni@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

2017-07-01

In order to project a nuclear reactor, the neutronic calculus must be validated, so that its thermal limits and safety parameters are respected. Considering this issue, this research aims to evaluate the APTh-100 reactor thermal limits. This PWR is a project developed in Universidade Federal do ABC (UFABC) using fuel composed of Uranium and Thorium oxide mixed (U,Th)O{sub 2}. For this purpose, a simplified, although conservative, code was developed in a MATLAB environment named STC-MOX-Th 'Simplified Thermal-hydraulics Code-Mixed Oxide Thorium'. This code provides axial and radial temperature distribution, as well as DNBR distribution over the hottest channel of the reactor core. Moreover, it brings other hydraulic quantities, such as pressure drop over the fuel rod, considering any fuel proportion of (U,Th)O{sub 2}.The software uses basic laws of conservation of mass, momentum and energy, it also calculates the thermal conduction equation, considering the thermal conductive coefficient as a temperature function. In order to solve this equation, the finite elements method was used. Furthermore, the proportion of 36% of UO{sub 2} was used to evaluate the temperature over the fuel rod and DNBR minimum in three burn conditions: beginning, middle and ending. The program has proven to be efficient in every condition and the results evidenced that the APTh-1000 reactor, in an initial analysis, has its thermal limits within the recommended security parameters. (author)

Fuel cycle cost comparisons with oxide and silicide fuels

Energy Technology Data Exchange (ETDEWEB)

Matos, J E; Freese, K E [RERTR Program, Argonne National Laboratory (United States)

1983-09-01

This paper addresses fuel cycle cost comparisons for a generic 10 MW reactor with HEU aluminide fuel and with LEU oxide and silicide fuels in several fuel element geometries. The intention of this study is to provide a consistent assessment of various design options from a cost point of view. The status of the development and demonstration of the oxide and silicide fuels are presented in several papers in these proceedings. Routine utilization of these fuels with the uranium densities considered here requires that they are successfully demonstrated and licensed. Thermal-hydraulic safety margins, shutdown margins, mixed cores, and transient analyses are not addressed here, but analyses of these safety issues are in progress for a limited number of the most promising design options. Fuel cycle cost benefits could result if a number of reactors were to utilize fuel elements with the same number or different numbers of the same standard fuel plate. Data is presented to quantify these potential cost benefits. This analysis shows that there are a number of fuel element designs using LEU oxide or silicide fuels that have either the same or lower total fuel cycle costs than the HEU design. Use of these fuels with the uranium densities considered requires that they are successfully demonstrated and licensed. All safety criteria for the reactor with these fuel element designs need to be satisfied as well. With LEU oxide fuel, 31 g U/cm{sup 3} 1 and 0.76 mm--thick fuel meat, elements with 18-22 plates 320-391 g {sup 235}U) result in the same or lower total costs than with the HEU element 23 plates, 280 g {sup 235}U). Higher LEU loadings (more plates per element) are needed for larger excess reactivity requirements. However, there is little cost advantage to using more than 20 of these plates per element. Increasing the fuel meat thickness from 0.76 mm to 1.0 mm with 3.1 g U/cm{sup 3} in the design with 20 plates per element could result in significant cost reductions if the

Simulation of thermal stresses in anode-supported solid oxide fuel cell stacks. Part II: Loss of gas-tightness, electrical contact and thermal buckling

Science.gov (United States)

Nakajo, Arata; Wuillemin, Zacharie; Van herle, Jan; Favrat, Daniel

Structural stability issues in planar solid oxide fuel cells arise from the mismatch between the coefficients of thermal expansion of the components. The stress state at operating temperature is the superposition of several contributions, which differ depending on the component. First, the cells accumulate residual stresses due to the sintering phase during the manufacturing process. Further, the load applied during assembly of the stack to ensure electric contact and flatten the cells prevents a completely stress-free expansion of each component during the heat-up. Finally, thermal gradients cause additional stresses in operation. The temperature profile generated by a thermo-electrochemical model implemented in an equation-oriented process modelling tool (gPROMS) was imported into finite-element software (ABAQUS) to calculate the distribution of stress and contact pressure on all components of a standard solid oxide fuel cell repeat unit. The different layers of the cell in exception of the cathode, i.e. anode, electrolyte and compensating layer were considered in the analysis to account for the cell curvature. Both steady-state and dynamic simulations were performed, with an emphasis on the cycling of the electrical load. The study includes two different types of cell, operation under both thermal partial oxidation and internal steam-methane reforming and two different initial thicknesses of the air and fuel compressive sealing gaskets. The results generated by the models are presented in two papers: Part I focuses on cell cracking. In the present paper, Part II, the occurrences of loss of gas-tightness in the compressive gaskets and/or electrical contact in the gas diffusion layer were identified. In addition, the dependence on temperature of both coefficients of thermal expansion and Young's modulus of the metallic interconnect (MIC) were implemented in the finite-element model to compute the plastic deformation, while the possibilities of thermal buckling

Sensitivity analysis of characteristics of spent mixed oxide fuel

International Nuclear Information System (INIS)

Hagura, Naoto; Yoshida, Tadashi

2008-01-01

Prediction error was evaluated for decay heat and nuclide generation in spent mixed oxide (MOX) fuels on the basis of error files in JENDL-3.3. This computational analysis was performed using SWAT code system, ORIGEN2 code, and ERRORJ code. The results of nuclide generation error evaluation were compared with some discrepancies in the calculated values to experimental values (C/E ratio) which were already published and were obtained by analysis of post irradiated experiments (PIE) data. Though the discrepancies of some C/E values, especially those of americium and curium isotopes, ranged from a half to twice, the present error evaluation based on the error file of nuclide generation became 10% or less. We conclude that the discrepancy between calculation and the PIE data is almost factor 5 larger than that evaluated from the covariance data in JENDL-3.3. Therefore the practical error value of total decay heat should be 20% or more on 1 σ basis. (authors)

Sensitivity analysis of characteristics of spent mixed oxide fuel

Energy Technology Data Exchange (ETDEWEB)

Hagura, Naoto; Yoshida, Tadashi [Musashi Institute of Technology, 1-28-1 Tamazutsumi, Setagaya, Tokyo 158-8557 (Japan)

2008-07-01

Prediction error was evaluated for decay heat and nuclide generation in spent mixed oxide (MOX) fuels on the basis of error files in JENDL-3.3. This computational analysis was performed using SWAT code system, ORIGEN2 code, and ERRORJ code. The results of nuclide generation error evaluation were compared with some discrepancies in the calculated values to experimental values (C/E ratio) which were already published and were obtained by analysis of post irradiated experiments (PIE) data. Though the discrepancies of some C/E values, especially those of americium and curium isotopes, ranged from a half to twice, the present error evaluation based on the error file of nuclide generation became 10% or less. We conclude that the discrepancy between calculation and the PIE data is almost factor 5 larger than that evaluated from the covariance data in JENDL-3.3. Therefore the practical error value of total decay heat should be 20% or more on 1 sigma basis. (authors)

Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

International Nuclear Information System (INIS)

Ozdemir, Levent; Acar, Banu Bulut; Zabunoglu, Okan H.

2011-01-01

When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of 239 Pu and 241 Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can be recycled as mixed-oxide (MOX) fuel after being blended with a fertile U makeup in a MOX fabrication plant. The burnup that can be obtained from MOX fuel depends on: (1) isotopic composition of Pu, which is closely related to the discharge burnup of spent fuel from which Pu is recovered; (2) the type of fertile U makeup material used (depleted U, natural U, or recovered U); and (3) fraction of makeup material in the mix (blending ratio), which in turn determines the total fissile fraction of MOX. Using the Non-linear Reactivity Model and the code MONTEBURNS, a step-by-step procedure for computing the total fissile content of MOX is introduced. As was intended, the resulting expression is simple enough for quick/hand calculations of total fissile content of MOX required to reach a desired burnup for a given discharge burnup of spent fuel and for a specified fertile U makeup. In any case, due to non-fissile (parasitic) content of recovered Pu, a greater fissile fraction in MOX than that in fresh U is required to obtain the same burnup as can be obtained by the fresh U fuel.

An energetically consistent vertical mixing parameterization in CCSM4

DEFF Research Database (Denmark)

Nielsen, Søren Borg; Jochum, Markus; Eden, Carsten

2018-01-01

An energetically consistent stratification-dependent vertical mixing parameterization is implemented in the Community Climate System Model 4 and forced with energy conversion from the barotropic tides to internal waves. The structures of the resulting dissipation and diffusivity fields are compared......, however, depends greatly on the details of the vertical mixing parameterizations, where the new energetically consistent parameterization results in low thermocline diffusivities and a sharper and shallower thermocline. It is also investigated if the ocean state is more sensitive to a change in forcing...

Mechanisms Underpinning Degradation of Protective Oxides and Thermal Barrier Coatings in High Hydrogen Content (HHC) - Fueled Turbines

Energy Technology Data Exchange (ETDEWEB)

Mumm, Daniel

2013-08-31

The overarching goal of this research program has been to evaluate the potential impacts of coal-derived syngas and high-hydrogen content fuels on the degradation of turbine hot-section components through attack of protective oxides and thermal barrier coatings. The primary focus of this research program has been to explore mechanisms underpinning the observed degradation processes, and connections to the combustion environments and characteristic non-combustible constituents. Based on the mechanistic understanding of how these emerging fuel streams affect materials degradation, the ultimate goal of the program is to advance the goals of the Advanced Turbine Program by developing materials design protocols leading to turbine hot-section components with improved resistance to service lifetime degradation under advanced fuels exposures. This research program has been focused on studying how: (1) differing combustion environments – relative to traditional natural gas fired systems – affect both the growth rate of thermally grown oxide (TGO) layers and the stability of these oxides and of protective thermal barrier coatings (TBCs); and (2) how low levels of fuel impurities and characteristic non-combustibles interact with surface oxides, for instance through the development of molten deposits that lead to hot corrosion of protective TBC coatings. The overall program has been comprised of six inter-related themes, each comprising a research thrust over the program period, including: (i) evaluating the role of syngas and high hydrogen content (HHC) combustion environments in modifying component surface temperatures, heat transfer to the TBC coatings, and thermal gradients within these coatings; (ii) understanding the instability of TBC coatings in the syngas and high hydrogen environment with regards to decomposition, phase changes and sintering; (iii) characterizing ash deposition, molten phase development and infiltration, and associated corrosive

Alternative fuels for the French fast breeder reactors programme

International Nuclear Information System (INIS)

Bailly, H.; Bernard, H.; Mansard, B.

