Irradiation effects on thermal properties of LWR hydride fuel

NASA Astrophysics Data System (ADS)

Terrani, Kurt; Balooch, Mehdi; Carpenter, David; Kohse, Gordon; Keiser, Dennis; Meyer, Mitchell; Olander, Donald

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%)ZrH1.6) 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.

AN EVALUATION OF POTENTIAL LINER MATERIALS FOR ELIMINATING FCCI IN IRRADIATED METALLIC NUCLEAR FUEL ELEMENTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. D. Keiser; J. I. Cole

2007-09-01

Metallic nuclear fuels are being looked at as part of the Global Nuclear Energy Program for transmuting longlive transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products. In order to optimize the performance of these fuels, the concept of using liners to eliminate the fuel/cladding chemical interactions that can occur during irradiation of a fuel element has been investigated. The potential liner materials Zr and V have been tested using solid-solid diffusion couples, consisting of liner materials butted against fuel alloys and against cladding materials. The couples were annealed at the relatively high temperature of 700°C. Thismore » temperature would be the absolute maximum temperature present at the fuel/cladding interface for a fuel element in-reactor. Analysis was performed using a scanning electron microscope equipped with energy-dispersive and wavelengthdispersive spectrometers (SEM/EDS/WDS) to evaluate any developed diffusion structures. At 700°C, minimal interaction was observed between the metallic fuels and either Zr or V. Similarly, limited interaction was observed between the Zr and V and the cladding materials. The best performing liner material appeared to be the V, based on amounts of interaction.« less

Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

DOE PAGES

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.; ...

2016-07-15

The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. In the most challenging cases, U–(7–10wt%)Mo monolithic plate fuel are proposed. The chosen design includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction in service. We investigated zircaloy cladding, specifically Zry–4as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo havemore » similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly between roll passes. Our final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction, either from fabrication or in-reactor testing, and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.54E+21« less

Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.

The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. In the most challenging cases, U–(7–10wt%)Mo monolithic plate fuel are proposed. The chosen design includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction in service. We investigated zircaloy cladding, specifically Zry–4as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo havemore » similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly between roll passes. Our final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction, either from fabrication or in-reactor testing, and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.54E+21« less

Fuel cladding behavior under rapid loading conditions

NASA Astrophysics Data System (ADS)

Yueh, K.; Karlsson, J.; Stjärnsäter, J.; Schrire, D.; Ledergerber, G.; Munoz-Reja, C.; Hallstadius, L.

2016-02-01

A modified burst test (MBT) was used in an extensive test program to characterize fuel cladding failure behavior under rapid loading conditions. The MBT differs from a normal burst test with the use of a driver tube to simulate the expansion of a fuel pellet, thereby producing a partial strain driven deformation condition similar to that of a fuel pellet expansion in a reactivity insertion accident (RIA). A piston/cylinder assembly was used to pressurize the driver tube. By controlling the speed and distance the piston travels the loading rate and degree of sample deformation could be controlled. The use of a driver tube with a machined gauge section localizes deformation and allows for continuous monitoring of the test sample diameter change at the location of maximum hoop strain, during each test. Cladding samples from five irradiated fuel rods were tested between 296 and 553 K and loading rates from 1.5 to 3.5/s. The test rods included variations of Zircaloy-2 with different liners and ZIRLO, ranging in burn-up from 41 to 74 GWd/MTU. The test results show cladding ductility is strongly temperature and loading rate dependent. Zircaloy-2 cladding ductility degradation due to operational hydrogen pickup started to recover at approximately 358 K for test condition used in the study. This recovery temperature is strongly loading rate dependent. At 373 K, ductility recovery was small for loading rates less than 8 ms equivalent RIA pulse width, but longer than 8 ms the ductility recovery increased exponentially with increasing pulse width, consistent with literature observations of loading rate dependent brittle-to-ductile (BTD) transition temperature. The cladding ductility was also observed to be strongly loading rate/pulse width dependent for BWR cladding below the BTD temperature and Pressurized Water Reactor (PWR) cladding at both 296 and 553 K.

Electrical heating tests of uranium dioxide external fuel configuration at emitter temperature of 1900 K

NASA Technical Reports Server (NTRS)

Diianni, D. C.; Mayer, J. T.

1974-01-01

Testing of two fuel clad specimens for thermionic reactor application is described. The annular UO2 fuel was clad on both sides with tungsten; heat rejection was radially inward. The tests were intended to study inner clad stability, fuel redistribution, and fuel melting problems. The specimens were tested in a vacuum chamber using electron bombardment heating. Fuel structural changes were studied using periodic gammagraphs and posttest metallography. The first specimen test was terminated at 50 hours because of a braze failure. The second specimen was tested for 240 hours when an outer clad leak developed due to a tungsten-water reaction. The fuel developed numerous cracks on cooldown but the inner clad remained dimensionally stable. The fuel cover gas did not impede the rate of fuel redistribution. Posttest examination showed the fuel had not melted during operation.

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

NASA Astrophysics Data System (ADS)

Karahan, Aydın; Kazimi, Mujid S.

2013-10-01

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.

Container for reprocessing and permanent storage of spent nuclear fuel assemblies

DOEpatents

Forsberg, Charles W.

1992-01-01

A single canister process container for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining their integrity at temperature necessary to oxide the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ingham, J.G.

Maximum cladding temperatures occur when the IDENT 1578 fuel pin shipping container is installed in the T-3 Cask. The maximum allowable cladding temperature of 800/sup 0/F is reached when the rate of energy deposited in the 19-pin basket reaches 400 watts. Since 45% of the energy which is generated in the fuel escapes the 19-pin basket without being deposited, mostly gamma energy, the maximum allowable rate of heat generation is 400/.55 = 727 watts. Similarly, the maximum allowable cladding temperature of 800/sup 0/F is reached when the rate of energy deposited in the 40-pin basket reaches 465 watts. Since 33%more » of the energy which is generated in the fuel escapes the 40-pin basket without being deposited, mostly gamma energy, the maximum allowable rate of heat generation is 465/.66 = 704 watts. The IDENT 1578 fuel pin shipping container therefore meets its thermal design criteria. IDENT 1578 can handle fuel pins with a decay heat load of 600 watts while maintaining the maximum fuel pin cladding temperature below 800/sup 0/F. The emissivities which were determined from the test results for the basket tubes and container are relatively low and correspond to new, shiny conditions. As the IDENT 1578 container is exposed to high temperatures for extended periods of time during the transportation of fuel pins, the emissivities will probably increase. This will result in reduced temperatures.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Konyashov, Vadim V.; Krasnov, Alexander M.

Results are provided of the experimental investigation of radioactive fission product (RFP) release, i.e., krypton, xenon, and iodine radionuclides from fuel elements with initial defects during long-term (3 to 5 yr) irradiation under low linear power (5 to 12 kW/m) and during special experiments in the VK-50 vessel-type boiling water reactor.The calculation model for the RFP release from the fuel-to-cladding gap of the defective fuel element into coolant was developed. It takes into account the convective transport in the fuel-to-cladding gap and RFP sorption on the internal cladding surface and is in good agreement with the available experimental data. Anmore » approximate analytical solution of the transport equation is given. The calculation dependencies of the RFP release coefficients on the main parameters such as defect size, fuel-to-cladding gap, temperature of the internal cladding surface, and radioactive decay constant were analyzed.It is shown that the change of the RFP release from the fuel elements with the initial defects during long-term irradiation is, mainly, caused by fuel swelling followed by reduction of the fuel-to-cladding gap and the fuel temperature. The calculation model for the RFP release from defective fuel elements applicable to light water reactors (LWRs) was developed. It takes into account the change of the defective fuel element parameters during long-term irradiation. The calculation error according to the program does not exceed 30% over all the linear power change range of the LWR fuel elements (from 5 to 26 kW/m)« less

Fabrication and testing of U-7Mo monolithic plate fuel with Zircaloy cladding

NASA Astrophysics Data System (ADS)

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.; Wachs, D. M.; Finlay, M. R.

2016-10-01

Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U-(7-10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry-4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry-4 clad U-7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry-4 and U-(7-10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction-either from fabrication or in-reactor testing-and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm3, 3.8E+21 (peak).

FRAPCON analysis of cladding performance during dry storage operations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Richmond, David J.; Geelhood, Kenneth J.

There is an increasing need in the U.S. and around the world to move used nuclear fuel from wet storage in fuel pools to dry storage in casks stored at independent spent fuel storage installations (ISFSI) or interim storage sites. The NRC limits cladding temperature to 400°C while maintaining cladding hoop stress below 90 MPa in an effort to avoid radial hydride reorientation. An analysis was conducted with FRAPCON-4.0 on three modern fuel designs with three representative used nuclear fuel storage temperature profiles that peaked at 400 °C. Results were representative of the majority of U.S. LWR fuel. They conservativelymore » showed that hoop stress remains below 90 MPa at the licensing temperature limit. Results also show that the limiting case for hoop stress may not be at the highest rod internal pressure in all cases but will be related to the axial temperature and oxidation profiles of the rods at the end of life and in storage.« less

BISON Fuel Performance Analysis of IFA-796 Rod 3 & 4 and Investigation of the Impact of Fuel Creep

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wirth, Brian; Terrani, Kurt A.; Sweet, Ryan T.

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace the currently used zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromiumaluminum (FeCrAl) alloys because they exhibit slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and slow cladding consumption in the presence of high temperature steam. These alloys should also exhibit increased “coping time” in the event of an accident scenario by improving the mechanical performance at high temperatures, allowing greater flexibility to achieve core cooling.more » As a continuation of the development of these alloys, in-reactor irradiation testing of FeCrAl cladded fuel rods has started. In order to provide insight on the possible behavior of these fuel rods as they undergo irradiation in the Halden Boiling Water Reactor, engineering analysis has been performed using FeCrAl material models implemented into the BISON fuel performance code. This milestone report provides an update on the ongoing development of modeling capability to predict FeCrAl cladding fuel performance and to provide an early look at the possible behavior of planned in-reactor FeCrAl cladding experiments. In particular, this report consists of two separate analyses. The first analysis consists of fuel performance simulations of IFA-796 rod 4 and two segments of rod 3. These simulations utilize previously implemented material models for the C35M FeCrAl alloy and UO2 to provide a bounding behavior analysis corresponding to variation of the initial fuel cladding gap thickness within the fuel rod. The second analysis is an assessment of the fuel and cladding stress states after modification of the fuel creep model that is currently implemented in the BISON fuel performance code. Effects from modifying the fuel creep model were identified for the BISON simulations of the IFA-796 rod 4 experiment, but show that varying the creep model (within the range investigated here) only provide a minimal increase in the fuel radius and maximum cladding hoop stress. Continued investigation of fuel behavioral models will include benchmarking the modified fuel creep model against available experimental data, as well as an investigation of the role that fuel cracking will play in the compliance of the fuel. Correctly calculating stress evolution in the fuel is key to assessing fuel behavior up to gap closure and the subsequent deformation of the cladding due to PCMI. The inclusion of frictional contact should also be investigated to determine the axial elongation of the fuel rods for comparison with data from this experiment.« less

Container for reprocessing and permanent storage of spent nuclear fuel assemblies

DOEpatents

Forsberg, C.W.

1992-03-24

A single canister process container is described for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining its integrity at a temperature necessary to oxidize the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container. 10 figs.

Blister Threshold Based Thermal Limits for the U-Mo Monolithic Fuel System

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. M. Wachs; I. Glagolenko; F. J. Rice

2012-10-01

Fuel failure is most commonly induced in research and test reactor fuel elements by exposure to an under-cooled or over-power condition that results in the fuel temperature exceeding a critical threshold above which blisters form on the plate. These conditions can be triggered by normal operational transients (i.e. temperature overshoots that may occur during reactor startup or power shifts) or mild upset events (e.g., pump coastdown, small blockages, mis-loading of fuel elements into higher-than-planned power positions, etc.). The rise in temperature has a number of general impacts on the state of a fuel plate that include, for example, stress relaxationmore » in the cladding (due to differential thermal expansion), softening of the cladding, increased mobility of fission gases, and increased fission-gas pressure in pores, all of which can encourage the formation of blisters on the fuel-plate surface. These blisters consist of raised regions on the surface of fuel plates that occur when the cladding plastically deforms in response to fission-gas pressure in large pores in the fuel meat and/or mechanical buckling of the cladding over damaged regions in the fuel meat. The blister temperature threshold decreases with irradiation because the mechanical properties of the fuel plate degrade while under irradiation (due to irradiation damage and fission-product accumulation) and because the fission-gas inventory progressively increases (and, thus, so does the gas pressure in pores).« less

Spent fuel behavior under abnormal thermal transients during dry storage

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stahl, D.; Landow, M.P.; Burian, R.J.

1986-01-01

This study was performed to determine the effects of abnormally high temperatures on spent fuel behavior. Prior to testing, calculations using the CIRFI3 code were used to determine the steady-state fuel and cask component temperatures. The TRUMP code was used to determine transient heating rates under postulated abnormal events during which convection cooling of the cask surfaces was obstructed by a debris bed covering the cask. The peak rate of temperature rise during the first 6 h was calculated to be about 15/sup 0/C/h, followed by a rate of about 1/sup 0/C/h. A Turkey Point spent fuel rod segment wasmore » heated to approx. 800/sup 0/C. The segment deformed uniformly with an average strain of 17% at failure and a local strain of 60%. Pretest characterization of the spent fuel consisted of visual examination, profilometry, eddy-current examination, gamma scanning, fission gas collection, void volume measurement, fission gas analysis, hydrogen analysis of the cladding, burnup analysis, cladding metallography, and fuel ceramography. Post-test characterization showed that the failure was a pinhole cladding breach. The results of the tests showed that spent fuel temperatures in excess of 700/sup 0/C are required to produce a cladding breach in fuel rods pressurized to 500 psing (3.45 MPa) under postulated abnormal thermal transient cask conditions. The pinhole cladding breach that developed would be too small to compromise the confinement of spent fuel particles during an abnormal event or after normal cooling conditions are restored. This behavior is similar to that found in other slow ramp tests with irradiated and nonirradiated rod sections and nonirradiated whole rods under conditions that bracketed postulated abnormal heating rates. This similarity is attributed to annealing of the irradiation-strengthened Zircaloy cladding during heating. In both cases, the failure was a benign, ductile pinhole rupture.« less

Innovative coating of nanostructured vanadium carbide on the F/M cladding tube inner surface for mitigating the fuel cladding chemical interactions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yang, Yong; Phillpot, Simon

Fuel cladding chemical interactions (FCCI) have been acknowledged as a critical issue in a metallic fuel/steel cladding system due to the formation of low melting intermetallic eutectic compounds between the fuel and cladding steel, resulting in reduction in cladding wall thickness as well as a formation of eutectic compounds that can initiate melting in the fuel at lower temperature. In order to mitigate FCCI, diffusion barrier coatings on the cladding inner surface have been considered. In order to generate the required coating techniques, pack cementation, electroplating, and electrophoretic deposition have been investigated. However, these methods require a high processing temperaturemore » of above 700 oC, resulting in decarburization and decomposition of the martensites in a ferritic/martensitic (F/M) cladding steel. Alternatively, organometallic chemical vapor deposition (OMCVD) can be a promising process due to its low processing temperature of below 600 oC. The aim of the project is to conduct applied and fundamental research towards the development of diffusion barrier coatings on the inner surface of F/M fuel cladding tubes. Advanced cladding steels such as T91, HT9 and NF616 have been developed and extensively studied as advanced cladding materials due to their excellent irradiation and corrosion resistance. However, the FCCI accelerated by the elevated temperature and high neutron exposure anticipated in fast reactors, can have severe detrimental effects on the cladding steels through the diffusion of Fe into fuel and lanthanides towards into the claddings. To test the functionality of developed coating layer, the diffusion couple experiments were focused on using T91 as cladding and Ce as a surrogate lanthanum fission product. By using the customized OMCVD coating equipment, thin and compact layers with a few micron between 1.5 µm and 8 µm thick and average grain size of 200 nm and 5 µm were successfully obtained at the specimen coated between 300oC and 500 oC, respectively. The coating layer contains both carbon and vanadium elements as quantified by WED, and the phases mainly consist of a mixture of V2C and VC, which was confirmed using X-ray diffraction patterns. In addition, the ratio between V and C varies with processing temperature, and it was observed that a higher temperature promotes the carbon adsorption and increases thickness of the coating. With optimized deposition conditions, we can apply the coating technique toward the actual T91 cladding materials, and provide the possibilities for the real application in sodium-cooled fast reactors (SFRs). Diffusion couple experiments were performed at both 550 oC and 690 oC, which corresponds to normal and aggressive operating temperatures, respectively. The results show that vanadium carbide coating with wider thickness (8 µm) and lower carbon concentration (27 at.%) reduced the width of the inter diffusion region, indicating that vanadium carbide coating can mitigate FCCI effectively. In specific, inter-diffusion between Fe and Ce was prohibited over most area, but Ce diffusion occurred toward the coating and the Fe substrate through thinner coating layer, which needs further optimization in terms of uniform coating thickness. Overall, it is concluded that this coating process can be successfully applied onto the inner surface of HT9 cladding tubes and the FCCI can be effectively mitigated if not totally eliminated.« less

Simulation of irradiation hardening of Zircaloy within plate-type dispersion nuclear fuel elements

NASA Astrophysics Data System (ADS)

Jiang, Yijie; Wang, Qiming; Cui, Yi; Huo, Yongzhong; Ding, Shurong

2011-06-01

Within plate-type dispersion nuclear fuel elements, the metal matrix and cladding attacked continuously by fast neutrons undergo irradiation hardening, which might have remarkable effects upon the mechanical behaviors within fuel elements. In this paper, with the irradiation hardening effect of metal materials mainly considered together with irradiation growth effect of the cladding, the three-dimensional large-deformation constitutive relations for the metal matrix and cladding are developed. The method of virtual temperature increase in the previous studies is further developed to model the irradiation swelling of fuel particles; the method of anisotropic thermal expansion is introduced to model irradiation growth of the cladding; and a method of multi-step-temperature loading is proposed to simulate the coupling features of irradiation-induced swelling of the fuel particles together with irradiation growth of the cladding. Above all, based on the developed relationship between irradiation growth at certain burnup and the loaded virtual temperatures, with considering that certain burnup corresponds to certain fast neutron fluence, the time-dependent constitutive relation due to irradiation hardening effect is replaced by the virtual-temperature-dependent one which is introduced into the commercial software to simulate the irradiation hardening effects of the matrix and cladding. Numerical simulations of the irradiation-induced mechanical behaviors are implemented with the finite element method in consideration of the micro-structure of the fuel meat. The obtained results indicate that when the irradiation hardening effects are introduced into the constitutive relations of the metal matrix and cladding: (1) higher maximum Mises stresses for certain burnup at the matrix exist with the equivalent plastic strains remaining almost the same at lower burnups; (2) the maximum Mises stresses for certain burnup at the cladding are enhanced while the maximum equivalent plastic strains are reduced; and (3) the maximum first principal stresses for certain burnup at the matrix or the cladding are lower than the ones without the hardening effect, and the differences are found to increase with burnup; and the variation rules of the interfacial stresses are similar.

Fabrication of (U, Zr) C-fueled/tungsten-clad specimens for irradiation in the Plum Brook Reactor Facility

NASA Technical Reports Server (NTRS)

1972-01-01

Fuel samples, 90UC - 10 ZrC, and chemically vapor deposited tungsten fuel cups were fabricated for the study of the long term dimensional stability and compatibility of the carbide-tungsten fuel-cladding systems under irradiation. These fuel samples and fuel cups were assembled into the fuel pins of two capsules, designated as V-2E and V-2F, for irradiation in NASA Plum Brook Reactor Facility at a fission power density of 172 watts/c.c. and a miximum cladding temperature of 1823 K. Fabrication methods and characteristics of the fuel samples and fuel cups prepared are described.

Carbide fuel pin and capsule design for irradiations at thermionic temperatures

NASA Technical Reports Server (NTRS)

Siegel, B. L.; Slaby, J. G.; Mattson, W. F.; Dilanni, D. C.

1973-01-01

The design of a capsule assembly to evaluate tungsten-emitter - carbide-fuel combinations for thermionic fuel elements is presented. An inpile fuel pin evaluation program concerned with clad temperture, neutron spectrum, carbide fuel composition, fuel geometry,fuel density, and clad thickness is discussed. The capsule design was a compromise involving considerations between heat transfer, instrumentation, materials compatibility, and test location. Heat-transfer calculations were instrumental in determining the method of support of the fuel pin to minimize axial temperature variations. The capsule design was easily fabricable and utilized existing state-of-the-art experience from previous programs.

Development of Advanced Ods Ferritic Steels for Fast Reactor Fuel Cladding

NASA Astrophysics Data System (ADS)

Ukai, S.; Oono, N.; Ohtsuka, S.; Kaito, T.

Recent progress of the 9CrODS steel development is presented focusing on their microstructure control to improve sufficient high-temperature strength as well as cladding manufacturing capability. The martensitic 9CrODS steel is primarily candidate cladding materials for the Generation IV fast reactor fuel. They are the attractive composite-like materials consisting of the hard residual ferrite and soft tempered martensite, which are able to be easily controlled by α-γ phase transformation. The residual ferrite containing extremely nanosized oxide particles leads to significantly improved creep rupture strength in 9CrODS cladding. The creep strength stability at extended time of 60,000 h at 700 ºC is ascribed to the stable nanosized oxide particles. It was also reviewed that 9CrODS steel has well irradiation stability and fuel pin irradiation test was conducted up to 12 at% burnup and 51 dpa at the cladding temperature of 700ºC.

Thermal analysis of conceptual designs for GPHS/FPSE power systems of 250 We and 500 We

NASA Technical Reports Server (NTRS)

Mccomas, Thomas J.; Dugan, Edward T.

1991-01-01

Thermal analyses were performed for two distinct configurations of a proposed space nuclear power system which combines General Purpose Heat Source (GPHS) modules with the state of the art Free-Piston Stirling Engines (FPSEs). The two configurations correspond to systems with power levels of 250 and 500 W(sub e). The 250 W(sub e) GPHS/FPSE power system utilizes four GPHS modules and one FPSE, and the 500 W(sub e) contains eight GPHS modules and two FPSEs. The configurations of the systems and the bases for selecting the configurations are described. Brief introductory sections are included to describe the GPHS modules and free piston Stirling engines. The primary focus of the thermal analyses is on the temperature of the iridium fuel clad within the GPHS modules. A design goal temperature of 1573 K was selected as the upper limit for the fuel clad during normal operating conditions. The basis for selecting this temperature limit is discussed in detail. Results obtained from thermal analysis of the 250 W(sub e) GPHS/FPSE power system indicate fuel clad temperatures which slightly exceed the design goal temperature of 1573 K. The results are considered favorable due to the numerous conservative assumptions used in developing the thermal model and performing the thermal analysis. To demonstrate the effects of the conservatism, a brief sensitivity analysis is performed in which a few of the key system parameters are varied to determine their effect on the fuel clad temperatures. It is concluded that thermal analysis of a more detailed thermal model would be expected to yield fuel clad temperatures below the design foal temperature limiy 1573 K.

Development of monolithic nuclear fuels for RERTR by hot isostatic pressing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jue, J.-F.; Park, Blair; Chapple, Michael

2008-07-15

The RERTR Program (Reduced Enrichment for Research and Test Reactors) is developing advanced nuclear fuels for high power test reactors. Monolithic fuel design provides a higher uranium loading than that of the traditional dispersion fuel design. In order to bond monolithic fuel meat to aluminum cladding, several bonding methods such as roll bonding, friction stir bonding and hot isostatic pressing, have been explored. Hot isostatic pressing is a promising process for low cost, batch fabrication of monolithic RERTR fuel plates. The progress on the development of this process at the Idaho National Laboratory will be presented. Due to the relativelymore » high processing temperature used, the reaction between fuel meat and aluminum cladding to form brittle intermetallic phases may be a concern. The effect of processing temperature and time on the fuel/cladding reaction will be addressed. The influence of chemical composition on the reaction will also be discussed. (author)« less

Preliminary analysis of hot spot factors in an advanced reactor for space electric power systems

NASA Technical Reports Server (NTRS)

Lustig, P. H.; Holms, A. G.; Davison, H. W.

1973-01-01

The maximum fuel pin temperature for nominal operation in an advanced power reactor is 1370 K. Because of possible nitrogen embrittlement of the clad, the fuel temperature was limited to 1622 K. Assuming simultaneous occurrence of the most adverse conditions a deterministic analysis gave a maximum fuel temperature of 1610 K. A statistical analysis, using a synthesized estimate of the standard deviation for the highest fuel pin temperature, showed probabilities of 0.015 of that pin exceeding the temperature limit by the distribution free Chebyshev inequality and virtually nil assuming a normal distribution. The latter assumption gives a 1463 K maximum temperature at 3 standard deviations, the usually assumed cutoff. Further, the distribution and standard deviation of the fuel-clad gap are the most significant contributions to the uncertainty in the fuel temperature.

Risk Assessment of Structural Integrity of Transportation Casks after Extended Storage

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ibarra, Luis; Medina, Ricardo; Yang, Haori

This study assessed the risk of loss of structural integrity of transportation casks and fuel cladding after extended storage. Although it is known that fuel rods discharged from NPPs have a small percentage of rod cladding defects, the behavior of fuel cladding and the structural elements of assemblies during transportation after long-term storage is not well understood. If the fuel degrades during extended storage, it could be susceptible to damage from vibration and impact loads during transport operations, releasing fission-product gases into the canister or the cask interior (NWTRB 2010). Degradation of cladding may occur due to mechanisms associated withmore » hydrogen embrittlement, delayed hydride cracking, low temperature creep, and stress corrosion cracking (SCC) that may affect fuel cladding and canister components after extended storage of hundreds of years. Over extended periods at low temperatures, these mechanisms affect the ductility, strength, and fracture toughness of the fuel cladding, which becomes brittle. For transportation purposes, the fuel may be transferred from storage to shipping casks, or dual-purpose casks may be used for storage and transportation. Currently, most of the transportation casks will be the former case. A risk assessment evaluation is conducted based on results from experimental tests and simulations with advanced numerical models. A novel contribution of this study is the evaluation of the combined effect of component aging and vibration/impact loads in transportation scenarios. The expected levels of deterioration will be obtained from previous and current studies on the effect of aging on fuel and cask components. The emphasis of the study is placed on the structural integrity of fuel cladding and canisters.« less

An experimental platform for real-time measurement of the deformation of nuclear fuel rod claddings submitted to thermal transients

NASA Astrophysics Data System (ADS)

Gallais, L.; Burla, R.; Martin, F.; Richaud, J. C.; Volle, G.; Pontillon, M.; Capdevila, H.; Pontillon, Y.

2018-01-01

We report on experimental development and qualification of a system developed to detect and quantify the deformations of the cladding surface of nuclear fuel pellet assemblies submitted to heat transient conditions. The system consists of an optical instrument, based on 2 wavelengths speckle interferometry, associated with an induction furnace and a model pellet assembly used to simulate the radial thermal gradient experienced by fuel pellets in pressurized water reactors. We describe the concept, implementation, and first results obtained with this system. We particularly demonstrate that the optical system is able to provide real time measurements of the cladding surface shape during the heat transients from ambient to high temperatures (up to a cladding surface temperature of 600 °C) with micrometric resolution.

An experimental platform for real-time measurement of the deformation of nuclear fuel rod claddings submitted to thermal transients.

PubMed

Gallais, L; Burla, R; Martin, F; Richaud, J C; Volle, G; Pontillon, M; Capdevila, H; Pontillon, Y

2018-01-01

We report on experimental development and qualification of a system developed to detect and quantify the deformations of the cladding surface of nuclear fuel pellet assemblies submitted to heat transient conditions. The system consists of an optical instrument, based on 2 wavelengths speckle interferometry, associated with an induction furnace and a model pellet assembly used to simulate the radial thermal gradient experienced by fuel pellets in pressurized water reactors. We describe the concept, implementation, and first results obtained with this system. We particularly demonstrate that the optical system is able to provide real time measurements of the cladding surface shape during the heat transients from ambient to high temperatures (up to a cladding surface temperature of 600 °C) with micrometric resolution.

Experimental evaluation of thermal ratcheting behavior in UO2 fuel elements

NASA Technical Reports Server (NTRS)

Phillips, W. M.

1973-01-01

The effects of thermal cycling of UO2 at high temperatures has been experimentally evaluated to determine the rates of distortion of UO2/clad fuel elements. Two capsules were rested in the 1500 C range, one with a 50 C thermal cycle, the other with a 100 C thermal cycle. It was observed that eight hours at the lower cycle temperature produced sufficient UO2 redistribution to cause clad distortion. The amount of distortion produced by the 100 C cycle was less than double that produced by the 50 C, indicating smaller thermal cycles would result in clad distortion. An incubation period was observed to occur before the onset of distortion with cycling similar to fuel swelling observed in-pile at these temperatures.

Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

NASA Astrophysics Data System (ADS)

Carmack, W. J.; Chichester, H. M.; Porter, D. L.; Wootan, D. W.

2016-05-01

The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The MFF fuel operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in EBR-II experiments. Data from the MFF-3 and MFF-5 assemblies are most comparable to the data obtained from the EBR-II X447 experiment. The two X447 pin breaches were strongly influenced by fuel/cladding chemical interaction (FCCI) at the top of the fuel column. Post irradiation examination data from MFF-3 and MFF-5 are presented and compared to historical EBR-II data.

Models for the Configuration and Integrity of Partially Oxidized Fuel Rod Cladding at High Temperatures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Siefken, L.J.

1999-01-01

Models were designed to resolve deficiencies in the SCDAP/RELAP5/MOD3.2 calculations of the configuration and integrity of hot, partially oxidized cladding. These models are expected to improve the calculations of several important aspects of fuel rod behavior. First, an improved mapping was established from a compilation of PIE results from severe fuel damage tests of the configuration of melted metallic cladding that is retained by an oxide layer. The improved mapping accounts for the relocation of melted cladding in the circumferential direction. Then, rules based on PIE results were established for calculating the effect of cladding that has relocated from abovemore » on the oxidation and integrity of the lower intact cladding upon which it solidifies. Next, three different methods were identified for calculating the extent of dissolution of the oxidic part of the cladding due to its contact with the metallic part. The extent of dissolution effects the stress and thus the integrity of the oxidic part of the cladding. Then, an empirical equation was presented for calculating the stress in the oxidic part of the cladding and evaluating its integrity based on this calculated stress. This empirical equation replaces the current criterion for loss of integrity which is based on temperature and extent of oxidation. Finally, a new rule based on theoretical and experimental results was established for identifying the regions of a fuel rod with oxidation of both the inside and outside surfaces of the cladding. The implementation of these models is expected to eliminate the tendency of the SCDAP/RELAP5 code to overpredict the extent of oxidation of the upper part of fuel rods and to underpredict the extent of oxidation of the lower part of fuel rods and the part with a high concentration of relocated material. This report is a revision and reissue of the report entitled, Improvements in Modeling of Cladding Oxidation and Meltdown.« less

Fuel Performance Calculations for FeCrAl Cladding in BWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

George, Nathan; Sweet, Ryan; Maldonado, G. Ivan

2015-01-01

This study expands upon previous neutronics analyses of the reactivity impact of alternate cladding concepts in boiling water reactor (BWR) cores and directs focus toward contrasting fuel performance characteristics of FeCrAl cladding against those of traditional Zircaloy. Using neutronics results from a modern version of the 3D nodal simulator NESTLE, linear power histories were generated and supplied to the BISON-CASL code for fuel performance evaluations. BISON-CASL (formerly Peregrine) expands on material libraries implemented in the BISON fuel performance code and the MOOSE framework by providing proprietary material data. By creating material libraries for Zircaloy and FeCrAl cladding, the thermomechanical behaviormore » of the fuel rod (e.g., strains, centerline fuel temperature, and time to gap closure) were investigated and contrasted.« less

Development and Experimental Benchmark of Simulations to Predict Used Nuclear Fuel Cladding Temperatures during Drying and Transfer Operations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Greiner, Miles

Radial hydride formation in high-burnup used fuel cladding has the potential to radically reduce its ductility and suitability for long-term storage and eventual transport. To avoid this formation, the maximum post-reactor temperature must remain sufficiently low to limit the cladding hoop stress, and so that hydrogen from the existing circumferential hydrides will not dissolve and become available to re-precipitate into radial hydrides under the slow cooling conditions during drying, transfer and early dry-cask storage. The objective of this research is to develop and experimentallybenchmark computational fluid dynamics simulations of heat transfer in post-pool-storage drying operations, when high-burnup fuel cladding ismore » likely to experience its highest temperature. These benchmarked tools can play a key role in evaluating dry cask storage systems for extended storage of high-burnup fuels and post-storage transportation, including fuel retrievability. The benchmarked tools will be used to aid the design of efficient drying processes, as well as estimate variations of surface temperatures as a means of inferring helium integrity inside the canister or cask. This work will be conducted effectively because the principal investigator has experience developing these types of simulations, and has constructed a test facility that can be used to benchmark them.« less

Modelling of LOCA Tests with the BISON Fuel Performance Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williamson, Richard L; Pastore, Giovanni; Novascone, Stephen Rhead

2016-05-01

BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculationsmore » are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.« less

EXPERIMENTAL STUDIES OF TRANSIENT EFFECTS IN FAST REACTOR FUELS. SERIES I. UO$sub 2$ IRRADIATIONS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, J.H.