1989-01-01

French fast breeder reactors use mixed oxide as reference fuel. A great deal of experience has been gained in the behaviour and manufacture of oxide fuel, which has proved to be the most suitable fuel for future commercial breeder reactors. However, France is maintaining long-term alternative fuels programme, in order to be in a position to satisfy eventually new future reactor design and operational requirements. Initially, the CEA in France developed a carbide-based, sodium-bonded fuel designed for a high specific power. The new objective of the alternative fuels programme is to define a fuel which could replace the oxide without requiring any significant changes to the operating conditions, fuel cycle processes or facilities. The current program concentrates on a nitride-based, helium-bonded fuel, bearing in mind the carbide solution. The paper describes the main characteristics required, the manufacturing process as developed, the inspection methods, and the results obtained. Present indications are that the industrial manufacture of mixed nitride is feasible and that production costs for nitride and oxide fuels would be not significantly different. (author) 8 refs., 2 figs

HETEROGENEOUS REBURNING BY MIXED FUELS

Energy Technology Data Exchange (ETDEWEB)

Wei-Yin Chen; Benson B. Gathitu

2005-01-14

Recent studies of heterogeneous reburning, i.e., reburning involving a coal-derived char, have elucidated its variables, kinetics and mechanisms that are valuable to the development of a highly efficient reburning process. Young lignite chars contain catalysts that not only reduce NO, but they also reduce HCN that is an important intermediate that recycles to NO in the burnout zone. Gaseous CO scavenges the surface oxides that are formed during NO reduction, regenerating the active sites on the char surface. Based on this mechanistic information, cost-effective mixed fuels containing these multiple features has been designed and tested in a simulated reburning apparatus. Remarkably high reduction of NO and HCN has been observed and it is anticipated that mixed fuel will remove 85% of NO in a three-stage reburning process.

Survey of post-irradiation examinations made of mixed carbide fuels

International Nuclear Information System (INIS)

Coquerelle, M.

1997-01-01

Post-irradiation examinations on mixed carbide, nitride and carbonitride fuels irradiated in fast flux reactors Rapsodie and DFR were carried out during the seventies and early eighties. In this report, emphasis was put on the fission gas release, cladding carburization and head-end gaseous oxidation process of these fuels, in particular, of mixed carbides. (author). 8 refs, 16 figs, 3 tabs

Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

International Nuclear Information System (INIS)

Karahan, Aydın; Kazimi, Mujid S.

2013-01-01

The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors

Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

Energy Technology Data Exchange (ETDEWEB)

Karahan, Aydın, E-mail: karahan@alum.mit.edu; Kazimi, Mujid S.

2013-10-15

The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.

Mixing ratio sensor of alcohol mixed fuel

Energy Technology Data Exchange (ETDEWEB)

Miyata, Shigeru; Matsubara, Yoshihiro

1987-08-07

In order to improve combustion efficiency of an internal combustion engine using gasoline-alcohol mixed fuel and to reduce harmful substance in its exhaust gas, it is necessary to control strictly the air-fuel ratio to be supplied and the ignition timing and change the condition of control depending upon the mixing ratio of the mixed fuel. In order to detect the mixing ratio of the mixed fuel, the above mixing ratio has so far been detected by casting a ray of light to the mixed fuel and utilizing a change of critical angle associated with the change of the composition of the fluid of the mixed fuel. However, in case when a light emitting diode is used for the light source above, two kinds of sensors are further needed. Concerning the two kinds of sensors above, this invention offers a mixing ratio sensor for the alcohol mixed fuel which can abolish a temperature sensor to detect the environmental temperature by making a single compensatory light receiving element deal with the compensation of the amount of light emission of the light emitting element due to the temperature change and the compensation of the critical angle caused by the temperature change. (6 figs)

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

Sodium cooled fast breeder reactors (FBRs) may experience accidental leakage of hot liquid sodium in the inert equipment cells and reactor cavity. The leaked sodium at temperature ranging from 120degC to 550degC can come in contact with the sacrificial layer of limestone concrete. In order to study the thermal and chemical impact of sodium on the limestone concrete, five experimental runs were carried out under different test conditions simulating accident scenarios as realistically as possible. In each experimental run, a given mass of liquid sodium preheated to a specified temperature was dumped on the surface of concrete specimen housed in a test vessel with argon atmosphere. The sodium pool formed on the concrete was heated with an immersion heater to maintain the pool temperature at pre-selected level. The temperatures at various strategic locations were continuously monitored throughout the test run. Online measurement of pressure, hydrogen gas and oxygen gas in argon atmosphere was conducted. The solid samples of sodium debris were retrieved from the posttest concrete specimen by manual core drilling device for chemical analysis of reacted and un-reacted sodium. After cleaning the sodium debris, a power-drilling machine was employed to collect powder samples at regular depth interval from the concrete block floor to determine residual free and bound water. This paper presents some of the dominant thermal and chemical features related to structural safety of the concrete. Among the thermal parameters, on-set time and residence period for Energetic Thermal Transients (ETT) along with peak and average heat generation rates are evaluated. Chemical parameters such as rate and extent of water release from concrete, sodium consumption, sodium hydroxide production and sodium emission into argon atmosphere are also elucidated. Physicochemical characteristics of post-test sodium and concrete debris were investigated. Moreover spatial distribution of sodium, free and

Mixing ratio sensor for alcohol mixed fuel

Energy Technology Data Exchange (ETDEWEB)

Miyata, Shigeru; Matsubara, Yoshihiro

1987-08-24

In order to improve the combustion efficiency of an internal combustion engine using gasoline-alcohol mixed fuel and to reduce harmful substance in its exhaust gas, it is necessary to control strictly the air-fuel ratio to be supplied and the ignition timing. In order to detect the mixing ratio of the mixed fuel, a mixing ratio sensor has so far been proposed to detect the above mixing ratio by casting a ray of light to the mixed fuel and utilizing a change of critical angle associated with the change of the composition of the fluid of the mixed fuel. However, because of the arrangement of its transparent substance in the fuel passage with the sealing material in between, this sensor invited the leakage of the fluid due to deterioration of the sealing material, etc. and its cost became high because of too many parts to be assembled. In view of the above, in order to reduce the number of parts, to lower the cost of parts and the assembling cost and to secure no fluid leakage from the fuel passage, this invention formed the above fuel passage and the above transparent substance both concerning the above mixing ratio sensor in an integrated manner using light transmitting resin. (3 figs)

Case study application of the IAEA safeguards assessment methodology to a mixed oxide fuel fabrication facility

International Nuclear Information System (INIS)

Swartz, J.; McDaniel, T.

1981-01-01

Science Applications, Inc. has prepared a case study illustrating the application of an assessment methodology to an international system for safeguarding mixed oxide (MOX) fuel fabrication facilities. This study is the second in a series of case studies which support an effort by the International Atomic Energy Agency (IAEA) and an international Consultant Group to develop a methodology for assessing the effectiveness of IAEA safeguards. 3 refs

Performance of IN-706 and PE-16 cladding in mixed-oxide fuel pins

International Nuclear Information System (INIS)

Makenas, B.J.; Lawrence, L.A.; Jensen, B.W.

1982-05-01

Iron-nickel base, precipitation-strengthened alloys, IN-706 and PE-16, advanced alloy cladding considered for breeder reactor applications, were irradiated in mixed-oxide fuel pins in the HEDL-P-60 subassembly in EBR-II. Initial selection of candidate advanced alloys was done using only nonfueled materials test results. However, to establish the performance characteristics of the candidate cladding alloys, i.e., dimensional stability and structural integrity under conditions of high neutron flux, elevated temperature, and applied stress, it was necessary to irradiate fuel pins under typical operating conditions. Fuel pins were clad with solution treated IN-706 and PE-16 and irradiated to peak fluences of 6.1 x 10 22 n/cm 2 (E > .1 MeV) and 8.8 x 10 22 n/cm 2 (E > .1 MeV) respectively. Fabrication and operating parameters for the fuel pins with the advanced cladding alloy candidates are summarized. Irradiation of HEDL-P-60 was interrupted with the breach of a pin with IN-706 cladding at 5.1 at % and the test was terminated with cladding breach in a pin with PE-16 cladding at 7.6 at %

FFTF vertical sodium storage tank preliminary thermal analysis

International Nuclear Information System (INIS)

Irwin, J.J.

1995-01-01

In the FFTF Shutdown Program, sodium from the primary and secondary heat transport loops, Interim Decay Storage (IDS), and Fuel Storage Facility (FSF) will be transferred to four large storage tanks for temporary storage. Three of the storage tanks will be cylindrical vertical tanks having a diameter of 28 feet, height of 22 feet and fabricated from carbon steel. The fourth tank is a horizontal cylindrical tank but is not the subject of this report. The storage tanks will be located near the FFTF in the 400 Area and rest on a steel-lined concrete slab in an enclosed building. The purpose of this work is to document the thermal analyses that were performed to ensure that the vertical FFTF sodium storage tank design is feasible from a thermal standpoint. The key criterion for this analysis is the time to heat up the storage tank containing frozen sodium at ambient temperature to 400 F. Normal operating conditions include an ambient temperature range of 32 F to 120 F. A key parameter in the evaluation of the sodium storage tank is the type of insulation. The baseline case assumed six inches of calcium silicate insulation. An alternate case assumed refractory fiber (Cerablanket) insulation also with a thickness of six inches. Both cases assumed a total electrical trace heat load of 60 kW, with 24 kW evenly distributed on the bottom head and 36 kW evenly distributed on the tank side wall

Licensing issues associated with the use of mixed-oxide fuel in U.S. commercial nuclear reactors

International Nuclear Information System (INIS)

Williams, D.L. Jr.

1997-04-01

On January 14, 1997, the Department of Energy, as part of its Record of Decision on the storage and disposition of surplus nuclear weapons materials, committed to pursue the use of excess weapons-usable plutonium in the fabrication of mixed-oxide (MOX) fuel for consumption in existing commercial nuclear power plants. Domestic use of MOX fuel has been deferred since the late 1970s, principally due to nuclear proliferation concerns. This report documents a review of past and present literature (i.e., correspondence, reports, etc.) on the domestic use of MOX fuel and provides discussion on the technical and regulatory issues that must be addressed by DOE (and the utility/consortia selected by DOE to effect the MOX fuel consumption strategy) in obtaining approval from the Nuclear Regulatory Commission to use MOX fuel in one or a group of existing commercial nuclear power plants

An Innovative High Thermal Conductivity Fuel Design

Energy Technology Data Exchange (ETDEWEB)

Jamil A. Khan

2009-11-21

Thermal conductivity of the fuel in today's Light Water Reactors, Uranium dioxide, can be improved by incorporating a uniformly distributed heat conducting network of a higher conductivity material, Silicon Carbide. The higher thermal conductivity of SiC along with its other prominent reactor-grade properties makes it a potential material to address some of the related issues when used in UO2 [97% TD]. This ongoing research, in collaboration with the University of Florida, aims to investigate the feasibility and develop a formal methodology of producing the resultant composite oxide fuel. Calculations of effective thermal conductivity of the new fuel as a function of %SiC for certain percentages and as a function of temperature are presented as a preliminary approach. The effective thermal conductivities are obtained at different temperatures from 600K to 1600K. The corresponding polynomial equations for the temperature-dependent thermal conductivities are given based on the simulation results. Heat transfer mechanism in this fuel is explained using a finite volume approach and validated against existing empirical models. FLUENT 6.1.22 was used for thermal conductivity calculations and to estimate reduction in centerline temperatures achievable within such a fuel rod. Later, computer codes COMBINE-PC and VENTURE-PC were deployed to estimate the fuel enrichment required, to maintain the same burnup levels, corresponding to a volume percent addition of SiC.

An Innovative High Thermal Conductivity Fuel Design

International Nuclear Information System (INIS)

Khan, Jamil A.

2009-01-01

Thermal conductivity of the fuel in today's Light Water Reactors, Uranium dioxide, can be improved by incorporating a uniformly distributed heat conducting network of a higher conductivity material, Silicon Carbide. The higher thermal conductivity of SiC along with its other prominent reactor-grade properties makes it a potential material to address some of the related issues when used in UO2 (97% TD). This ongoing research, in collaboration with the University of Florida, aims to investigate the feasibility and develop a formal methodology of producing the resultant composite oxide fuel. Calculations of effective thermal conductivity of the new fuel as a function of %SiC for certain percentages and as a function of temperature are presented as a preliminary approach. The effective thermal conductivities are obtained at different temperatures from 600K to 1600K. The corresponding polynomial equations for the temperature-dependent thermal conductivities are given based on the simulation results. Heat transfer mechanism in this fuel is explained using a finite volume approach and validated against existing empirical models. FLUENT 6.1.22 was used for thermal conductivity calculations and to estimate reduction in centerline temperatures achievable within such a fuel rod. Later, computer codes COMBINE-PC and VENTURE-PC were deployed to estimate the fuel enrichment required, to maintain the same burnup levels, corresponding to a volume percent addition of SiC.

Study of reactions between fuel (mixed oxide (UPu)Osub(2-x)) and cladding (stainless-steel) in reactors: influence of iodine compounds

International Nuclear Information System (INIS)

Aubert, Michel.

1976-03-01

The influence of iodine compounds on the development of the oxide-cladding reaction was examined. The action of iodine, cesium and cesium iodide on type 316 stainless was determined in the presence or absence of uranium oxide or mixed uranium-plutonium oxide type fuel in a closed system, isothermal or with a temperature gradient. The study of the stainless steel iodine reactions was developed in particular. These experiments showed that cesium combines with uranium oxide to give cesium uranate Cs 2 U 2 O 7 ; it is not unreasonable to suppose that cesium urano-plutonate Cs 2 (U,Pu) 2 O 7 could be formed inside the pile. It was then shown that cesium iodide in the presence of sufficiently non-stoichiometric mixed oxide could contribute towards the degradation of the stainless steel cladding. Under these conditions the reaction is accompained by a transport of manganese, chromium and iron into the hot parts of the fuel by a Van-Arkel type mechanism. This might explain the presence of metallic precipitates in the fuel, but the role assigned to molybdenum iodide in the same phenomenon is considered unlikely. Finally it is proposed to deposit a thin layer of manganese metal on the inner surface of the cladding in order to minimize the action of fission products (CsI, Te) [fr

Development of materials for use in solid oxid fuel cells anodes using renewable fuels in direct operation

International Nuclear Information System (INIS)

Lima, D.B.P.L. de; Florio, D.Z. de; Bezerra, M.E.O.

2016-01-01

Fuel cells produce electrical current from the electrochemical combustion of a gas or liquid (H2, CH4, C2H5OH, CH3OH, etc.) inserted into the anode cell. An important class of fuel cells is the SOFC (Solid Oxide Cell Fuel). It has a ceramic electrolyte that transports protons (H +) or O-2 ions and operating at high temperatures (500-1000 °C) and mixed conductive electrodes (ionic and electronic) ceramics or cermets. This work aims to develop anodes for fuel cells of solid oxide (SOFC) in order to direct operations with renewable fuels and strategic for the country (such as bioethanol and biogas). In this context, it becomes important to study in relation to the ceramic materials, especially those that must be used in high temperatures. Some types of double perovskites such as Sr2MgMoO6 (or simply SMMO) have been used as anodes in SOFC. In this study were synthesized by the polymeric precursor method, analyzed and characterized different ceramic samples of families SMMO, doped with Nb, this is: Sr2 (MgMo)1-xNbxO6 with 0 ≤ x ≤ 0.2. The materials produced were characterized by various techniques such as, thermal analysis, X-ray diffraction and scanning electron microscopy, and electrical properties determined by dc and ac measurements in a wide range of temperature, frequency and partial pressure of oxygen. The results of this work will contribute to a better understanding of advanced ceramic properties with mixed driving (electronic and ionic) and contribute to the advancement of SOFC technology operating directly with renewable fuels. (author)

Infrared and x-ray photoelectron spectroscopic studies on sodium borosilicate glass interacted with thermally oxidized aluminides formed on alloy 690

International Nuclear Information System (INIS)

Yusufali, C.; Dutta, R.S.; Dey, G.K.; Kshirsagar, R.J.; Jagannath; Mishra, R.K.

2012-01-01

Thermally oxidized aluminides formed on Ni-Cr-Fe based superalloy 690 substrates were subjected to interaction with sodium borosilicate melt (used as matrices for immobilization of high-level radioactive liquid waste) at 1248 K for 192 hours. After the interaction, Fourier-transform infrared (FT-IR) spectroscopy analysis of glass samples indicated the incorporation of Al in the glass network. X-ray photoelectron spectroscopy (XPS) of glass specimens revealed modified glass structure. (author)

Energetics of edge oxidization of graphene nanoribbons

Science.gov (United States)

Yasuma, Airi; Yamanaka, Ayaka; Okada, Susumu

2018-06-01

On the basis of the density functional theory, we studied the geometries and energetics of O atoms adsorbed on graphene edges for simulating the initial stage of the edge oxidization of graphene. Our calculations showed that oxygen atoms are preferentially adsorbed onto the graphene edges with the zigzag portion, resulting in a large adsorption energy of about 5 eV. On the other hand, the edges with armchair shape are rarely oxidized, or the oxidization causes substantial structural reconstructions, because of the stable covalent bond at the armchair edge with the triple bond nature. Furthermore, the energetics sensitively depends on the edge angles owing to the inhomogeneity of the charge density at the edge atomic sites.

Management of spent oxide fuel from thermal reactors: the environmental and radiological effects of alternative approaches

International Nuclear Information System (INIS)

Martin, A.; Fry, T.M.; Edmunds, J.

1983-07-01

The report presents the results of a study of the environmental and radiological effects of alternative approaches to the management of spent oxide fuel from the UK nuclear power programme. The two main alternatives are extended storage of the fuel, pending decisions on fissile material recycle, and reprocessing of the fuel commencing in the early 1990s. A review is made of experience in spent fuel storage, of the technical issues in long-term storage and of the possible needs of the UK nuclear programme. The main environmental and radiological effects of reprocessing considered in the study are the generation of radioactive wastes, the release of radioactivity in effluents with the resulting radiation exposure of the public, and occupational radiation exposure. The influence of the delay between reprocessing and fabrication of mixed oxide fuel on the environmental and radiological effects of plutonium recycle is investigated and it is shown that the effects would be minimised if reprocessing were delayed until there was a requirement for recycle. Strategies for the management of heavy elements other than plutonium are discussed and results are presented illustrating aspects of actinide management. (author)

Pilot-scale equipment development for lithium-based reduction of spent oxide fuel

International Nuclear Information System (INIS)

Herrmann, S. D.

1998-01-01

An integral function of the electrometallurgical conditioning of DOE spent nuclear fuel is the standardization of waste forms. Argonne National Laboratory (ANL) has developed and is presently demonstrating the electrometallurgical conditioning of sodium-bonded metal fuel from Experimental Breeder Reactor II, resulting in uranium, ceramic waste, and metal waste forms. Engineering studies are underway at ANL in support of pilot-scale equipment development, which would precondition irradiated oxide fuel and likewise demonstrate the application of electrometallurgical conditioning to such non-metallic fuels. This paper highlights the integration of proposed spent oxide fuel conditioning with existing electrometallurgical processes. Additionally, technical bases for engineering activities to support a scale up of an oxide reduction process are described

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)

Thermo-hydraulic-mechanical analysis of the SS-050 sodium loop during a thermal shock of 2000C/s

International Nuclear Information System (INIS)

Jesus Miranda, C.A. de; Gebrin, A.N.

1988-01-01

An analytical thermo-hydraulic model was developed to obtain the temperature of the sodium flowing between the mixing tank TM of constant volume and the drain tank of the SS-050 sodium test facility. The piping connecting these two tanks is considered in the analysis. The sodium enters in the TM through a tube with lateral holes immersed in the TM's sodium. The model and relative computer program were tested and a typical situation was studied: a thermal shock with -200 0 C/s of thermal gradient in the test section. The sodium temperature time-histories along the piping length are presented. For the thermal shock situation, the temperature field in the TM bottom and outlet nozzle was calculated and the stresses were evaluated. The final thermal stresses will allow a detailed verification of the circuit design. (author) [pt

CP ESFR: Collaborative Project for a European Sodium Fast Reactor Core studies

International Nuclear Information System (INIS)

Buiron, L.; Vasile, A.; Sunderland, R.

2013-01-01

• Significant progress has been made in optimizing both the oxide and carbide ESFR cores; • For the oxide core the optimisation process concentrated on the reduction of the sodium void reactivity effect and on the evaluation of MA burning performances. The CONF2 axial configuration has provided a significant overall reduction of the sodium void reactivity effect. • The carbide core had a significantly higher reactivity loss over the fuel cycle compared to the oxide one. By increasing slightly the fuel pin diameter, whilst still retaining the advantages of lower fuel temperatures of carbide fuel, and making changes in the core layout, the reactivity loss over the cycle has been reduced to a level similar to that of the oxide core. By adopting the CONF2 axial configuration initially developed for the oxide core, the sodium void reactivity of the carbide core has also been reduced appreciably. • The MA transmutation performances of the optimized ESFR oxide core have been investigated with respect to two boundary configurations. The HET2 configuration shows a low MA transmutation rate sufficient to burn the MA produced by the ESFR core without affecting the safety parameters. The HOM4 configuration (where 4%wt. MA are loaded homogeneously in each core SA) is the most challenging configuration due to its impact on safety coefficients but it shows an high MA burning rate suitable for burning also MA accumulated by a thermal reactor fleet

Comparison of open cycles of uranium and mixed oxides of thorium-uranium using advanced reactors

International Nuclear Information System (INIS)

Gonçalves, Letícia C.; Maiorino, José R.