1962-11-15

An experimental program to evaluate the performance of FCR and EFCR fuel during transient operation is outlined, and the initial series of tests are described in some detail. Test results from five experiments in the TREAT reactor, using 1-in. OD SS-clad UO/sub 2/ fuel specimens, are compared with regard to fuel temperatures, mechanical integrity, and post-irradiation appearance. Incipient fuel pin failure limits for transients are identified with maximum fuel temperatures in the range of 7000 deg F. Multiple transient damage to the cladding is likely for transients above the melting point of the fuel. (auth)

Development of Cold Spray Coatings for Accident-Tolerant Fuel Cladding in Light Water Reactors

NASA Astrophysics Data System (ADS)

Maier, Benjamin; Yeom, Hwasung; Johnson, Greg; Dabney, Tyler; Walters, Jorie; Romero, Javier; Shah, Hemant; Xu, Peng; Sridharan, Kumar

2018-02-01

The cold spray coating process has been developed at the University of Wisconsin-Madison for the deposition of oxidation-resistant coatings on zirconium alloy light water reactor fuel cladding with the goal of improving accident tolerance during loss of coolant scenarios. Coatings of metallic (Cr), alloy (FeCrAl), and ceramic (Ti2AlC) materials were successfully deposited on zirconium alloy flats and cladding tube sections by optimizing the powder size, gas preheat temperature, pressure and composition, and other process parameters. The coatings were dense and exhibited excellent adhesion to the substrate. Evaluation of the samples after high-temperature oxidation tests at temperatures up to 1300°C showed that the cold spray coatings significantly mitigate oxidation kinetics because of the formation of thin passive oxide layers on the surface. The results of the study indicate that the cold spray coating process is a viable near-term option for developing accident-tolerant zirconium alloy fuel cladding.

Fabrication of Monolithic RERTR Fuels by Hot Isostatic Pressing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jan-Fong Jue; Blair H. Park; Curtis R. Clark

2010-11-01

The RERTR (Reduced Enrichment for Research and Test Reactors) Program is developing advanced nuclear fuels for high-power test reactors. Monolithic fuel design provides higher uranium loading than that of the traditional dispersion fuel design. Hot isostatic pressing is a promising process for low-cost batch fabrication of monolithic RERTR fuel plates for these high-power reactors. Bonding U Mo fuel foil and 6061 Al cladding by hot isostatic press bonding was successfully developed at Idaho National Laboratory. Due to the relatively high processing temperature, the interaction between fuel meat and aluminum cladding is a concern. Two different methods were employed to mitigatemore » this effect: (1) a diffusion barrier and (2) a doping addition to the interface. Both types of fuel plates have been fabricated by hot isostatic press bonding. Preliminary results show that the direct fuel/cladding interaction during the bonding process was eliminated by introducing a thin zirconium diffusion barrier layer between the fuel and the cladding. Fuel plates were also produced and characterized with a silicon-rich interlayer between fuel and cladding. This paper reports the recent progress of this developmental effort and identifies the areas that need further attention.« less

75 FR 77017 - Nextera Energy Seabrook, LLC Seabrook Station Independent Spent Fuel Storage Installation; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-10

... April 2011. The NUHOMS[supreg] HD-32PTH system is designed to passively remove decay heat to assure integrity of the concrete HSM and fuel cladding, and the thermal monitoring requirements for the system are... to prevent conditions that could lead to exceeding the concrete and fuel cladding temperature...

Full-length U-xPu-10Zr (x = 0, 8, 19 wt.%) fast reactor fuel test in FFTF

NASA Astrophysics Data System (ADS)

Porter, D. L.; Tsai, Hanchung

2012-08-01

The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt.%) metallic fast reactor test with commercial-length (91.4-cm active fuel-column length) conducted to date. With few remaining test reactors, there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning-of-life (BOL) peak cladding temperature of the hottest pin was 608 °C, cooling to 522 °C at end-of-life (EOL). Selected fuel pins were examined non-destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta-gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3-cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at ˜0.7 X/L axial location along the fuel column. This resulted from a higher production of rare-earth fission products at this location and a higher ΔT between fuel center and cladding than at core center, together providing more rare earths at the cladding and more FCCI. This behavior could actually help extend the life of a fuel pin in a "long pin" reactor design to a higher peak fuel burnup.

Development of Self-Healing Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sridharan, Kumar; Mariani, Robert; Bai, Xianming

Zirconium-alloy fuel claddings have been used successfully in Light Water Reactors (LWR) for over four decades. However, under high temperature accident conditions, zirconium-alloys fuel claddings exhibit profuse exothermic oxidation accompanied by release of hydrogen gas due to the reaction with water/steam. Additionally, the ZrO 2 layer can undergo monoclinic to tetragonal to cubic phase transformations at high temperatures which can induce stresses and cracking. These events were unfortunately borne out in the Fukushima-Daiichi accident in in Japan in 2011. In reaction to such accident, protective oxidation-resistant coatings for zirconium-alloy fuel claddings has been extensively investigated to enhance safety margins inmore » accidents as well as fuel performance under normal operation conditions. Such surface modification could also beneficially affect fuel rod heat transfer characteristics. Zirconium-silicide, a candidate coating material, is particularly attractive because zirconium-silicide coating is expected to bond strongly to zirconium-alloy substrate. Intermetallic compound phases of zirconium-silicide have high melting points and oxidation of zirconium silicide produces highly corrosion resistant glassy zircon (ZrSiO 4) and silica (SiO 2) which possessing self-healing qualities. Given the long-term goal of developing such coatings for use with nuclear reactor fuel cladding, this work describes results of oxidation and corrosion behavior of bulk zirconium-silicide and fabrication of zirconium-silicide coatings on zirconium-alloy test flats, tube configurations, and SiC test flats. In addition, boiling heat transfer of these modified surfaces (including ZrSi 2 coating) during clad quenching experiments is discussed in detail.« less

Fabrication and life testing of thermionic converters

NASA Technical Reports Server (NTRS)

Yang, L.; Bruce, R.

1973-01-01

An unfueled converter containing a chloride-fluoride duplex tungsten emitter of 4.78 eV vacuum work function was tested for 46,647 hours at an emitter temperature of 1973 K and an electrode power output of about 8 watts/sq cm. The test demonstrated the superior and stable performance of the (110) oriented tungsten emitter at high temperatures. Three 90 UC-10 ZrC(C/U = 1.04, tungsten additive = 4 wt %) fueled converters were fabricated and tested at an emitter temperature of 1873 K. Converter containing chloride-arc-cast duplex tungsten cladding showed temperature thermionic performance and slower rate of performance drop than converter containing chloride-fluoride duplex tungsten cladding. This is believed to be due to the superior fuel component diffusion resistance of the arc-cast tungsten substrate used in the fuel cladding. It was shown that a converter containing a carbide fueled chloride-arc-cast duplex tungsten emitter with an initial electrode power output of 6.80 watts/sq cm could still deliver an electrode power output of 6.16 watts/sq cm after 18,632 hours of operation at an emitter temperature of 1873 K.

Thermal Analysis of ZPPR High Pu Content Stored Fuel

DOE PAGES

Solbrig, Charles W.; Pope, Chad L.; Andrus, Jason P.

2014-09-17

The Zero Power Physics Reactor (ZPPR) operated from April 18, 1969, until 1990. ZPPR operated at low power for testing nuclear reactor designs. This paper examines the temperature of Pu content ZPPR fuel while it is in storage. Heat is generated in the fuel due to Pu and Am decay and is a concern for possible cladding damage. Damage to the cladding could lead to fuel hydriding and oxidizing. A series of computer simulations were made to determine the range of temperatures potentially occuring in the ZPPR fuel. The maximum calculated fuel temperature is 292°C (558°F). Conservative assumptions in themore » model intentionally overestimate temperatures. The stored fuel temperatures are dependent on the distribution of fuel in the surrounding storage compartments, the heat generation rate of the fuel, and the orientation of fuel. Direct fuel temperatures could not be measured but storage bin doors, storage sleeve doors, and storage canister temperatures were measured. Comparison of these three temperatures to the calculations indicates that the temperatures calculated with conservative assumptions are, as expected, higher than the actual temperatures. The maximum calculated fuel temperature with the most conservative assumptions is significantly below the fuel failure criterion of 600°C (1,112°F).« less

Laser and Pressure Resistance Weld of Thin-Wall Cladding for LWR Accident-Tolerant Fuels

NASA Astrophysics Data System (ADS)

Gan, J.; Jerred, N.; Perez, E.; Haggard, D. C.

2017-12-01

FeCrAl alloy with typical composition of approximately Fe-15Cr-5Al is considered a primary candidate cladding material for light water reactor accident-tolerant fuel because of its superior resistance to oxidation in high-temperature steam compared with Zircaloy cladding. Thin-walled FeCrAl cladding at 350 μm wall thickness is required, and techniques for joining endplug to cladding need to be developed. Fusion-based laser weld and solid-state joining with pressure resistance weld were investigated in this study. The results of microstructural characterization, mechanical property evaluation by tensile testing, and hydraulic pressure burst testing of the welds for the cladding-endplug specimen are discussed.

Laser and Pressure Resistance Weld of Thin-Wall Cladding for LWR Accident-Tolerant Fuels

NASA Astrophysics Data System (ADS)

Gan, J.; Jerred, N.; Perez, E.; Haggard, D. C.

2018-02-01

FeCrAl alloy with typical composition of approximately Fe-15Cr-5Al is considered a primary candidate cladding material for light water reactor accident-tolerant fuel because of its superior resistance to oxidation in high-temperature steam compared with Zircaloy cladding. Thin-walled FeCrAl cladding at 350 μm wall thickness is required, and techniques for joining endplug to cladding need to be developed. Fusion-based laser weld and solid-state joining with pressure resistance weld were investigated in this study. The results of microstructural characterization, mechanical property evaluation by tensile testing, and hydraulic pressure burst testing of the welds for the cladding-endplug specimen are discussed.

Fuel clad chemical interactions in fast reactor MOX fuels

NASA Astrophysics Data System (ADS)

Viswanathan, R.

2014-01-01

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

Method of bonding

DOEpatents

Saller, deceased, Henry A.; Hodge, Edwin S.; Paprocki, Stanley J.; Dayton, Russell W.

1987-12-01

1. A method of making a fuel-containing structure for nuclear reactors, comprising providing an assembly comprising a plurality of fuel units; each fuel unit consisting of a core plate containing thermal-neutron-fissionable material, sheets of cladding metal on its bottom and top surfaces, said cladding sheets being of greater width and length than said core plates whereby recesses are formed at the ends and sides of said core plate, and end pieces and first side pieces of cladding metal of the same thickness as the core plate positioned in said recesses, the assembly further comprising a plurality of second side pieces of cladding metal engaging the cladding sheets so as to space the fuel units from one another, and a plurality of filler plates of an acid-dissolvable nonresilient material whose melting point is above 2000.degree. F., each filler plate being arranged between a pair of said second side pieces and the cladding plates of two adjacent fuel units, the filler plates having the same thickness as the second side pieces; the method further comprising enclosing the entire assembly in an envelope; evacuating the interior of the entire assembly through said envelope; applying inert gas under a pressure of about 10,000 psi to the outside of said envelope while at the same time heating the assembly to a temperature above the flow point of the cladding metal but below the melting point of any material of the assembly, whereby the envelope is pressed against the assembly and integral bonds are formed between plates, sheets, first side pieces, and end pieces and between the sheets and the second side pieces; slowly cooling the assembly to room temperature; removing the envelope; and dissolving the filler plates without attacking the cladding metal.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wood, Elizabeth Sooby; Parker, Stephen Scott; Nelson, Andrew Thomas

The Fuel Cycle Research and Development program’s Advanced Fuels Campaign is currently supporting a range of experimental efforts aimed at the development and qualification of ‘accident tolerant’ nuclear fuel forms. One route to enhance the accident tolerance of nuclear fuel is to replace the zirconium alloy cladding, which is prone to rapid oxidation in steam at elevated temperatures, with a more oxidation-resistant cladding. Several cladding replacement solutions have been envisaged. The cladding can be completely replaced with a more oxidation resistant alloy, a layered approach can be used to optimize the strength, creep resistance, and oxidation tolerance of various materials,more » or the existing zirconium alloy cladding can be coated with a more oxidation-resistant material. Molybdenum is one candidate cladding material favored due to its high temperature creep resistance. However, it performs poorly under autoclave testing and suffers degradation under high temperature steam oxidation exposure. Development of composite cladding architectures consisting of a molybdenum core shielded by a molybdenum disilicide (MoSi 2) coating is hypothesized to improve the performance of a Mo-based cladding system. MoSi 2 was identified based on its high temperature oxidation resistance in O 2 atmospheres (e.g. air and “wet air”). However, its behavior in H 2O is less known. This report presents thermogravimetric analysis (TGA), scanning electron microscopy (SEM), and x-ray diffraction (XRD) results for MoSi 2 exposed to 670-1498 K water vapor. Synthetic air (80-20%, Ar-O 2) exposures were also performed, and those results are presented here for a comparative analysis. It was determined that MoSi 2 displays drastically different oxidation behavior in water vapor than in dry air. In the 670-1498 K temperature range, four distinct behaviors are observed. Parabolic oxidation is exhibited in only 670-773 K water vapor, a temperature range in which the material pests in dry O 2 environments. From 877-1084 K in water vapor, MoSi 2 undergoes rapid mass gain resulting in oxidation throughout the bulk of the sample at 980 K and 1084 K. The resulting material displays swelling and warping after the 980-1084 K exposures. A pre-passivation heat treatment performed at 1395 K was found capable of producing a coarse SiO 2 layer that limited pesting at lower temperatures in water vapor over the time periods investigated.« less

Applicability of out-of-pile fretting wear tests to in-reactor fretting wear-induced failure time prediction

NASA Astrophysics Data System (ADS)

Kim, Kyu-Tae

2013-02-01

In order to investigate whether or not the grid-to-rod fretting wear-induced fuel failure will occur for newly developed spacer grid spring designs for the fuel lifetime, out-of-pile fretting wear tests with one or two fuel assemblies are to be performed. In this study, the out-of-pile fretting wear tests were performed in order to compare the potential for wear-induced fuel failure in two newly-developed, Korean PWR spacer grid designs. Lasting 20 days, the tests simulated maximum grid-to-rod gap conditions and the worst flow induced vibration effects that might take place over the fuel life time. The fuel rod perforation times calculated from the out-of-pile tests are greater than 1933 days for 2 μm oxidized fuel rods with a 100 μm grid-to-rod gap, whereas those estimated from in-reactor fretting wear failure database may be about in the range of between 60 and 100 days. This large discrepancy in fuel rod perforation may occur due to irradiation-induced cladding oxide microstructure changes on the one hand and a temperature gradient-induced hydrogen content profile across the cladding metal region on the other hand, which may accelerate brittleness in the grid-contacting cladding oxide and metal regions during the reactor operation. A three-phase grid-to-rod fretting wear model is proposed to simulate in-reactor fretting wear progress into the cladding, considering the microstructure changes of the cladding oxide and the hydrogen content profile across the cladding metal region combined with the temperature gradient. The out-of-pile tests cannot be directly applicable to the prediction of in-reactor fretting wear-induced cladding perforations but they can be used only for evaluating a relative wear resistance of one grid design against the other grid design.

Low temperature chemical processing of graphite-clad nuclear fuels

DOEpatents

Pierce, Robert A.

2017-10-17

A reduced-temperature method for treatment of a fuel element is described. The method includes molten salt treatment of a fuel element with a nitrate salt. The nitrate salt can oxidize the outer graphite matrix of a fuel element. The method can also include reduced temperature degradation of the carbide layer of a fuel element and low temperature solubilization of the fuel in a kernel of a fuel element.

Multispectral pyrometry for surface temperature measurement of oxidized Zircaloy claddings

NASA Astrophysics Data System (ADS)

Bouvry, B.; Cheymol, G.; Ramiandrisoa, L.; Javaudin, B.; Gallou, C.; Maskrot, H.; Horny, N.; Duvaut, T.; Destouches, C.; Ferry, L.; Gonnier, C.

2017-06-01

Non-contact temperature measurement in a nuclear reactor is still a huge challenge because of the numerous constraints to consider, such as the high temperature, the steam atmosphere, and irradiation. A device is currently developed at CEA to study the nuclear fuel claddings behavior during a Loss-of-Coolant Accident. As a first step of development, we designed and tested an optical pyrometry procedure to measure the surface temperature of nuclear fuel claddings without any contact, under air, in the temperature range 700-850 °C. The temperature of Zircaloy-4 cladding samples was retrieved at various temperature levels. We used Multispectral Radiation Thermometry with the hypothesis of a constant emissivity profile in the spectral ranges 1-1.3 μm and 1.45-1.6 μm. To allow for comparisons, a reference temperature was provided by a thermocouple welded on the cladding surface. Because of thermal losses induced by the presence of the thermocouple, a heat transfer simulation was also performed to estimate the bias. We found a good agreement between the pyrometry measurement and the temperature reference, validating the constant emissivity profile hypothesis used in the MRT estimation. The expanded measurement uncertainty (k = 2) of the temperature obtained by the pyrometry method was ±4 °C, for temperatures between 700 and 850 °C. Emissivity values, between 0.86 and 0.91 were obtained.

Irradiation of TZM: Uranium dioxide fuel pin at 1700 K

NASA Technical Reports Server (NTRS)

Mcdonald, G. E.

1973-01-01

A fuel pin clad with TZM and containing solid pellets of uranium dioxide was fission heated in a static helium-cooled capsule at a maximum surface temperature of 1700 K for approximately 1000 hr and to a total burnup of 2.0 percent of the uranium-235. The results of the postirradiation examination indicated: (1) A transverse, intergranular failure of the fuel pin occurred when the fuel pin reached 2.0-percent burnup. This corresponds to 1330 kW-hr/cu cm, where the volume is the sum of the fuel, clad, and void volumes in the fuel region. (2) The maximum swelling of the fuel pin was less than 1.5 percent on the fuel-pin diameter. (3) There was no visible interaction between the TZM clad and the UO2. (4) Irradiation at 1700 K produced a course-grained structure, with an average grain diameter of 0.02 centimeter and with some of the grains extending one-half of the thickness of the clad. (5) Below approximately 1500 K, the irradiation of the clad produced a moderately fine-grained structure, with an average grain diameter of 0.004 centimeter.

Status Report on the Fabrication of Fuel Cladding Chemical Interaction Test Articles for ATR Irradiations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G.; Howard, Richard H.

FeCrAl alloys are a promising new class of alloys for light water reactor (LWR) applications due to their superior oxidation and corrosion resistance in high temperature environments. The current R&D efforts have focused on the alloy composition and processing routes to generate nuclear grade FeCrAl alloys with optimized properties for enhanced accident tolerance while maintaining properties needed for normal operation conditions. Therefore, the composition and processing routes must be optimized to maintain the high temperature steam oxidation (typically achieved by increasing the Cr and Al content) while still exhibiting properties conducive to normal operation in a LWR (such as radiationmore » tolerance where reducing Cr content is favorable). Within this balancing act is the addition of understanding the influence on composition and processing routes on the FeCrAl alloys for fuel-cladding chemical interactions (FCCI). Currently, limited knowledge exists on FCCI for the FeCrAl-UO 2 clad-fuel system. To overcome the knowledge gaps on the FCCI for the FeCrAl-UO2 clad-fuel system a series of fueled irradiation tests have been developed for irradiation in the Advanced Test Reactor (ATR) housed at the Idaho National Laboratory (INL). The first series of tests has already been reported. These tests used miniaturized 17x17 PWR fuel geometry rodlets of second-generation FeCrAl alloys fueled with industrial Westinghouse UO 2 fuel. These rodlets were encapsulated within a stainless steel housing.To provide high fidelity experiments and more robust testing, a new series of rodlets have been developed deemed the Accident Tolerant Fuel Experiment #1 Oak Ridge National Laboratory FCCI test (ATF-1 ORNL FCCI). The main driving factor, which is discussed in detail, was to provide a radiation environment where prototypical fuel-clad interface temperatures are met while still maintaining constant contact between industrial fuel and the candidate cladding alloys, hence promoting FCCI between the fuel-clad systems. The other factor was to develop a test bed where multiple candidate alloys could be evaluated within a single irradiation test train, thereby reducing overall costs and increasing efficiency in alloy screening efforts. A collaboration between ORNL and INL was developed to facilitate the completion of the test bed for FCCI testing. The report highlights the activities related to the development of the ATF-1 ORNL FCCI rodlets for irradiation in INL’s ATR as part of the on-going ATF-1 experiments.« less

Mechanical behavior of aluminum-bearing ferritic alloys for accident-tolerant fuel cladding applications

NASA Astrophysics Data System (ADS)

Guria, Ankan

Nuclear power currently provides about 13% of electrical power worldwide. Nuclear reactors generating this power traditionally use Zirconium (Zr) based alloys as the fuel cladding material. Exothermic reaction of Zr with steam under accident conditions may lead to production of hydrogen with the possibility of catastrophic consequences. Following the Fukushima-Daiichi incident, the exploration of accident-tolerant fuel cladding materials accelerated. Aluminum-rich (around 5 wt. %) ferritic steels such as Fecralloy, APMT(TM) and APM(TM) are considered as potential materials for accident-tolerant fuel cladding applications. These materials create an aluminum-based oxide scale protecting the alloy at elevated temperatures. Tensile deformation behavior of the above alloys was studied at different temperatures (25-500 °C) at a strain rate of 10-3 s-1 and correlated with microstructural characteristics. Higher strength and decent ductility of APMT(TM) led to further investigation of the alloy at various combination of strain rates and temperatures followed by fractography and detailed microscopic analyses. Serrations appeared in the stress-strain curves of APMT(TM) and Fecralloy steel tested in a limited temperature range (250-400 °C). The appearance of serrations is explained on the basis of dynamic strain aging (DSA) effect due to solute-dislocation interactions. The research in this study is being performed using the funds received from the US DOE Office of Nuclear Energy's Nuclear Energy University Programs (NEUP).

Hot Cell Installation and Demonstration of the Severe Accident Test Station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Linton, Kory D.; Burns, Zachary M.; Terrani, Kurt A.

A Severe Accident Test Station (SATS) capable of examining the oxidation kinetics and accident response of irradiated fuel and cladding materials for design basis accident (DBA) and beyond design basis accident (BDBA) scenarios has been successfully installed and demonstrated in the Irradiated Fuels Examination Laboratory (IFEL), a hot cell facility at Oak Ridge National Laboratory. The two test station modules provide various temperature profiles, steam, and the thermal shock conditions necessary for integral loss of coolant accident (LOCA) testing, defueled oxidation quench testing and high temperature BDBA testing. The installation of the SATS system restores the domestic capability to examinemore » postulated and extended LOCA conditions on spent fuel and cladding and provides a platform for evaluation of advanced fuel and accident tolerant fuel (ATF) cladding concepts. This document reports on the successful in-cell demonstration testing of unirradiated Zircaloy-4. It also contains descriptions of the integral test facility capabilities, installation activities, and out-of-cell benchmark testing to calibrate and optimize the system.« less

Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Billone, M. C.; Burtseva, T. A.

2016-08-30

The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

Novel Accident-Tolerant Fuel Meat and Cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robert D. Mariani; Pavel G Medvedev; Douglas L Porter

A novel accident-tolerant fuel meat and cladding are here proposed. The fuel meat design incorporates annular fuel with inserts and discs that are fabricated from a material having high thermal conductivity, for example niobium. The inserts are rods or tubes. Discs separate the fuel pellets. Using the BISON fuel performance code it was found that the peak fuel temperature can be lowered by more than 600 degrees C for one set of conditions with niobium metal as the thermal conductor. In addition to improved safety margin, several advantages are expected from the lower temperature such as decreased fission gas releasemore » and fuel cracking. Advantages and disadvantages are discussed. An enrichment of only 7.5% fully compensates the lost reactivity of the displaced UO2. Slightly higher enrichments, such as 9%, allow uprates and increased burnups to offset the initial costs for retooling. The design has applications for fast reactors and transuranic burning, which may accelerate its development. A zirconium silicide coating is also described for accident tolerant applications. A self-limiting degradation behavior for this coating is expected to produce a glassy, self-healing layer that becomes more protective at elevated temperature, with some similarities to MoSi2 and other silicides. Both the fuel and coating may benefit from the existing technology infrastructure and the associated wide expertise for a more rapid development in comparison to other, more novel fuels and cladding.« less

Estimation of ring tensile properties of steam oxidized Zircaloy-4 fuel cladding under simulated LOCA condition

NASA Astrophysics Data System (ADS)

Shriwastaw, R. S.; Sawarn, Tapan K.; Banerjee, Suparna; Rath, B. N.; Dubey, J. S.; Kumar, Sunil; Singh, J. L.; Bhasin, Vivek

2017-09-01

The present study involves the estimation of ring tensile properties of Indian Pressurised Heavy Water Reactor (IPHWR) fuel cladding made of Zircaloy-4, subjected to experiments under a simulated loss-of-coolant-accident (LOCA) condition. Isothermal steam oxidation experiments were conducted on clad tube specimens at temperatures ranging from 900 to 1200 °C at an interval of 50 °C for different soaking periods with subsequent quenching in water at ambient temperature. The specimens, which survived quenching, were then subjected to ambient temperature ring tension test (RTT). The microstructure was correlated with the mechanical properties. The yield strength (YS) and ultimate tensile strength (UTS) increased initially with rise in oxidation temperature and time duration but then decreased with further increase in oxidation. Ductility is adversely affected with rising oxidation temperature and longer holding time. A higher fraction of load bearing phase and lower oxygen content in it ensures higher residual ductility. Cladding shows almost zero ductility behavior in RIT when load bearing phase fraction is less than 0.72 and its average oxygen concentration is greater than 0.58 wt%.

Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCulloch, R.W.; MacPherson, R.E.

1983-03-01

The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through cladmore » melting at 1370/sup 0/C.« less

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

NASA Astrophysics Data System (ADS)

Karahan, Aydın; 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 UO2-PuO2 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. FEAST-METAL was benchmarked against the open-literature EBR-II database for steady state and furnace tests (transients). The results show that the code is able to predict important phenomena such as clad strain, fission gas release, clad wastage, clad failure time, axial fuel slug deformation and fuel constituent redistribution, satisfactorily.

Off-design temperature effects on nuclear fuel pins for an advanced space-power-reactor concept

NASA Technical Reports Server (NTRS)

Bowles, K. J.

1974-01-01

An exploratory out-of-reactor investigation was made of the effects of short-time temperature excursions above the nominal operating temperature of 990 C on the compatibility of advanced nuclear space-power reactor fuel pin materials. This information is required for formulating a reliable reactor safety analysis and designing an emergency core cooling system. Simulated uranium mononitride (UN) fuel pins, clad with tungsten-lined T-111 (Ta-8W-2Hf) showed no compatibility problems after heating for 8 hours at 2400 C. At 2520 C and above, reactions occurred in 1 hour or less. Under these conditions free uranium formed, redistributed, and attacked the cladding.

Migration of Point Defects in the Field of a Temperature Gradient

NASA Astrophysics Data System (ADS)

Kozlov, A. V.; Portnykh, I. A.; Pastukhov, V. I.

2018-04-01

The influence of the temperature gradient over the thickness of the cladding of a fuel element of a fast-neutron reactor on the migration of point defects formed in the cladding material due to neutron irradiation has been studied. It has been shown that, under the action of the temperature gradient, the flux of vacancies onto the inner surface of the cladding is higher than the flux of interstitial atoms, which leads to the formation of a specific concentration profile in the cladding with a vacancy-depleted zone near the inner surface. The experimental results on the spatial distribution of pores over the cladding thickness have been presented with which the data on the concentration profiles and vacancy fluxes have been compared.

Analysis of LOCA Scenarios in the NIST Research Reactor Before and After Fuel Conversion

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baek, J. S.; Cheng, L. Y.; Diamond, D.

An analysis has been done of hypothetical loss-of-coolant-accidents (LOCAs) in the research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The purpose of the analysis is to determine if the peak clad temperature remains below the Safety Limit, which is the blister temperature for the fuel. The configuration of the NBSR considered in the analysis is that projected for the future when changes will be made so that shutdown pumps do not operate when a LOCA signal is detected. The analysis was done for the present core with high-enriched uranium (HEU) fuel and with the proposed low-enrichedmore » uranium (LEU) fuel that would be used when the NBSR is converted from one to the other. The analysis consists of two parts. The first examines how the water would drain from the primary system following a break and the possibility for the loss of coolant from within the fuel element flow channels. This work is performed using the TRACE system thermal-hydraulic code. The second looks at the fuel clad temperature as a function of time given that the water may have drained from many of the flow channels and the water in the vessel is in a quasi-equilibrium state. The temperature behavior is investigated using the three-dimensional heat conduction code HEATING7.3. The results in all scenarios considered for both HEU and LEU fuel show that the peak clad temperature remains below the blister temperature.« less

The effect of silicon on the interaction between metallic uranium and aluminum: A 50 year long diffusion experiment

NASA Astrophysics Data System (ADS)

Leenaers, A.; Detavernier, C.; Van den Berghe, S.

2008-11-01

The core of the BR1 research reactor at SCK•CEN, Mol (Belgium) has a graphite matrix loaded with fuel rods consisting of a natural uranium slug in aluminum cladding. The BR1 reactor has been in operation since 1956 and still contains its original fuel rods. After more than 50 years irradiation at low temperature, some of the fuel rods have been examined. Fabrication reports indicate that a so-called AlSi bonding layer and an U(Al,Si) 3 anti-diffusion layer on the natural uranium fuel slug were applied to limit the interaction between the uranium fuel and aluminum cladding. The microstructure of the fuel, bonding and anti-diffusion layer and cladding were analysed using optical microscopy, scanning electron microscopy and electron microprobe analysis. It was found that the AlSi bonding layer does provide a tight bond between fuel and cladding but that it is a thin USi layer that acts as effective anti-diffusion layer and not the intended U(Al,Si) 3 layer.

Neutron diffraction measurement of residual stresses, dislocation density and texture in Zr-bonded U-10Mo “mini” fuel foils and plates

DOE PAGES

Brown, Donald William; Okuniewski, Maria A.; Sisneros, Thomas A.; ...

2016-12-01

Here, Al clad U-10Mo fuel plates are being considered for conversion of several research reactors from high-enriched to low-enriched U fuel. Neutron diffraction measurements of the textures, residual phase stresses, and dislocation densities in the individual phases of the mini-foils throughout several processing steps and following hot-isostatic pressing to the Al cladding, have been completed. Recovery and recrystallization of the bare U-10Mo fuel foil, as indicated by the dislocation density and texture, are observed depending on the state of the material prior to annealing and the duration and temperature of the annealing process. In general, the cladding procedure significantly reducesmore » the dislocation density, but the final state of the clad plate, both texture and dislocation density, depends strongly on the final processing step of the fuel foil. In contrast, the residual stress state of the final plate is dominated by the thermal expansion mismatch of the constituent materials.« less

Neutron diffraction measurement of residual stresses, dislocation density and texture in Zr-bonded U-10Mo “mini” fuel foils and plates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Donald William; Okuniewski, Maria A.; Sisneros, Thomas A.

Here, Al clad U-10Mo fuel plates are being considered for conversion of several research reactors from high-enriched to low-enriched U fuel. Neutron diffraction measurements of the textures, residual phase stresses, and dislocation densities in the individual phases of the mini-foils throughout several processing steps and following hot-isostatic pressing to the Al cladding, have been completed. Recovery and recrystallization of the bare U-10Mo fuel foil, as indicated by the dislocation density and texture, are observed depending on the state of the material prior to annealing and the duration and temperature of the annealing process. In general, the cladding procedure significantly reducesmore » the dislocation density, but the final state of the clad plate, both texture and dislocation density, depends strongly on the final processing step of the fuel foil. In contrast, the residual stress state of the final plate is dominated by the thermal expansion mismatch of the constituent materials.« less

NUCLEAR REACTOR COMPENENT CLADDING MATERIAL

DOEpatents

Draley, J.E.; Ruther, W.E.

1959-01-27

Fuel elements and coolant tubes used in nuclear reactors of the heterogeneous, water-cooled type are described, wherein the coolant tubes extend through the moderator and are adapted to contain the fuel elements. The invention comprises forming the coolant tubes and the fuel element cladding material from an alloy of aluminum and nickel, or an alloy of aluminum, nickel, alloys are selected to prevent intergranular corrosion of these components by water at temperatures up to 35O deg C.

Welding fixture for nuclear fuel pin cladding assemblies

DOEpatents

Oakley, David J.; Feld, Sam H.

1986-01-01

A welding fixture for locating a driver sleeve about the open end of a nuclear fuel pin cladding. The welding fixture includes a holder provided with an open cavity having shoulders for properly positioning the driver sleeve, the end cap, and a soft, high temperature resistant plastic protective sleeve that surrounds a portion of the end cap stem. Ejected contaminant particles spewed forth by closure of the cladding by pulsed magnetic welding techniques are captured within a contamination trap formed in the holder for ultimate removal and disposal of contaminating particles along with the holder.

Fuel Retention Improvement at High Temperatures in Tungsten-Uranium Dioxide Dispersion Fuel Elements by Plasma-Spray Cladding

NASA Technical Reports Server (NTRS)

Grisaffe, Salvatore J.; Caves, Robert M.

1964-01-01

An investigation was undertaken to determine the feasibility of depositing integrally bonded plasma-sprayed tungsten coatings onto 80-volume-percent tungsten - 20-volume-percent uranium dioxide composites. These composites were face clad with thin tungsten foil to inhibit uranium dioxide loss at elevated temperatures, but loss at the unclad edges was still significant. By preheating the composite substrates to approximately 3700 degrees F in a nitrogen environment, metallurgically bonded tungsten coatings could be obtained directly by plasma spraying. Furthermore, even though these coatings were thin and somewhat porous, they greatly inhibited the loss of uranium dioxide. For example, a specimen that was face clad but had no edge cladding lost 5.8 percent uranium dioxide after 2 hours at 4750 dgrees F in flowing hydrogen. A similar specimen with plasma-spray-coated edges, however, lost only 0.75 percent uranium dioxide under the same testing conditions.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gleicher, Frederick; Ortensi, Javier; DeHart, Mark

Accurate calculation of desired quantities to predict fuel behavior requires the solution of interlinked equations representing different physics. Traditional fuels performance codes often rely on internal empirical models for the pin power density and a simplified boundary condition on the cladding edge. These simplifications are performed because of the difficulty of coupling applications or codes on differing domains and mapping the required data. To demonstrate an approach closer to first principles, the neutronics application Rattlesnake and the thermal hydraulics application RELAP-7 were coupled to the fuels performance application BISON under the master application MAMMOTH. A single fuel pin was modeledmore » based on the dimensions of a Westinghouse 17x17 fuel rod. The simulation consisted of a depletion period of 1343 days, roughly equal to three full operating cycles, followed by a station blackout (SBO) event. The fuel rod was depleted for 1343 days for a near constant total power loading of 65.81 kW. After 1343 days the fission power was reduced to zero (simulating a reactor shut-down). Decay heat calculations provided the time-varying energy source after this time. For this problem, Rattlesnake, BISON, and RELAP-7 are coupled under MAMMOTH in a split operator approach. Each system solves its physics on a separate mesh and, for RELAP-7 and BISON, on only a subset of the full problem domain. Rattlesnake solves the neutronics over the whole domain that includes the fuel, cladding, gaps, water, and top and bottom rod holders. Here BISON is applied to the fuel and cladding with a 2D axi-symmetric domain, and RELAP-7 is applied to the flow of the circular outer water channel with a set of 1D flow equations. The mesh on the Rattlesnake side can either be 3D (for low order transport) or 2D (for diffusion). BISON has a matching ring structure mesh for the fuel so both the power density and local burn up are copied accurately from Rattlesnake. At each depletion time step, Rattlesnake calculates a power density, fission density rate, burn-up distribution and fast flux based on the current water density and fuel temperature. These are then mapped to the BISON mesh for a fuels performance solve. BISON calculates the fuel temperature and cladding surface temperature based upon the current power density and bulk fluid temperature. RELAP-7 then calculates the fluid temperature, water density fraction and water phase velocity based upon the cladding surface temperature. The fuel temperature and the fluid density are then passed back to Rattlesnake for another neutronics calculation. Six Picard or fixed-point style iterations are preformed in this manner to obtain consistent tightly coupled and stable results. For this paper a set of results from the detailed calculation are provided for both during depletion and the SBO event. We demonstrate that a detailed calculation closer to first principles can be done under MAMMOTH between different applications on differing domains.« less

Overview of the multifaceted activities towards development and deployment of nuclear-grade FeCrAl Alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G; Yamamoto, Yukinori; Pint, Bruce A

2016-01-01

A large effort is underway under the leadership of US DOE Fuel Cycle R&D program to develop advanced FeCrAl alloys as accident tolerant fuel (ATF) cladding to replace Zr-based alloys in light water reactors. The primary motivation is the excellent oxidation resistance of these alloys in high-temperature steam environments right up to their melting point (roughly three orders of magnitude slower oxidation kinetics than zirconium). A multifaceted effort is ongoing to rapidly advance FeCrAl alloys as a mature ATF concept. The activities span the broad spectrum of alloy development, environmental testing (high-temperature high-pressure water and elevated temperature steam), detailed mechanicalmore » characterization, material property database development, neutron irradiation, thin tube production, and multiple integral fuel test campaigns. Instead of off-the-shelf commercial alloys that might not prove optimal for the LWR fuel cladding application, a large amount of effort has been placed on the alloy development to identify the most optimum composition and microstructure for this application. The development program is targeting a cladding that offers performance comparable to or better than modern Zr-based alloys under normal operating and off-normal conditions. This paper provides a comprehensive overview of the systematic effort to advance nuclear-grade FeCrAl alloys as an ATF cladding in commercial LWRs.« less

Vanadium diffusion coating on HT-9 cladding for mitigating the fuel cladding chemical interactions

NASA Astrophysics Data System (ADS)

Lo, Wei-Yang; Yang, Yong

2014-08-01

Fuel cladding chemical interaction (FCCI) has been identified as one of the crucial issues for developing Ferritic/Martensitic (F/M) stainless steel claddings for metallic fuels in a fast reactor. The anticipated elevated temperature and high neutron flux can significantly aggravate the FCCI, in terms of formation of inter-diffusion and lower melting point eutectic phases. To mitigate the FCCI, vanadium carbide coating as a diffusion barrier was deposited on the HT-9 substrate using a pack cementation diffusion coating (PCDC) method, and the processing temperature was optimized down to 730 °C. A solid metallurgical bonding between the coating layer and substrate was achieved, and the coating is free from through depth cracks. The microstructural characterizations using SEM and TEM show a nanostructured grain structure. EDS/WDS and XRD analysis confirm the phase of coating layer as V2C. Diffusion couple tests at 660 °C for 100 h demonstrate that V2C layer with a thickness of less than 5 μm can effectively eliminate the inter-diffusion between the lanthanide cerium and HT-9 steel.

COMPARTMENTED REACTOR FUEL ELEMENT

DOEpatents

Cain, F.M. Jr.

1962-09-11

A method of making a nuclear reactor fuel element of the elongated red type is given wherein the fissionable fuel material is enclosed within a tubular metal cladding. The method comprises coating the metal cladding tube on its inside wall with a brazing alloy, inserting groups of cylindrical pellets of fissionable fuel material into the tube with spacing members between adjacent groups of pellets, sealing the ends of the tubes to leave a void space therewithin, heating the tube and its contents to an elevated temperature to melt the brazing alloy and to expand the pellets to their maximum dimensions under predetermined operating conditions thereby automatically positioning the spacing members along the tube, and finally cooling the tube to room temperature whereby the spacing disks become permanently fixed at their edges in the brazing alloy and define a hermetically sealed compartment for each fl group of fuel pellets. Upon cooling, the pellets contract thus leaving a space to accommodate thermal expansion of the pellets when in use in a reactor. The spacing members also provide lateral support for the tubular cladding to prevent collapse thereof when subjected to a reactor environment. (AEC)

First overpower tests of metallic IFR [Integral Fast Reactor] fuel in TREAT [Transient Reactor Test Facility]: Data and analysis from tests M5, M6, and M7

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

1989-12-01

Results and analyses of margin to cladding failure and pre-failure axial expansion of metallic fuel are reported for TREAT in-pile transient overpower tests M5--M7. These are the first such tests on reference binary and ternary alloy fuel of the Integral Fast Reactor (IFR) concept with burnup ranging from 1 to 10 at. %. In all cases, test fuel was subjected to an exponential power rise on an 8 s period until either incipient or actual cladding failure was achieved. Objectives, designs and methods are described with emphasis on developments unique to metal fuel safety testing. The resulting database for claddingmore » failure threshold and prefailure fuel expansion is presented. The nature of the observed cladding failure and resultant fuel dispersals is described. Simple models of cladding failures and pre-failure axial expansions are described and compared with experimental results. Reported results include: temperature, flow, and pressure data from test instrumentation; fuel motion diagnostic data principally from the fast neutron hodoscope; and test remains described from both destructive and non-destructive post-test examination. 24 refs., 144 figs., 17 tabs.« less

STUDIES OF FAST REACTOR FUEL ELEMENT BEHAVIOR UNDER TRANSIENT HEATING TO FAILURE. I. INITIAL EXPERIMENTS ON METALLIC SAMPLES IN THE ABSENCE OF COOLANT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dickerman, C. E.; Sowa, E. S.; Okrent, D.

1961-08-01

Meltdown tests on single metallic unirradiated fuel elements in TREAT are described. The fuel elements (EBRII Mark I fuel pins, EBR-II fuel pins with retractory Nb or Ta cladding, and Fermi-I fuel pins) are tested in an inert atmosphere, with no coolant. The fuel elements are exposed to reactor power bursts of 200 msec to 25 sec duration, under conditions simulating fast reactor operations. For these tests, the type of power burst, the integrated power, the fuel enrichment, the maximum cladding temperature, and the effects of the test on the fuel element are recorded. ( T.F.H.)

Pu-Zr alloy for high-temperature foil-type fuel

DOEpatents

McCuaig, Franklin D.

1977-01-01

A nuclear reactor fuel alloy consists essentially of from slightly greater than 7 to about 4 w/o zirconium, balance plutonium, and is characterized in that the alloy is castable and is rollable to thin foils. A preferred embodiment of about 7 w/o zirconium, balance plutonium, has a melting point substantially above the melting point of plutonium, is rollable to foils as thin as 0.0005 inch thick, and is compatible with cladding material when repeatedly cycled to temperatures above 650.degree. C. Neutron reflux densities across a reactor core can be determined with a high-temperature activation-measurement foil which consists of a fuel alloy foil core sandwiched and sealed between two cladding material jackets, the fuel alloy foil core being a 7 w/o zirconium, plutonium foil which is from 0.005 to 0.0005 inch thick.

Corrosion-resistant fuel cladding allow for liquid metal fast breeder reactors

DOEpatents

Brehm, Jr., William F.; Colburn, Richard P.

1982-01-01

An aluminide coating for a fuel cladding tube for LMFBRs (liquid metal fast breeder reactors) such as those using liquid sodium as a heat transfer agent. The coating comprises a mixture of nickel-aluminum intermetallic phases and presents good corrosion resistance to liquid sodium at temperatures up to 700.degree. C. while additionally presenting a barrier to outward diffusion of .sup.54 Mn.

Fuel pin cladding

DOEpatents

Vaidyanathan, Swaminathan; Adamson, Martyn G.

1986-01-01

An improved fuel pin cladding, particularly adapted for use in breeder reactors, consisting of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel and/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients.

Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

DOE PAGES

George, Nathan Michael; Terrani, Kurt A.; Powers, Jeffrey J.; ...

2014-09-29

A study analyzed the neutronics of alternate cladding materials in a pressurized water reactor (PWR) environment. Austenitic type 310 (310SS) and 304 stainless steels, ferritic Fe-20Cr-5Al (FeCrAl) and APMT™ alloys, and silicon carbide (SiC)-based materials were considered and compared with Zircaloy-4. SCALE 6.1 was used to analyze the associated neutronics penalty/advantage, changes in reactivity coefficients, and spectral variations once a transition in the cladding was made. In the cases examined, materials containing higher absorbing isotopes invoked a reduction in reactivity due to an increase in neutron absorption in the cladding. Higher absorbing materials produced a harder neutron spectrum in themore » fuel pellet, leading to a slight increase in plutonium production. A parametric study determined the geometric conditions required to match cycle length requirements for each alternate cladding material in a PWR. A method for estimating the end of cycle reactivity was implemented to compare each model to that of standard Zircaloy-4 cladding. By using a thinner cladding of 350 μm and keeping a constant outer diameter, austenitic stainless steels require an increase of no more than 0.5 wt% enriched 235U to match fuel cycle requirements, while the required increase for FeCrAl was about 0.1%. When modeling SiC (with slightly lower thermal absorption properties than that of Zircaloy), a standard cladding thickness could be implemented with marginally less enriched uranium (~0.1%). Moderator temperature and void coefficients were calculated throughout the depletion cycle. Nearly identical reactivity responses were found when coolant temperature and void properties were perturbed for each cladding material. By splitting the pellet into 10 equal areal sections, relative fission power as a function of radius was found to be similar for each cladding material. FeCrAl and 310SS cladding have a slightly higher fission power near the pellet’s periphery due to the harder neutron spectrum in the system, causing more 239Pu breeding. An economic assessment calculated the change in fuel pellet production costs for use of each cladding. Furthermore, implementing FeCrAl alloys would increase fuel pellet production costs about 15% because of increased 235U enrichment and the additional UO 2 pellet volume enabled by using thinner cladding.« less

MODELLING OF FUEL BEHAVIOUR DURING LOSS-OF-COOLANT ACCIDENTS USING THE BISON CODE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pastore, G.; Novascone, S. R.; Williamson, R. L.

2015-09-01

This work presents recent developments to extend the BISON code to enable fuel performance analysis during LOCAs. This newly developed capability accounts for the main physical phenomena involved, as well as the interactions among them and with the global fuel rod thermo-mechanical analysis. Specifically, new multiphysics models are incorporated in the code to describe (1) transient fission gas behaviour, (2) rapid steam-cladding oxidation, (3) Zircaloy solid-solid phase transition, (4) hydrogen generation and transport through the cladding, and (5) Zircaloy high-temperature non-linear mechanical behaviour and failure. Basic model characteristics are described, and a demonstration BISON analysis of a LWR fuel rodmore » undergoing a LOCA accident is presented. Also, as a first step of validation, the code with the new capability is applied to the simulation of experiments investigating cladding behaviour under LOCA conditions. The comparison of the results with the available experimental data of cladding failure due to burst is presented.« less

Welding fixture for nuclear fuel pin cladding assemblies

DOEpatents

Oakley, D.J.; Feld, S.H.

1984-02-22

A welding fixture is described for locating a driver sleeve about the open end of a nuclear fuel pin cladding. The welding fixture includes a holder provided with an open cavity having shoulders for properly positioning the driver sleeve, the end cap, and a soft, high temperature resistant plastic protective sleeve that surrounds a portion of the end cap stem. Ejected contaminant particles spewed forth by closure of the cladding by pulsed magnetic welding techniques are captured within a contamination trap formed in the holder for ultimate removal and disposal of contaminating particles along with the holder.

Unrestrained swelling of uranium-nitride fuel irradiated at temperatures ranging from 1100 to 1400 K (1980 to 2520 R)

NASA Technical Reports Server (NTRS)

Rohal, R. G.; Tambling, T. N.

1973-01-01

Six fuel pins were assembled, encapsulated, and irradiated in the Plum Brook Reactor. The fuel pins employed uranium mononitride (UN) in a stainless steel (type 304L) clad. The pins were irradiated for approximately 4000 hours to burnups of about 2.0 atom percent uranium. The average clad surface temperature during irradiation was about 1100 K (1980 deg R). Since stainless steel has a very low creep strength relative to that of UN at this temperature, these tests simulated unrestrained swelling of UN. The tests indicated that at 1 percent uranium atom burnup the unrestrained diametrical swelling of UN is about 0.5, 0.8, and 1.0 percent at 1223, 1264, and 1306 K (2200, deg 2273 deg, and 2350 deg R), respectively. The tests also indicated that the irradiation induced swelling of unrestrained UN fuel pellets appears to be isotropic.

Fuel pin cladding

DOEpatents

Vaidyanathan, S.; Adamson, M.G.

1986-01-28

Disclosed is an improved fuel pin cladding, particularly adapted for use in breeder reactors, consisting of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel and/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients. 2 figs.

Fuel pin cladding

DOEpatents

Vaidyanathan, S.; Adamson, M.G.

1983-12-16

An improved fuel pin cladding, particularly adapted for use in breeder reactors, is described which consist of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel an/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients.

Uranium nitride fuel fabrication for SP-100 reactors

NASA Technical Reports Server (NTRS)

Mason, Richard E.; Chidester, Kenneth M.; Hoth, Carl W.; Matthews, Bruce R.

1987-01-01

Fuel pins of uranium mononitride clad in Nb-1 percent Zr were fabricated for irradiation tests in EBR-II. Laboratory scale process parameters to synthesize UN powders and fabricate UN pellets were developed. Uranium mononitride was prepared by converting UO2 to UN. Fuel pellets were prepared by communition of UN briquettes, uniaxial pressing, and high temperature sintering. Techniques for machining, cleaning, and welding Nb-1 percent Zr cladding components were developed. End caps were electron beam welded to the tubing. Helium back-fill holes were sealed with a laser weld.

Uranium nitride fuel fabrication for SP-100 reactors

NASA Astrophysics Data System (ADS)

Mason, Richard E.; Chidester, Kenneth M.; Hoth, Carl W.; Matthews, Bruce R.

Fuel pins of uranium mononitride clad in Nb-1 percent Zr were fabricated for irradiation tests in EBR-II. Laboratory scale process parameters to synthesize UN powders and fabricate UN pellets were developed. Uranium mononitride was prepared by converting UO2 to UN. Fuel pellets were prepared by communition of UN briquettes, uniaxial pressing, and high temperature sintering. Techniques for machining, cleaning, and welding Nb-1 percent Zr cladding components were developed. End caps were electron beam welded to the tubing. Helium back-fill holes were sealed with a laser weld.

Design, fabrication, and operation of capsules for the irradiation testing of candidate advanced space reactor fuel pins

NASA Technical Reports Server (NTRS)

Thoms, K. R.

1975-01-01

Fuel irradiation experiments were designed, built, and operated to test uranium mononitride (UN) fuel clad in tungsten-lined T-111 and uranium dioxide fuel clad in both tungsten-lined T-111 and tungsten-lined Nb-1% Zr. A total of nine fuel pins was irradiated at average cladding temperatures ranging from 931 to 1015 C. The UN experiments, capsules UN-4 and -5, operated for 10,480 and 10,037 hr, respectively, at an average linear heat generation rate of 10 kW/ft. The UO2 experiment, capsule UN-6, operated for 8333 hr at an average linear heat generation rate of approximately 5 kW/ft. Following irradiation, the nine fuel pins were removed from their capsules, externally examined, and sent to the NASA Plum Brook Facility for more detailed postirradiation examination. During visual examination, it was discovered that the cladding of the fuel pin containing dense UN in each of capsules UN-4 and -5 had failed, exposing the UN fuel to the NaK in which the pins were submerged and permitting the release of fission gas from the failed pins. A rough analysis of the fission gas seen in samples of the gas in the fuel pin region indicated fission gas release-to-birth rates from these fuel pins in the range of .00001.

Analysis of pellet cladding interaction and creep of U 3SIi2 fuel for use in light water reactors

NASA Astrophysics Data System (ADS)

Metzger, Kathryn E.

Following the accident at the Fukushima plant, enhancing the accident tolerance of the light water reactor (LWR) fleet became a topic of serious discussion. Under the direction of congress, the DOE office of Nuclear Energy added accident tolerant fuel development as a primary component to the existing Advanced Fuels Program. The DOE defines accident tolerant fuels as fuels that "in comparison with the standard UO2- Zircaloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, as well as design-basis and beyond design-basis events." To be economically viable, proposed accident tolerant fuels and claddings should be backward compatible with LWR designs, provide significant operating cost improvements such as power uprates, increased fuel burnup, or increased cycle length. In terms of safety, an alternative fuel pellet must have resistance to water corrosion comparable to UO2, thermal conductivity equal to or larger than that of UO2, and a melting temperature that allows the material to remain solid under power reactor conditions. Among the candidates, U3Si2 has a number of advantageous thermophysical properties, including; high density, high thermal conductivity at room temperature, and a high melting temperature. These properties support its use as an accident tolerant fuel while its high uranium density is capable of supporting uprates to the LWR fleet. This research characterizes U3Si2 pellets and analyzes U3Si2 under light water reactor conditions using the fuel performance code BISON. While some thermophysical properties for U3Si2 have been found in the literature, the irradiation behavior is sparse and limited to experience with dispersion fuels. Accordingly, the creep behavior for U3Si2 has been unknown, making it difficult to predict fuel-cladding mechanical behavior. This information is essential for designing accident tolerant fuel systems where ceramic claddings, like silicon carbide (SiC) are proposed. This research provides a model for both the thermal and irradiation creep behavior for U3Si2. This body of research is comprised of both experimental and modeling components. Characterization of the fuel microstructure includes; optical microscopy with pore and grain size analysis, helium pycnometry for density determination, mercury intrusion porosimetry, compositional analysis in the form of XRD, second phase identification using EDX, electrical resistance measurement via four point probe, determination of hardness and toughness through Vickers indentation testing, and determination of elastic properties using the impulse excitation method. Post-sintering grain size data allowed for the determination of grain boundary activation energy and diffusion coefficients, which were used to develop creep models. This was extended to lattice and irradiation enhanced diffusion in order to develop a U3Si2 creep model over thermal and irradiation creep regimes. In addition to the creep model, thermal and swelling behavior models for U3Si2 were implemented into the BISON fuel performance code. A series of simulations evaluated the performance and behavior of U3Si2 under typical light water reactor conditions with advanced SiC ceramic cladding. Simulation results show that fuel creep relieves stress in the ceramic cladding and postpones the. moment of fuel-clad contact. However, the stress reduction to the cladding is minimal because the fuel creep rate is low while the swelling rate is high. Future work should include the investigation of monolithic U3Si2 irradiation swelling since the current model relies upon the swelling data of U3Si2 particles in a metallic dispersion fuel. Additionally, planned thermal creep testing at the University of South Carolina can provide confirmation of the U3Si2 creep model contained herein.

A pulse-controlled modified-burst test instrument for accident-tolerant fuel cladding

DOE PAGES

Cinbiz, M. Nedim; Brown, Nicholas R.; Terrani, Kurt A.; ...

2017-06-03

Pellet-cladding mechanical interaction due to thermal expansion of nuclear fuel pellets during a reactivity-initiated accident (RIA) is a potential mechanism for failure of nuclear fuel cladding. To investigate the mechanical behavior of cladding during an RIA, we developed a mechanical pulse-controlled modified burst test instrument that simulates transient events with a pulse width from 10 to 300 ms. This paper includes validation tests of unirradiated and prehydrided ZIRLO cladding tubes. A ZIRLO cladding sample with a hydrogen content of 168 wt. ppm showed ductile behavior and failed at the maximum limits of the test setup with hoop strain to failuremore » greater than 9.2%. ZIRLO samples showed high resistance to failure even at very high hydrogen contents (1,466 wt. ppm). When the hydrogen content was increased to 1,554 wt. ppm, brittle-like behavior was observed at a hoop strain of 2.5%. Preliminary scoping tests at room temperature with FeCrAl tubes were conducted to imitate the pulse behavior of transient test reactors during integral tests. The preliminary FeCrAl tests are informative from the perspective of characterizing the test rig and supporting the design of integral tests for current and potentially accident tolerant cladding materials.« less

A pulse-controlled modified-burst test instrument for accident-tolerant fuel cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cinbiz, M. Nedim; Brown, Nicholas R.; Terrani, Kurt A.

Pellet-cladding mechanical interaction due to thermal expansion of nuclear fuel pellets during a reactivity-initiated accident (RIA) is a potential mechanism for failure of nuclear fuel cladding. To investigate the mechanical behavior of cladding during an RIA, we developed a mechanical pulse-controlled modified burst test instrument that simulates transient events with a pulse width from 10 to 300 ms. This paper includes validation tests of unirradiated and prehydrided ZIRLO cladding tubes. A ZIRLO cladding sample with a hydrogen content of 168 wt. ppm showed ductile behavior and failed at the maximum limits of the test setup with hoop strain to failuremore » greater than 9.2%. ZIRLO samples showed high resistance to failure even at very high hydrogen contents (1,466 wt. ppm). When the hydrogen content was increased to 1,554 wt. ppm, brittle-like behavior was observed at a hoop strain of 2.5%. Preliminary scoping tests at room temperature with FeCrAl tubes were conducted to imitate the pulse behavior of transient test reactors during integral tests. The preliminary FeCrAl tests are informative from the perspective of characterizing the test rig and supporting the design of integral tests for current and potentially accident tolerant cladding materials.« less

Pu-ZR Alloy high-temperature activation-measurement foil

DOEpatents

McCuaig, Franklin D.

1977-08-02

A nuclear reactor fuel alloy consists essentially of from slightly greater than 7 to about 4 w/o zirconium, balance plutonium, and is characterized in that the alloy is castable and is rollable to thin foils. A preferred embodiment of about 7 w/o zirconium, balance plutonium, has a melting point substantially above the melting point of plutonium, is rollable to foils as thin as 0.0005 inch thick, and is compatible with cladding material when repeatedly cycled to temperatures above 650.degree. C. Neutron flux densities across a reactor core can be determined with a high-temperature activation-measurement foil which consists of a fuel alloy foil core sandwiched and sealed between two cladding material jackets, the fuel alloy foil core being a 7 w/o zirconium, plutonium foil which is from 0.005 to 0.0005 inch thick.

Validating the BISON fuel performance code to integral LWR experiments

DOE PAGES

Williamson, R. L.; Gamble, K. A.; Perez, D. M.; ...

2016-03-24

BISON is a modern finite element-based nuclear fuel performance code that has been under development at the Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON’s computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to datemore » for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Our results demonstrate that 1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, 2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and 3) comparison of rod diameter results indicates a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. In the initial rod diameter comparisons they were unsatisfactory and have lead to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.« less

Surface modification techniques for increased corrosion tolerance of zirconium fuel cladding

NASA Astrophysics Data System (ADS)

Carr, James Patrick, IV

Corrosion is a major issue in applications involving materials in normal and severe environments, especially when it involves corrosive fluids, high temperatures, and radiation. Left unaddressed, corrosion can lead to catastrophic failures, resulting in economic and environmental liabilities. In nuclear applications, where metals and alloys, such as steel and zirconium, are extensively employed inside and outside of the nuclear reactor, corrosion accelerated by high temperatures, neutron radiation, and corrosive atmospheres, corrosion becomes even more concerning. The objectives of this research are to study and develop surface modification techniques to protect zirconium cladding by the incorporation of a specific barrier coating, and to understand the issues related to the compatibility of the coatings examined in this work. The final goal of this study is to recommend a coating and process that can be scaled-up for the consideration of manufacturing and economic limits. This dissertation study builds on previous accident tolerant fuel cladding research, but is unique in that advanced corrosion methods are tested and considerations for implementation by industry are practiced and discussed. This work will introduce unique studies involving the materials and methods for accident tolerant fuel cladding research by developing, demonstrating, and considering materials and processes for modifying the surface of zircaloy fuel cladding. This innovative research suggests that improvements in the technique to modify the surface of zirconium fuel cladding are likely. Three elements selected for the investigation of their compatibility on zircaloy fuel cladding are aluminum, silicon, and chromium. These materials are also currently being investigated at other labs as alternate alloys and coatings for accident tolerant fuel cladding. This dissertation also investigates the compatibility of these three elements as surface modifiers, by comparing their microstructural and mechanical properties. To test their application for use in corrosive atmospheres, the corrosion behaviors are also compared in steam, water, and boric-acid environments. Various methods of surface modification were attempted in this investigation, including dip coating, diffusion bonding, casting, sputtering, and evaporation. The benefits and drawbacks of each method are discussed with respect to manufacturing and economic limits. Characterization techniques utilized in this work include optical microscopy, scanning electron microscopy, energy-dispersive spectroscopy, X-ray diffraction, nanoindentation, adhesion testing, and atomic force microscopy. The composition, microstructure, hardness, modulus, and coating adhesion were studied to provide encompassing properties to determine suitable comparisons and to choose an ideal method to scale to industrial applications. The experiments, results, and detailed discussions are presented in the following chapters of this dissertation research.

Multiphysics modeling of two-phase film boiling within porous corrosion deposits

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jin, Miaomiao, E-mail: mmjin@mit.edu; Short, Michael, E-mail: hereiam@mit.edu

2016-07-01

Porous corrosion deposits on nuclear fuel cladding, known as CRUD, can cause multiple operational problems in light water reactors (LWRs). CRUD can cause accelerated corrosion of the fuel cladding, increase radiation fields and hence greater exposure risk to plant workers once activated, and induce a downward axial power shift causing an imbalance in core power distribution. In order to facilitate a better understanding of CRUD's effects, such as localized high cladding surface temperatures related to accelerated corrosion rates, we describe an improved, fully-coupled, multiphysics model to simulate heat transfer, chemical reactions and transport, and two-phase fluid flow within these deposits.more » Our new model features a reformed assumption of 2D, two-phase film boiling within the CRUD, correcting earlier models' assumptions of single-phase coolant flow with wick boiling under high heat fluxes. This model helps to better explain observed experimental values of the effective CRUD thermal conductivity. Finally, we propose a more complete set of boiling regimes, or a more detailed mechanism, to explain recent CRUD deposition experiments by suggesting the new concept of double dryout specifically in thick porous media with boiling chimneys. - Highlights: • A two-phase model of CRUD's effects on fuel cladding is developed and improved. • This model eliminates the formerly erroneous assumption of wick boiling. • Higher fuel cladding temperatures are predicted when accounting for two-phase flow. • Double-peaks in thermal conductivity vs. heat flux in experiments are explained. • A “double dryout” mechanism in CRUD is proposed based on the model and experiments.« less

Nanocrystalline diamond protects Zr cladding surface against oxygen and hydrogen uptake: Nuclear fuel durability enhancement.

PubMed

Škarohlíd, Jan; Ashcheulov, Petr; Škoda, Radek; Taylor, Andrew; Čtvrtlík, Radim; Tomáštík, Jan; Fendrych, František; Kopeček, Jaromír; Cháb, Vladimír; Cichoň, Stanislav; Sajdl, Petr; Macák, Jan; Xu, Peng; Partezana, Jonna M; Lorinčík, Jan; Prehradná, Jana; Steinbrück, Martin; Kratochvílová, Irena

2017-07-25

In this work, we demonstrate and describe an effective method of protecting zirconium fuel cladding against oxygen and hydrogen uptake at both accident and working temperatures in water-cooled nuclear reactor environments. Zr alloy samples were coated with nanocrystalline diamond (NCD) layers of different thicknesses, grown in a microwave plasma chemical vapor deposition apparatus. In addition to showing that such an NCD layer prevents the Zr alloy from directly interacting with water, we show that carbon released from the NCD film enters the underlying Zr material and changes its properties, such that uptake of oxygen and hydrogen is significantly decreased. After 100-170 days of exposure to hot water at 360 °C, the oxidation of the NCD-coated Zr plates was typically decreased by 40%. Protective NCD layers may prolong the lifetime of nuclear cladding and consequently enhance nuclear fuel burnup. NCD may also serve as a passive element for nuclear safety. NCD-coated ZIRLO claddings have been selected as a candidate for Accident Tolerant Fuel in commercially operated reactors in 2020.