2017-01-01

A comparative study of the mass balance and production costs of uranium oxide fuels was carried out for an AP1000 reactor and thorium-uranium mixed oxide in a reactor proposal using thorium called AP-Th1000. Assuming the input mass values for a fuel load the average enrichment for both reactors as well as their feed mass was determined. With these parameters, the costs were calculated in each fuel preparation process, assuming the prices provided by the World Nuclear Association. The total fuel costs for the two reactors were quantitatively compared with 18-month open cycle. Considering enrichment of 20% for the open cycle of mixed U-Th oxide fuel, the total uranium consumption of this option was 50% higher and the cost due to the enrichment was 70% higher. The results show that the use of U-Th mixed oxide fuels can be advantageous considering sustainability issues. In this case other parameters and conditions should be investigated, especially those related to fuel recycling, spent fuel storage and reduction of the amount of transuranic radioactive waste

Thermal-hydraulic calculations for KUHFR with reduced enrichment uranium fuel

International Nuclear Information System (INIS)

Mishima, Kaichiro; Shibata, Toshikazu.

1982-01-01

This report provides the preliminary results of the thermal-hydraulic calculations to study the safety aspects in fueling the KUHFR with reduced enrichment uranium. The calculations were based on what was outlined in the Safety Analysis Report for the KUHFR and the guidebook for research reactor core conversion, IAEA-TECDOC-233, published by the International Atomic Energy Agency. No significant differences in the thermal-hydraulic operating conditions have been found between HEU and MEU fuels. However, in LEU cases, the combination of three factors - larger power peaking with LEU fuel, smaller thermal conductivity of U 3 O 8 -Al fuel with high uranium densities, and thicker fuel meat - resulted in higher maximum fuel and surface temperatures with the LEU oxide fuel. (author)

Evaluation of the characteristics of uranium and plutonium Mixed Oxide (MOX) fuel

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

MOX fuel irradiation test up to high burnup has been performed for five years. Irradiation test of MOX fuel having high plutonium content has also been performed from JFY 2007 and it still continues. A lot of irradiation data have been obtained through these tests. The activities done in JFY 2012 are mainly focused on Post Irradiation Examination (PIE) data analysis concerning thermal property change and fission gas release. In the former work thermal conductivity degradation due to burnup is examined and in the latter work the dependence of fission gas release mechanism on fuel pellet microstructure is examined. This report mainly covers the result of analysis. It is found that thermal conductivity degradation of MOX fuel due to burnup is less than that of UO{sub 2} fuel and that fission gas release mechanism of high enriched fissile zone (so called Pu spot) is much different from that of low enriched fissile zone (so called Matrix). (author)

Dissolution behavior of irradiated mixed oxide fuel with short stroke shearing for fast reactor reprocessing

International Nuclear Information System (INIS)

Ikeuchi, Hirotomo; Sano, Yuichi; Shibata, Atsuhiro; Koizumi, Tsutomu; Washiya, Tadahiro

2013-01-01

An efficient dissolution process was established for future reprocessing in which mixed-oxide (MOX) fuels with high plutonium contents and dissolver solution with high heavy-metal (HM) concentrations (more than 500 g dm -3 ) will be treated. This dissolution process involves short stroke shearing of fuels (∼10 mm in length). The dissolution kinetics of irradiated MOX fuels and the effects of the Pu content, HM concentration, and fuel form on the dissolution rate were investigated. Irradiated fuel was found to dissolve as 10 2 -10 3 times fast as non-irradiated fuel, but the rate decreased with increasing Pu content. Kinetic analysis based on the fragmentation model, which considers the penetration and diffusion of nitric acid through fuel matrices prior to chemical reaction, indicated that the dissolution rate of irradiated fuel was affected not only by the volume ratio of liquid to solid (L/S ratio) but also by the exposed surface area per unit mole of nitric acid (A/m ratio). The penetration rate of nitric acid is expected to be decreased at high HM concentrations by a reduction in the L/S ratio, but enhanced by shearing the fuel pieces with short strokes and thus enlarging the A/m ratio. (author)

Green hypergolic combination: Diethylenetriamine-based fuel and hydrogen peroxide

Science.gov (United States)

Kang, Hongjae; Kwon, Sejin

2017-08-01

The present research dealt with the concept of green hypergolic combination to replace the toxic hypergolic combinations. Hydrogen peroxide was selected as a green oxidizer. A novel recipe for the non-toxic hypergolic fuel (Stock 3) was suggested. Sodium borohydride was blended into the mixture of energetic hydrocarbon solvents as an ignition source for hypergolic ignition. The main ingredient of the mixture was diethylenetriamine. By mixing some amount of tetrahydrofuran with diethylenetriamine, the mixture became more flammable and volatile. The mixture of Stock 3 fuel remained stable for four months in the lab scale storability test. Through a simple drop test, the hypergolicity of the green hypergolic combination was verified. Comparing to the toxic hypergolic combination MMH/NTO as the reference, the theoretical performance of the green hypergolic combination would be achieved about 96.7% of the equilibrium specific impulse and about 105.7% of the density specific impulse. The applicability of the green hypergolic combination was successfully confirmed through the static hot-fire tests using 500 N scale hypergolic thruster.

Iron oxide/aluminum/graphene energetic nanocomposites synthesized by atomic layer deposition: Enhanced energy release and reduced electrostatic ignition hazard

Science.gov (United States)

Yan, Ning; Qin, Lijun; Hao, Haixia; Hui, Longfei; Zhao, Fengqi; Feng, Hao

2017-06-01

Nanocomposites consisting of iron oxide (Fe2O3) and nano-sized aluminum (Al), possessing outstanding exothermic redox reaction characteristics, are highly promising nanothermite materials. However, the reactant diffusion inhibited in the solid state system makes the fast and complete energy release very challenging. In this work, Al nanoparticles anchored on graphene oxide (GO/Al) was initially prepared by a solution assembly approach. Fe2O3 was deposited on GO/Al substrates by atomic layer deposition (ALD). Simultaneously thermal reduction of GO occurs, resulting in rGO/Al@Fe2O3 energetic composites. Differential scanning calorimetry (DSC) analysis reveals that rGO/Al@Fe2O3 composite containing 4.8 wt% of rGO exhibits a 50% increase of the energy release compared to the Al@Fe2O3 nanothermite synthesized by ALD, and an increase of about 130% compared to a random mixture of rGO/Al/Fe2O3 nanoparticles. The enhanced energy release of rGO/Al@Fe2O3 is attributed to the improved spatial distribution as well as the increased interfacial intimacy between the oxidizer and the fuel. Moreover, the rGO/Al@Fe2O3 composite with an rGO content of 9.6 wt% exhibits significantly reduced electrostatic discharge sensitivity. These findings may inspire potential pathways for engineering energetic nanocomposites with enhanced energy release and improved safety characteristics.

Minor Actinide Transmutation Physics for Low Conversion Ratio Sodium Fast Reactors

International Nuclear Information System (INIS)

Mehdi Asgari; Samuel E. Bays; Benoit Forget; Rodolfo Ferrer

2007-01-01

The effects of varying the reprocessing strategy used in the closed cycle of a Sodium Fast Reactor (SNF) prototype are presented in this paper. The isotopic vector from the aqueous separation of transuranic (TRU) elements in Light Water Reactor (LWR) spent nuclear fuel (SNF) is assumed to also vary according to the reprocessing strategy of the closed fuel cycle. The decay heat, gamma energy, and neutron emission of the fuel discharge at equilibrium are found to vary depending on the separation strategy. The SFR core used in this study corresponds to a burner configuration with a conversion ratio of ∼0.5 based on the Super-PRISM design. The reprocessing strategies stemming from the choice of either metal or oxide fuel for the SFR are found to have a large impact on the equilibrium discharge decay heat, gamma energy, and neutron emission. Specifically, metal fuel SFR with pyroprocessing of the discharge produces the largest amount of TRU consumption (166 kg per Effective Full Power Year or EFPY), but also the highest decay heat, gamma energy, and neutron emission. On the other hand, an oxide fuel SFR with PUREX reprocessing minimizes the decay heat and related parameters of interest to a minimum, even when compared to thermal Mixed Oxide (MOX) or Inert Matrix Fuel (IMF) on a per mass basis. On an assembly basis, however, the metal SFR discharge has a lower decay heat than an equivalent oxide SFR assembly for similar minor actinide consumptions (∼160 kg/EFPY.) Another disadvantage in the oxide PUREX reprocessing scenario is that there is no consumption of americium and curium, since PUREX reprocessing separates these minor actinides (MA) and requires them to be disposed of externally

Pilot-scale equipment development for pyrochemical treatment of spent oxide fuel

International Nuclear Information System (INIS)

Herrmann, S. D.

1999-01-01

Fundamental objectives regarding spent nuclear fuel treatment technologies include, first, the effective distribution of spent fuel constituents among product and stable waste forms and, second, the minimization and standardization of waste form types and volumes. Argonne National Laboratory (ANL) has developed and is presently demonstrating the electrometallurgical treatment of sodium-bonded metal fuel from Experimental Breeder Reactor II, resulting in an uranium product and two stable waste forms, i.e. ceramic and metallic. Engineering efforts are underway at ANL to develop pilot-scale equipment which would precondition irradiated oxide fuel via pyrochemical processing and subsequently allow for electrometallurgical treatment of such non-metallic fuels into standard product and waste forms. This paper highlights the integration of proposed spent oxide fuel treatment with existing electrometallurgical processes. System designs and technical bases for development of pilot-scale oxide reduction equipment are also described

GAPCON-THERMAL-2: a computer program for calculating the thermal behavior of an oxide fuel rod

International Nuclear Information System (INIS)

Beyer, C.E.; Hann, C.R.; Lanning, D.D.; Panisko, F.E.; Parchen, L.J.

1975-11-01

A description is presented of the computer code GAPCON THERMAL-2, a light water reactor (LWR) fuel thermal performance prediction code. GAPCON-THERMAL-2, is intended to be used as a calculational tool for reactor fuel steady-state thermal performance and to provide input for accident analyses. Some models used in the code provide best estimate as well as conservative predictions. Each of the individual models in the code is based on the best available data

Safety aspects related to burnup increase and mixed oxide fuel

International Nuclear Information System (INIS)

Thomas, W.

1992-01-01

The dominant factor presently limiting the fuel burnup is the response of the cladding hulls. To maintain the excellent record of very low fuel failure rates for increased burnups further technical development is underway and necessary. In the nuclear fuel cycle increased burnups lead to a remarkable reduction of spent fuel arisings and corresponding economic savings. Thermal recycling of plutonium presently provides an opportunity to reduce the rising accumulation of plutunium in a situation where there is no demand for this fissile material in Fast Breeder Reactors. (orig.) [de

Sustained Recycle in Light Water and Sodium-Cooled Reactors

International Nuclear Information System (INIS)

Piet, Steven J.; Bays, Samuel E.; Pope, Michael A.; Youinou, Gilles J.