Assessment of Silicon Carbide Composites for Advanced Salt-Cooled Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

2007-09-01

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

Axisymmetric whole pin life modelling of advanced gas-cooled reactor nuclear fuel

NASA Astrophysics Data System (ADS)

Mella, R.; Wenman, M. R.

2013-06-01

Thermo-mechanical contributions to pellet-clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS's well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used. The usability of a FE based fuel performance code would be an enhancement over past codes. Pre- and post-processors have lowered the entry barrier for the development of a fuel performance model to permit the ability to model complicated systems. Typical runtimes for a 5 year axisymmetric model takes less than one hour on a single core workstation. The current model has implemented: Non-linear fuel thermal behaviour, including a complex description of heat flow in the fuel. Coupled with a variety of different FE and finite difference models. Non-linear mechanical behaviour of the fuel and cladding including, fuel creep and swelling and cladding creep and plasticity each with dependencies on a variety of different properties. A fission gas release model which takes inputs from first principles calculations. Explicitly integrated inventory calculations performed in a coupled manner. Freedom to model steady state and transient behaviour using implicit time integration. The whole pin geometry is considered over an entire typical fuel life. The model showed by examination of normal operation and a subsequent transient chosen for software demonstration purposes: ABAQUS may be a sufficiently flexible platform to develop a complete and verified fuel performance code. The importance and effectiveness of the geometry of the fuel spacer pellets was characterised. The fuels performance under normal conditions (high friction no power spikes) would not suggest serious degradation of the cladding in fuel life. Large plastic strains were found when pellet bonding was strong, these would appear at all pellets cladding triple points and all pellet radial crack and cladding interfaces thus showing a possible axial direction to cracks forming from ductility exhaustion.

Behavior of an improved Zr fuel cladding with oxidation resistant coating under loss-of-coolant accident conditions

NASA Astrophysics Data System (ADS)

Park, Dong Jun; Kim, Hyun Gil; Jung, Yang Il; Park, Jung Hwan; Yang, Jae Ho; Koo, Yang Hyun

2016-12-01

This study investigates protective coatings for improving the high temperature oxidation resistance of Zr fuel claddings for light water nuclear reactors. FeCrAl alloy and Cr layers were deposited onto Zr plates and tubes using cold spraying. For the FeCrAl/Zr system, a Mo layer was introduced between the FeCrAl coating and the Zr matrix to prevent inter-diffusion at high temperatures. Both the FeCrAl and Cr coatings improved the oxidation resistance compared to that of the uncoated Zr alloy when exposed to a steam environment at 1200 °C. The ballooning behavior and mechanical properties of the coated cladding samples were studied under simulated loss-of-coolant accident conditions. The coated samples showed higher burst temperatures, lower circumferential strain, and smaller rupture openings compared to the uncoated Zr. Although 4-point bend tests of the coated samples showed a small increase in the maximum load, ring compression tests of a sectioned sample showed increased ductility.

Modeling of laser cladding with application to fuel cell manufacturing.

DOT National Transportation Integrated Search

2010-01-01

Polymer electrolyte membrane (PEM) fuel cells have many advantages such as compactness, : lightweight, high power density, low temperature operation and near zero emissions. Although : many research organizations have intensified their efforts toward...

Nuclear fuel elements having a composite cladding

DOEpatents

Gordon, Gerald M.; Cowan, II, Robert L.; Davies, John H.

1983-09-20

An improved nuclear fuel element is disclosed for use in the core of nuclear reactors. The improved nuclear fuel element has a composite cladding of an outer portion forming a substrate having on the inside surface a metal layer selected from the group consisting of copper, nickel, iron and alloys of the foregoing with a gap between the composite cladding and the core of nuclear fuel. The nuclear fuel element comprises a container of the elongated composite cladding, a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container and a nuclear fuel material retaining means positioned in the cavity. The metal layer of the composite cladding prevents perforations or failures in the cladding substrate from stress corrosion cracking or from fuel pellet-cladding interaction or both. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy.

FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Davidson, J.K.

1963-11-19

A fuel element structure particularly useful in high temperature nuclear reactors is presented. Basically, the structure comprises two coaxial graphite sleeves integrally joined together by radial fins. Due to the high structural strength of graphite at high temperatures and the rigidity of this structure, nuclear fuel encased within the inner sleeve in contiguous relation therewith is supported and prevented from expanding radially at high temperatures. Thus, the necessity of relying on the usual cladding materials with relatively low temperature limitations for structural strength is removed. (AEC)

Anisotropic Azimuthal Power and Temperature distribution on FuelRod. Impact on Hydride Distribution

DOE Office of Scientific and Technical Information (OSTI.GOV)

Motta, Arthur; Ivanov, Kostadin; Arramova, Maria

2015-04-29

The degradation of the zirconium cladding may limit nuclear fuel performance. In the high temperature environment of a reactor, the zirconium in the cladding corrodes, releasing hydrogen in the process. Some of this hydrogen is absorbed by the cladding in a highly inhomogeneous manner. The distribution of the absorbed hydrogen is extremely sensitive to temperature and stress concentration gradients. The absorbed hydrogen tends to concentrate near lower temperatures. This hydrogen absorption and hydride formation can cause cladding failure. This project set out to improve the hydrogen distribution prediction capabilities of the BISON fuel performance code. The project was split intomore » two primary sections, first was the use of a high fidelity multi-physics coupling to accurately predict temperature gradients as a function of r, θ , and z, and the second was to use experimental data to create an analytical hydrogen precipitation model. The Penn State version of thermal hydraulics code COBRA-TF (CTF) was successfully coupled to the DeCART neutronics code. This coupled system was verified by testing and validated by comparison to FRAPCON data. The hydrogen diffusion and precipitation experiments successfully calculated the heat of transport and precipitation rate constant values to be used within the hydrogen model in BISON. These values can only be determined experimentally. These values were successfully implemented in precipitation, diffusion and dissolution kernels that were implemented in the BISON code. The coupled output was fed into BISON models and the hydrogen and hydride distributions behaved as expected. Simulations were conducted in the radial, axial and azimuthal directions to showcase the full capabilities of the hydrogen model.« less

Large-break LOCA, in-reactor fuel bundle Materials Test MT-6A

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wilson, C.L.; Hesson, G.M.; Pilger, J.P.

1993-09-01

This is a report on one of a series of experiments to simulates a loss-of-coolant accident (LOCA) using full-length fuel rods for pressurized water reactors (PWR). The experiments were conducted by Pacific Northwest Laboratory (PNL) under the LOCA simulation Program sponsored by the US Nuclear Regulatory Commission (NRC). The major objective of this program was causing the maximum possible expansion of the cladding on the fuel rods from a short-term adiabatic temperature transient to 1200 K (1700 F) leading to the rupture of the cladding; and second, by reflooding the fuel rods to determine the rate at which the fuelmore » bundle is cooled.« less

Screening of advanced cladding materials and UN-U3Si5 fuel

NASA Astrophysics Data System (ADS)

Brown, Nicholas R.; Todosow, Michael; Cuadra, Arantxa

2015-07-01

In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO2) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO2 fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO2-Zr fuel-cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN-U3Si5 fuels with Kanthal AF or APMT cladding. The objective of the U3Si5 phase in the UN-U3Si5 fuel concept is to shield the nitride phase from water. It was shown that UN-U3Si5 fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO2-Zr fuel-cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to 14N content in UN ceramic composites is high. Analysis of the rim effect due to self-shielding in the fuel shows that the UN-based ceramic fuels are not expected to have significantly different relative burn-up distributions at discharge relative to the UO2 reference fuel. However, the overall harder spectrum in the UN ceramic composite fuels increases transuranic build-up, which will increase long-term activity in a once-thru fuel cycle but is expected to be a significant advantage in a fuel cycle with continuous recycling of transuranic material. It is recognized that the fuel and cladding properties assumed in these assessments are preliminary, and that additional data are necessary for these materials, most significantly under irradiation.

75 FR 3876 - Mark Edward Leyse; Receipt of Petition for Rulemaking

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-25

... (assembly) severe fuel damage experiments. The petitioner also requests that the NRC promulgate a regulation... aware that data from multi-rod (assembly) severe fuel damage experiments indicates that the current... fuel damage experiments indicates that the current peak cladding temperature limit contained in 10 CFR...

The R&D PERFROI Project on Thermal Mechanical and Thermal Hydraulics Behaviors of a Fuel Rod Assembly during a Loss of Coolant Accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Repetto, G.; Dominguez, C.; Durville, B.

The safety principle in case of a LOCA is to preserve the short and long term coolability of the core. The associated safety requirements are to ensure the resistance of the fuel rods upon quench and post-quench loads and to maintain a coolable geometry in the core. An R&D program has been launched by IRSN with the support of EDF, to perform both experimental and modeling activities in the frame of the LOCA transient, on technical issues such as: - flow blockage within a fuel rods bundle and its potential impact on coolability, - fuel fragment relocation in the balloonedmore » areas: its potential impact on cladding PCT (Peak Cladding Temperature) and on the maximum oxidation rate, - potential loss of cladding integrity upon quench and post-quench loads. The PERFROI project (2014-2019) focusing on the first above issue, is structured in two axes: 1. axis 1: thermal mechanical behavior of deformation and rupture of cladding taking into account the contact between fuel rods; specific research at LaMCoS laboratory focus on the hydrogen behavior in cladding alloys and its impact on the mechanical behavior of the rod; and, 2. axis 2: thermal hydraulics study of a partially blocked region of the core (ballooned area taking into account the fuel relocation with local over power), during cooling phase by water injection; More detailed activities foreseen in collaboration with LEMTA laboratory will focus on the characterization of two phase flows with heat transfer in deformed structures.« less

Severe Accident Test Station Activity Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pint, Bruce A.; Terrani, Kurt A.

2015-06-01

Enhancing safety margins in light water reactor (LWR) severe accidents is currently the focus of a number of international R&D programs. The current UO2/Zr-based alloy fuel system is particularly susceptible since the Zr-based cladding experiences rapid oxidation kinetics in steam at elevated temperatures. Therefore, alternative cladding materials that offer slower oxidation kinetics and a smaller enthalpy of oxidation can significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. In the U.S. program, the high temperature steam oxidation performance of accident tolerant fuel (ATF) cladding solutions has been evaluated in the Severe Accidentmore » Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012. This report summarizes the capabilities of the SATS and provides an overview of the oxidation kinetics of several candidate cladding materials. A suggested baseline for evaluating ATF candidates is a two order of magnitude reduction in the steam oxidation resistance above 1000ºC compared to Zr-based alloys. The ATF candidates are categorized based on the protective external oxide or scale that forms during exposure to steam at high temperature: chromia, alumina, and silica. Comparisons are made to literature and SATS data for Zr-based alloys and other less-protective materials.« less

Sensitivity analysis of FeCrAl cladding and U3Si2 fuel under accident conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gamble, Kyle Allan Lawrence; Hales, Jason Dean

2016-08-01

The purpose of this milestone report is to highlight the results of sensitivity analyses performed on two accident tol- erant fuel concepts: U3Si2 fuel and FeCrAl cladding. The BISON fuel performance code under development at Idaho National Laboratory was coupled to Sandia National Laboratories’ DAKOTA software to perform the sensitivity analyses. Both Loss of Coolant (LOCA) and Station blackout (SBO) scenarios were analyzed using main effects studies. The results indicate that for FeCrAl cladding the input parameters with greatest influence on the output metrics of interest (fuel centerline temperature and cladding hoop strain) during the LOCA were the isotropic swellingmore » and fuel enrichment. For U3Si2 the important inputs were found to be the intergranular diffusion coefficient, specific heat, and fuel thermal conductivity. For the SBO scenario, Young’s modulus was found to be influential in FeCrAl in addition to the isotropic swelling and fuel enrichment. Contrarily to the LOCA case, the specific heat of U3Si2 was found to have no effect during the SBO. The intergranular diffusion coefficient and fuel thermal conductivity were still found to be of importance. The results of the sensitivity analyses have identified areas where further research is required including fission gas behavior in U3Si2 and irradiation swelling in FeCrAl. Moreover, the results highlight the need to perform the sensitivity analyses on full length fuel rods for SBO scenarios.« less

Implementation and evaluation of fuel creep using advanced light-water reactor materials in FRAPCON 3.5

NASA Astrophysics Data System (ADS)

Carroll, Spencer

As current reactors approach the end of their operable lifetime, new reactors are needed if nuclear power is to continue being generated in the United States. Some utilities have already began construction on newer, more advanced LWR reactors, which use the same fuel as current reactors and have a similar but updated design. Others are researching next generation (GEN-IV) reactors which have new designs that utilize alternative fuel, coolants and other reactor materials. Many of these alternative fuels are capable of achieving higher burnups and are designed to be more accident tolerant than the currently used UO2 fuel. However, before these new materials can be used, extensive research must be done in order to obtain a detailed understanding of how the new fuels and other materials will interact. New fuels, such as uranium nitride (UN) and uranium carbide (UC) have several advantages over UO2, such as increased burnup capabilities and higher thermal conductivities. However, there are issues with each that prevent UC and UN from being used as direct replacements for UO2. Both UC and UN swell at a significantly higher rate than UO2 and neither fuel reacts favorably when exposed to water. Due to this, UC and UN are being considered more for GEN-IV reactors that use alternative coolant rather than for current LWRs. In an effort to increase accident tolerance, silicon carbide (SiC) is being considered for use as an alternative cladding. The high strength, high melting point and low oxidation of SiC make it an attractive cladding choice, especially in an accident scenario. However, as a ceramic, SiC is not ductile and will not creep outwards upon pellet-clad mechanical interaction (PCMI) which can cause a large build up in interfacial pressure. In order to understand the interaction between the high swelling fuels and unyielding SiC cladding, data on the properties and behaviors of these materials must be gathered and incorporated into FRAPCON. FRAPCON is a fuel performance code developed by PNNL and used by the Nuclear Regulatory Commission (NRC) as a licensing code for US reactors. FRAPCON will give insight into how these new fuel-cladding combinations will affect cladding hoop stress and help determine if the new materials are feasible for use in a reactor. To accurately simulate the interaction between the new materials, a soft pellet model that allows for stresses on the pellet to affect pellet deformation will have to be implemented. Currently, FRAPCON uses a rigid pellet model that does not allow for feedback of the cladding onto the pellet. Since SiC does not creep at the temperatures being considered and is not ductile, any PCMI create a much higher interfacial pressure than is possible with Zircaloy. Because of this, it is necessary to implement a model that allows for pellet creep to alleviate some of these cladding stresses. These results will then be compared to FEMAXI-6, a Japanese fuel performance code that already calculates pellet stress and allows for cladding feedback onto the pellet. This research is intended to be a continuation and verification of previous work done by USC on the analysis of accident tolerant fuels with alternative claddings and is intended to prove that a soft pellet model is necessary to accurately model any fuel with SiC cladding.

Status Report on Irradiation Capsules Designed to Evaluate FeCrAl-UO 2 Interactions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G.; Howard, Richard H.

This status report provides the background and current status of a series of irradiation capsules that were designed and are being built to test the interactions between candidate FeCrAl cladding for enhanced accident tolerant applications and prototypical enriched commercial UO 2 fuel in a neutron radiation environment. These capsules will test the degree, if any, of fuel cladding chemical interactions (FCCI) between FeCrAl and UO 2. The capsules are to be irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory to burn-ups of 10, 30, and 50 GWd/MT with a nominal target temperature at the interfaces between themore » pellets and clad of 350°C.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, C. S.; Zhang, Hongbin

Uncertainty quantification and sensitivity analysis are important for nuclear reactor safety design and analysis. A 2x2 fuel assembly core design was developed and simulated by the Virtual Environment for Reactor Applications, Core Simulator (VERA-CS) coupled neutronics and thermal-hydraulics code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). An approach to uncertainty quantification and sensitivity analysis with VERA-CS was developed and a new toolkit was created to perform uncertainty quantification and sensitivity analysis with fourteen uncertain input parameters. Furthermore, the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel center-line temperature, and maximum outer clad surfacemore » temperature were chosen as the selected figures of merit. Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis and coolant inlet temperature was consistently the most influential parameter. We used parameters as inputs to the critical heat flux calculation with the W-3 correlation were shown to be the most influential on the MDNBR, maximum fuel center-line temperature, and maximum outer clad surface temperature.« less

Uncertainty quantification and sensitivity analysis with CASL Core Simulator VERA-CS

DOE PAGES

Brown, C. S.; Zhang, Hongbin

2016-05-24

Uncertainty quantification and sensitivity analysis are important for nuclear reactor safety design and analysis. A 2x2 fuel assembly core design was developed and simulated by the Virtual Environment for Reactor Applications, Core Simulator (VERA-CS) coupled neutronics and thermal-hydraulics code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). An approach to uncertainty quantification and sensitivity analysis with VERA-CS was developed and a new toolkit was created to perform uncertainty quantification and sensitivity analysis with fourteen uncertain input parameters. Furthermore, the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel center-line temperature, and maximum outer clad surfacemore » temperature were chosen as the selected figures of merit. Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis and coolant inlet temperature was consistently the most influential parameter. We used parameters as inputs to the critical heat flux calculation with the W-3 correlation were shown to be the most influential on the MDNBR, maximum fuel center-line temperature, and maximum outer clad surface temperature.« less

Development of data base with mechanical properties of un- and pre-irradiated VVER cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Asmolov, V.; Yegorova, L.; Kaplar, E.

1998-03-01

Analysis of recent RIA test with PWR and VVER high burnup fuel, performed at CABRI, NSRR, IGR reactors has shown that the data base with mechanical properties of the preirradiated cladding is necessary to interpret the obtained results. During 1997 the corresponding cycle of investigations for VVER clad material was performed by specialists of NSI RRC KI and RIAR in cooperation with NRC (USA), IPSN (France) in two directions: measurements of mechanical properties of Zr-1%Nb preirradiated cladding versus temperature and strain rate; measurements of failure parameters for gas pressurized cladding tubes. Preliminary results of these investigations are presented in thismore » paper.« less

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

NASA Astrophysics Data System (ADS)

Karahan, Aydın; Buongiorno, Jacopo

2010-01-01

An engineering code to model the irradiation behavior of UO2-PuO2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

An Innovative Accident Tolerant LWR Fuel Rod Design Based on Uranium-Molybdenum Metal Alloy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Montgomery, Robert O.; Bennett, Wendy D.; Henager, Charles H.

2016-09-12

The US Department of Energy is developing a uranium-molybdenum metal alloy Enhanced Accident Tolerant Fuel concept for Light Water Reactor applications that provides improved fuel performance during normal operation, anticipated operational occurrences, and postulated accidents. The high initial uranium atom density, the high thermal conductivity, and a low heat capacity permit a U-Mo-based fuel assembly to meet important design and safety requirements. These attributes also result in a fuel design that can satisfy increased fuel utilization demands and allow for improved accident tolerance in LWRs. This paper summarizes the results obtained from the on-going activities to; 1) evaluate the impactmore » of the U-10wt%Mo thermal properties on operational and accident safety margins, 2) produce a triple extrusion of stainless steel cladding/niobium liner/U-10Mo fuel rod specimen and 3) test the high temperature water corrosion of rodlet samples containing a drilled hole in the cladding. Characterization of the cladding and liner thickness uniformity, microstructural features of the U-Mo gamma phase, and the metallurgical bond between the component materials will be presented. The results from corrosion testing will be discussed which yield insights into the resistance to attack by water ingress during high temperature water exposure for the triple extruded samples containing a drilled hole. These preliminary evaluations find that the U-10Mo fuel design concept has many beneficial features that can meet or improve conventional LWR fuel performance requirements under normal operation, AOOs, and postulated accidents. The viability of a deployable U-Mo fuel design hinges on demonstrating that fabrication processes and alloying additions can produce acceptable irradiation stability during normal operation and accident conditions and controlled metal-water reaction rates in the unlikely event of a cladding perforation. In the area of enhanced accident tolerance, a key objective is to establish that the lower stored energy of the U-Mo fuel design can provide the emergency core cooling systems the opportunity to maintain the reactor core in a coolable geometry following an accident.« less

U.S. Department of Energy Accident Resistant SiC Clad Nuclear Fuel Development

DOE Office of Scientific and Technical Information (OSTI.GOV)

George W. Griffith

2011-10-01

A significant effort is being placed on silicon carbide ceramic matrix composite (SiC CMC) nuclear fuel cladding by Light Water Reactor Sustainability (LWRS) Advanced Light Water Reactor Nuclear Fuels Pathway. The intent of this work is to invest in a high-risk, high-reward technology that can be introduced in a relatively short time. The LWRS goal is to demonstrate successful advanced fuels technology that suitable for commercial development to support nuclear relicensing. Ceramic matrix composites are an established non-nuclear technology that utilizes ceramic fibers embedded in a ceramic matrix. A thin interfacial layer between the fibers and the matrix allows formore » ductile behavior. The SiC CMC has relatively high strength at high reactor accident temperatures when compared to metallic cladding. SiC also has a very low chemical reactivity and doesn't react exothermically with the reactor cooling water. The radiation behavior of SiC has also been studied extensively as structural fusion system components. The SiC CMC technology is in the early stages of development and will need to mature before confidence in the developed designs can created. The advanced SiC CMC materials do offer the potential for greatly improved safety because of their high temperature strength, chemical stability and reduced hydrogen generation.« less

ANALYSIS AND EXAMINATION OF MOX FUEL FROM NONPROLIFERATION PROGRAMS

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCoy, Kevin; Machut, Dr McLean; Morris, Robert Noel

The U.S. Department of Energy has decided to dispose of a portion of the nation s surplus plutonium by reconstituting it into mixed oxide (MOX) fuel and irradiating it in commercial power reactors. Four lead assemblies were manufactured and irradiated to a maximum fuel rod burnup of 47.3 MWd/kg heavy metal. This was the first commercial irradiation of MOX fuel with a 240Pu/239Pu ratio of less than 0.10. Five fuel rods with varying burnups and plutonium contents were selected from one of the assemblies and shipped to Oak Ridge National Laboratory for hot cell examination. The performance of the rodsmore » was analyzed with AREVA s next-generation GALILEO code. The results of the analysis confirmed that the fuel rods had performed safely and predictably, and that GALILEO is applicable to MOX fuel with a low 240Pu/239Pu ratio as well as to standard MOX. The results are presented and compared to the GALILEO database. In addition, the fuel cladding was tested to confirm that traces of gallium in the fuel pellets had not affected the mechanical properties of the cladding. The irradiated cladding was found to remain ductile at both room temperature and 350 C for both the axial and circumferential directions.« less

Studies of thermionic materials for space power applications

NASA Technical Reports Server (NTRS)

1971-01-01

Service life tests of LC-8 and LC-9 carbide-fueled thermionic converters are discussed. Post operational tests of the converters to show emitter diametric change, microstructures of cladding and fuel, and analysis of fuel composition are described. The fabrication and performance of high temperature thermocouples used in the test procedures are included.

Effect of Fuel Temperature Profile on Eigenvalue Calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Greifenkamp, Tom E; Clarno, Kevin T; Gehin, Jess C

2008-01-01

Use of an average fuel temperature is a current practice when modeling fuel for eigenvalue (k-inf) calculations. This is an approximation, as it is known from Heat-transfer methods that a fuel pin having linear power q', will have a temperature that varies radially and has a maximum temperature at the center line [1]. This paper describes an investigation into the effects on k-inf and isotopic concentrations of modeling a fuel pin using a single average temperature versus a radially varying fuel temperature profile. The axial variation is not discussed in this paper. A single fuel pin was modeled having 1,more » 3, 5, 8, or 10 regions of equal volumes (areas). Fig. 1 shows a model of a 10-ring fuel pin surrounded by a gap and then cladding.« less

Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

DOE PAGES

Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; ...

2015-09-03

Low-enrichment (U-235 < 20%) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing was comprised of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates that were tested inmore » INL's Advanced Test Reactor (ATR) were subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. Adjacent to the AA6061 cladding were Mg-rich precipitates, which was in close proximity to the region where Xe is observed to be enriched. In samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface were possible indications of porosity/debonding, which suggested that the interface in this location is relatively weak.« less

Nuclear reactor fuel element having improved heat transfer

DOEpatents

Garnier, J.E.; Begej, S.; Williford, R.E.; Christensen, J.A.

1982-03-03

A nuclear reactor fuel element having improved heat transfer between fuel material and cladding is described. The element consists of an outer cladding tube divided into an upper fuel section containing a central core of fissionable or mixed fissionable and fertile fuel material, slightly smaller in diameter than the inner surface of the cladding tube and a small lower accumulator section, the cladding tube being which is filled with a low molecular weight gas to transfer heat from fuel material to cladding during irradiation. A plurality of essentially vertical grooves in the fuel section extend downward and communicate with the accumulator section. The radial depth of the grooves is sufficient to provide a thermal gradient between the hot fuel surface and the relatively cooler cladding surface to allow thermal segregation to take place between the low molecular weight heat transfer gas and high molecular weight fission product gases produced by the fuel material during irradiation.

Compatibility studies on Mo-coating systems for nuclear fuel cladding applications

NASA Astrophysics Data System (ADS)

Koh, Huan Chin; Hosemann, Peter; Glaeser, Andreas M.; Cionea, Cristian

2017-12-01

To improve the safety factor of nuclear power plants in accident scenarios, molybdenum (Mo), with its high-temperature strength, is proposed as a potential fuel-cladding candidate. However, Mo undergoes rapid oxidation and sublimation at elevated temperatures in oxygen-rich environments. Thus, it is necessary to coat Mo with a protective layer. The diffusional interactions in two systems, namely, Zircaloy-2 (Zr2) on a Mo tube, and iron-chromium-aluminum (FeCrAl) on a Mo rod, were studied by aging coated Mo substrates in high vacuum at temperatures ranging from 650 °C to 1000° for 1000 h. The specimens were characterized using scanning electron microscopy (SEM), energy-dispersive spectrometry (EDS) and nanoindentation. In both systems, pores in the coating increased in size and number with increasing temperature over time, and cracks were also observed; intermetallic phases formed between the Mo and its coatings.

Benefits of barrier fuel on fuel cycle economics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crowther, R.L.; Kunz, C.L.

1988-01-01

Barrier fuel rod cladding was developed to eliminate fuel rod failures from pellet/cladding stress/corrosion interaction and to eliminate the associated need to restrict the rate at which fuel rod power can be increased. The performance of barrier cladding has been demonstrated through extensive testing and through production application to many boiling water reactors (BWRs). Power reactor data have shown that barrier fuel rod cladding has a significant beneficial effect on plant capacity factor and plant operating costs and significantly increases fuel reliability. Independent of the fuel reliability benefit, it is less obvious that barrier fuel has a beneficial effect ofmore » fuel cycle costs, since barrier cladding is more costly to fabricate. Evaluations, measurements, and development activities, however, have shown that the fuel cycle cost benefits of barrier fuel are large. This paper is a summary of development activities that have shown that application of barrier fuel significantly reduces BWR fuel cycle costs.« less

Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Medvedev, Pavel; Madden, James; Wachs, Dan; Clark, Curtis; Meyer, Mitch

2015-09-01

Low-enrichment (235U < 20 pct) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing consisted of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates were fabricated using a friction bonding process, tested in INL's advanced test reactor (ATR), and then subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. In the samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface, possible indications of porosity/debonding were found, which suggested that the interface in this location is relatively weak.

Hydrogen permeation in FeCrAl alloys for LWR cladding application

NASA Astrophysics Data System (ADS)

Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; Snead, Lance L.

2015-06-01

FeCrAl, an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In this study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. The total tritium inventory inside the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.

Clad metals by roll bonding for SOFC interconnects

NASA Astrophysics Data System (ADS)

Chen, L.; Jha, B.; Yang, Zhenguo; Xia, Guang-Guang; Stevenson, Jeffry W.; Singh, Prabhakar

2006-08-01

High-temperature oxidation-resistant alloys are currently considered as a candidate material for construction of interconnects in intermediate-temperature solid oxide fuel cells. Among these alloys, however, different groups of alloys demonstrate different advantages and disadvantages, and few, if any, can completely satisfy the stringent requirements for the application. To integrate the advantages and avoid the disadvantages of different groups of alloys, cladding has been proposed as one approach in fabricating metallic layered interconnect structures. To examine the feasibility of this approach, the austenitic Ni-base alloy Haynes 230 and the ferritic stainless steel AL 453 were selected as examples and manufactured into a clad metal. Its suitability as an interconnect construction material was investigated. This paper provides a brief overview of the cladding approach and discusses the viability of this technology to fabricate the metallic layered-structure interconnects.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maloy, Stuart Andrew; Pestovich, Kimberly Shay; Anderoglu, Osman

The Fuel Cycle Research and Development program is investigating methods of transmuting minor actinides in various fuel cycle options. To achieve this goal, new fuels and cladding materials must be developed and tested to high burnup levels (e.g. >20%) requiring cladding to withstand very high doses (greater than 200 dpa) while in contact with the coolant and the fuel. To develop and qualify materials to a total fluence greater than 200 dpa requires development of advanced alloys and irradiations in fast reactors to test these alloys. Recent results from testing numerous ferritic/martensitic steels at low temperatures suggest that improvements inmore » low temperature radiation tolerance can be achieved through carefully controlling the nitrogen content in these alloys. Thus, four new heats of HT-9 were produced with controlled nitrogen content: two by Metalwerks and two by Sophisticated Alloys. Initial results on these new alloys are presented including microstructural analysis and hardness testing. Future testing will include irradiation testing with ions and in reactor.« less

Nuclear fuels for very high temperature applications

NASA Astrophysics Data System (ADS)

Lundberg, L. B.; Hobbins, R. R.

The success of the development of nuclear thermal propulsion devices and thermionic space nuclear power generation systems depends on the successful utilization of nuclear fuel materials at temperatures in the range 2000 to 3500 K. Problems associated with the utilization of uranium bearing fuel materials at these very high temperatures while maintaining them in the solid state for the required operating times are addressed. The critical issues addressed include evaporation, melting, reactor neutron spectrum, high temperature chemical stability, fabrication, fission induced swelling, fission product release, high temperature creep, thermal shock resistance, and fuel density, both mass and fissile atom. Candidate fuel materials for this temperature range are based on UO2 or uranium carbides. Evaporation suppression, such as a sealed cladding, is required for either fuel base. Nuclear performance data needed for design are sparse for all candidate fuel forms in this temperature range, especially at the higher temperatures.

Nuclear fuel alloys or mixtures and method of making thereof

DOEpatents

Mariani, Robert Dominick; Porter, Douglas Lloyd

2016-04-05

Nuclear fuel alloys or mixtures and methods of making nuclear fuel mixtures are provided. Pseudo-binary actinide-M fuel mixtures form alloys and exhibit: body-centered cubic solid phases at low temperatures; high solidus temperatures; and/or minimal or no reaction or inter-diffusion with steel and other cladding materials. Methods described herein through metallurgical and thermodynamics advancements guide the selection of amounts of fuel mixture components by use of phase diagrams. Weight percentages for components of a metallic additive to an actinide fuel are selected in a solid phase region of an isothermal phase diagram taken at a temperature below an upper temperature limit for the resulting fuel mixture in reactor use. Fuel mixtures include uranium-molybdenum-tungsten, uranium-molybdenum-tantalum, molybdenum-titanium-zirconium, and uranium-molybdenum-titanium systems.

Hydride reorientation and its impact on ambient temperature mechanical properties of high burn-up irradiated and unirradiated recrystallized Zircaloy-2 nuclear fuel cladding with an inner liner

NASA Astrophysics Data System (ADS)

Auzoux, Q.; Bouffioux, P.; Machiels, A.; Yagnik, S.; Bourdiliau, B.; Mallet, C.; Mozzani, N.; Colas, K.