2010-01-01

From a physics standpoint, it is feasible to sustain recycle of used fuel in either thermal or fast reactors. This paper examines multi-recycle potential performance by considering three recycling approaches and calculating several fuel cycle parameters, including heat, gamma, and neutron emission of fresh fuel; radiotoxicity of waste; and uranium utilization. The first recycle approach is homogeneous mixed oxide (MOX) fuel assemblies in a light water reactor (LWR). The transuranic portion of the MOX was varied among Pu, NpPu, NpPuAm, or all-TRU. (All-TRU means all isotopes through Cf-252.) The Pu case was allowed to go to 10% Pu in fresh fuel, but when the minor actinides were included, the transuranic enrichment was kept below 8% to satisfy the expected void reactivity constraint. The uranium portion of the MOX was enriched uranium. That enrichment was increased (to as much as 6.5%) to keep the fuel critical for a typical LWR irradiation. The second approach uses heterogeneous inert matrix fuel (IMF) assemblies in an LWR - a mix of IMF and traditional UOX pins. The uranium-free IMF fuel pins were Pu, NpPu, NpPuAm, or all-TRU. The UOX pins were limited to 4.95% U-235 enrichment. The number of IMF pins was set so that the amount of TRU in discharged fuel from recycle N (from both IMF and UOX pins) was made into the new IMF pins for recycle N+1. Up to 60 of the 264 pins in a fuel assembly were IMF. The assembly-average TRU content was 1-6%. The third approach uses fast reactor oxide fuel in a sodium-cooled fast reactor with transuranic conversion ratio of 0.50 and 1.00. The transuranic conversion ratio is the production of transuranics divided by destruction of transuranics. The FR at CR=0.50 is similar to the CR for the MOX case. The fast reactor cases had a transuranic content of 33-38%, higher than IMF or MOX.

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)

Conversion of highly enriched uranium in thorium-232 based oxide fuel for light water reactors: MOX-T fuel

Energy Technology Data Exchange (ETDEWEB)

Vapirev, E I; Jordanov, T; Christoskov, I [Sofia Univ. (Bulgaria). Fizicheski Fakultet

1994-12-31

The idea of conversion of highly enriched uranium (HEU) from warheads without mixing it with natural uranium as well as the utilization of plutonium as fuel component is discussed. A nuclear fuel which is a mixture of 4% {sup 235}U (HEU) as a fissile isotope and 96 % {sup 232}Th (ThO{sub 2}) as a non-fissile isotope in a mixed oxide with thorium fuel is proposed. It is assumed that plutonium can also be used in the proposed fuel in a mixture with {sup 235}U. The following advantages of the use of HEU in LWRs in mixed {sup 235}U - Th fuel are pointed out: (1) No generation of long-living plutonium and americium isotopes (in case of reprocessing the high level radioactive wastes will contain only fission fragments and uranium); (2) The high conversion ratio of Th extends the expected burnup by approximately 1/3 without higher initial enrichment (the same initial enrichment simplifies the problem for compensation of the excess reactivity in the beginning with burnable poison and boric acid); (3) The high conversion ratio of Th allows the fuel utilization with less initial enrichment (by approx. 1/3) for the same burnup; thus less excess reactivity has to be compensated after reloading; in case of fuel reprocessing all fissile materials ({sup 235}U + {sup 233}U) could be chemically extracted. Irrespectively to the optimistic expectations outlined, further work including data on optimal loading and reloading schemes, theoretical calculations of thermal properties of {sup 235}U + Th fuel rods, manufacturing of several test fuel assemblies and investigations of their operational behaviour in a reactor core is still needed. 1 fig., 7 refs.

Technology, safety and costs of decommissioning a reference small mixed oxide fuel fabrication plant. Volume 1. Main report

Energy Technology Data Exchange (ETDEWEB)

Jenkins, C. E.; Murphy, E. S.; Schneider, K J

1979-01-01

Detailed technology, safety and cost information are presented for the conceptual decommissioning of a reference small mixed oxide fuel fabrication plant. Alternate methods of decommissioning are described including immediate dismantlement, safe storage for a period of time followed by dismantlement and entombment. Safety analyses, both occupational and public, and cost evaluations were conducted for each mode.

Present status of uranium-plutonium mixed carbide fuel development for LMFBRs

International Nuclear Information System (INIS)

Handa, Muneo; Suzuki, Yasufumi

1984-01-01

The feature of carbide fuel is that it has the doubling time as short as about 13 years, that is, close to one half as compared with oxide fuel. The development of the carbide fuel in the past 10 years has been started in amazement. Especially in the program of new fuel development in USA started in 1974, He and Na bond fuel attained the burnup of 16 a/o without causing the breaking of cladding tubes. In 1984, the irradiation of the assembly composed of 91 fuel pins in the FFTF is expected. On the other hand in Japan, the fuel research laboratory was constructed in 1974 in the Oarai Laboratory, Japan Atomic Energy Research Institute, to carry out the studies on carbide fuel. In the autumn of 1982, two carbide fuel pins with different chemical composition have been successfully made. Accordingly, the recent status of the development is explained. The uranium-plutonium mixed carbide fuel is suitable to liquid metal-cooled fast breeder reactors because of large heat conductivity and the high density of nuclear fission substances. The thermal and nuclear characteristics of carbide fuel, the features of the reactor core using carbide fuel, the chemical and mechanical interaction of fuel and cladding tubes, the selection of bond materials, the manufacturing techniques for the fuel, the development of the analysis code for fuel behavior, and the research and development of carbide fuel in Japan are described. (Kako, I.)

Iron oxide/aluminum/graphene energetic nanocomposites synthesized by atomic layer deposition: Enhanced energy release and reduced electrostatic ignition hazard

Energy Technology Data Exchange (ETDEWEB)

Yan, Ning; Qin, Lijun [Laboratory of Material Surface Engineering and Nanofabrication, Xi’an Modern Chemistry Research Institute, Shaanxi (China); Science and Technology on Combustion and Explosion Laboratory, Xi’an Modern Chemistry Research Institute, Shaanxi (China); Hao, Haixia [Science and Technology on Combustion and Explosion Laboratory, Xi’an Modern Chemistry Research Institute, Shaanxi (China); Hui, Longfei [Laboratory of Material Surface Engineering and Nanofabrication, Xi’an Modern Chemistry Research Institute, Shaanxi (China); Science and Technology on Combustion and Explosion Laboratory, Xi’an Modern Chemistry Research Institute, Shaanxi (China); Zhao, Fengqi [Science and Technology on Combustion and Explosion Laboratory, Xi’an Modern Chemistry Research Institute, Shaanxi (China); Feng, Hao, E-mail: fenghao98@hotmail.com [Laboratory of Material Surface Engineering and Nanofabrication, Xi’an Modern Chemistry Research Institute, Shaanxi (China); State Key Laboratory of Fluorine and Nitrogen Chemicals, Xi’an Modern Chemistry Research Institute, Shaanxi (China)

2017-06-30

Highlights: • Energetic rGO/Al@Fe{sub 2}O{sub 3}nanocompositeswerefabricatedbyatomiclayerdepositionapproach. • A novel Al@Fe{sub 2}O{sub 3} unit featuring core-shell structure was decorated on the graphene nanosheet. • RGO/Al@Fe{sub 2}O{sub 3} nanocomposite exhibits superior energy release and reduced electrostatic ignition hazard. - Abstract: Nanocomposites consisting of iron oxide (Fe{sub 2}O{sub 3}) and nano-sized aluminum (Al), possessing outstanding exothermic redox reaction characteristics, are highly promising nanothermite materials. However, the reactant diffusion inhibited in the solid state system makes the fast and complete energy release very challenging. In this work, Al nanoparticles anchored on graphene oxide (GO/Al) was initially prepared by a solution assembly approach. Fe{sub 2}O{sub 3} was deposited on GO/Al substrates by atomic layer deposition (ALD). Simultaneously thermal reduction of GO occurs, resulting in rGO/Al@Fe{sub 2}O{sub 3} energetic composites. Differential scanning calorimetry (DSC) analysis reveals that rGO/Al@Fe{sub 2}O{sub 3} composite containing 4.8 wt% of rGO exhibits a 50% increase of the energy release compared to the Al@Fe{sub 2}O{sub 3} nanothermite synthesized by ALD, and an increase of about 130% compared to a random mixture of rGO/Al/Fe{sub 2}O{sub 3} nanoparticles. The enhanced energy release of rGO/Al@Fe{sub 2}O{sub 3} is attributed to the improved spatial distribution as well as the increased interfacial intimacy between the oxidizer and the fuel. Moreover, the rGO/Al@Fe{sub 2}O{sub 3} composite with an rGO content of 9.6 wt% exhibits significantly reduced electrostatic discharge sensitivity. These findings may inspire potential pathways for engineering energetic nanocomposites with enhanced energy release and improved safety characteristics.

Iron oxide/aluminum/graphene energetic nanocomposites synthesized by atomic layer deposition: Enhanced energy release and reduced electrostatic ignition hazard

International Nuclear Information System (INIS)

Yan, Ning; Qin, Lijun; Hao, Haixia; Hui, Longfei; Zhao, Fengqi; Feng, Hao

2017-01-01

Highlights: • Energetic rGO/Al@Fe 2 O 3 nanocompositeswerefabricatedbyatomiclayerdepositionapproach. • A novel Al@Fe 2 O 3 unit featuring core-shell structure was decorated on the graphene nanosheet. • RGO/Al@Fe 2 O 3 nanocomposite exhibits superior energy release and reduced electrostatic ignition hazard. - Abstract: Nanocomposites consisting of iron oxide (Fe 2 O 3 ) and nano-sized aluminum (Al), possessing outstanding exothermic redox reaction characteristics, are highly promising nanothermite materials. However, the reactant diffusion inhibited in the solid state system makes the fast and complete energy release very challenging. In this work, Al nanoparticles anchored on graphene oxide (GO/Al) was initially prepared by a solution assembly approach. Fe 2 O 3 was deposited on GO/Al substrates by atomic layer deposition (ALD). Simultaneously thermal reduction of GO occurs, resulting in rGO/Al@Fe 2 O 3 energetic composites. Differential scanning calorimetry (DSC) analysis reveals that rGO/Al@Fe 2 O 3 composite containing 4.8 wt% of rGO exhibits a 50% increase of the energy release compared to the Al@Fe 2 O 3 nanothermite synthesized by ALD, and an increase of about 130% compared to a random mixture of rGO/Al/Fe 2 O 3 nanoparticles. The enhanced energy release of rGO/Al@Fe 2 O 3 is attributed to the improved spatial distribution as well as the increased interfacial intimacy between the oxidizer and the fuel. Moreover, the rGO/Al@Fe 2 O 3 composite with an rGO content of 9.6 wt% exhibits significantly reduced electrostatic discharge sensitivity. These findings may inspire potential pathways for engineering energetic nanocomposites with enhanced energy release and improved safety characteristics.

Fabrication of 0.5-inch diameter FBR mixed oxide fuel pellets

International Nuclear Information System (INIS)

Rasmussen, D.E.; Benecke, M.W.; McCord, R.B.