2017-10-01

Precipitation of radial hydrides in zirconium-based alloy cladding concomitant with the cooling of spent nuclear fuel during dry storage can potentially compromise cladding integrity during its subsequent handling and transportation. This paper investigates hydride reorientation and its impact on ductility in unirradiated and irradiated recrystallized Zircaloy-2 cladding with an inner liner (cladding for boiling water reactors) subjected to hydride reorientation treatments. Cooling from 400 °C, hydride reorientation occurs in recrystallized Zircaloy-2 with liner at a lower effective stress in irradiated samples (below 40 MPa) than in unirradiated specimens (between 40 and 80 MPa). Despite significant hydride reorientation, unirradiated recrystallized Zircaloy-2 with liner cladding containing ∼200 wppm hydrogen shows a high diametral strain at fracture (>15%) during burst tests at ambient temperature. This ductile behavior is due to (1) the lower yield stress of the recrystallized cladding materials in comparison to hydride fracture strength (corrected by the compression stress arising from the precipitation) and (2) the hydride or hydrogen-depleted zone as a result of segregation of hydrogen into the liner layer. In irradiated Zircaloy-2 with liner cladding containing ∼340 wppm hydrogen, the conservation of some ductility during ring tensile tests at ambient temperature after reorientation treatment at 400 °C with cooling rates of ∼60 °C/h is also attributed to the existence of a hydride-depleted zone. Treatments at lower cooling rates (∼6 °C/h and 0.6 °C/h) promote greater levels of hydrogen segregation into the liner and allow for increased irradiation defect annealing, both of which result in a significant increase in ductility. Based on this investigation, given the very low cooling rates typical of dry storage systems, it can be concluded that the thermal transients associated with dry storage should not degrade, and more likely should actually improve, ductility of recrystallized Zircaloy-2 cladding with inner liner with such hydrogen content.

Parametric and experimentally informed BWR Severe Accident Analysis Utilizing FeCrAl - M3FT-17OR020205041

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ott, Larry J.; Howell, Michael; Robb, Kevin R.

Iron-chromium-aluminum (FeCrAl) alloys are being considered as advanced fuel cladding concepts with enhanced accident tolerance. At high temperatures, FeCrAl alloys have slower oxidation kinetics and higher strength compared with zirconium-based alloys. FeCrAl could be used for fuel cladding and spacer or mixing vane grids in light water reactors and/or as channel box material in boiling water reactors (BWRs). There is a need to assess the potential gains afforded by the FeCrAl accident-tolerant-fuel (ATF) concept over the existing zirconium-based materials employed today. To accurately assess the response of FeCrAl alloys under severe accident conditions, a number of FeCrAl properties and characteristicsmore » are required. These include thermophysical properties as well as burst characteristics, oxidation kinetics, possible eutectic interactions, and failure temperatures. These properties can vary among different FeCrAl alloys. Oak Ridge National Laboratory has pursued refined values for the oxidation kinetics of the B136Y FeCrAl alloy (Fe-13Cr-6Al wt %). This investigation included oxidation tests with varying heating rates and end-point temperatures in a steam environment. The rate constant for the low-temperature oxidation kinetics was found to be higher than that for the commercial APMT FeCrAl alloy (Fe-21Cr-5Al-3Mo wt %). Compared with APMT, a 5 times higher rate constant best predicted the entire dataset (root mean square deviation). Based on tests following heating rates comparable with those the cladding would experience during a station blackout, the transition to higher oxidation kinetics occurs at approximately 1,500°C. A parametric study varying the low-temperature FeCrAl oxidation kinetics was conducted for a BWR plant using FeCrAl fuel cladding and channel boxes using the MELCOR code. A range of station blackout severe accident scenarios were simulated for a BWR/4 reactor with Mark I containment. Increasing the FeCrAl low-temperature oxidation rate constant (3 times and 10 times that of the rate constant for APMT) had a negligible impact on the early stages of the accident and minor impacts on the accident progression after the first relocation of the fuel. At temperatures below 1,500°C, increasing the rate constant for APMT by a factor of 10 still resulted in only minor FeCrAl oxidation. In general, the gains afforded by the FeCrAl enhanced ATF concept with respect to accident sequence timing and combustible gas generation are consistent with previous efforts. Compared with the traditional Zircaloy-based cladding and channel box system, the FeCrAl concept could provide a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. For example, a station blackout was simulated in which cooling water injection was lost 36 hours after shutdown. The timing to first fuel relocation was delayed by approximately 5 h for the FeCrAl ATF concept compared with that of the traditional Zircaloy-based cladding and channel box system.« less

Enhancing the ABAQUS Thermomechanics Code to Simulate Steady and Transient Fuel Rod Behavior

DOE Office of Scientific and Technical Information (OSTI.GOV)

R. L. Williamson; D. A. Knoll

2009-09-01

A powerful multidimensional fuels performance capability, applicable to both steady and transient fuel behavior, is developed based on enhancements to the commercially available ABAQUS general-purpose thermomechanics code. Enhanced capabilities are described, including: UO2 temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, fuel densification, fission gas release, cladding thermal and irradiation creep, cladding irradiation growth , gap heat transfer, and gap/plenum gas behavior during irradiation. The various modeling capabilities are demonstrated using a 2D axisymmetric analysis of the upper section of a simplified multi-pellet fuel rod, during both steady and transient operation. Computational results demonstrate the importancemore » of a multidimensional fully-coupled thermomechanics treatment. Interestingly, many of the inherent deficiencies in existing fuel performance codes (e.g., 1D thermomechanics, loose thermo-mechanical coupling, separate steady and transient analysis, cumbersome pre- and post-processing) are, in fact, ABAQUS strengths.« less

Evaluation of refractory-metal-clad uranium nitride and uranium dioxide fuel pins after irradiation for times up to 10 450 hours at 990 C

NASA Technical Reports Server (NTRS)

Bowles, K. J.; Gluyas, R. E.

1975-01-01

The effects of some materials variables on the irradiation performance of fuel pins for a lithium-cooled space power reactor design concept were examined. The variables studied were UN fuel density, fuel composition, and cladding alloy. All pins were irradiated at about 990 C in a thermal neutron environment to the design fuel burnup. An 85-percent dense UN fuel gave the best overall results in meeting the operational goals. The T-111 cladding on all specimens was embrittled, possibly by hydrogen in the case of the UN fuel and by uranium and oxygen in the case of the UO2 fuel. Tests with Cb-1Zr cladding indicate potential use of this cladding material. The UO2 fueled specimens met the operational goals of less than 1 percent cladding strain, but other factors make UO2 less attractive than low-density UN for the contemplated space power reactor use.

Hydrogen permeation in FeCrAl alloys for LWR cladding application

DOE PAGES

Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; ...

2015-03-19

FeCrAl is an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In our study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. Also, the total tritium inventory insidemore » the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.« less

Results of Uranium Dioxide-Tungsten Irradiation Test and Post-Test Examination

NASA Technical Reports Server (NTRS)

Collins, J. F.; Debogdan, C. E.; Diianni, D. C.

1973-01-01

A uranium dioxide (UO2) fueled capsule was fabricated and irradiated in the NASA Plum Brook Reactor Facility. The capsule consisted of two bulk UO2 specimens clad with chemically vapor deposited tungsten (CVD W) 0.762 and 0.1016 cm (0.030-and 0.040-in.) thick, respectively. The second specimen with 0.1016-cm (0.040-in.) thick cladding was irradiated at temperature for 2607 hours, corresponding to an average burnup of 1.516 x 10 to the 20th power fissions/cu cm. Postirradiation examination showed distortion in the bottom end cap, failure of the weld joint, and fracture of the central vent tube. Diametral growth was 1.3 percent. No evidence of gross interaction between CVD tungsten or arc-cast tungsten cladding and the UO2 fuel was observed. Some of the fission gases passed from the fuel cavity to the gas surrounding the fuel specimen via the vent tube and possibly the end-cap weld failure. Whether the UO2 loss rates through the vent tube were within acceptable limits could not be determined in view of the end-cap weld failure.

Assembly and Delivery of Rabbit Capsules for Irradiation of Silicon Carbide Cladding Tube Specimens in the High Flux Isotope Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koyanagi, Takaaki; Petrie, Christian M.

Neutron irradiation of silicon carbide (SiC)-based fuel cladding under a high radial heat flux presents a critical challenge for SiC cladding concepts in light water reactors (LWRs). Fission heating in the fuel provides a high heat flux through the cladding, which, combined with the degraded thermal conductivity of SiC under irradiation, results in a large temperature gradient through the thickness of the cladding. The strong temperature dependence of swelling in SiC creates a complex stress profile in SiCbased cladding tubes as a result of differential swelling. The Nuclear Science User Facilities (NSUF) Program within the US Department of Energy Officemore » of Nuclear Energy is supporting research efforts to improve the scientific understanding of the effects of irradiation on SiC cladding tubes. Ultimately, the results of this project will provide experimental validation of multi-physics models for SiC-based fuel cladding during LWR operation. The first objective of this project is to irradiate tube specimens using a previously developed design that allows for irradiation testing of miniature SiC tube specimens subjected to a high radial heat flux. The previous “rabbit” capsule design uses the gamma heating in the core of the High Flux Isotope Reactor (HFIR) to drive a high heat flux through the cladding tube specimens. A compressible aluminum foil allows for a constant thermal contact conductance between the cladding tubes and the rabbit housing despite swelling of the SiC tubes. To allow separation of the effects of irradiation from those due to differential swelling under a high heat flux, a new design was developed under the NSUF program. This design allows for irradiation of similar SiC cladding tube specimens without a high radial heat flux. This report briefly describes the irradiation experiment design concepts, summarizes the irradiation test matrix, and reports on the successful delivery of six rabbit capsules to the HFIR. Rabbits of both low and high heat flux configurations have been assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. These rabbits contain a wide variety of specimens including monolith tubes, SiC fiber SiC matrix (SiC/SiC) composites, duplex specimens (inner composite, outer monolith), and specimens with a variety of metallic or ceramic coatings on the outer surface. The rabbits are targeted for insertion during HFIR cycle 475, which is scheduled for September 2017.« less

LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carmack, William Jonathan; Barrett, Kristine Eloise; Chichester, Heather Jean MacLean

2015-09-01

The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirementsmore » for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.« less

Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Kurt R.; Howard, Richard H.; Daily, Charles R.

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designsmore » allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.« less

Accident tolerant fuel cladding development: Promise, status, and challenges

NASA Astrophysics Data System (ADS)

Terrani, Kurt A.

2018-04-01

The motivation for transitioning away from zirconium-based fuel cladding in light water reactors to significantly more oxidation-resistant materials, thereby enhancing safety margins during severe accidents, is laid out. A review of the development status for three accident tolerant fuel cladding technologies, namely coated zirconium-based cladding, ferritic alumina-forming alloy cladding, and silicon carbide fiber-reinforced silicon carbide matrix composite cladding, is offered. Technical challenges and data gaps for each of these cladding technologies are highlighted. Full development towards commercial deployment of these technologies is identified as a high priority for the nuclear industry.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rosenbaum, H.S.

1979-03-01

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

Clad metals, roll bonding and their applications for SOFC interconnects

NASA Astrophysics Data System (ADS)

Chen, Lichun; Yang, Zhenguo; Jha, Bijendra; Xia, Guanguang; Stevenson, Jeffry W.

Metallic interconnects have been becoming an increasingly interesting topic in the development in intermediate temperature solid oxide fuel cells (SOFC). High temperature oxidation resistant alloys are currently considered as candidate materials. Among these alloys however, different groups of alloys demonstrate different advantages and disadvantages, and few if any can completely satisfy the stringent requirements for the application. To integrate the advantages and avoid the disadvantages of different groups of alloys, clad metal has been proposed for SOFC interconnect applications and interconnect structures. This paper gives a brief overview of the cladding approach and its applications, and discuss the viability of this technology to fabricate the metallic layered-structure interconnects. To examine the feasibility of this approach, the austenitic Ni-base alloy Haynes 230 and the ferritic stainless steel AL 453 were selected as examples and manufactured into a clad metal. Its suitability as an interconnect construction material was investigated.

Top Ten Reasons for DEOX as a Front End to Pyroprocessing

DOE Office of Scientific and Technical Information (OSTI.GOV)

B.R. Westphal; K.J. Bateman; S.D. Herrmann

A front end step is being considered to augment chopping during the treatment of spent oxide fuel by pyroprocessing. The front end step, termed DEOX for its emphasis on decladding via oxidation, employs high temperatures to promote the oxidation of UO2 to U3O8 via an oxygen carrier gas. During oxidation, the spent fuel experiences a 30% increase in lattice structure volume resulting in the separation of fuel from cladding with a reduced particle size. A potential added benefit of DEOX is the removal of fission products, either via direct release from the broken fuel structure or via oxidation and volatilizationmore » by the high temperature process. Fuel element chopping is the baseline operation to prepare spent oxide fuel for an electrolytic reduction step. Typical chopping lengths range from 1 to 5 mm for both individual elements and entire assemblies. During electrolytic reduction, uranium oxide is reduced to metallic uranium via a lithium molten salt. An electrorefining step is then performed to separate a majority of the fission products from the recoverable uranium. Although DEOX is based on a low temperature oxidation cycle near 500oC, additional conditions have been tested to distinguish their effects on the process.[1] Both oxygen and air have been utilized during the oxidation portion followed by vacuum conditions to temperatures as high as 1200oC. In addition, the effects of cladding on fission product removal have also been investigated with released fuel to temperatures greater than 500oC.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stimpson, Shane G; Powers, Jeffrey J; Clarno, Kevin T

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity, multiphysics simulations of light water reactors (LWRs) by coupling a variety of codes within the Virtual Environment for Reactor Analysis (VERA). One of the primary goals of CASL is to predict local cladding failure through pellet-clad interaction (PCI). This capability is currently being pursued through several different approaches, such as with Tiamat, which is a simulation tool within VERA that more tightly couples the MPACT neutron transport solver, the CTF thermal hydraulics solver, and the MOOSE-based Bison-CASL fuel performance code. However, the process in this papermore » focuses on running fuel performance calculations with Bison-CASL to predict PCI using the multicycle output data from coupled neutron transport/thermal hydraulics simulations. In recent work within CASL, Watts Bar Unit 1 has been simulated over 12 cycles using the VERA core simulator capability based on MPACT and CTF. Using the output from these simulations, Bison-CASL results can be obtained without rerunning all 12 cycles, while providing some insight into PCI indicators. Multi-cycle Bison-CASL results are presented and compared against results from the FRAPCON fuel performance code. There are several quantities of interest in considering PCI and subsequent fuel rod failures, such as the clad hoop stress and maximum centerline fuel temperature, particularly as a function of time. Bison-CASL performs single-rod simulations using representative power and temperature distributions, providing high-resolution results for these and a number of other quantities. This will assist in identifying fuels rods as potential failure locations for use in further analyses.« less

Three-dimensional fuel pin model validation by prediction of hydrogen distribution in cladding and comparison with experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aly, A.; Avramova, Maria; Ivanov, Kostadin

To correctly describe and predict this hydrogen distribution there is a need for multi-physics coupling to provide accurate three-dimensional azimuthal, radial, and axial temperature distributions in the cladding. Coupled high-fidelity reactor-physics codes with a sub-channel code as well as with a computational fluid dynamics (CFD) tool have been used to calculate detailed temperature distributions. These high-fidelity coupled neutronics/thermal-hydraulics code systems are coupled further with the fuel-performance BISON code with a kernel (module) for hydrogen. Both hydrogen migration and precipitation/dissolution are included in the model. Results from this multi-physics analysis is validated utilizing calculations of hydrogen distribution using models informed bymore » data from hydrogen experiments and PIE data.« less

Capabilities to improve corrosion resistance of fuel claddings by using powerful laser and plasma sources

NASA Astrophysics Data System (ADS)

Borisov, V. M.; Trofimov, V. N.; Sapozhkov, A. Yu.; Kuzmenko, V. A.; Mikhaylov, V. B.; Cherkovets, V. Ye.; Yakushkin, A. A.; Yakushin, V. L.; Dzhumayev, P. S.

2016-12-01

The treatment conditions of fuel claddings of the E110 alloy by using powerful UV or IR laser radiation, which lead to the increase in the corrosion resistance at the high-temperature ( T = 1100°C) oxidation simulating a loss-of-coolant accident, are determined. The possibility of the complete suppression of corrosion under these conditions by using pulsed laser deposition of a Cr layer is demonstrated. The behavior of protective coatings of Al, Al2O3, and Cr planted on steel EP823 by pulsed laser deposition, which is planned to be used in the BREST-OD-300, is studied. The methods of the almost complete suppression of corrosion in liquid lead to the temperature of 720°C are shown.

Preliminary Modeling of Accident Tolerant Fuel Concepts under Accident Conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gamble, Kyle A.; Hales, Jason D.

2016-12-01

The catastrophic events that occurred at the Fukushima-Daiichi nuclear power plant in 2011 have led to widespread interest in research of alternative fuels and claddings that are proposed to be accident tolerant. Thus, the United States Department of Energy through its NEAMS (Nuclear Energy Advanced Modeling and Simulation) program has funded an Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The ATF HIP is funded for a three-year period. The purpose of the HIP is to perform research into two potential accident tolerant concepts and provide an in-depth report to the Advanced Fuels Campaign (AFC) describing the behavior of themore » concepts, both of which are being considered for inclusion in a lead test assembly scheduled for placement into a commercial reactor in 2022. The initial focus of the HIP is on uranium silicide fuel and iron-chromium-aluminum (FeCrAl) alloy cladding. Utilizing the expertise of three national laboratory participants (INL, LANL, and ANL) a comprehensive mulitscale approach to modeling is being used including atomistic modeling, molecular dynamics, rate theory, phase-field, and fuel performance simulations. In this paper, we present simulations of two proposed accident tolerant fuel systems: U3Si2 fuel with Zircaloy-4 cladding, and UO2 fuel with FeCrAl cladding. The simulations investigate the fuel performance response of the proposed ATF systems under Loss of Coolant and Station Blackout conditions using the BISON code. Sensitivity analyses are completed using Sandia National Laboratories’ DAKOTA software to determine which input parameters (e.g., fuel specific heat) have the greatest influence on the output metrics of interest (e.g., fuel centerline temperature). Early results indicate that each concept has significant advantages as well as areas of concern. Further work is required prior to formulating the proposition report for the Advanced Fuels Campaign.« less

Modelling Accident Tolerant Fuel Concepts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hales, Jason Dean; Gamble, Kyle Allan Lawrence

2016-05-01

The catastrophic events that occurred at the Fukushima-Daiichi nuclear power plant in 2011 have led to widespread interest in research of alternative fuels and claddings that are proposed to be accident tolerant. The United States Department of Energy (DOE) through its Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has funded an Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The ATF HIP is a three-year project to perform research on two accident tolerant concepts. The final outcome of the ATF HIP will be an in-depth report to the DOE Advanced Fuels Campaign (AFC) giving a recommendation on whether eithermore » of the two concepts should be included in their lead test assembly scheduled for placement into a commercial reactor in 2022. The two ATF concepts under investigation in the HIP are uranium silicide fuel and iron-chromium-aluminum (FeCrAl) alloy cladding. Utilizing the expertise of three national laboratory participants (Idaho National Laboratory, Los Alamos National Laboratory, and Argonne National Laboratory), a comprehensive multiscale approach to modeling is being used that includes atomistic modeling, molecular dynamics, rate theory, phase-field, and fuel performance simulations. Model development and fuel performance analysis are critical since a full suite of experimental studies will not be complete before AFC must prioritize concepts for focused development. In this paper, we present simulations of the two proposed accident tolerance fuel systems: U3Si2 fuel with Zircaloy-4 cladding, and UO2 fuel with FeCrAl cladding. Sensitivity analyses are completed using Sandia National Laboratories’ Dakota software to determine which input parameters (e.g., fuel specific heat) have the greatest influence on the output metrics of interest (e.g., fuel centerline temperature). We also outline the multiscale modelling approach being employed. Considerable additional work is required prior to preparing the recommendation report for the Advanced Fuels Campaign.« less

The Mechanical Response of Advanced Claddings during Proposed Reactivity Initiated Accident Conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cinbiz, Mahmut N; Brown, Nicholas R; Terrani, Kurt A

2017-01-01

This study investigates the failure mechanisms of advanced nuclear fuel cladding of FeCrAl at high-strain rates, similar to design basis reactivity initiated accidents (RIA). During RIA, the nuclear fuel cladding was subjected to the plane-strain to equibiaxial tension strain states. To achieve those accident conditions, the samples were deformed by the expansion of high strength Inconel alloy tube under pre-specified pressure pulses as occurring RIA. The mechanical response of the advanced claddings was compared to that of hydrided zirconium-based nuclear fuel cladding alloy. The hoop strain evolution during pressure pulses were collected in situ; the permanent diametral strains of bothmore » accident tolerant fuel (ATF) claddings and the current nuclear fuel alloys were determined after rupture.« less

Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1–3

DOE PAGES

Stimpson, Shane; Powers, Jeffrey; Clarno, Kevin; ...

2017-12-22

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity multiphysics simulations of light water nuclear reactors. To accomplish this, CASL is developing the Virtual Environment for Reactor Applications (VERA), which is a suite of code packages for thermal hydraulics, neutron transport, fuel performance, and coolant chemistry. As VERA continues to grow and expand, there has been an increased focus on incorporating fuel performance analysis methods. One of the primary goals of CASL is to estimate local cladding failure probability through pellet-clad interaction, which consists of both pellet-clad mechanical interaction (PCMI) and stress corrosion cracking. Estimatingmore » clad failure is important to preventing release of fission products to the primary system and accurate estimates could prove useful in establishing less conservative power ramp rates or when considering load-follow operations.While this capability is being pursued through several different approaches, the procedure presented in this article focuses on running independent fuel performance calculations with BISON using a file-based one-way coupling based on multicycle output data from high fidelity, pin-resolved coupled neutron transport–thermal hydraulics simulations. This type of approach is consistent with traditional fuel performance analysis methods, which are typically separate from core simulation analyses. A more tightly coupled approach is currently being developed, which is the ultimate target application in CASL.Recent work simulating 12 cycles of Watts Bar Unit 1 with VERA core simulator are capitalized upon, and quarter-core BISON results for parameters of interest to PCMI (maximum centerline fuel temperature, maximum clad hoop stress, and minimum gap size) are presented for Cycles 1–3. In conclusion, based on these results, this capability demonstrates its value and how it could be used as a screening tool for gathering insight into PCMI, singling out limiting rods for further, more detailed analysis.« less

Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1–3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stimpson, Shane; Powers, Jeffrey; Clarno, Kevin

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity multiphysics simulations of light water nuclear reactors. To accomplish this, CASL is developing the Virtual Environment for Reactor Applications (VERA), which is a suite of code packages for thermal hydraulics, neutron transport, fuel performance, and coolant chemistry. As VERA continues to grow and expand, there has been an increased focus on incorporating fuel performance analysis methods. One of the primary goals of CASL is to estimate local cladding failure probability through pellet-clad interaction, which consists of both pellet-clad mechanical interaction (PCMI) and stress corrosion cracking. Estimatingmore » clad failure is important to preventing release of fission products to the primary system and accurate estimates could prove useful in establishing less conservative power ramp rates or when considering load-follow operations.While this capability is being pursued through several different approaches, the procedure presented in this article focuses on running independent fuel performance calculations with BISON using a file-based one-way coupling based on multicycle output data from high fidelity, pin-resolved coupled neutron transport–thermal hydraulics simulations. This type of approach is consistent with traditional fuel performance analysis methods, which are typically separate from core simulation analyses. A more tightly coupled approach is currently being developed, which is the ultimate target application in CASL.Recent work simulating 12 cycles of Watts Bar Unit 1 with VERA core simulator are capitalized upon, and quarter-core BISON results for parameters of interest to PCMI (maximum centerline fuel temperature, maximum clad hoop stress, and minimum gap size) are presented for Cycles 1–3. In conclusion, based on these results, this capability demonstrates its value and how it could be used as a screening tool for gathering insight into PCMI, singling out limiting rods for further, more detailed analysis.« less

PATHFINDER ATOMIC POWER PLANT TECHNICAL PROGRESS REPORT FOR JULY 1, 1959- SEPTEMBER 30, 1959

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1960-10-31

ABS>Fuel Element Research and Development. Dynamic and static corrosion tests on 8001 Al were completed. Annealmmmg of 1100 cladding on 5083 and M400 cladding on X2219 were tested at 500 deg C, and investigation continued on producing X8101 Al alloy cladding in tube plates by extrusion. Boiler fuel element capsule irradiation tests and subassembly tests are described Heat transfer loop studies and fuel fabrication for the critical facility are reported. Boiler fuel element mechanical design and testing progress is desc ribed. and the superheater fuel element temperature evaluating routine is discussed. Low- enrichment superheater fuel element development included design studiesmore » and stainless steel powder and UO/sub 2/ powder fabrication studies Reactor Mechanical Studies. Research is reported on vessel and structure design, fabrication, and testing, recirculation system design, steam separator tests, and control rod studies. Nuclear Analysis. Reactor physics studies are reported on nuclear constants, baffle plate analysis, comparison of core representations, delayed neutron fraction. and shielding analysis of the reactor building. Reactor and system dynamics and critical experiments were also studied. Chemistry. Progress is reported on recombiner. radioactive gas removal and storage, ion exchanger and radiochemical processing. (For preceding period see ACNP-5915.) (T.R.H.)« less

Final report on accident tolerant fuel performance analysis of APMT-Steel Clad/UO₂ fuel and APMT-Steel Clad/UN-U₃Si₅ fuel concepts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Unal, Cetin; Galloway, Jack D.

2014-09-12

In FY2014 our group completed and documented analysis of new Accident Tolerant Fuel (ATF) concepts using BISON. We have modeled the viability of moving from Zircaloy to stainless steel cladding in traditional light water reactors (LWRs). We have explored the reactivity penalty of this change using the MCNP-based burnup code Monteburns, while attempting to minimize this penalty by increasing the fuel pellet radius and decreasing the cladding thickness. Fuel performance simulations using BISON have also been performed to quantify changes to structural integrity resulting from thinner stainless steel claddings. We account for thermal and irradiation creep, fission gas swelling, thermalmore » swelling and fuel relocation in the models for both Zircaloy and stainless steel claddings. Additional models that account for the lower oxidation stainless steel APMT are also invoked where available. Irradiation data for HT9 is used as a fallback in the absence of appropriate models. In this study the isotopic vectors within each natural element are varied to assess potential reactivity gains if advanced enrichment capabilities were levied towards cladding technologies. Recommendations on cladding thicknesses for a robust cladding as well as the constitutive components of a less penalizing composition are provided. In the first section (section 1-3), we present results accepted for publication in the 2014 TOPFUEL conference regarding the APMT/UO₂ ATF concept (J. Galloway & C. Unal, Accident Tolerant and Neutronically Favorable LWR Cladding, Proceedings of WRFPM 2014, Sendai, Japan, Paper No.1000050). Next we discuss our preliminary findings from the thermo-mechanical analysis of UN-U₃Si₅ fuel with APMT clad. In this analysis we used models developed from limited data that need to be updated when the irradiation data from ATF-1 test is available. Initial results indicate a swelling rate less than 1.5% is needed to prevent excessive clad stress.« less

Structure and Thermodynamical Properties of Zirconium Hydrides from First-Principle

NASA Astrophysics Data System (ADS)

Blomqvist, Jakob; Olofsson, Johan; Alvarez, Anna-Maria; Bjerkén, Christina

Zirconium alloys are used as nuclear fuel cladding material due to their mechanical and corrosion resistant properties together with their favorable cross-section for neutron scattering. At running conditions, however, there will be an increase of hydrogen in the vicinity of the cladding surface at the water side of the fuel. The hydrogen will diffuse into the cladding material and at certain conditions, such as lower temperatures and external load, hydrides will precipitate out in the material and cause well known embrittlement, blistering and other unwanted effects. Using phase-field methods it is now possible to model precipitation buildup in metals, for example as a function of hydrogen concentration, temperature and external load, but the technique relies on input of parameters, such as the formation energy of the hydrides and matrix. To that end, we have computed, using the density functional theory (DFT) code GPAW, the latent heat of fusion as well as solved the crystal structure for three zirconium hydride polymorphs: δ-ZrH1.6, γ-ZrH, and Є-ZrH2.

Efficient Geometry and Data Handling for Large-Scale Monte Carlo - Thermal-Hydraulics Coupling

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard

2014-06-01

Detailed coupling of thermal-hydraulics calculations to Monte Carlo reactor criticality calculations requires each axial layer of each fuel pin to be defined separately in the input to the Monte Carlo code in order to assign to each volume the temperature according to the result of the TH calculation, and if the volume contains coolant, also the density of the coolant. This leads to huge input files for even small systems. In this paper a methodology for dynamical assignment of temperatures with respect to cross section data is demonstrated to overcome this problem. The method is implemented in MCNP5. The method is verified for an infinite lattice with 3x3 BWR-type fuel pins with fuel, cladding and moderator/coolant explicitly modeled. For each pin 60 axial zones are considered with different temperatures and coolant densities. The results of the axial power distribution per fuel pin are compared to a standard MCNP5 run in which all 9x60 cells for fuel, cladding and coolant are explicitly defined and their respective temperatures determined from the TH calculation. Full agreement is obtained. For large-scale application the method is demonstrated for an infinite lattice with 17x17 PWR-type fuel assemblies with 25 rods replaced by guide tubes. Again all geometrical detailed is retained. The method was used in a procedure for coupled Monte Carlo and thermal-hydraulics iterations. Using an optimised iteration technique, convergence was obtained in 11 iteration steps.

The behavior of breached boiling water reactor fuel rods on long-term exposure to air and argon at 598 K

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kohli, R.; Gilbert, E.R.; Johnson, A.B.

1985-05-01

Two irradiated boiling water reactor fuel rods with breached cladding were exposed to argon and to air at 598 K for 7.56 Ms (2100 h). These tests were conducted to determine fuel swelling and cladding crack propagation under conditions that promote UO/sub 2/ fuel oxidation and to observe the behavior of water-logged breached fuel in an inert gas environment. The two rods were selected for testing after extensive hot cell examination had shown the cladding of both rods to be breached with several centimetres of open cracks; the cracks were characterized in detail before the test. As part of themore » experiment, the amount of the readily removable water contained in the fuel rods was determined. To oxidize the fuel to a significant level ( about10%), the air in the annealine capsule was replenished approximately daily. The depletion of oxygen available in the air capsule due to fuel oxidation occurred in about0.036 Ms (10 h). At the end of the test period, about6% of the fuel is estimated to have oxidized. Posttest examination of the rods showed that cladding degradation resulted from swelling due to oxidation of the fuel in the air environment. The cladding degradation was localized and fuel oxidation did not measurably extend beyond the cladding breach. No cladding degradation was measurable in the breached fuel rod tested in argon.« less

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

DOE PAGES

Kim, Taeil; Harbaruk, Dzmitry; Gerardi, Craig; ...

2017-07-10

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in themore » relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling« less

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Taeil; Harbaruk, Dzmitry; Gerardi, Craig

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in themore » relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Young-Ho; Byun, Thak Sang

Accident-tolerant fuels are expected to have considerably longer coping time to respond to the loss of active cooling under severe accidents and, at the same time, have comparable or improved fuel performance during normal operation. The wear resistance of accident tolerant fuels, therefore, needs to be examined to determine the applicability of these cladding candidates to the current operating PWRs because the most common failure of nuclear fuel claddings is still caused by grid-to-rod fretting during normal operations. In this study, reciprocating sliding wear tests on three kinds of cladding candidates for accident-tolerant fuels have been performed to investigate themore » tribological compatibilities of selfmated cladding candidates and to determine the direct applicability of conventional Zirconium-based alloys as supporting structural materials. The friction coefficients of the cladding candidates are strongly influenced by the test environments and coupled materials. The wear test results under water lubrication conditions indicate that the supporting structural materials for the cladding candidates of accident-tolerant fuels need to be replaced with the same cladding materials instead of using conventional Zirconium-based alloys.« less

Nuclear-powered pacemaker fuel cladding study. [Difficulty of dissolving cladding and /sup 238/PuO/sub 2/ for obtaining materials for acts of terrorism

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shoup, R.L.

1976-07-01

The fabrication of fuel capsules with refractory metal and alloy clads used in nuclear-powered cardiac pacemakers precludes the expedient dissolution of the clad in inorganic acid solutions. An experiment to measure penetration rates of acids on commonly used fuel pellet clads indicated that it is not impossible, but that it would be very difficult to dissolve the multiple cladding. This work was performed because of a suggestion that a /sup 238/PuO/sub 2/-powered pacemaker could be transformed into a terrorism weapon.

Ceramic Coatings for Clad (The C 3 Project): Advanced Accident-Tolerant Ceramic Coatings for Zr-Alloy Cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sickafus, Kurt E.; Wirth, Brian; Miller, Larry

The goal of this NEUP-IRP project is to develop a fuel concept based on an advanced ceramic coating for Zr-alloy cladding. The coated cladding must exhibit demonstrably improved performance compared to conventional Zr-alloy clad in the following respects: During normal service, the ceramic coating should decrease cladding oxidation and hydrogen pickup (the latter leads to hydriding and embrittlement). During a reactor transient (e.g., a loss of coolant accident), the ceramic coating must minimize or at least significantly delay oxidation of the Zr-alloy cladding, thus reducing the amount of hydrogen generated and the oxygen ingress into the cladding. The specific objectivesmore » of this project are as follows: To produce durable ceramic coatings on Zr-alloy clad using two possible routes: (i) MAX phase ceramic coatings or similar nitride or carbide coatings; and (ii) graded interface architecture (multilayer) ceramic coatings, using, for instance, an oxide such as yttria-stabilized zirconia (YSZ) as the outer protective layer. To characterize the structural and physical properties of the coated clad samples produced in 1. above, especially the corrosion properties under simulated normal and transient reactor operating conditions. To perform computational analyses to assess the effects of such coatings on fuel performance and reactor neutronics, and to perform fuel cycle analyses to assess the economic viability of modifying conventional Zr-alloy cladding with ceramic coatings. This project meets a number of the goals outlined in the NEUP-IRP call for proposals, including: Improve the fuel/cladding system through innovative designs (e.g. coatings/liners for zirconium-based cladding) Reduce or eliminate hydrogen generation Increase resistance to bulk steam oxidation Achievement of our goals and objectives, as defined above, will lead to safer light-water reactor (LWR) nuclear fuel assemblies, due to improved cladding properties and built-in accident resistance, as well as the possibilities for enhanced fuel/clad system performance and longevity.« less

Zirconium alloys with small amounts of iron and copper or nickel show improved corrosion resistance in superheated steam

NASA Technical Reports Server (NTRS)

Greenberg, S.; Youngdahl, C. A.

1967-01-01

Heat treating various compositions of zirconium alloys improve their corrosion resistance to superheated steam at temperatures higher than 500 degrees C. This increases their potential as fuel cladding for superheated-steam nuclear-fueled reactors as well as in autoclaves operating at modest pressures.

Severe accident modeling of a PWR core with different cladding materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, S. C.; Henry, R. E.; Paik, C. Y.

2012-07-01

The MAAP v.4 software has been used to model two severe accident scenarios in nuclear power reactors with three different materials as fuel cladding. The TMI-2 severe accident was modeled with Zircaloy-2 and SiC as clad material and a SBO accident in a Zion-like, 4-loop, Westinghouse PWR was modeled with Zircaloy-2, SiC, and 304 stainless steel as clad material. TMI-2 modeling results indicate that lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would result if SiC was substituted for Zircaloy-2 as cladding. SBO modeling results indicate that the calculated time to RCSmore » rupture would increase by approximately 20 minutes if SiC was substituted for Zircaloy-2. Additionally, when an extended SBO accident (RCS creep rupture failure disabled) was modeled, significantly lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would be generated by substituting SiC for Zircaloy-2 or stainless steel cladding. Because the rate of SiC oxidation reaction with elevated temperature H{sub 2}O (g) was set to 0 for this work, these results should be considered preliminary. However, the benefits of SiC as a more accident tolerant clad material have been shown and additional investigation of SiC as an LWR core material are warranted, specifically investigations of the oxidation kinetics of SiC in H{sub 2}O (g) over the range of temperatures and pressures relevant to severe accidents in LWR 's. (authors)« less

Pellet Cladding Mechanical Interaction Modeling Using the Extended Finite Element Method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spencer, Benjamin W.; Jiang, Wen; Dolbow, John E.

As a brittle material, the ceramic UO2 used as light water reactor fuel experiences significant fracturing throughout its life, beginning with the first rise to power of fresh fuel. This has multiple effects on the thermal and mechanical response of the fuel/cladding system. One such effect that is particularly important is that when there is mechanical contact between the fuel and cladding, cracks that extending from the outer surface of the fuel into the volume of the fuel cause elevated stresses in the adjacent cladding, which can potentially lead to cladding failure. Modeling the thermal and mechanical response of themore » cladding in the vicinity of these surface-breaking cracks in the fuel can provide important insights into this behavior to help avoid operating conditions that could lead to cladding failure. Such modeling has traditionally been done in the context of finite-element-based fuel performance analysis by modifying the fuel mesh to introduce discrete cracks. While this approach is effective in capturing the important behavior at the fuel/cladding interface, there are multiple drawbacks to explicitly incorporating the cracks in the finite element mesh. Because the cracks are incorporated in the original mesh, the mesh must be modified for cracks of specified location and depth, so it is difficult to account for crack propagation and the formation of new cracks at other locations. The extended finite element method (XFEM) has emerged in recent years as a powerful method to represent arbitrary, evolving, discrete discontinuities within the context of the finite element method. Development work is underway by the authors to implement XFEM in the BISON fuel performance code, and this capability has previously been demonstrated in simulations of fracture propagation in ceramic nuclear fuel. These preliminary demonstrations have included only the fuel, and excluded the cladding for simplicity. This paper presents initial results of efforts to apply XFEM to model stress concentrations induced by fuel fractures at the fuel/cladding interface during pellet cladding mechanical interaction (PCMI). This is accomplished by enhancing the thermal and mechanical contact enforcement algorithms employed by BISON to permit their use in conjunction with XFEM. The results from this methodology are demonstrated to be equivalent to those from using meshed discrete cracks. While the results of the two methods are equivalent for the case of a stationary crack, it is demonstrated that XFEM provides the additional flexibility of allowing arbitrary crack initiation and propagation during the analysis, and minimizes model setup effort for cases with stationary cracks.« less

Thermo-mechanical assessment of full SiC/SiC composite cladding for LWR applications with sensitivity analysis

NASA Astrophysics Data System (ADS)

Singh, Gyanender; Terrani, Kurt; Katoh, Yutai

2018-02-01

SiC/SiC composites are considered among leading candidates for accident tolerant fuel cladding in light water reactors. However, when SiC-based materials are exposed to neutron irradiation, they experience significant changes in dimensions and physical properties. Under a large heat flux application (i.e. fuel cladding), the non-uniform changes in the dimensions and physical properties will lead to build-up of stresses in the structure over the course of time. To ensure reliable and safe operation of such a structure it is important to assess its thermo-mechanical performance under in-reactor conditions of irradiation and elevated temperature. In this work, the foundation for 3D thermo-mechanical analysis of SiC/SiC cladding is put in place and a set of analyses with simplified boundary conditions has been performed. The analyses were carried out with two different codes that were benchmarked against one another and prior results in the literature. A constitutive model is constructed and solved numerically to predict the stress distribution and variation in the cladding under normal operating conditions. The dependence of dimensions and physical properties variation with irradiation and temperature has been incorporated. These robust models may now be modified to take into account the axial and circumferential variation in neutron and heat flux to fully account for 3D effects. The results from the simple analyses show the development of high tensile stresses especially in the circumferential and axial directions at the inner region of the cladding. Based on the results obtained, design guidelines are recommended. For lack of certainty in or tailor-ability for the physical and mechanical properties of SiC/SiC composite material a sensitivity analysis is conducted. The analysis results establish a precedence order of the properties based on the extent to which these properties influence the temperature and the stresses.

Capabilities to improve corrosion resistance of fuel claddings by using powerful laser and plasma sources

DOE Office of Scientific and Technical Information (OSTI.GOV)

Borisov, V. M., E-mail: borisov@triniti.ru; Trofimov, V. N.; Sapozhkov, A. Yu.

2016-12-15

The treatment conditions of fuel claddings of the E110 alloy by using powerful UV or IR laser radiation, which lead to the increase in the corrosion resistance at the high-temperature (T = 1100°C) oxidation simulating a loss-of-coolant accident, are determined. The possibility of the complete suppression of corrosion under these conditions by using pulsed laser deposition of a Cr layer is demonstrated. The behavior of protective coatings of Al, Al{sub 2}O{sub 3}, and Cr planted on steel EP823 by pulsed laser deposition, which is planned to be used in the BREST-OD-300, is studied. The methods of the almost complete suppressionmore » of corrosion in liquid lead to the temperature of 720°C are shown.« less

Cladding burst behavior of Fe-based alloys under LOCA

DOE PAGES

Terrani, Kurt A.; Dryepondt, Sebastien N.; Pint, Bruce A.; ...

2015-12-17

Burst behavior of austenitic and ferritic Fe-based alloy tubes has been examined under a simulated large break loss of coolant accident. Specifically, type 304 stainless steel (304SS) and oxidation resistant FeCrAl tubes were studied alongside Zircaloy-2 and Zircaloy-4 that are considered reference fuel cladding materials. Following the burst test, characterization of the cladding materials was carried out to gain insights regarding the integral burst behavior. Given the widespread availability of a comprehensive set of thermo-mechanical data at elevated temperatures for 304SS, a modeling framework was implemented to simulate the various processes that affect burst behavior in this Fe-based alloy. Themore » most important conclusion is that cladding ballooning due to creep is negligible for Fe-based alloys. Thus, unlike Zr-based alloys, cladding cross-sectional area remains largely unchanged up to the point of burst. Furthermore, for a given rod internal pressure, the temperature onset of burst in Fe-based alloys appears to be simply a function of the alloy's ultimate tensile strength, particularly at high rod internal pressures.« less

Enhancing the ABAQUS thermomechanics code to simulate multipellet steady and transient LWR fuel rod behavior

DOE Office of Scientific and Technical Information (OSTI.GOV)

R. L. Williamson

A powerful multidimensional fuels performance analysis capability, applicable to both steady and transient fuel behavior, is developed based on enhancements to the commercially available ABAQUS general-purpose thermomechanics code. Enhanced capabilities are described, including: UO2 temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, fuel densification, fission gas release, cladding thermal and irradiation creep, cladding irradiation growth, gap heat transfer, and gap/plenum gas behavior during irradiation. This new capability is demonstrated using a 2D axisymmetric analysis of the upper section of a simplified multipellet fuel rod, during both steady and transient operation. Comparisons are made between discrete andmore » smeared-pellet simulations. Computational results demonstrate the importance of a multidimensional, multipellet, fully-coupled thermomechanical approach. Interestingly, many of the inherent deficiencies in existing fuel performance codes (e.g., 1D thermomechanics, loose thermomechanical coupling, separate steady and transient analysis, cumbersome pre- and post-processing) are, in fact, ABAQUS strengths.« less

The fractalline properties of experimentally simulated PWR fuel crud

NASA Astrophysics Data System (ADS)

Dumnernchanvanit, I.; Mishra, V. K.; Zhang, N. Q.; Robertson, S.; Delmore, A.; Mota, G.; Hussey, D.; Wang, G.; Byers, W. A.; Short, M. P.

2018-02-01

The buildup of fouling deposits on nuclear fuel rods, known as crud, continues to challenge the worldwide fleet of light water reactors (LWRs). Crud may cause serious operational problems for LWRs, including axial power shifts, accelerated fuel clad corrosion, increased primary circuit radiation dose rates, and in some instances has led directly to fuel failure. Numerous studies continue to attempt to model and predict the effects of crud, but each makes critical assumptions regarding how to treat the complex, porous microstructure of crud and its resultant effects on temperature, pressure, and crud chemistry. In this study, we demonstrate that crud is indeed a fractalline porous medium using flowing loop experiments, validating the most recent models of its effects on LWR fuel cladding. This crud is shown to match that in other LWR-prototypical facilities through a porosity-fractal dimension scaling law. Implications of this result range from post-mortem analysis of the effects of crud on reactor fuel performance, to utilizing crud's fractalline dimensions to quantify the effectiveness of anti-fouling measures.

FISSILE MATERIAL AND FUEL ELEMENTS FOR NEUTRONIC REACTORS

DOEpatents

Shaner, B.E.

1961-08-15

The fissile material consists of about 64 to 70% (weight) zirconium dioxide, 15 to 19% uranium dioxide, and 8 to 17% calcium oxide. The fissile material is formed into sintered composites which are disposed in a compartmented fuel element, comprising essentially a flat filler plate having a plurality of compartments therein, enclosed in cladding plates of the same material as the filler plate. The resultant fuel has good resistance to corrosion in high temperature pressurized water, good dimensional stability to elevated temperatures, and good resistance to thermal shock. (AEC)

Nuclear reactor fuel element with vanadium getter on cladding

DOEpatents

Johnson, Carl E.; Carroll, Kenneth G.

1977-01-01

A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of vanadium as an oxygen getter on the inner surface of the cladding. The vanadium reacts with oxygen released by the fissionable material during irradiation of the core to prevent the oxygen from reacting with and corroding the cladding. Also described is a method for coating the inner surface of small diameter tubes of cladding with a layer of vanadium.

Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

NASA Astrophysics Data System (ADS)

Gan, J.; Miller, B. D.; Keiser, D. D.; Jue, J. F.; Madden, J. W.; Robinson, A. B.; Ozaltun, H.; Moore, G.; Meyer, M. K.

2017-08-01

Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (<20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. Different methods have been employed to fabricate monolithic fuel plates, including hot-rolling with no cold-rolling. L1P09T is a hot-rolled fuel plate irradiated to high fission density in the RERTR-9B experiment. This paper discusses the TEM characterization results for this U-10Mo/Zr/Al6061 monolithic fuel plate (∼59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 °C, respectively. TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (>1 μm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ∼30 at% and ∼7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

Microstructural characteristics of HIP-bonded monolithic nuclear fuels with a diffusion barrier

NASA Astrophysics Data System (ADS)

Jue, Jan-Fong; Keiser, Dennis D.; Breckenridge, Cynthia R.; Moore, Glenn A.; Meyer, Mitchell K.

2014-05-01

Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative is developing an advanced monolithic fuel to convert US high-performance research reactors to low-enriched uranium. Hot-isostatic-press (HIP) bonding was the single process down-selected to bond monolithic U-Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U-Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between the fuel meat, the cladding, and the diffusion barrier, as well as between the U-10Mo fuel meat and the Al-6061 cladding, were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are:

Microstructural Characteristics of HIP-bonded Monolithic Nuclear Fuels with a Diffusion Barrier

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jan-Fong Jue; Dennis D. Keiser, Jr.; Cynthia R. Breckenridge

Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative (GTRI) is developing an advanced monolithic fuel to convert US high performance research reactors to low-enriched uranium. Hot-isostatic-press bonding was the single process down-selected to bond monolithic U-Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U–Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between fuel meat, cladding, and diffusion barrier, as well as U–10Momore » fuel meat and Al–6061 cladding were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are • A typical Zr diffusion barrier of thickness 25 µm • Transverse cross section that exhibits relatively equiaxed grains with an average grain diameter of 10 µm • Chemical banding, in some areas more than 100 µm in length, that is very pronounced in longitudinal (i.e., rolling) direction with Mo concentration varying from 7–13 wt% • Decomposed areas containing plate-shaped low-Mo phase • A typical Zr/cladding interaction layer of thickness 1-2 µm • A visible UZr2 bearing layer of thickness 1-2 µm • Mo-rich precipitates (mainly Mo2Zr, forming a layer in some areas) followed by a Mo-depleted sub-layer between the visible UZr2-bearing layer and the U–Mo matrix • No excessive interaction between cladding and the uncoated fuel edge • Cladding-to-cladding bonding that exhibits no cracks or porosity with second phases high in Mg, Si, and O decorating the bond line. • Some of these attributes might be critical to the irradiation performance of monolithic U-10Mo nuclear fuel. There are several issues or concerns that warrant more detailed study, such as precipitation along cladding-to-cladding bond line, chemical banding, uncovered fuel-zone edge, and interaction layer between U–Mo fuel meat and zirconium. Future post-irradiation examination results will focus, among other things, on identifying in-reactor failure mechanisms and, eventually, directing further fresh fuel characterization efforts.« less

Prognosis and comparison of performances of composite CERCER and CERMET fuels dedicated to transmutation of TRU in an EFIT ADS

NASA Astrophysics Data System (ADS)

Sobolev, V.; Uyttenhove, W.; Thetford, R.; Maschek, W.

2011-07-01

The neutronic and thermomechanical performances of two composite fuel systems: CERCER with (Pu,Np,Am,Cm)O 2-x fuel particles in ceramic MgO matrix and CERMET with metallic Mo matrix, selected for transmutation of minor actinides in the European Facility for Industrial Transmutation (EFIT), were analysed aiming at their optimisation. The ALEPH burnup code system, based on MNCPX and ORIGEN codes and JEFF3.1 nuclear data library, and the modern version of the fuel rod performance code TRAFIC were used for this analysis. Because experimental data on the properties of the mixed minor-actinide oxides are scarce, and the in-reactor behaviour of the T91 steel chosen as cladding, as well as of the corrosion protective layer, is still not well-known, a set of "best estimates" provided the properties used in the code. The obtained results indicate that both fuel candidates, CERCER and CERMET, can satisfy the fuel design and safety criteria of EFIT. The residence time for both types of fuel elements can reach about 5 years with the reactivity swing within ±1000 pcm, and about 22% of the loaded MA is transmuted during this period. However, the fuel centreline temperature in the hottest CERCER fuel rod is close to the temperature above which MgO matrix becomes chemically instable. Moreover, a weak PCMI can appear in about 3 years of operation. The CERMET fuel can provide larger safety margins: the fuel temperature is more than 1000 K below the permitted level of 2380 K and the pellet-cladding gap remains open until the end of operation.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lombardo, N.J.; Marseille, T.J.; White, M.D.

TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic inmore » form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000{degree}F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion ( bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled.« less

Development of new ferritic steels as cladding material for metallic fuel fast breeder reactor

NASA Astrophysics Data System (ADS)

Tokiwai, Moriyasu; Horie, Masaaki; Kako, Kenji; Fujiwara, Masayuki

1993-09-01

The excellent thermal, chemical and neutronic properties of metallic fuel (U-Pu-Zr alloy) will lead to drastic improvements in fast reactor safety and the related fuel cycle economy. Some new high molybdenum 12Cr ferritic stainless steel candidate cladding alloys have been designed to achieve the mechanical properties required for high performance metallic fuel elements. These candidate claddings were irradiated by ion bombardment and tested to determine their strength and creep rupture properties. A 12Cr-8Mo and a 12Cr-8Mo-0.1Y 2O 3 steel were fabricated into cladding via a powder metallurgy process and by a mechanical alloying process, respectively. These claddings had two and three times the creep rupture strength (pressurized at 650°C for 10000 h) of a conventional 12Cr ferritic steel (HT-9). These two steels also showed no void formation up to 350 dpa by Ni 3+ irradiation. A zircaloy-2 lined steel cladding tube has also been fabricated for the purpose of reducing fuel-cladding interdiffusion and chemical interaction.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jensen, Colby B.; Folsom, Charles P.; Davis, Cliff B.

Experimental testing in the Multi-Static Environment Rodlet Transient Test Apparatus (SERTTA) will lead the rebirth of transient fuel testing in the United States as part of the Accident Tolerant Fuels (ATF) progam. The Multi-SERTTA is comprised of four isolated pressurized environments capable of a wide variety of working fluids and thermal conditions. Ultimately, the TREAT reactor as well as the Multi-SERTTA test vehicle serve the purpose of providing desired thermal-hydraulic boundary conditions to the test specimen. The initial ATF testing in TREAT will focus on reactivity insertion accident (RIA) events using both gas and water environments including typical PWR operatingmore » pressures and temperatures. For the water test environment, a test configuration is envisioned using the expansion tank as part of the gas-filled expansion volume seen by the test to provide additional pressure relief. The heat transfer conditions during the high energy power pulses of RIA events remains a subject of large uncertainty and great importance for fuel performance predictions. To support transient experiments, the Multi-SERTTA vehicle has been modeled using RELAP5 with a baseline test specimen composed of UO2 fuel in zircaloy cladding. The modeling results show the influence of the designs of the specimen, vehicle, and transient power pulses. The primary purpose of this work is to provide input and boundary conditions to fuel performance code BISON. Therefore, studies of parameters having influence on specimen performance during RIA transients are presented including cladding oxidation, power pulse magnitude and width, cladding-to-coolant heat fluxes, fuel-to-cladding gap, transient boiling effects (modified CHF values), etc. The results show the great flexibility and capacity of the TREAT Multi-SERTTA test vehicle to provide testing under a wide range of prototypic thermal-hydraulic conditions as never done before.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dionne, B.; Tzanos, C. P.

To support the safety analyses required for the conversion of the Belgian Reactor 2 (BR2) from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, the simulation of a number of loss-of-flow tests, with or without loss of pressure, has been undertaken. These tests were performed at BR2 in 1963 and used instrumented fuel assemblies (FAs) with thermocouples (TC) imbedded in the cladding as well as probes to measure the FAs power on the basis of their coolant temperature rise. The availability of experimental data for these tests offers an opportunity to better establish the credibility of the RELAP5-3D model andmore » methodology used in the conversion analysis. In order to support the HEU to LEU conversion safety analyses of the BR2 reactor, RELAP simulations of a number of loss-of-flow/loss-of-pressure tests have been undertaken. Preliminary analyses showed that the conservative power distributions used historically in the BR2 RELAP model resulted in a significant overestimation of the peak cladding temperature during the transient. Therefore, it was concluded that better estimates of the steady-state and decay power distributions were needed to accurately predict the cladding temperatures measured during the tests and establish the credibility of the RELAP model and methodology. The new approach ('best estimate' methodology) uses the MCNP5, ORIGEN-2 and BERYL codes to obtain steady-state and decay power distributions for the BR2 core during the tests A/400/1, C/600/3 and F/400/1. This methodology can be easily extended to simulate any BR2 core configuration. Comparisons with measured peak cladding temperatures showed a much better agreement when power distributions obtained with the new methodology are used.« less

Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gan, J.; Miller, B. D.; Keiser, D. D.

Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (< 20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. This paper discusses the TEM results of the U-10Mo/Zr/Al6061 monolithic fuel plate (Plate ID: L1P09T, ~ 59% U-235 enrichment) irradiated in Advancedmore » Test Reactor at Idaho National Laboratory as part of RERTR-9B irradiation campaign with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 C, respectively. A total of 5 TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (> 1 µm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ~ 30 at% and ~ 7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.« less

Microstructural Characterization of High Burn-up Mixed Oxide Fast Reactor Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Melissa C. Teague; Brian P. Gorman; Steven L. Hayes

2013-10-01

High burn-up mixed oxide fuel with local burn-ups of 3.4–23.7% FIMA (fissions per initial metal atom) were destructively examined as part of a research project to understand the performance of oxide fuel at extreme burn-ups. Optical metallography of fuel cross-sections measured the fuel-to-cladding gap, clad thickness, and central void evolution in the samples. The fuel-to-cladding gap closed significantly in samples with burn-ups below 7–9% FIMA. Samples with burn-ups in excess of 7–9% FIMA had a reopening of the fuel-to-cladding gap and evidence of joint oxide-gain (JOG) formation. Signs of axial fuel migration to the top of the fuel column weremore » observed in the fuel pin with a peak burn-up of 23.7% FIMA. Additionally, high burn-up structure (HBS) was observed in the two highest burn-up samples (23.7% and 21.3% FIMA). The HBS layers were found to be 3–5 times thicker than the layers found in typical LWR fuel. The results of the study indicate that formation of JOG and or HBS prevents any significant fuel-cladding mechanical interaction from occurring, thereby extending the potential life of the fuel elements.« less

Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Qu, Jun; Cooley, Kevin M.; Shaw, Austin H.

In the cores of pressurized water nuclear reactors, water-flow induced vibration is known to cause claddings on the fuel rods to rub against their supporting grids. Such grid-to-rod-fretting (GTRF) may lead to fretting wear-through and the leakage of radioactive species. The surfaces of actual zirconium alloy claddings in a reactor are inevitably oxidized in the high-temperature pressurized water, and some claddings are even pre-oxidized. As a result, the wear process of the surface oxide film is expected to be quite different from the zirconium alloy substrate. In this paper, we attempt to measure the wear coefficients of zirconium claddings withoutmore » and with pre-oxidation rubbing against grid samples using a bench-scale fretting tribometer. Results suggest that the volumetric wear coefficient of the pre-oxidized cladding is 50 to 200 times lower than that of the untreated cladding. In terms of the linear rate of wear depth, the pre-oxidized alloy wears about 15 times more slowly than the untreated cladding. Finally, fitted with the experimentally-determined wear rates, a stage-wise GTRF engineering wear model demonstrates good agreement with in-reactor experience in predicting the trend of cladding lives.« less

Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation

DOE PAGES

Qu, Jun; Cooley, Kevin M.; Shaw, Austin H.; ...

2016-03-16

In the cores of pressurized water nuclear reactors, water-flow induced vibration is known to cause claddings on the fuel rods to rub against their supporting grids. Such grid-to-rod-fretting (GTRF) may lead to fretting wear-through and the leakage of radioactive species. The surfaces of actual zirconium alloy claddings in a reactor are inevitably oxidized in the high-temperature pressurized water, and some claddings are even pre-oxidized. As a result, the wear process of the surface oxide film is expected to be quite different from the zirconium alloy substrate. In this paper, we attempt to measure the wear coefficients of zirconium claddings withoutmore » and with pre-oxidation rubbing against grid samples using a bench-scale fretting tribometer. Results suggest that the volumetric wear coefficient of the pre-oxidized cladding is 50 to 200 times lower than that of the untreated cladding. In terms of the linear rate of wear depth, the pre-oxidized alloy wears about 15 times more slowly than the untreated cladding. Finally, fitted with the experimentally-determined wear rates, a stage-wise GTRF engineering wear model demonstrates good agreement with in-reactor experience in predicting the trend of cladding lives.« less

2nd Gen FeCrAl ODS Alloy Development For Accident-Tolerant Fuel Cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dryepondt, Sebastien N.; Massey, Caleb P.; Edmondson, Philip D.

Extensive research at ORNL aims at developing advanced low-Cr high strength FeCrAl alloys for accident tolerant fuel cladding. One task focuses on the fabrication of new low Cr oxide dispersion strengthened (ODS) FeCrAl alloys. The first Fe-12Cr-5Al+Y 2O 3 (+ ZrO 2 or TiO 2) ODS alloys exhibited excellent tensile strength up to 800 C and good oxidation resistance in steam up to 1400 C, but very limited plastic deformation at temperature ranging from room to 800 C. To improve alloy ductility, several fabrication parameters were considered. New Fe-10-12Cr-6Al gas-atomized powders containing 0.15 to 0.5wt% Zr were procured and ballmore » milled for 10h, 20h or 40h with Y2O3. The resulting powder was then extruded at temperature ranging from 900 to 1050 C. Decreasing the ball milling time or increasing the extrusion temperature changed the alloy grain size leading to lower strength but enhanced ductility. Small variations of the Cr, Zr, O and N content did not seem to significantly impact the alloy tensile properties, and, overall, the 2nd gen ODS FeCrAl alloys showed significantly better ductility than the 1st gen alloys. Tube fabrication needed for fuel cladding will require cold or warm working associated with softening heat treatments, work was therefore initiated to assess the effect of these fabrications steps on the alloy microstructure and properties. This report has been submitted as fulfillment of milestone M3FT 16OR020202091 titled, Report on 2nd Gen FeCrAl ODS Alloy Development for the Department of Energy Office of Nuclear Energy, Advanced Fuel Campaign of the Fuel Cycle R&D program.« less

Nuclear reactor fuel element

DOEpatents

Johnson, Carl E.; Crouthamel, Carl E.

1980-01-01

A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of oxygen gettering material on the inner surface of the cladding. The gettering material reacts with oxygen released by the fissionable material during irradiation of the core thereby preventing the oxygen from reacting with and corroding the cladding. Also described is an improved method for coating the inner surface of the cladding with a layer of gettering material.

Pre-irradiation testing and analysis to support the LWRS Hybrid SiC-CMC-Zircaloy-04 unfueled rodlet irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Isabella J van Rooyen

2012-09-01

Nuclear fuel performance is a significant driver of nuclear power plant operational performance, safety, economics and waste disposal requirements. The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Pathway focuses on improving the scientific knowledge basis to enable the development of high-performance, high burn-up fuels with improved safety and cladding integrity and improved nuclear fuel cycle economics. To achieve significant improvements, fundamental changes are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction.

Pre-irradiation testing and analysis to support the LWRS Hybrid SiC-CMC-Zircaloy-04 unfueled rodlet irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Isabella J van Rooyen

2013-01-01

Nuclear fuel performance is a significant driver of nuclear power plant operational performance, safety, economics and waste disposal requirements. The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Pathway focuses on improving the scientific knowledge basis to enable the development of high-performance, high burn-up fuels with improved safety and cladding integrity and improved nuclear fuel cycle economics. To achieve significant improvements, fundamental changes are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ingham, J.G.

The IDENT 1578 container, which is a 110-in. long 5.5-in. OD tube, is designed for shipping FFTF fuel elements in T-3 casks between HEDL, HFEF, and other laboratories. The thermal analysis was conducted to evaluate whether or not the container satisfies its thermal design criteria (handle a decay heat load of 600 watts, max fuel pin cladding temperature not exceeding 800/sup 0/F).

77 FR 24539 - Virginia Electric and Power Company; Surry Power Station Units 1 and 2; Independent Spent Fuel...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-04-24

... bounding thermal analysis using ANSYS finite element software to evaluate the misloading events. The ANSYS analysis consists of a half-symmetric, three-dimensional model of a 32PTH DSC with a number of conservative... the maximum fuel cladding temperature presented in the UFSAR analysis dated October 2, 2009, with the...

History of Resistance Welding Oxide Dispersion Strengthened Cladding and other High Temperature Materials at Center for Advanced Energy Studies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Larry Zirker; Nathan Jerred; Dr. Indrajit Charit

2012-03-01

Research proposal 08-1079, 'A Comparative Study of Welded ODS Cladding Materials for AFCI/GNEP,' was funded in 2008 under an Advanced Fuel Cycle Initiative (AFCI) Research and Development Funding Opportunity, number DE-PS07-08ID14906. Th proposal sought to conduct research on joining oxide dispersion strengthen (ODS) tubing material to a solid end plug. This document summarizes the scientific and technical progress achieved during the project, which ran from 2008 to 2011.

Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

NASA Astrophysics Data System (ADS)

Venkiteswaran, C. N.; Jayaraj, V. V.; Ojha, B. K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B. P. C.; Kasiviswanathan, K. V.; Jayakumar, T.

2014-06-01

The 500 MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100 GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450 W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GWd/t. Fission product distribution, swelling and fuel-clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel-clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.

Use of the Hugoniot elastic limit in laser shockwave experiments to relate velocity measurements

NASA Astrophysics Data System (ADS)

Smith, James A.; Lacy, Jeffrey M.; Lévesque, Daniel; Monchalin, Jean-Pierre; Lord, Martin

2016-02-01

The US National Nuclear Security Agency has a Global Threat Reduction Initiative (GTRI) with the goal of reducing the worldwide use of high-enriched uranium (HEU). A salient component of that initiative is the conversion of research reactors from HEU to low enriched uranium (LEU) fuels. An innovative fuel is being developed to replace HEU in high-power research reactors. The new LEU fuel is a monolithic fuel made from a U-Mo alloy foil encapsulated in Al-6061 cladding. In order to support the fuel qualification process, the Laser Shockwave Technique (LST) is being developed to characterize the clad-clad and fuel-clad interface strengths in fresh and irradiated fuel plates. This fuel-cladding interface qualification will ensure the survivability of the fuel plates in the harsh reactor environment even under abnormal operating conditions. One of the concerns of the project is the difficulty of calibrating and standardizing the laser shock technique. An analytical study under development and experimental testing supports the hypothesis that the Hugoniot Elastic Limit (HEL) in materials can be a robust and simple benchmark to compare stresses generated by different laser shock systems.

Effects of hydrogen on thermal creep behaviour of Zircaloy fuel cladding

NASA Astrophysics Data System (ADS)

Suman, Siddharth; Khan, Mohd Kaleem; Pathak, Manabendra; Singh, R. N.

2018-01-01

Zirconium alloys are extensively used for nuclear fuel cladding. Creep is one of the most likely degradation mechanisms for fuel cladding during reactor operating and repository conditions. Fuel cladding tubes undergo waterside corrosion during service and hydrogen is produced as a result of it-a fraction of which is picked up by cladding. Hydrogen remains in solid solution up to terminal solid solubility and it precipitates as brittle hydride phase in the zirconium metal matrix beyond this limiting concentration. Hydrogen, either in solid solution or as precipitated hydride, alters the creep behaviour of zirconium alloys. The present article critically reviews the influence of hydrogen on thermal creep behaviour of zirconium alloys, develops the systematic understanding of this multifaceted phenomenon, and delineates the thrust areas which require further investigations.