1979-01-01

Large diameter (0.535 inch) mixed oxide fuel pellets for Fast Breeder Reactor application were successfully fabricated by the cold-press-and-sinter technique. Enriched UO 2 , PuO 2 -UO 2 , and PuO 2 -ThO 2 compositions were fabricated into nominally 90% theoretical density pellets for the UO 2 and PuO 2 -UO 2 compositions, and 88% and 93% T.D. for the PuO 2 -ThO 2 compositions. Some processing adjustments were required to achieve satisfactory pellet quality and density. Furnace heating rate was reduced from 200 to 50 0 C/h for the organic binder burnout cycle for the large, 0.535-inch diameter pellets to eliminate pellet cracking during sintering. Additional preslugging steps and die wall lubrication during pressing were used to eliminate pressing cracks in the PuO 2 -ThO 2 pellets

Hot vacuum outgassing to ensure low hydrogen content in MOX fuel pellets for thermal reactors

International Nuclear Information System (INIS)

Majumdar, S.; Nair, M.R.; Kumar, Arun

1983-01-01

Hot vacuum outgassing treatment to ensure low hydrogen content in Mixed Oxide Fuel (MOX) pellets for thermal reactors has been described. Hypostoichiometric sintered MOX pellets retain more hydrogen than UO 2 pellets. The hydrogen content further increases with the addition of admixed lubricant and pore formers. However, low hydrogen content in the MOX pellets can be ensured by a hot vacuum outgassing treatment at a temperature between 773K to 823K for 2 hrs. (author)

Mixed Solvent Reactive Recrystallization of Sodium Carbonate

NARCIS (Netherlands)

Gaertner, R.S.

2005-01-01

Investigation of the reactive recrystallization of trona (sodium sesquicarbonate) and sodium bicarbonate to sodium carbonate (soda) in a mixed solvent led to the design of several alternative, less energy consumptive, economically very attractive process routes for the production of soda from all

Alternative oxidation technologies for organic mixed waste

International Nuclear Information System (INIS)

Borduin, L.C.; Fewell, T.

1998-01-01

The Mixed Waste Focus Area (MWFA) is currently supporting the development and demonstration of several alternative oxidation technology (AOT) processes for treatment of combustible mixed low-level wastes. AOTs have been defined as technologies that destroy organic material without using open-flame reactions. AOTs include both thermal and nonthermal processes that oxidize organic wastes but operate under significantly different physical and chemical conditions than incinerators. Nonthermal processes currently being studied include Delphi DETOX and acid digestion at the Savannah River Site (SRS), and direct chemical oxidation at Lawrence Livermore National Laboratory (LLNL). All three technologies are at advanced stages of development or are entering the demonstration phase. Nonflame thermal processes include catalytic chemical oxidation, which is being developed and deployed at Lawrence Berkeley National Laboratory (LBNL), and steam reforming, a commercial process being supported by the Department of Energy (DOE). Although testing is complete on some AOT technologies, most require additional support to complete some or all of the identified development objectives. Brief descriptions, status, and planned paths forward for each of the technologies are presented

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

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

Molten carbonate fuel cell cathode with mixed oxide coating

Science.gov (United States)

Hilmi, Abdelkader; Yuh, Chao-Yi

2013-05-07

A molten carbonate fuel cell cathode having a cathode body and a coating of a mixed oxygen ion conductor materials. The mixed oxygen ion conductor materials are formed from ceria or doped ceria, such as gadolinium doped ceria or yttrium doped ceria. The coating is deposited on the cathode body using a sol-gel process, which utilizes as precursors organometallic compounds, organic and inorganic salts, hydroxides or alkoxides and which uses as the solvent water, organic solvent or a mixture of same.

International safeguards for a modern MOX [mixed-oxide] fuel fabrication facility

International Nuclear Information System (INIS)

Pillay, K.K.S.; Stirpe, D.; Picard, R.R.

1987-03-01

Bulk-handling facilities that process plutonium for commercial fuel cycles offer considerable challenges to nuclear materials safeguards. Modern fuel fabrication facilities that handle mixed oxides of plutonium and uranium (MOX) often have large inventories of special nuclear materials in their process lines and in storage areas for feed and product materials. In addition, the remote automated processing prevalent at new MOX facilities, which is necessary to minimize radiation exposures to personnel, tends to limit access for measurements and inspections. The facility design considered in this study incorporates all these features as well as state-of-the-art measurement technologies for materials accounting. Key elements of International Atomic Energy Agency (IAEA) safeguards for such a fuel-cycle facility have been identified in this report, and several issues of primary importance to materials accountancy and IAEA verifications have been examined. We have calculated detection sensitivities for abrupt and protracted diversions of plutonium assuming a single materials balance area for all processing areas. To help achieve optimal use of limited IAEA inspection resources, we have calculated sampling plans for attributes/variables verification. In addition, we have demonstrated the usefulness of calculating σ/sub (MUF-D)/ and detection probabilities corresponding to specified material-loss scenarios and resource allocations. The data developed and the analyses performed during this study can assist both the facility operator and the IAEA in formulating necessary safeguards approaches and verification procedures to implement international safeguards for special nuclear materials

International safeguards for a modern MOX (mixed-oxide) fuel fabrication facility

Energy Technology Data Exchange (ETDEWEB)

Pillay, K.K.S.; Stirpe, D.; Picard, R.R.

1987-03-01

Bulk-handling facilities that process plutonium for commercial fuel cycles offer considerable challenges to nuclear materials safeguards. Modern fuel fabrication facilities that handle mixed oxides of plutonium and uranium (MOX) often have large inventories of special nuclear materials in their process lines and in storage areas for feed and product materials. In addition, the remote automated processing prevalent at new MOX facilities, which is necessary to minimize radiation exposures to personnel, tends to limit access for measurements and inspections. The facility design considered in this study incorporates all these features as well as state-of-the-art measurement technologies for materials accounting. Key elements of International Atomic Energy Agency (IAEA) safeguards for such a fuel-cycle facility have been identified in this report, and several issues of primary importance to materials accountancy and IAEA verifications have been examined. We have calculated detection sensitivities for abrupt and protracted diversions of plutonium assuming a single materials balance area for all processing areas. To help achieve optimal use of limited IAEA inspection resources, we have calculated sampling plans for attributes/variables verification. In addition, we have demonstrated the usefulness of calculating sigma/sub (MUF-D)/ and detection probabilities corresponding to specified material-loss scenarios and resource allocations. The data developed and the analyses performed during this study can assist both the facility operator and the IAEA in formulating necessary safeguards approaches and verification procedures to implement international safeguards for special nuclear materials.

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

Sol-gel route to synthesis of microporous ceramic membranes: Thermal stability of TiO2-ZrO2 mixed oxides

International Nuclear Information System (INIS)

Qunyin Xu; Anderson, M.A.

1993-01-01

In this paper concerning the synthesis of microporous ceramic membranes, the authors focus on the preparation and thermal stability of unsupported microporous TiO 2 -ZrO 2 mixed-oxide membranes. It has been observed that, by adding up to 20% ZrO 2 into TiO 2 or up to 10% TiO 2 into ZrO 2 , these microporous membranes display improved thermal stability. They can be fired up to 500 C for 0.5 h without closing micropores. However, membranes containing almost equal percentages of each component have lost microporous features and have low surface areas and low porosities. A phase diagram of a two-component TiO 2 -ZrO 2 mixed-oxide membrane has been prepared based on DTA and X-ray diffraction data in order to better understand the microstructure changes upon firing

Review of direct electrical heating experiments on irradiated mixed-oxide fuel

International Nuclear Information System (INIS)

Fenske, G.R.; Bandyopadhyay, G.

1982-01-01

Results of approximately 50 out-of-reactor experiments that simulated various stages of a loss-of-flow event with irradiated fuel are presented. The tests, which utilized the direct electrical heating technique to simulate nuclear heating, were performed either on fuel segments with their original cladding intact or on fuel segments that were extruded into quartz tubes. The test results demonstrated that the macro- and microscopic fuel behavior was dependent on a number of variables including fuel heating rate, thermal history prior to a transient, the number of heating cycles, type of cladding (quartz vs stainless steel), and fuel burnup

Detailed description of an SSAC at the facility level for mixed oxide fuel fabrication facilities

International Nuclear Information System (INIS)

Jones, R.J.

1985-09-01

The purpose of this document is to provide a detailed description of a system for the accounting for and control of nuclear material in a mixed oxide fuel fabrication facility which can be used by a facility operator to establish his own system to comply with a national system for nuclear material accounting and control and to facilitate application of IAEA safeguards. The scope of this document is limited to descriptions of the following SSAC elements: (1) Nuclear Material Measurements; (2) Measurement Quality; (3) Records and Reports; (4) Physical Inventory Taking; (5) Material Balance Closing

Enhanced photocatalytic activity of titania-silica mixed oxide prepared via basic hydrolyzation

International Nuclear Information System (INIS)

Xie Chao; Xu Zili; Yang Qiujing; Xue Baoyong; Du Yaoguo; Zhang Jiahua

2004-01-01

Two different synthesis routes were applied to prepare TiO 2 -XSiO 2 (X denotes mol% of silica in titania-silica mixed oxides) with different silica concentrations by using ammonia water as hydrolysis catalyst. Through comparing the photocatalytic performance of two sets of mixed oxides, we found that the photocatalytic activity of mixed oxides prepared via the route which can promote homogeneity, was significantly enhanced as compared with that of counterparts prepared via the another route, and the highest photocatalytic activity obtained by adding about 9.1 mol% silica into titania was much higher than that of pure TiO 2 . The mixed oxides were investigated by means of XRD, thermal analysis, UV-vis, FT-IR and XPS. The characterization results suggest that, in comparison with pure TiO 2 , the mixed oxides exhibit smaller crystallite size and higher thermal stability which can elevate the temperature of anatase to rutile phase transformation due to the addition of silica. Furthermore, Broensted acidity, which is associated with the formation of Ti-O-Si hetero linkages where tetrahedrally coordinated silica is chemically mixed with the octahedral titania matrix, may be a very important contribution to the enhanced photocatalytic activity of titania-silica mixed oxides as well

Spent oxide fuel regeneration by crystallization in molybdate melts

International Nuclear Information System (INIS)

Ustinov, O.A.; Sukhanov, L.P.; Yakunin, S.A.

2006-01-01

Paper describes a procedure to regenerate spent oxide fuel by its crystallization in molybdate melts. Paper presents the process procedures to regenerate spent fuel of both fast and thermal neutron reactors. One analyzes the advantages of the elaborated procedure [ru

Dynamic thermal baffle on lower head of FBR sodium-sodium intermediate heat exchanger

International Nuclear Information System (INIS)

Charbonnel, A.; Foussat, C.

1981-01-01

The cover head of the heat exchanger is bathed on the one side by the primary sodium of the 'cold' header of the vessel and on the other side by the secondary sodium which feeds the heat exchange tube bank through the lower tubesheet. In the case of transient or permanent operating conditions at partial ratings, there are large temperature differences between the inner sodium (inlet temperature conditions of secondary sodium) and the outer sodium (mean temperature conditions in the primary sodium outlet port), hence the necessity of designing a thermal baffle which protects the head and its connection to the tubesheet. A 'static' thermal baffle consisting of a thick steel plate enclosing static sodium around the head proves inadequate during transient operating conditions. This is why a 'dynamic' thermal baffle is used whose design is based on the fact that the primary sodium in the lower part of the outlet port is always at a temperature close to that of the secondary sodium in the inlet header and the head. The primary sodium is taken from the bottom of the outlet port by a ring deflector and circulates in an annulus created by a double housing and the head. It flows out through openings in the lower part of the housing. (orig./GL)

Formation and destruction of nitrogen oxides at elevated pressures with mixed fuels; Typenoksidimuodostus ja tuhoaminen paineistetuissa olosuhteissa ja ongelmapolttoaineilla

Energy Technology Data Exchange (ETDEWEB)

Aho, M.; Haemaelaeinen, J.; Paakkinen, K.; Rantanen, J. [VTT Energy, Jyvaeskylae (Finland)

1997-10-01

Destruction of NO with NH{sub 3} (Thermal DeNO{sub x}) and formation of nitrogen oxides (especially NO{sub 2}) from fuel nitrogen were studied at elevated pressure (up to 15 bar) with a pressurized entrained flow reactor (PEFR) at conditions simulating freeboard area of pressurized fluidized bed boiler. Effect of HCl on the oxidation of CO was studied at atmospheric pressure. These results give information about emission formation during combustion of chlorine-containing wastes. N{sub x}O{sub y} formation from fuel mixtures will be studied with a new fluidized bed reactor (FBR) in 1997. Thermal DeNox-experiments were performed at p= 2-15 bar, T= 700-950 deg C. Concentrations of NO, N{sub 2}O, NO{sub 2} and NH{sub 3} were measured at different residence times (0.2-2s). After the experiments with the bare NH{sub 3}/NO mixture, the effects of two additional gases (N{sub 2}O and CO) were measured. A new reaction tube made of quartz was employed to prevent catalytic destruction of NH{sub 3}. Formation of NO{sub 2} was studied with eleven solid fuels. The first experiments were carried out at 12 bar, O{sub 2}=20-19 %, PO{sub 2}=2.4 bar and the additional ones at p=8 bar, O{sub 2}=4-5 % => PO{sub 2}=0.4 bar at 800- 900 deg C. PCA analysis was used for finding dependency between fuel properties and the convention of fuel-N to NO{sub 2}

Protocol Fuel Mix reporting

International Nuclear Information System (INIS)

2002-07-01

The protocol in this document describes a method for an Electricity Distribution Company (EDC) to account for the fuel mix of electricity that it delivers to its customers, based on the best available information. Own production, purchase and sale of electricity, and certificates trading are taken into account. In chapter 2 the actual protocol is outlined. In the appendixes additional (supporting) information is given: (A) Dutch Standard Fuel Mix, 2000; (B) Calculation of the Dutch Standard fuel mix; (C) Procedures to estimate and benchmark the fuel mix; (D) Quality management; (E) External verification; (F) Recommendation for further development of the protocol; (G) Reporting examples

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

SIEX: a correlated code for the prediction of liquid metal fast breeder reactor (LMFBR) fuel thermal performance

International Nuclear Information System (INIS)

Dutt, D.S.; Baker, R.B.

1975-06-01

The SIEX computer program is a steady state heat transfer code developed to provide thermal performance calculations for a mixed-oxide fuel element in a fast neutron environment. Fuel restructuring, fuel-cladding heat conduction and fission gas release are modeled to provide assessment of the temperature. Modeling emphasis has been placed on correlations to measurable quantities from EBR-II irradiation tests and the inclusion of these correlations in a physically based computational scheme. SIEX is completely modular in construction allowing the user options for material properties and correlated models. Required code input is limited to geometric and environmental parameters, with a ''consistent'' set of material properties and correlated models provided by the code. 24 references. (U.S.)

Thermal degradation of ligno-cellulosic fuels. DSC and TGA studies

Energy Technology Data Exchange (ETDEWEB)

Leroy, V.; Cancellieri, D.; Leoni, E. [SPE-CNRS UMR 6134, University of Corsica, Campus Grossetti, BP 52, 20250 Corti (France)

2006-12-01

The scope of this work was to show the utility of thermal analysis and calorimetric experiments to study the thermal oxidative degradation of Mediterranean scrubs. We investigated the thermal degradation of four species; DSC and TGA were used under air sweeping to record oxidative reactions in dynamic conditions. Heat released and mass loss are important data to be measured for wildland fires modelling purpose and fire hazard studies on ligno-cellulosic fuels. Around 638 and 778K, two dominating and overlapped exothermic peaks were recorded in DSC and individualized using a experimental and numerical separation. This stage allowed obtaining the enthalpy variation of each exothermic phenomenon. As an application, we propose to classify the fuels according to the heat released and the rate constant of each reaction. TGA experiments showed under air two successive mass loss around 638 and 778K. Both techniques are useful in order to measure ignitability, combustibility and sustainability of forest fuels. (author)

Fuel fragmentation model advances using TEXAS-V

Energy Technology Data Exchange (ETDEWEB)

Corradini, M.L.; El-Beshbeeshy, M.; Nilsuwankowsit, S.; Tang, J. [Wisconsin Univ., Madison, WI (United States). Dept. of Nuclear Engineering and Engineering Physics

1998-01-01

Because an energetic fuel-coolant interaction may be a safety hazard, experiments are being conducted to investigate the fuel-coolant mixing/quenching process (FARO) as well as the energetics of vapor explosion propagation for high temperature fuel melt simulants (KROTOS, WFCI, ZrEX). In both types of experiments, the dynamic breakup of the fuel is one of the key aspects that must be fundamentally understood to better estimate the magnitude of the mixing/quenching process or the explosion energetics. To aid our understanding the TEXAS fuel-coolant interaction computer model has been developed and is being used to analyze these experiments. Recently, the models for dynamic fuel fragmentation during the mixing and explosion phases of the FCI have been improved by further insights into these processes. The purpose of this paper is to describe these enhancements and to demonstrate their improvements by analysis of particular JRC FCI data. (author)

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

Results of recent reactor-material tests on dispersal of oxide fuel from a disrupted core

International Nuclear Information System (INIS)

Spencer, B.W.; Wilson, R.J.; Vetter, D.L.; Erickson, E.G.; Dewey, G.

1985-01-01

The results of experimental investigations and related analyses are reported addressing the dispersal of molten oxide fuel from a disrupted core via various available pathways for the CRBR system. These investigations included the GAPFLOW tests in which pressure-driven and gravity drainage tests were performed using dispersal pathways mocking up the intersubassembly gaps, the CAMEL C6 and C7 tests in which molten fuel entered sodium-filled control assembly ducts under prototypic thermal-hydraulic conditions, and the Lower Internals Drainage (LID) tests in which molten fuel drained downward through simulated below-core structure (orifice plate stacks) as the bottom of control assembly ducts. The results of SHOTGUN tests addressing basic freezing of molten UO 2 and UO 2 /metal mixtures flowing through circular tubes are also reported. Test results have invariably shown the existance of stable UO 2 crusts on the inside surfaces of the flow paths. Appreciable removal of fuel was indicated prior to freezing-induced immobilization. Application of heat transfer models based upon the presence of stable, insulating fuel crusts tends to overpredict the removal process

Results of the post-irradiation examination of a highly-rated mixed oxide fuel rod from the Mol 7B experiment

International Nuclear Information System (INIS)

Coquerelle, M.; Walker, C.T.; Whitlow, W.H.

1980-01-01

The experiment MOL 7B was carried out in a epithermal flux in the Belgian reactor BR2. The pin examined contained fuel of initial composition (Usub(0.7)Pusub(0.3))Osub(1.98). It had been irradiated to a maximum burn-up of 13.2 at% at a maximum linear power of 568Wcm -1 . The fuel was clad with coldworked stainless steel. Electron microprobe analysis indicated that a Cr 2 O 3 type oxide was the main constituent of the grey phases in the gap. A metallic phase on the fuel surface had apparently resulted from the mechanical compaction of fragments of cladding that had been depleted in chromium by oxidation. Thus the main components of the phase were iron and nickel. Chromium loss from the inner cladding surface was significant only in the upper regions of the pin. In pin sections that were metallographically examined sigma phase and carbides of the type M 23 C 6 were present at the grain boundaries of the cladding. Cladding corrosion was not an Arrhenius function of the cladding temperature: the amount of metal lost from the inner cladding surface decreased with rise in cladding temperature above 910 K. A contributor to metal loss was the mechanical detachment of fragments of cladding which reformed as a metallic layer on the surface of the fuel. Chromium depletion and sigma phase formation at grain boundaries lowered the cohesive forces between grains which were then mechanically detached. Chromium loss from grain boundaries is mainly the results of oxidation of the cladding by the mixed oxide fuel. Data are presented to support the view that the local average O/M of the fuel determined the rate of oxidation and consequently the extent of chromium depletion. Fuel-cladding mechanical interactions were weak in the upper regions of the pin where metal loss was small

Improving the Thermal Conductivity of UO2 Fuel with the Addition of Graphene

International Nuclear Information System (INIS)

Cho, Byoung Jin; Kim, Young Jin; Sohn, Dong Seong

2012-01-01

Improvement of fuel performances by increasing the fuel thermal conductivity using the BeO or W were reported elsewhere. In this paper, some major fuel performances of improved thermal conductivity oxide (ICO) nuclear fuel with the addition of 10 v/o graphene have been compared to those of standard UO 2 fuel. The fuel thermal conductivity affects many performance parameters and thus is an important parameter to determine the fuel performance. Furthermore, it also affects the performance of the fuel during reactor accidents. The improved thermal conductivity of the fuel would reduce the fuel temperature at the same power condition and would improve the fission gas release, rod internal pressure and fuel stored energy. Graphene is well known for its excellent electrical conductivity, strength and thermal conductivity. The addition of graphene to the UO 2 fuel could increase the thermal conductivity of the ICO fuel. Although the graphene material is extensively studied recently, the characteristics of the graphene material, especially the thermal properties, are not well-known yet. In this study, we used the Light Water Reactor fuel performance analysis code FRAPCON-3.2 to analyze the performance of standard UO 2 and ICO fuel

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.

Solid oxide fuel cells fueled with reducible oxides

Science.gov (United States)

Chuang, Steven S.; Fan, Liang Shih

2018-01-09

A direct-electrochemical-oxidation fuel cell for generating electrical energy includes a cathode provided with an electrochemical-reduction catalyst that promotes formation of oxygen ions from an oxygen-containing source at the cathode, a solid-state reduced metal, a solid-state anode provided with an electrochemical-oxidation catalyst that promotes direct electrochemical oxidation of the solid-state reduced metal in the presence of the oxygen ions to produce electrical energy, and an electrolyte disposed to transmit the oxygen ions from the cathode to the solid-state anode. A method of operating a solid oxide fuel cell includes providing a direct-electrochemical-oxidation fuel cell comprising a solid-state reduced metal, oxidizing the solid-state reduced metal in the presence of oxygen ions through direct-electrochemical-oxidation to obtain a solid-state reducible metal oxide, and reducing the solid-state reducible metal oxide to obtain the solid-state reduced metal.