Overview of the U.S. DOE Accident Tolerant Fuel Development Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jon Carmack; Frank Goldner; Shannon M. Bragg-Sitton

2013-09-01

The United States Fuel Cycle Research and Development Advanced Fuels Campaign has been given the responsibility to conduct research and development on enhanced accident tolerant fuels with the goal of performing a lead test assembly or lead test rod irradiation in a commercial reactor by 2022. The Advanced Fuels Campaign has defined fuels with enhanced accident tolerance as those that, in comparison with the standard UO2-Zircaloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining ormore » improving the fuel performance during normal operations and operational transients, as well as design-basis and beyond design-basis events. This paper provides an overview of the FCRD Accident Tolerant Fuel program. The ATF attributes will be presented and discussed. Attributes identified as potentially important to enhance accident tolerance include reduced hydrogen generation (resulting from cladding oxidation), enhanced fission product retention under severe accident conditions, reduced cladding reaction with high-temperature steam, and improved fuel-cladding interaction for enhanced performance under extreme conditions. To demonstrate the enhanced accident tolerance of candidate fuel designs, metrics must be developed and evaluated using a combination of design features for a given LWR design, potential improvements to that design, and the design of an advanced fuel/cladding system. The aforementioned attributes provide qualitative guidance for parameters that will be considered for fuels with enhanced accident tolerance. It may be unnecessary to improve in all attributes and it is likely that some attributes or combination of attributes provide meaningful gains in accident tolerance, while others may provide only marginal benefits. Thus, an initial step in program implementation will be the development of quantitative metrics. A companion paper in these proceedings provides an update on the status of establishing these quantitative metrics for accident tolerant LWR fuel.1 The United States FCRD Advanced Fuels Campaign has embarked on an aggressive schedule for development of enhanced accident tolerant LWR fuels. The goal of developing such a fuel system that can be deployed in the U.S. LWR fleet in the next 10 to 20 years supports the sustainability of clean nuclear power generation in the United States.« less

Oxide particle size distribution from shearing irradiated and unirradiated LWR fuels in Zircaloy and stainless steel cladding: significance for risk assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Davis, W. Jr.; West, G.A.; Stacy, R.G.

1979-03-22

Sieve fractionation was performed with oxide particles dislodged during shearing of unirradiated or irradiated fuel bundles or single rods of UO/sub 2/ or 96 to 97% ThO/sub 2/--3 to 4% UO/sub 2/. Analyses of these data by nonlinear least-squares techniques demonstrated that the particle size distribution is lognormal. Variables involved in the numerical analyses include lognormal median size, lognormal standard deviation, and shear cut length. Sieve-fractionation data are presented for unirradiated bundles of stainless-steel-clad or Zircaloy-2-clad UO/sub 2/ or ThO/sub 2/--UO/sub 2/ sheared into lengths from 0.5 to 2.0 in. Data are also presented for irradiated single rods (sheared intomore » lengths of 0.25 to 2.0 in.) of Zircaloy-2-clad UO/sub 2/ from BWRs and of Zircaloy-4-clad UO/sub 2/ from PWRs. Median particle sizes of UO/sub 2/ from shearing irradiated stainless-steel-clad fuel ranged from 103 to 182 ..mu..m; particle sizes of ThO/sub 2/--UO/sub 2/, under these same conditions, ranged from 137 to 202 ..mu..m. Similarly, median particle sizes of UO/sub 2/ from shearing unirradiated Zircaloy-2-clad fuel ranged from 230 to 957 ..mu..m. Irradiation levels of fuels from reactors ranged from 9,000 to 28,000 MWd/MTU. In general, particle sizes from shearing these irradiated fuels are larger than those from the unirradiated fuels; however, unirradiated fuel from vendors was not available for performing comparative shearing experiments. In addition, variations in particle size parameters pertaining to samples of a single vendor varied as much as those between different vendors. The fraction of fuel dislodged from the cladding is nearly proportional to the reciprocal of the shear cut length, until the cut length attains some minimum value below which all fuel is dislodged. Particles of fuel are generally elongated with a long-to-short axis ratio usually less than 3. Using parameters of the lognormal distribution estimates can be made of fractions of dislodged fuel having dimensions less than specified values.« less

Uranium dioxide fuel cladding strain investigation with the use of CYGRO-2 computer program

NASA Technical Reports Server (NTRS)

Smith, J. R.

1973-01-01

Previously irradiated UO2 thermionic fuel pins in which gross fuel-cladding strain occurred were modeled with the use of a computer program to define controlling parameters which may contribute to cladding strain. The computed strain was compared with measured strain, and the computer input data were studied in an attempt to get agreement with measured strain. Because of the limitations of the program and uncertainties in input data, good agreement with measured cladding strain was not attained. A discussion of these limitations is presented.

Reactor Physics Assessment of Thick Silicon Carbide Clad PWR Fuels

DTIC Science & Technology

2013-06-01

Densities ............................................................................................................ 21 2.3 Fuel Mass (Core Total...70 7.1 Geometry, Material Density, and Mass Summary for All Cores...21 Table 3: Fuel Rod Masses for Different Clads

Microstructure Instability of Candidate Fuel Cladding Alloys: Corrosion and Stress Corrosion Cracking Implications

NASA Astrophysics Data System (ADS)

Jiao, Yinan; Zheng, Wenyue; Guzonas, David; Kish, Joseph

2016-02-01

This paper addresses some of the overarching aspects of microstructure instability expected from both high temperature and radiation exposure that could affect the corrosion and stress corrosion cracking (SCC) resistance of the candidate austenitic Fe-Cr-Ni alloys being considered for the fuel cladding of the Canadian supercritical water-cooled reactor (SCWR) concept. An overview of the microstructure instability expected by both exposures is presented prior to turning the focus onto the implications of such instability on the corrosion and SCC resistance. Results from testing conducted using pre-treated (thermally-aged) Type 310S stainless steel to shed some light on this important issue are included to help identify the outstanding corrosion resistance assessment needs.

Between-cycle laser system for depressurization and resealing of modified design nuclear fuel assemblies

DOEpatents

Bradley, John G.

1982-01-01

A laser beam is used to puncture fuel cladding for release of contained pressurized fission gas from plenum sections or irradiated fuel pins. Exhausted fission gases are collected and trapped for safe disposal. The laser beam, adjusted to welding mode, is subsequently used to reseal the puncture holes. The fuel assembly is returned to additional irradiation or, if at end of reactivity lifetime, is routed to reprocess. The fuel assembly design provides graded cladding lengths, by rows or arrays, such that the cladding of each component fuel element of the assembly is accessible to laser beam reception.

Early implementation of SiC cladding fuel performance models in BISON

DOE Office of Scientific and Technical Information (OSTI.GOV)

Powers, Jeffrey J.

2015-09-18

SiC-based ceramic matrix composites (CMCs) [5–8] are being developed and evaluated internationally as potential LWR cladding options. These development activities include interests within both the DOE-NE LWR Sustainability (LWRS) Program and the DOE-NE Advanced Fuels Campaign. The LWRS Program considers SiC ceramic matrix composites (CMCs) as offering potentially revolutionary gains as a cladding material, with possible benefits including more efficient normal operating conditions and higher safety margins under accident conditions [9]. Within the Advanced Fuels Campaign, SiC-based composites are a candidate ATF cladding material that could achieve several goals, such as reducing the rates of heat and hydrogen generation duemore » to lower cladding oxidation rates in HT steam [10]. This work focuses on the application of SiC cladding as an ATF cladding material in PWRs, but these work efforts also support the general development and assessment of SiC as an LWR cladding material in a much broader sense.« less

Double-clad nuclear fuel safety rod

DOEpatents

McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

1984-01-01

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

Double-clad nuclear-fuel safety rod

DOEpatents

McCarthy, W.H.; Atcheson, D.B.

1981-12-30

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

78 FR 40200 - Duke Energy Carolinas, LLC, Oconee Nuclear Station Units 1, 2, and 3; Independent Spent Fuel...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-03

... breaches.'' Zircaloy is a type of zirconium alloy which includes both Zircaloy-2 and Zircaloy-4 cladding, but does not include M5 cladding. The M5 is a different type of zirconium alloy, which does not... ``zirconium alloy'' clad spent fuel assemblies in the 24PHB DSC, which would include both the ``zircaloy clad...

Advanced Steels for Accident Tolerant Fuel Cladding in Current Light Water Reactors

NASA Astrophysics Data System (ADS)

Rebak, Raul B.

After the March 2011 Fukushima events, the U.S. Congress directed the Department of Energy (DOE) to focus efforts on the development of fuel cladding materials with enhanced accident tolerance. In comparison with the stand-ard UO2-Zirconium based system, the new fuels need to tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operation conditions. Advanced steels such as iron-chromium-aluminum (FeCrAl) alloys are being investigated for degradation behavior both under normal operation conditions in high temperature water (e.g. 288°C) and under accident conditions for reaction with steam up to 1400°C. Commercial and experimental alloys were tested for several periods of time in 100% superheated steam from 800°C to 1475°C. Results show that FeCrAl alloys significantly outperform the resistance in steam of the current zirconium alloys.

MTR plates modeling with MAIA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marelle, V.; Dubois, S.; Ripert, M.

2008-07-15

MAIA is a thermo-mechanical code dedicated to the modeling of MTR fuel plates. The main physical phenomena modeled in the code are the cladding oxidation, the interaction between fuel and Al-matrix, the swelling due to fission products and the Al/fuel particles interaction. The creeping of the plate can be modeled in the mechanical calculation. MAIA has been validated on U-Mo dispersion fuel experiments such as IRIS 1 and 2 and FUTURE. The results are in rather good agreement with post-irradiation examinations. MAIA can also be used to calculate in-pile behavior of U{sub 3}Si{sub 2} plates as in the SHARE experimentmore » irradiated in the SCK/Mol BR2 reactor. The main outputs given by MAIA throughout the irradiation are temperatures, cladding oxidation thickness, interaction thickness, volume fraction of meat constituents, swelling, displacements, strains and stresses. MAIA is originally a two-dimensional code but a three-dimensional version is currently under development. (author)« less

FUEL ELEMENT CONSTRUCTION

DOEpatents

Zumwalt, L.R.

1961-08-01

Fuel elements having a solid core of fissionable material encased in a cladding material are described. A conversion material is provided within the cladding to react with the fission products to form stable, relatively non- volatile compounds thereby minimizing the migration of the fission products into the coolant. The conversion material is preferably a metallic fluoride, such as lead difluoride, and may be in the form of a coating on the fuel core or interior of the cladding, or dispersed within the fuel core. (AEC)

Comparison of thermal compatibility between atomized and comminuted U{sub 3}Si dispersion fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ryu, Woo-Seog; Park, Jong-Man; Kim, Chang-Kyu

1997-08-01

Thermal compatibility of atomized U{sub 3}Si dispersion fuels were evaluated up to 2600 hours in the temperature range from 250 to 500{degrees}C, and compared with that of comminuted U{sub 3}Si. Atomized U{sub 3}Si showed better performance in terms of volume expansion of fuel meats. The reaction zone of U{sub 3}Si and Al occurred along the grain boundaries and deformation bands in U{sub 3}Si particles. Pores around fuel particles appeared at high temperature or after long-term annealing tests to remain diffusion paths over the trench of the pores. The constraint effects of cladding on fuel rod suppressed the fuel meat, andmore » reduced the volume expansion.« less

Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Heidet, F.; Kim, T.; Grandy, C.

2012-07-30

Although thorium has long been considered as an alternative to uranium-based fuels, most of the reactors built to-date have been fueled with uranium-based fuel with the exception of a few reactors. The decision to use uranium-based fuels was initially made based on the technology maturity compared to thorium-based fuels. As a result of this experience, lot of knowledge and data have been accumulated for uranium-based fuels that made it the predominant nuclear fuel type for extant nuclear power. However, following the recent concerns about the extent and availability of uranium resources, thorium-based fuels have regained significant interest worldwide. Thorium ismore » more abundant than uranium and can be readily exploited in many countries and thus is now seen as a possible alternative. As thorium-based fuel technologies mature, fuel conversion from uranium to thorium is expected to become a major interest in both thermal and fast reactors. In this study the feasibility of fuel conversion in a fast reactor is assessed and several possible approaches are proposed. The analyses are performed using the Advanced Fast Reactor (AFR-100) design, a fast reactor core concept recently developed by ANL. The AFR-100 is a small 100 MW{sub e} reactor developed under the US-DOE program relying on innovative fast reactor technologies and advanced structural and cladding materials. It was designed to be inherently safe and offers sufficient margins with respect to the fuel melting temperature and the fuel-cladding eutectic temperature when using U-10Zr binary metal fuel. Thorium-based metal fuel was preferred to other thorium fuel forms because of its higher heavy metal density and it does not need to be alloyed with zirconium to reduce its radiation swelling. The various approaches explored cover the use of pure thorium fuel as well as the use of thorium mixed with transuranics (TRU). Sensitivity studies were performed for the different scenarios envisioned in order to determine the best core performance characteristics for each of them. With the exception of the fuel type and enrichment, the reference AFR-100 core design characteristics were kept unchanged, including the general core layout and dimensions, assembly dimensions, materials and power rating. In addition, the mass of {sup 235}U required was kept within a reasonable range from that of the reference AFR-100 design. The core performance characteristics, kinetics parameters and reactivity feedback coefficients were calculated using the ANL suite of fast reactor analysis code systems. Orifice design calculations and the steady-state thermal-hydraulic analyses were performed using the SE2-ANL code. The thermal margins were evaluated by comparing the peak temperatures to the design limits for parameters such as the fuel melting temperature and the fuel-cladding eutectic temperature. The inherent safety features of AFR-100 cores proposed were assessed using the integral reactivity parameters of the quasi-static reactivity balance analysis. The design objectives and requirements, the computation methods used as well as a description of the core concept are provided in Section 2. The three major approaches considered are introduced in Section 3 and the neutronics performances of those approaches are discussed in the same section. The orifice zoning strategies used and the steady-state thermal-hydraulic performance are provided in Section 4. The kinetics and reactivity coefficients, including the inherent safety characteristics, are provided in Section 5, and the Conclusions in Section 6. Other scenarios studied and sensitivity studies are provided in the Appendix section.« less

Nuclear fuel element

DOEpatents

Armijo, Joseph S.; Coffin, Jr., Louis F.

1983-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has a composite cladding having a substrate and a metal barrier metallurgically bonded on the inside surface of the substrate so that the metal barrier forms a shield between the substrate and the nuclear fuel material held within the cladding. The metal barrier forms about 1 to about 30 percent of the thickness of the cladding and is comprised of a low neutron absorption metal of substantially pure zirconium. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the substrate from contact and reaction with such impurities and fission products. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy. Methods of manufacturing the composite cladding are also disclosed.

Stress corrosion crack initiation of Zircaloy-4 cladding tubes in an iodine vapor environment during creep, relaxation, and constant strain rate tests

NASA Astrophysics Data System (ADS)

Jezequel, T.; Auzoux, Q.; Le Boulch, D.; Bono, M.; Andrieu, E.; Blanc, C.; Chabretou, V.; Mozzani, N.; Rautenberg, M.

2018-02-01

During accidental power transient conditions with Pellet Cladding Interaction (PCI), the synergistic effect of the stress and strain imposed on the cladding by thermal expansion of the fuel, and corrosion by iodine released as a fission product, may lead to cladding failure by Stress Corrosion Cracking (SCC). In this study, internal pressure tests were conducted on unirradiated cold-worked stress-relieved Zircaloy-4 cladding tubes in an iodine vapor environment. The goal was to investigate the influence of loading type (constant pressure tests, constant circumferential strain rate tests, or constant circumferential strain tests) and test temperature (320, 350, or 380 °C) on iodine-induced stress corrosion cracking (I-SCC). The experimental results obtained with different loading types were consistent with each other. The apparent threshold hoop stress for I-SCC was found to be independent of the test temperature. SEM micrographs of the tested samples showed many pits distributed over the inner surface, which tended to coalesce into large pits in which a microcrack could initiate. A model for the time-to-failure of a cladding tube was developed using finite element simulations of the viscoplastic mechanical behavior of the material and a modified Kachanov's damage growth model. The times-to-failure predicted by this model are consistent with the experimental data.

Basic elements of light water reactor fuel rod design. [FUELROD code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weisman, J.; Eckart, R.

1981-06-01

Basic design techniques and equations are presented to allow students to understand and perform preliminary fuel design for normal reactor conditions. Each of the important design considerations is presented and discussed in detail. These include the interaction between fuel pellets and cladding and the changes in fuel and cladding that occur during the operating lifetime of the fuel. A simple, student-oriented, fuel rod design computer program, called FUELROD, is described. The FUELROD program models the in-pile pellet cladding interaction and allows a realistic exploration of the effect of various design parameters. By use of FUELROD, the student can gain anmore » appreciation of the fuel rod design process. 34 refs.« less

BISON Modeling of Reactivity-Initiated Accident Experiments in a Static Environment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Folsom, Charles P.; Jensen, Colby B.; Williamson, Richard L.

2016-09-01

In conjunction with the restart of the TREAT reactor and the design of test vehicles, modeling and simulation efforts are being used to model the response of Accident Tolerant Fuel (ATF) concepts under reactivity insertion accident (RIA) conditions. The purpose of this work is to model a baseline case of a 10 cm long UO2-Zircaloy fuel rodlet using BISON and RELAP5 over a range of energy depositions and with varying reactor power pulse widths. The results show the effect of varying the pulse width and energy deposition on both thermal and mechanical parameters that are important for predicting failure ofmore » the fuel rodlet. The combined BISON/RELAP5 model captures coupled thermal and mechanical effects on the fuel-to-cladding gap conductance, cladding-to-coolant heat transfer coefficient and water temperature and pressure that would not be capable in each code individually. These combined effects allow for a more accurate modeling of the thermal and mechanical response in the fuel rodlet and thermal-hydraulics of the test vehicle.« less

A fission gas release correlation for uranium nitride fuel pins

NASA Technical Reports Server (NTRS)

Weinstein, M. B.; Davison, H. W.

1973-01-01

A model was developed to predict fission gas releases from UN fuel pins clad with various materials. The model was correlated with total release data obtained by different experimentors, over a range of fuel temperatures primarily between 1250 and 1660 K, and fuel burnups up to 4.6 percent. In the model, fission gas is transported by diffusion mechanisms to the grain boundaries where the volume grows and eventually interconnects with the outside surface of the fuel. The within grain diffusion coefficients are found from fission gas release rate data obtained using a sweep gas facility.

Preliminary posttest analysis of LOFT loss-of-coolant experiment L2-2

DOE Office of Scientific and Technical Information (OSTI.GOV)

White, J.R.; Grush, W.H.; Keeler, C.D.

A preliminary posttest analysis of Loss-of-Coolant Experiment (LOCE) L2-2, which was conducted in the Loss-of-Fluid Test (LOFT) facility, was performed to gain an understanding of the cause of the disparity between predicted and measured fuel rod cladding temperature responses in the LOFT core. LOCE L2-2 is the first experiment in the LOFT Power Ascension Series L2 (first series of LOFT nuclear experiments), which was designed to investigate the response of the LOFT nuclear core to the blowdown, refill, and reflood transients during LOCEs conducted at gradually increasing power levels. LOCE L2-2 was conducted at 50% power (25 MW, 26.38 kW/m).more » Results show that a core-wide rewet occurred early in the transient (during blowdown starting at about 7 s after rupture) which was not calculated in the pretest prediction analysis. This early core-wide rewet resulted in the peak fuel rod cladding temperatures being lower (by a mean value of 166/sup 0/K for 24 thermocouples) than had been calculated. This preliminary posttest analysis was concerned solely with determining why the early core-wide rewet was not predicted by the RELAP4/MOD6 pretest analysis and be no means is it a complete posttest analysis of LOCE L2-2 results. However, during this analysis, several errors made in the prettest analysis were found, and their impact on the predicted results is assessed. Three factors were postulated to have caused the disparity between predicted and measured fuel rod cladding temperatures for LOCE L2-2: (a) the initial fuel rod stored energy, (b) the heat transfer surface, and (c) the hydraulics calculation. These factors were examined and are discussed in this report. It was determined that core hydraulics, as influenced by the calculation of broken loop cold leg break flow, was the major factor causing the disparity.« less

Mechanistic Considerations Used in the Development of the PROFIT PCI Failure Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pankaskie, P. J.

A fuel Pellet-Zircaloy Cladding (thermo-mechanical-chemical) Interactions (PC!) failure model for estimating the probability of failure in !ransient increases in power (PROFIT) was developed. PROFIT is based on 1) standard statistical methods applied to available PC! fuel failure data and 2) a mechanistic analysis of the environmental and strain-rate-dependent stress versus strain characteristics of Zircaloy cladding. The statistical analysis of fuel failures attributable to PCI suggested that parameters in addition to power, transient increase in power, and burnup are needed to define PCI fuel failures in terms of probability estimates with known confidence limits. The PROFIT model, therefore, introduces an environmentalmore » and strain-rate dependent strain energy absorption to failure (SEAF) concept to account for the stress versus strain anomalies attributable to interstitial-disloction interaction effects in the Zircaloy cladding. Assuming that the power ramping rate is the operating corollary of strain-rate in the Zircaloy cladding, then the variables of first order importance in the PCI fuel failure phenomenon are postulated to be: 1. pre-transient fuel rod power, P{sub I}, 2. transient increase in fuel rod power, {Delta}P, 3. fuel burnup, Bu, and 4. the constitutive material property of the Zircaloy cladding, SEAF.« less

Studies of Lanthanide Transport in Metallic Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, Jinsuo; Taylor, Christopher

Metallic nuclear fuels were tested in fast reactor programs and performed well. However, metallic fuels have shown the phenomenon of FCCI that are due to deleterious reactions between lanthanide fission products and cladding material. As the burnup is increased, lanthanide fission products that contact with the cladding could react with cladding constituents such as iron and chrome. These reactions produce higher-melting intermetallic compounds and low-melting alloys, and weaken the mechanical integrity.

Irradiation test of tungsten clad uranium carbide-zirconium carbide ((U,Zr)C) specimens for thermionic reactor application at conditions conductive to long-term performance

NASA Technical Reports Server (NTRS)

Creagh, J. W. R.; Smith, J. R.

1973-01-01

Uranium carbide fueled, thermionic emitter configurations were encapsulated and irradiated. One capsule contained a specimen clad with fluoride derived chemically vapor deposited (CVD) tungsten. The other capsule used a duplex clad specimen consisting of chloride derived on floride derived CVD tungsten. Both fuel pins were 16 millimeters in diameter and contained a 45.7-millimeter length of fuel.

Oxidation of SUS-316 stainless steel for fast breeder reactor fuel cladding under oxygen pressure controlled by Ni/NiO oxygen buffer

NASA Astrophysics Data System (ADS)

Saito, Minoru; Furuya, Hirotaka; Sugisaki, Masayasu

1985-09-01

Oxidation of SUS-316 stainless steel for a fast breeder reactor fuel cladding was examined in the temperature range of 843-1010 K under the oxygen pressure of 1017 t - 10 t-13 Pa hy use of an experimental technique of a Ni/NiO oxygen buffer. The formation of the duplex oxide layer, i.e. an outer Fe 3O 4 layer and an inner (Fe, Cr, Ni)-spinel layer, was observed and the oxidation kinetics was found to obey the parabolic rate law. The oxygen pressure and temperature dependence of the parabolic rate constant kp( PO2, T) was determined as follows: kp( PO2, T)/ kg2 · m-1 · s-1 = 0.170( PO2/ Pa) 0.141exp[-114 × 10 3/( RT/ J)]. On the basis of the oxidation kinetics and the metallographic information, the outward diffusion of Fe in the outer oxide layer was assigned to be the rate-determining process.

Process for making a martensitic steel alloy fuel cladding product

DOEpatents

Johnson, Gerald D.; Lobsinger, Ralph J.; Hamilton, Margaret L.; Gelles, David S.

1990-01-01

This is a very narrowly defined martensitic steel alloy fuel cladding material for liquid metal cooled reactors, and a process for making such a martensitic steel alloy material. The alloy contains about 10.6 wt. % chromium, about 1.5 wt. % molybdenum, about 0.85 wt. % manganese, about 0.2 wt. % niobium, about 0.37 wt. % silicon, about 0.2 wt. % carbon, about 0.2 wt. % vanadium, 0.05 maximum wt. % nickel, about 0.015 wt. % nitrogen, about 0.015 wt. % sulfur, about 0.05 wt. % copper, about 0.007 wt. % boron, about 0.007 wt. % phosphorous, and with the remainder being essentially iron. The process utilizes preparing such an alloy and homogenizing said alloy at about 1000.degree. C. for 16 hours; annealing said homogenized alloy at 1150.degree. C. for 15 minutes; and tempering said annealed alloy at 700.degree. C. for 2 hours. The material exhibits good high temperature strength (especially long stress rupture life) at elevated temperature (500.degree.-760.degree. C.).

Complete Non-Radioactive Operability Tests for Cladding Hull Chlorination

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collins, Emory D; Johnson, Jared A.; Hylton, Tom D.

2016-04-01

Non-radioactive operability tests were made to test the metal chlorination reactor and condenser and their accessories using batch chlorinations of non-radioactive cladding samples and to identify optimum operating practices and components that need further modifications prior to installation of the equipment into the hot cell for tests on actual used nuclear fuel (UNF) cladding. The operability tests included (1) modifications to provide the desired heating and reactor temperature profile; and (2) three batch chlorination tests using, respectively, 100, 250, and 500 g of cladding. During the batch chlorinations, metal corrosion of the equipment was assessed, pressurization of the gas inletmore » was examined and the best method for maintaining solid salt product transfer through the condenser was determined. Also, additional accessing equipment for collection of residual ash and positioning of the unit within the hot cell were identified, designed, and are being fabricated.« less

Investigation of Zircaloy-2 oxidation model for SFP accident analysis

NASA Astrophysics Data System (ADS)

Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Ogawa, Chihiro; Kondo, Keietsu; Nakashima, Kazuo; Kanazawa, Toru; Tojo, Masayuki

2017-05-01

The authors previously conducted thermogravimetric analyses on Zircaloy-2 in air. By using the thermogravimetric data, an oxidation model was constructed in this study so that it can be applied for the modeling of cladding degradation in spent fuel pool (SFP) severe accident condition. For its validation, oxidation tests of long cladding tube were conducted, and computational fluid dynamics analyses using the constructed oxidation model were proceeded to simulate the experiments. In the oxidation tests, high temperature thermal gradient along the cladding axis was applied and air flow rates in testing chamber were controlled to simulate hypothetical SFP accidents. The analytical outputs successfully reproduced the growth of oxide film and porous oxide layer on the claddings in oxidation tests, and validity of the oxidation model was proved. Influence of air flow rate for the oxidation behavior was thought negligible in the conditions investigated in this study.

Nuclear fuels - Present and future

NASA Astrophysics Data System (ADS)

Olander, D.

2009-06-01

The important developments in nuclear fuels and their problems are reviewed and compared with the status of present light-water reactor fuels. The limitations of LWR fuels are reviewed with respect to important recent concerns, namely provision of outlet coolant temperatures high enough for use in H 2 production, destruction of plutonium to eliminate proliferation concerns, and burning of the minor actinides to reduce the waste repository heat load and long-term radiation hazard. In addition to current oxide-based fuel rod designs, the hydride fuel with liquid-metal thermal bonding of the fuel-cladding gap is covered. Finally, two of the most promising Generation IV reactor concepts, the very high temperature reactor and the sodium fast reactor, and the accompanying reprocessing technologies, aqueous-based UREX+1a and pyrometallurgical, are summarized. In all of the topics covered, the thermodynamics involved in the fuel's behavior under irradiation and in the reprocessing schemes are emphasized.

Fuel pin

DOEpatents

Christiansen, D.W.; Karnesky, R.A.; Leggett, R.D.; Baker, R.B.

1987-11-24

A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

Fuel pin

DOEpatents

Christiansen, David W.; Karnesky, Richard A.; Leggett, Robert D.; Baker, Ronald B.

1989-10-03

A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

Fuel pin

DOEpatents

Christiansen, David W.; Karnesky, Richard A.; Leggett, Robert D.; Baker, Ronald B.

1989-01-01

A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

Reactivity Initiated Accident Simulation to Inform Transient Testing of Candidate Advanced Cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas R; Wysocki, Aaron J; Terrani, Kurt A

2016-01-01

Abstract. Advanced cladding materials with potentially enhanced accident tolerance will yield different light water reactor performance and safety characteristics than the present zirconium-based cladding alloys. These differences are due to different cladding material properties and responses to the transient, and to some extent, reactor physics, thermal, and hydraulic characteristics. Some of the differences in reactors physics characteristics will be driven by the fundamental properties (e.g., absorption in iron for an iron-based cladding) and others will be driven by design modifications necessitated by the candidate cladding materials (e.g., a larger fuel pellet to compensate for parasitic absorption). Potential changes in thermalmore » hydraulic limits after transition from the current zirconium-based cladding to the advanced materials will also affect the transient response of the integral fuel. This paper leverages three-dimensional reactor core simulation capabilities to inform on appropriate experimental test conditions for candidate advanced cladding materials in a control rod ejection event. These test conditions are using three-dimensional nodal kinetics simulations of a reactivity initiated accident (RIA) in a representative state-of-the-art pressurized water reactor with both nuclear-grade iron-chromium-aluminum (FeCrAl) and silicon carbide based (SiC-SiC) cladding materials. The effort yields boundary conditions for experimental mechanical tests, specifically peak cladding strain during the power pulse following the rod ejection. The impact of candidate cladding materials on the reactor kinetics behavior of RIA progression versus reference zirconium cladding is predominantly due to differences in: (1) fuel mass/volume/specific power density, (2) spectral effects due to parasitic neutron absorption, (3) control rod worth due to hardened (or softened) spectrum, and (4) initial conditions due to power peaking and neutron transport cross sections in the equilibrium cycle cores due to hardened (or softened) spectrum. This study shows minimal impact of SiC-based cladding configurations on the transient response versus reference zirconium-based cladding. However, the FeCrAl cladding response indicates similar energy deposition, but with significantly shorter pulses of higher magnitude. Therefore the FeCrAl-based cases have a more rapid fuel thermal expansion rate and the resultant pellet-cladding interaction occurs more rapidly.« less

Nuclear fuel element

DOEpatents

Armijo, Joseph S.; Coffin, Jr., Louis F.

1980-04-29

A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has an improved composite cladding comprised of a moderate purity metal barrier of zirconium metallurgically bonded on the inside surface of a zirconium alloy tube. The metal barrier forms a shield between the alloy tube and a core of nuclear fuel material enclosed in the composite cladding. There is a gap between the cladding and the core. The metal barrier forms about 1 to about 30 percent of the thickness of the composite cladding and has low neutron absorption characteristics. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the alloy tube from contact and reaction with such impurities and fission products. Methods of manufacturing the composite cladding are also disclosed.

Design and Laboratory Evaluation of Future Elongation and Diameter Measurements at the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

K. L. Davis; D. L. Knudson; J. L. Rempe

New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. Such materials can undergo significant dimensional and physical changes during high temperature irradiations. In order to accurately predict these changes, real-time data must be obtained under prototypic irradiation conditions for model development and validation. To provide such data, researchers at the Idaho National Laboratory (INL) High Temperature Test Laboratory (HTTL) are developing several instrumented test rigs to obtain data real-time from specimens irradiated in well-controlled pressurized water reactor (PWR) coolant conditions in the Advanced Test Reactor (ATR). This paper reports the status ofmore » INL efforts to develop and evaluate prototype test rigs that rely on Linear Variable Differential Transformers (LVDTs) in laboratory settings. Although similar LVDT-based test rigs have been deployed in lower flux Materials Testing Reactors (MTRs), this effort is unique because it relies on robust LVDTs that can withstand higher temperatures and higher fluxes than often found in other MTR irradiations. Specifically, the test rigs are designed for detecting changes in length and diameter of specimens irradiated in ATR PWR loops. Once implemented, these test rigs will provide ATR users with unique capabilities that are sorely needed to obtain measurements such as elongation caused by thermal expansion and/or creep loading and diameter changes associated with fuel and cladding swelling, pellet-clad interaction, and crud buildup.« less

Microstructural characterization of annealed U-12Zr-4Pd and U-12Zr-4Pd-5Ln: Investigating Pd as a metallic fuel additive

NASA Astrophysics Data System (ADS)

Benson, Michael T.; He, Lingfeng; King, James A.; Mariani, Robert D.