Limits to fuel/coolant mixing

International Nuclear Information System (INIS)

Corradini, M.L.; Moses, G.A.

1985-01-01

The vapor explosion process involves the mixing of fuel with coolant prior to the explosion. A number of analysts have identified limits to the amount of fuel/coolant mixing that could occur within the reactor vessel following a core melt accident. Past models are reviewed and a sim plified approach is suggested to estimate the upper limit on the amount of fuel/coolant mixing pos sible. The approach uses concepts first advanced by Fauske in a different way. The results indicat that water depth is an important parameter as well as the mixing length scale D /SUB mix/ , and for large values of D /SUB mix/ the fuel mass mixed is limited to <7% of the core mass

Plutonium-containing aerosols found within containment enclosures in industrial mixed-oxide reactor fuel fabrication

International Nuclear Information System (INIS)

Newton, G.J.; Yeh, H.C.; Stanley, J.A.

1977-01-01

Mixed oxide (PuO 2 and UO 2 ) nuclear reactor fuel pellets are fabricated within safety enclosures at Babcock and Wilcox's Park Township site near Apollo, PA. Forty-two sample runs of plutonium-containing aerosols were taken from within glove boxes during routine industrial operations. A small, seven-stage cascade impactor and the Lovelace Aerosol Particle Separator (LAPS) were used to determine aerodynamic size distribution and gross alpha aerosol concentration. Powder comminution and blending produced aerosols with lognormal size distributions characterized by activity median aerodynamic diameters (AMAD) of 1.89 +- 0.33 μm, sigma/sub g/ = 1.62 +- 0.09 and a gross alpha aerosol concentration range of 0.1 to 150 nCi/l. Slug pressing and grinding produced aerosols of AMAD = 3.08 +- 0.1 μm, sigma/sub g/ = 1.53 +- 0.01 and AMAD = 2.26 +- 0.16 μm, sigma/sub g/ = 1.68 +- 0.20, respectively. Gross alpha aerosol concentrations ranged from 3.4 to 450 nCi/l. Centerless grinding produced similar-sized aerosols but the gross alpha concentration ranged from 220 to 1690 nCi/l. In vitro solubility studies on selected LAPS samples in a lung fluid simulant indicate that plutonium mixed-oxide aerosols are more soluble than laboratory-produced plutonium aerosols

Irradiation effects on thermal properties of LWR hydride fuel

Energy Technology Data Exchange (ETDEWEB)

Terrani, Kurt, E-mail: terrani@berkeley.edu [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Balooch, Mehdi [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Carpenter, David; Kohse, Gordon [Massachusetts Institute of Technology, 138 Albany St., Cambridge, MA 02139 (United States); Keiser, Dennis; Meyer, Mitchell [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Olander, Donald [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States)

2017-04-01

Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or ∼5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U (30 wt%)ZrH{sub 1.6}) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup.

The modeling and simulation of thermal based modified solid oxide fuel cell (SOFC for grid-connected systems

Directory of Open Access Journals (Sweden)

Ayetül Gelen

2015-05-01

Full Text Available This paper presents a thermal based modified dynamic model of a Solid Oxide Fuel Cell (SOFC for grid-connected systems. The proposed fuel cell model involves ohmic, activation and concentration voltage losses, thermal dynamics, methanol reformer, fuel utilization factor and power limiting module. A power conditioning unit (PCU, which consists of a DC-DC boost converter and a DC-AC voltage-source inverter (VSI, their controller, transformer and filter, is designed for grid-connected systems. The voltage-source inverter with six Insulated Gate Bipolar Transistor (IGBT switches inverts the DC voltage that comes from the converter into a sinusoidal voltage synchronized with the grid. The simulations and modeling of the system are developed on Matlab/Simulink environment. The performance of SOFC with converter is examined under step and random load conditions. The simulation results show that the designed boost converter for the proposed thermal based modified SOFC model has fairly followed different DC load variations. Finally, the AC bus of 400 Volt and 50 Hz is connected to a single-machine infinite bus (SMIB through a transmission line. The real and reactive power managements of the inverter are analyzed by an infinite bus system. Thus, the desired nominal values are properly obtained by means of the inverter controller.

Charge distribution on plutonium-containing aerosols produced in mixed-oxide reactor fuel fabrication and the laboratory

International Nuclear Information System (INIS)

Yeh, H.C.; Newton, G.J.; Teague, S.V.

1976-01-01

The inhalation toxicity of potentially toxic aerosols may be affected by the electrostatic charge on the particles. Charge may influence the deposition site during inhalation and therefore its subsequent clearance and dose patterns. The electrostatic charge distributions on plutonium-containing aerosols were measured with a miniature, parallel plate, aerosol electrical mobility spectrometer. Two aerosols were studied: a laboratory-produced 238 PuO 2 aerosol (15.8 Ci/g) and a plutonium mixed-oxide aerosol (PU-MOX, natural UO 2 plus PuO 2 , 0.02 Ci/g) formed during industrial centerless grinding of mixed-oxide reactor fuel pellets. Plutonium-238 dioxide particles produced in the laboratory exhibited a small net positive charge within a few minutes after passing through a 85 Kr discharger due to alpha particle emission removal of valence electrons. PU-MOX aerosols produced during centerless grinding showed a charge distribution essentially in Boltzmann equilibrium. The gross alpha aerosol concentrations (960-1200 nCi/l) within the glove box were sufficient to provide high ion concentrations capable of discharging the charge induced by mechanical and/or nuclear decay processes

FRAPCON-2: A Computer Code for the Calculation of Steady State Thermal-Mechanical Behavior of Oxide Fuel Rods

Energy Technology Data Exchange (ETDEWEB)

Berna, G. A; Bohn, M. P.; Rausch, W. N.; Williford, R. E.; Lanning, D. D.

1981-01-01

FRAPCON-2 is a FORTRAN IV computer code that calculates the steady state response of light Mater reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, deformation, and tai lure histories of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (a) heat conduction through the fuel and cladding, (b) cladding elastic and plastic deformation, (c) fuel-cladding mechanical interaction, (d) fission gas release, (e} fuel rod internal gas pressure, (f) heat transfer between fuel and cladding, (g) cladding oxidation, and (h) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat transfer correlations. FRAPCON-2 is programmed for use on the CDC Cyber 175 and 176 computers. The FRAPCON-2 code Is designed to generate initial conditions for transient fuel rod analysis by either the FRAP-T6 computer code or the thermal-hydraulic code, RELAP4/MOD7 Version 2.

Thermal Design for Extra-Terrestrial Regenerative Fuel Cell System

Science.gov (United States)

Gilligan, R.; Guzik, M.; Jakupca, I.; Bennett, W.; Smith, P.; Fincannon, J.

2017-01-01

The Advanced Exploration Systems (AES) Advanced Modular Power Systems (AMPS) Project is investigating different power systems for various lunar and Martian mission concepts. The AMPS Fuel Cell (FC) team has created two system-level models to evaluate the performance of regenerative fuel cell (RFC) systems employing different fuel cell chemistries. Proton Exchange Membrane fuel cells PEMFCs contain a polymer electrolyte membrane that separates the hydrogen and oxygen cavities and conducts hydrogen cations (protons) across the cell. Solid Oxide fuel cells (SOFCs) operate at high temperatures, using a zirconia-based solid ceramic electrolyte to conduct oxygen anions across the cell. The purpose of the modeling effort is to down select one fuel cell chemistry for a more detailed design effort. Figures of merit include the system mass, volume, round trip efficiency, and electrolyzer charge power required. PEMFCs operate at around 60 C versus SOFCs which operate at temperatures greater than 700 C. Due to the drastically different operating temperatures of the two chemistries the thermal control systems (TCS) differ. The PEM TCS is less complex and is characterized by a single pump cooling loop that uses deionized water coolant and rejects heat generated by the system to the environment via a radiator. The solid oxide TCS has its own unique challenges including the requirement to reject high quality heat and to condense the steam produced in the reaction. This paper discusses the modeling of thermal control systems for an extraterrestrial RFC that utilizes either a PEM or solid oxide fuel cell.

Summary of the physical chemical analyses of mixed oxide nuclear fuel as they might influence biological behavior and internal dose

International Nuclear Information System (INIS)

Eidson, A.F.; Mewhinney, J.A.

1987-01-01

Twelve representative materials that might be accidentally released during the fabrication of mixed-oxide nuclear fuel pellets were studied using x-ray diffraction, infrared spectroscopy, energy dispersive x-ray fluorescence, alpha spectroscopy and in vitro dissolution methods. The results are related to a postulated exposure accident and to inhalation experiments using laboratory animals. 19 refs., 5 figs., 19 tabs

Alternative anode materials for solid oxide fuel cells

Energy Technology Data Exchange (ETDEWEB)

Goodenough, John B.; Huang, Yun-Hui [Texas Materials Institute, ETC 9.102, 1 University Station, C2200, The University of Texas at Austin, Austin, TX 78712 (United States)

2007-11-08

The electrolyte of a solid oxide fuel cell (SOFC) is an O{sup 2-}-ion conductor. The anode must oxidize the fuel with O{sup 2-} ions received from the electrolyte and it must deliver electrons of the fuel chemisorption reaction to a current collector. Cells operating on H{sub 2} and CO generally use a porous Ni/electrolyte cermet that supports a thin, dense electrolyte. Ni acts as both the electronic conductor and the catalyst for splitting the H{sub 2} bond; the oxidation of H{sub 2} to H{sub 2}O occurs at the Ni/electrolyte/H{sub 2} triple-phase boundary (TPB). The CO is oxidized at the oxide component of the cermet, which may be the electrolyte, yttria-stabilized zirconia, or a mixed oxide-ion/electron conductor (MIEC). The MIEC is commonly a Gd-doped ceria. The design and fabrication of these anodes are evaluated. Use of natural gas as the fuel requires another strategy, and MIECs are being explored for this application. The several constraints on these MIECs are outlined, and preliminary results of this on-going investigation are reviewed. (author)

Emission computer tomography on a Dodewaard mixed oxide fuel pin. Comparative PIE work with non-destructive and destructive techniques

Energy Technology Data Exchange (ETDEWEB)

Buurveld, H.A.; Dassel, G.

1993-12-01

A nondestructive technique as well as a destructive PIE technique have been used to verify the results obtained with a newly 8-e computer tomography (GECT) system. Multi isotope Scanning (MIS), electron probe micro analysis (EPMA) and GECT were used on a mixed oxide (MOX) fuel rod from the Dodewaard reactor with an average burnup of 24 MWd/kg fuel. GECT shows migration of Cs to the periphery of fuel pellets and to radial cracks and pores in the fuel, whereas MIS shows Cs migration to pellet interfaces. The EPMA technique appeared not to be useful to show migration of Cs but, it shows the distribution of fission products from Pu. EPMA clearly shows the distribution of fission products from Pu, but did not reveal the Cs-migration. (orig./HP)

Assessment of bio-fuel options for solid oxide fuel cell applications