2018-04-01

Palladium is being investigated as a potential additive to metallic fuel to control fuel-cladding chemical interaction (FCCI). A primary cause of FCCI is the lanthanide fission products moving to the fuel periphery and interacting with the cladding. This interaction will lead to wastage of the cladding and, given enough time or burn-up, eventually to a cladding breach. The current study is a scanning electron microscopy (SEM) and transmission electron microscopy (TEM) characterization of annealed U-12Zr-4Pd and U-12Zr-4Pd-5Ln, where Ln = 53Nd-25Ce-16Pr-6La. The present study shows that Pd preferentially binds the lanthanides over other fuel constituents, which may prevent lanthanide migration and interaction with the cladding during irradiation. The SEM analysis indicates the 1:1 Pd-Ln compound is being formed, while the TEM analysis, due to higher resolution, found the 1:1 compound, as well as Pd-rich compounds Pd2Ln and Pd3Ln2.

Posttest examination results of recent treat tests on metal fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holland, J.W.; Wright, A.E.; Bauer, T.H.

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

Recycle of Zirconium from Used Nuclear Fuel Cladding: A Major Element of Waste Reduction

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collins, Emory D; DelCul, Guillermo D; Terekhov, Dmitri

2011-01-01

Feasibility tests were initiated to determine if the zirconium in commercial used nuclear fuel (UNF) cladding can be recovered in sufficient purity to permit re-use, and if the recovery process can be operated economically. Initial tests are being performed with unirradiated, non-radioactive samples of various types of Zircaloy materials that are used in UNF cladding to develop the recovery process and determine the degree of purification that can be obtained. Early results indicate that quantitative recovery can be accomplished and product contamination with alloy constituents can be controlled sufficiently to meet purification requirements. Future tests with actual radioactive UNF claddingmore » are planned. The objective of current research is to determine the feasibility of recovery and recycle of zirconium from used fuel cladding wastes. Zircaloy cladding, which contains 98+% of hafnium-free zirconium, is the second largest mass, on average {approx}25 wt %, of the components in used U.S. light-water-reactor fuel assemblies. Therefore, recovery and recycle of the zirconium would enable a large reduction in geologic waste disposal for advanced fuel cycles. Current practice is to compact or grout the cladding waste and store it for subsequent disposal in a geologic repository. This paper describes results of initial tests being performed with unirradiated, non-radioactive samples of various types of Zircaloy materials that are used in UNF cladding to develop the recovery process and determine the degree of purification that can be obtained. Future tests with actual radioactive UNF cladding are planned.« less

Steam Oxidation Testing in the Severe Accident Test Station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pint, Bruce A.; McMurray, Jake W.

2016-08-01

Since 2011, Oak Ridge National Laboratory (ORNL) has been conducting high temperature steam oxidation testing of candidate alloys for accident tolerant fuel (ATF) cladding. These concepts are designed to enhance safety margins in light water reactors (LWR) during severe accident scenarios. In the US ATF community, the Severe Accident Test Station (SATS) has been evaluating candidate materials (including coatings) since 2012. Compared to the current UO 2/Zr-based alloy fuel system, alternative cladding materials need to offer slower oxidation kinetics and a smaller enthalpy of oxidation in order to significantly reduce the rate of heat and hydrogen generation in the coremore » during a coolant-limited severe accident. The steam oxidation behavior of candidate materials is a key metric in the evaluation of ATF concepts and also an important input into models. However, prior modeling work of FeCrAl cladding has used incomplete information on the physical properties of FeCrAl. Also, the steam oxidation data being collected at 1200°-1700°C is unique as no prior work has considered steam oxidation of alloys at such high temperatures. In some cases, the results have been difficult to interpret and more fundamental information is needed such as the stability of alumina in flowing steam at 1400°-1500°C. This report summarizes recent work to measure the steam oxidation kinetics of candidate alloys, the evaporation rate of alumina in steam and the development of integral data on FeCrAl compared to conventional Zr-based cladding.« less

ORNL Interim Progress Report on Hydride Reorientation CIRFT Tests

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Yan, Yong; Wang, Hong

A systematic study of H. B. Robinson (HBR) high burnup spent nuclear fuel (SNF) vibration integrity was performed in Phase I project under simulated transportation environments, using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot cell testing technology developed at Oak Ridge National Laboratory in 2013–14. The data analysis on the as-irradiated HBR SNF rods demonstrated that the load amplitude is the dominant factor that controls the fatigue life of bending rods. However, previous studies have shown that the hydrogen content and hydride morphology has an important effect on zirconium alloy mechanical properties. To address the effect of radial hydridesmore » in SNF rods, in Phase II a test procedure was developed to simulate the effects of elevated temperatures, pressures, and stresses during transfer-drying operations. Pressurized and sealed fuel segments were heated to the target temperature for a preset hold time and slow-cooled at a controlled rate. The procedure was applied to both non-irradiated/prehydrided and high-burnup Zircaloy-4 fueled cladding segments using the Nuclear Regulatory Commission-recommended 400°C maximum temperature limit at various cooling rates. Before testing high-burnup cladding, four out-of-cell tests were conducted to optimize the hydride reorientation (R) test condition with pre-hydride Zircaloy-4 cladding, which has the same geometry as the high burnup fuel samples. Test HR-HBR#1 was conducted at the maximum hoop stress of 145 MPa, at a 400°C maximum temperature and a 5°C/h cooling rate. On the other hand, thermal cycling was performed for tests HR-HBR#2, HR-HBR#3, and HR-HBR#4 to generate more radial hydrides. It is clear that thermal cycling increases the ratio of the radial hydride to circumferential hydrides. The internal pressure also has a significant effect on the radial hydride morphology. This report describes a procedure and experimental results of the four out-of-cell hydride reorientation tests of hydrided Zircaloy-4 cladding, which served as a guideline to prepare in-cell hydride reorientation samples with high burnup HBR fuel segments. This report also provides the Phase II CIRFT test data for the hydride reorientation irradiated samples. The variations in fatigue life are provided in terms of moment, equivalent stress, curvature, and equivalent strain for the tested SNFs. The CIRFT results appear to indicate that hydride reoriented treatment (HRT) have a negative effect on fatigue life, in addition to hydride reorientation effect. For HR4 specimen that had no pressurization procedure applied, the thermal annealing treatment alone showed a negative impact on the fatigue life compared to the HBR rod.« less

Developing a laser shockwave model for characterizing diffusion bonded interfaces

NASA Astrophysics Data System (ADS)

Lacy, Jeffrey M.; Smith, James A.; Rabin, Barry H.

2015-03-01

The US National Nuclear Security Agency has a Global Threat Reduction Initiative (GTRI) with the goal of reducing the worldwide use of high-enriched uranium (HEU). A salient component of that initiative is the conversion of research reactors from HEU to low enriched uranium (LEU) fuels. An innovative fuel is being developed to replace HEU in high-power research reactors. The new LEU fuel is a monolithic fuel made from a U-Mo alloy foil encapsulated in Al-6061 cladding. In order to support the fuel qualification process, the Laser Shockwave Technique (LST) is being developed to characterize the clad-clad and fuel-clad interface strengths in fresh and irradiated fuel plates. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves to characterize interfaces in nuclear fuel plates. However, because the deposition of laser energy into the containment layer on a specimen's surface is intractably complex, the shock wave energy is inferred from the surface velocity measured on the backside of the fuel plate and the depth of the impression left on the surface by the high pressure plasma pulse created by the shock laser. To help quantify the stresses generated at the interfaces, a finite element method (FEM) model is being utilized. This paper will report on initial efforts to develop and validate the model by comparing numerical and experimental results for back surface velocities and front surface depressions in a single aluminum plate representative of the fuel cladding.

TESTING AND ACCEPTANCE OF FUEL PLATES FOR RERTR FUEL DEVELOPMENT EXPERIMENTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

J.M. Wight; G.A. Moore; S.C. Taylor

2008-10-01

This paper discusses how candidate fuel plates for RERTR Fuel Development experiments are examined and tested for acceptance prior to reactor insertion. These tests include destructive and nondestructive examinations (DE and NDE). The DE includes blister annealing for dispersion fuel plates, bend testing of adjacent cladding, and microscopic examination of archive fuel plates. The NDE includes Ultrasonic (UT) scanning and radiography. UT tests include an ultrasonic scan for areas of “debonds” and a high frequency ultrasonic scan to determine the "minimum cladding" over the fuel. Radiography inspections include identifying fuel outside of the maximum fuel zone and measurements and calculationsmore » for fuel density. Details of each test are provided and acceptance criteria are defined. These tests help to provide a high level of confidence the fuel plate will perform in the reactor without a breach in the cladding.« less

The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

NASA Astrophysics Data System (ADS)

Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

2017-04-01

The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world's highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding during fabrication and are enhanced during irradiation. One aspect of fuel development and qualification is to demonstrate an appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 °C). The mechanisms responsible for fission gas release events are discussed.

The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

2017-04-01

The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world’s highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form during fabrication and are enhanced during irradiation between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding. One aspect of fuel development and qualification is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding andmore » Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 oC). The mechanisms responsible for fission gas release events are discussed.« less

In-situ tube burst testing and high-temperature deformation behavior of candidate materials for accident tolerant fuel cladding

DOE PAGES

Byun, Thak Sang; Yamamoto, Yukinori; Maloy, Stuart A.; ...

2015-08-25

Here, one of the most essential properties of accident tolerant fuel (ATF) for maintaining structural integrity during a loss-of-coolant accident (LOCA) is high resistance of the cladding to plastic deformation and burst failure, since the deformation and burst behavior governs the cooling efficiency of flow channels and the process of fission product release. To simulate and evaluate the deformation and burst process of thin-walled cladding, an in-situ testing and evaluation method has been developed on the basis of visual imaging and image analysis techniques. The method uses a specialized optics system consisting of a high-resolution video camera, a light filteringmore » unit, and monochromatic light sources. The in-situ testing is performed using a 50 mm long pressurized thin-walled tubular specimen set in a programmable furnace. As the first application, ten (10) candidate cladding materials for ATF, i.e., five FeCrAl alloys and five nanostructured steels, were tested using the newly developed method, and the time-dependent images were analyzed to produce detailed deformation and burst data such as true hoop stress, strain (creep) rate, and failure stress. Relatively soft FeCrAl alloys deformed and burst below 800 °C, while negligible strain rates were measured for higher strength alloys.« less

Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bradley K. Heath

2014-03-01

This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show thatmore » fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.« less

Chemical compatibility between UO2 fuel and SiC cladding for LWRs. Application to ATF (Accident-Tolerant Fuels)

NASA Astrophysics Data System (ADS)

Braun, James; Guéneau, Christine; Alpettaz, Thierry; Sauder, Cédric; Brackx, Emmanuelle; Domenger, Renaud; Gossé, Stéphane; Balbaud-Célérier, Fanny

2017-04-01

Silicon carbide-silicon carbide (SiC/SiC) composites are considered to replace the current zirconium-based cladding materials thanks to their good behavior under irradiation and their resistance under oxidative environments at high temperature. In the present work, a thermodynamic analysis of the UO2±x/SiC system is performed. Moreover, using two different experimental methods, the chemical compatibility of SiC towards uranium dioxide, with various oxygen contents (UO2±x) is investigated in the 1500-1970 K temperature range. The reaction leads to the formation of mainly uranium silicides and carbides phases along with CO and SiO gas release. Knudsen Cell Mass Spectrometry is used to measure the gas release occurring during the reaction between UO2+x and SiC powders as function of time and temperature. These experimental conditions are representative of an open system. Diffusion couple experiments with pellets are also performed to study the reaction kinetics in closed system conditions. In both cases, a limited chemical reaction is observed below 1700 K, whereas the reaction is enhanced at higher temperature due to the decomposition of SiC leading to Si vaporization. The temperature of formation of the liquid phase is found to lie between 1850 < T < 1950 K.

The impact of interface bonding efficiency on high-burnup spent nuclear fuel dynamic performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiang, Hao; Wang, Jy-An John; Wang, Hong

Finite element analysis (FEA) was used to investigate the impact of interfacial bonding efficiency at pellet-pellet and pellet-clad interfaces of high-burnup (HBU) spent nuclear fuel (SNF) on system dynamic performance. Bending moments M were applied to FEA model to evaluate the system responses. From bending curvature, κ, flexural rigidity EI can be estimated as EI = M/κ. The FEA simulation results were benchmarked with experimental results from cyclic integrated reversal bending fatigue test (CIRFT) of HBR fuel rods. The consequence of interface debonding between fuel pellets and cladding is a redistribution of the loads carried by the fuel pellets tomore » the clad, which results in a reduction in composite rod system flexural rigidity. Furthermore, the interface bonding efficiency at the pellet-pellet and pellet-clad interfaces can significantly dictate the SNF system dynamic performance. With the consideration of interface bonding efficiency, the HBU SNF fuel property was estimated with CIRFT test data.« less

The impact of interface bonding efficiency on high-burnup spent nuclear fuel dynamic performance

DOE PAGES

Jiang, Hao; Wang, Jy-An John; Wang, Hong

2016-09-26

Finite element analysis (FEA) was used to investigate the impact of interfacial bonding efficiency at pellet-pellet and pellet-clad interfaces of high-burnup (HBU) spent nuclear fuel (SNF) on system dynamic performance. Bending moments M were applied to FEA model to evaluate the system responses. From bending curvature, κ, flexural rigidity EI can be estimated as EI = M/κ. The FEA simulation results were benchmarked with experimental results from cyclic integrated reversal bending fatigue test (CIRFT) of HBR fuel rods. The consequence of interface debonding between fuel pellets and cladding is a redistribution of the loads carried by the fuel pellets tomore » the clad, which results in a reduction in composite rod system flexural rigidity. Furthermore, the interface bonding efficiency at the pellet-pellet and pellet-clad interfaces can significantly dictate the SNF system dynamic performance. With the consideration of interface bonding efficiency, the HBU SNF fuel property was estimated with CIRFT test data.« less

Oxidation of aluminum alloy cladding for research and test reactor fuel

NASA Astrophysics Data System (ADS)

Kim, Yeon Soo; Hofman, G. L.; Robinson, A. B.; Snelgrove, J. L.; Hanan, N.

2008-08-01

The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

Irradiation performance of U-Mo monolithic fuel

DOE PAGES

Meyer, M. K.; Gan, J.; Jue, J. F.; ...

2014-04-01

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties.more » Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.« less

IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL

DOE Office of Scientific and Technical Information (OSTI.GOV)

M.K. Meyer; J. Gan; J.-F. Jue

2014-04-01

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. UMo alloys represent the best known tradeoff in these properties.more » Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.« less

Neutronic fuel element fabrication

DOEpatents

Korton, George

2004-02-24

This disclosure describes a method for metallurgically bonding a complete leak-tight enclosure to a matrix-type fuel element penetrated longitudinally by a multiplicity of coolant channels. Coolant tubes containing solid filler pins are disposed in the coolant channels. A leak-tight metal enclosure is then formed about the entire assembly of fuel matrix, coolant tubes and pins. The completely enclosed and sealed assembly is exposed to a high temperature and pressure gas environment to effect a metallurgical bond between all contacting surfaces therein. The ends of the assembly are then machined away to expose the pin ends which are chemically leached from the coolant tubes to leave the coolant tubes with internal coolant passageways. The invention described herein was made in the course of, or under, a contract with the U.S. Atomic Energy Commission. It relates generally to fuel elements for neutronic reactors and more particularly to a method for providing a leak-tight metal enclosure for a high-performance matrix-type fuel element penetrated longitudinally by a multiplicity of coolant tubes. The planned utilization of nuclear energy in high-performance, compact-propulsion and mobile power-generation systems has necessitated the development of fuel elements capable of operating at high power densities. High power densities in turn require fuel elements having high thermal conductivities and good fuel retention capabilities at high temperatures. A metal clad fuel element containing a ceramic phase of fuel intimately mixed with and bonded to a continuous refractory metal matrix has been found to satisfy the above requirements. Metal coolant tubes penetrate the matrix to afford internal cooling to the fuel element while providing positive fuel retention and containment of fission products generated within the fuel matrix. Metal header plates are bonded to the coolant tubes at each end of the fuel element and a metal cladding or can completes the fuel-matrix enclosure by encompassing the sides of the fuel element between the header plates.

Valve for fuel pin loading system

DOEpatents

Christiansen, David W.

1985-01-01

A cyclone valve surrounds a wall opening through which cladding is projected. An axial valve inlet surrounds the cladding. Air is drawn through the inlet by a cyclone stream within the valve. An inflatable seal is included to physically engage a fuel pin subassembly during loading of fuel pellets.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spencer, B. W.; Williamson, R. L.; Stafford, D. S.

One of the important roles of cladding in light water reactor fuel rods is to prevent the release of fission products. To that end, it is essential that the cladding maintain its integrity under a variety of thermal and mechanical loading conditions. Local geometric irregularities in fuel pellets caused by manufacturing defects known as missing pellet surfaces (MPS) can in some circumstances lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. The BISON nuclear fuel performance code developed at Idaho National Laboratory can bemore » used to simulate the global thermo-mechanical fuel rod behavior, as well as the local response of regions of interest, in either 2D or 3D. In either case, a full set of models to represent the thermal and mechanical properties of the fuel, cladding and plenum gas is employed. A procedure for coupling 2D full-length fuel rod models to detailed 3D models of the region of the rod containing a MPS defect is detailed in this paper. The global and local model each contain appropriate physics and behavior models for nuclear fuel. This procedure is demonstrated on a simulation of a boiling water reactor (BWR) fuel rod containing a pellet with an MPS defect, subjected to a variety of transient events, including a control blade withdrawal and a ramp to high power. The importance of modeling the local defect using a 3D model is highlighted by comparing 3D and 2D representations of the defective pellet region. Finally, parametric studies demonstrate the effects of the choice of gaseous swelling model and of the depth and geometry of the MPS defect on the response of the cladding adjacent to the defect.« less

Core power and decay time limits for a disabled LOFT ECCS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Atkinson, S.A.

1978-01-09

An analysis was done to determine at what LOFT total core power (nuclear plus decay power) the ECCS could be inoperable. The criteria used for the analysis was that the maximum fuel clad temperature should not exceed 1650/sup 0/F given a loss of coolant. Calculations for natural convection cooling of the fuel by air with an inlet temperature of 580/sup 0/F determined that the limiting core power is 25 kW (discounted by 15 percent to 20 percent for potential uncertainties). Shutdown times are listed for when the LOFT ECCS can be safely bypassed or disabled.

Manufacturing process to reduce large grain growth in zirconium alloys

DOEpatents

Rosecrans, P.M.

1984-08-01

It is an object of the present invention to provide a procedure for desensitizing zirconium-based alloys to large grain growth (LGG) during thermal treatment above the recrystallization temperature of the alloy. It is a further object of the present invention to provide a method for treating zirconium-based alloys which have been cold-worked in the range of 2 to 8% strain to reduce large grain growth. It is another object of the present invention to provide a method for fabricating a zirconium alloy clad nuclear fuel element wherein the zirconium clad is resistant to large grain growth.

Measurement of carbon distribution in nuclear fuel pin cladding specimens by means of a secondary ion mass spectrometer

NASA Astrophysics Data System (ADS)

Bart, Gerhard; Aerne, Ernst Tino; Burri, Martin; Zwicky, Hans-Urs

1986-11-01

Cladding carburization during irradiation of advanced mixed uranium plutonium carbide fast breeder reactor fuel is possibly a life limiting fuel pin factor. The quantitative assessment of such clad carbon embrittlement is difficult to perform by electron microprobe analysis because of sample surface contamination, and due to the very low energy of the carbon K α X-ray transition. The work presented here describes a method developed at the Swiss Federal Institute for Reactor Research (EIR) to use shielded secondary ion mass spectrometry (SIMS) as an accurate tool to determine radial distribution profiles of carbon in radioactive stainless steel fuel pin cladding. Compared with nuclear microprobe analysis (NMA) [1], which is also an accurate method for carbon analysis, the SIMS method distinguishes itself by its versatility for simultaneous determination of additional impurities.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jackson, Timothy D; Hollenbach, Daniel F; Shedlock, Daniel

Radiography by Selective Detection (RSD), was investigated for its ability to determine the presence and types of defects in a UO{sub 2} fuel rod surrounded by zirconium cladding. Images created using a Monte Carlo model compared favorably with actual X-ray backscatter images from mock fuel rods. A fuel rod was modeled as a rectangular parallelepiped with zirconium cladding, and pencil beam X-ray sources of 160 kVp (79 keV avg) and 480 kVp (218 keV avg) were generated using the Monte Carlo N-Particle Transport Code to attempt to image void and palladium (Pd) defects in the interior and on the surfacemore » of the fuel pellet. It was found that the 160 kVp spectrum was unable to detect the presence of interior defects, whereas the 480 kVp spectrum detected them with both the standard and the RSD backscatter methods, though the RSD method was very inefficient. It was also found that both energy spectra were able to detect void and Pd defects on the surface using both imaging methods. Additionally, two mock fuel rods were imaged using a backscatter X-ray imaging system, one consisting of hafnium pellets in a Zircaloy-4 cladding and the other consisting of steel pellets in a Zircalloy-4 cladding which was then encased in a steel cladding (a double encapsulation configuration employed in irradiation and experiments). It was found that the system was capable of detecting individual HfO{sub 2} pellets in a Zircaloy-4 cladding and may be capable of detecting individual steel pellets in the double-encapsulated sample. It is expected that the system would also be capable of detecting individual UO{sub 2} pellets in a Zircaloy-4 cladding, though no UO{sub 2} fuel rod was available for imaging.« less

A Historical Review of Cermet Fuel Development and the Engine Performance Implications

NASA Technical Reports Server (NTRS)

Stewart, Mark E. M.

2015-01-01

This paper reviews test data for cermet fuel samples developed in the 1960's to better quantify Nuclear Thermal Propulsion (NTP) cermet engine performance, and to better understand contemporary fuel testing results. Over 200 cermet (W-UO2) samples were tested by thermally cycling to 2500 deg (2770 K) in hydrogen. The data indicates two issues at high temperatures: the vaporization rate of UO2 and the chemical stability of UO2. The data show that cladding and chemical stabilizers each result in large, order of magnitude improvements in high temperature performance, while other approaches yield smaller, incremental improvements. Data is very limited above 2770 K, and this complicates predictions of engine performance at high Isp. The paper considers how this material performance data translates into engine performance. In particular, the location of maximum temperature within the fuel element and the effect of heat deposition rate are examined.

Developing a laser shockwave model for characterizing diffusion bonded interfaces

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lacy, Jeffrey M., E-mail: Jeffrey.Lacy@inl.gov; Smith, James A., E-mail: Jeffrey.Lacy@inl.gov; Rabin, Barry H., E-mail: Jeffrey.Lacy@inl.gov

2015-03-31

The US National Nuclear Security Agency has a Global Threat Reduction Initiative (GTRI) with the goal of reducing the worldwide use of high-enriched uranium (HEU). A salient component of that initiative is the conversion of research reactors from HEU to low enriched uranium (LEU) fuels. An innovative fuel is being developed to replace HEU in high-power research reactors. The new LEU fuel is a monolithic fuel made from a U-Mo alloy foil encapsulated in Al-6061 cladding. In order to support the fuel qualification process, the Laser Shockwave Technique (LST) is being developed to characterize the clad-clad and fuel-clad interface strengthsmore » in fresh and irradiated fuel plates. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves to characterize interfaces in nuclear fuel plates. However, because the deposition of laser energy into the containment layer on a specimen's surface is intractably complex, the shock wave energy is inferred from the surface velocity measured on the backside of the fuel plate and the depth of the impression left on the surface by the high pressure plasma pulse created by the shock laser. To help quantify the stresses generated at the interfaces, a finite element method (FEM) model is being utilized. This paper will report on initial efforts to develop and validate the model by comparing numerical and experimental results for back surface velocities and front surface depressions in a single aluminum plate representative of the fuel cladding.« less

The effect of stress state on zirconium hydride reorientation

NASA Astrophysics Data System (ADS)

Cinbiz, Mahmut Nedim

Prior to storage in a dry-cask facility, spent nuclear fuel must undergo a vacuum drying cycle during which the spent fuel rods are heated up to elevated temperatures of ≤ 400°C to remove moisture the canisters within the cask. As temperature increases during heating, some of the hydride particles within the cladding dissolve while the internal gas pressure in fuel rods increases generating multi-axial hoop and axial stresses in the closed-end thin-walled cladding tubes. As cool-down starts, the hydrogen in solid solution precipitates as hydride platelets, and if the multiaxial stresses are sufficiently large, the precipitating hydrides reorient from their initial circumferential orientation to radial orientation. Radial hydrides can severely embrittle the spent nuclear fuel cladding at low temperature in response to hoop stress loading. Because the cladding can experience a range of stress states during the thermo-mechanical treatment induced during vacuum drying, this study has investigated the effect of stress state on the process of hydride reorientation during controlled thermo-mechanical treatments utilizing the combination of in situ X-ray diffraction and novel mechanical testing analyzed by the combination of metallography and finite element analysis. The study used cold worked and stress relieved Zircaloy-4 sheet containing approx. 180 wt. ppm hydrogen as its material basis. The failure behavior of this material containing radial hydrides was also studied over a range of temperatures. Finally, samples from reactor-irradiated cladding tubes were examined by X-ray diffraction using synchrotron radiation. To reveal the stress state effect on hydride reorientation, the critical threshold stress to reorient hydrides was determined by designing novel mechanical test samples which produce a range of stress states from uniaxial to "near-equibiaxial" tension when a load is applied. The threshold stress was determined after thermo-mechanical treatments by correlating the finite element stress-state results with the spatial distribution of hydride microstructures observed within the optical micrographs for each sample. Experiments showed that the hydride reorientation was enhanced as the stress biaxiality increased. The threshold stress decreased from 150 MPa to 80 MPa when stress biaxiality ratio increased from uniaxial tension to near-equibiaxial tension. This behavior was also predicted by classical nucleation theory based on the Gibbs free energy of transformation being assisted by the far-field stress. An analysis of in situ X-ray diffraction data obtained during a thermo-mechanical cycle typical of vacuum drying showed a complex lattice-spacing behavior of the hydride phase during the dissolution and precipitation. The in-plane hydrides showed bilinear lattice expansion during heating with the intrinsic thermal expansion rate of the hydrides being observed only at elevated temperatures as they dissolve. For radial hydrides that precipitate during cooling under stress, the spacing of the close-packed {111} planes oriented normal to the maximum applied stress was permanently higher than the corresponding {111} plane spacing in the other directions. This behavior is believed to be a result of a complex stress state within the precipitating plate-like hydrides that induces a strain component within the hydrides normal to its "plate" face (i.e., the applied stress direction) that exceeds the lattice spacing strains in the other directions. During heat-up, the lattice spacing of these same "plate" planes actually contract due to the reversion of the stress state within the plate-like hydrides as they dissolve. The presence of radial hydrides and their connectivity with in-plane hydrides was shown to increase the ductile-to-brittle transition temperature during tensile testing. This behavior can be understood in terms of the role of radial hydrides in promoting the initiation of a long crack that subsequently propagates under fracture mechanics conditions. Finally, the d-spacing of irradiated Zircaloy-4 and M5 cladding tubes was measured at room temperature and compared to that of unirradiated samples.

Standalone BISON Fuel Performance Results for Watts Bar Unit 1, Cycles 1-3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Clarno, Kevin T.; Pawlowski, Roger; Stimpson, Shane

2016-03-07

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is moving forward with more complex multiphysics simulations and increased focus on incorporating fuel performance analysis methods. The coupled neutronics/thermal-hydraulics capabilities within the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) have become relatively stable, and major advances have been made in analysis efforts, including the simulation of twelve cycles of Watts Bar Nuclear Unit 1 (WBN1) operation. While this is a major achievement, the VERA-CS approaches for treating fuel pin heat transfer have well-known limitations that could be eliminated through better integration with the BISON fuel performance code. Severalmore » approaches are being implemented to consider fuel performance, including a more direct multiway coupling with Tiamat, as well as a more loosely coupled one-way approach with standalone BISON cases. Fuel performance typically undergoes an independent analysis using a standalone fuel performance code with manually specified input defined from an independent core simulator solution or set of assumptions. This report summarizes the improvements made since the initial milestone to execute BISON from VERA-CS output. Many of these improvements were prompted through tighter collaboration with the BISON development team at Idaho National Laboratory (INL). A brief description of WBN1 and some of the VERA-CS data used to simulate it are presented. Data from a small mesh sensitivity study are shown, which helps justify the mesh parameters used in this work. The multi-cycle results are presented, followed by the results for the first three cycles of WBN1 operation, particularly the parameters of interest to pellet-clad interaction (PCI) screening (fuel-clad gap closure, maximum centerline fuel temperature, maximum/minimum clad hoop stress, and cumulative damage index). Once the mechanics of this capability are functioning, future work will target cycles with known or suspected PCI failures to determine how well they can be estimated.« less

Preliminary calculations related to the accident at Three Mile Island

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kirchner, W.L.; Stevenson, M.G.

This report discusses preliminary studies of the Three Mile Island Unit 2 (TMI-2) accident based on available methods and data. The work reported includes: (1) a TRAC base case calculation out to 3 hours into the accident sequence; (2) TRAC parametric calculations, these are the same as the base case except for a single hypothetical change in the system conditions, such as assuming the high pressure injection (HPI) system operated as designed rather than as in the accident; (3) fuel rod cladding failure, cladding oxidation due to zirconium metal-steam reactions, hydrogen release due to cladding oxidation, cladding ballooning, cladding embrittlement,more » and subsequent cladding breakup estimates based on TRAC calculated cladding temperatures and system pressures. Some conclusions of this work are: the TRAC base case accident calculation agrees very well with known system conditions to nearly 3 hours into the accident; the parametric calculations indicate that, loss-of-core cooling was most influenced by the throttling of High-Pressure Injection (HPI) flows, given the accident initiating events and the pressurizer electromagnetic-operated valve (EMOV) failing to close as designed; failure of nearly all the rods and gaseous fission product gas release from the failed rods is predicted to have occurred at about 2 hours and 30 minutes; cladding oxidation (zirconium-steam reaction) up to 3 hours resulted in the production of approximately 40 kilograms of hydrogen.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kristine Barrett; Shannon Bragg-Sitton

The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R&D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental improvements are required in the areas of nuclear fuel composition, cladding integrity, and the fuel/cladding interaction to allow power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an “accident tolerant” fuel system thatmore » would offer improved coping time under accident scenarios. With a development time of about 20 – 25 years, advanced fuel designs must be started today and proven in current reactors if future reactor designs are to be able to use them with confidence.« less

Production test IP-376-D, Supplement B Irradiation of MGCR-HDR-3 Test Element

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baars, R.E.

The objective of this supplement to PT-IP-376-D, Irradiation of MGCR-HDR-3 Test Element is to authorize 1000 hours of operation at a maximum test specimen surface temperature of 1700 F. The original production test authorized a test duration of four months at a maximum specimen surface temperature of 1500 F; supplement A authorized extension of the test duration to ten months. The desired increase in surface temperature is requested to demonstrate the general feasibility of operation of the fuel element at 1700 F, and to obtain specific information on the performance of Hastelloy-X cladding and fuel bodies. The increased temperature hasmore » been approved by the Atomic Energy Commission.« less

Cladding material, tube including such cladding material and methods of forming the same

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garnier, John E.; Griffith, George W.

A multi-layered cladding material including a ceramic matrix composite and a metallic material, and a tube formed from the cladding material. The metallic material forms an inner liner of the tube and enables hermetic sealing of thereof. The metallic material at ends of the tube may be exposed and have an increased thickness enabling end cap welding. The metallic material may, optionally, be formed to infiltrate voids in the ceramic matrix composite, the ceramic matrix composite encapsulated by the metallic material. The ceramic matrix composite includes a fiber reinforcement and provides increased mechanical strength, stiffness, thermal shock resistance and highmore » temperature load capacity to the metallic material of the inner liner. The tube may be used as a containment vessel for nuclear fuel used in a nuclear power plant or other reactor. Methods for forming the tube comprising the ceramic matrix composite and the metallic material are also disclosed.« less

3D modeling of missing pellet surface defects in BWR fuel

DOE PAGES

Spencer, B. W.; Williamson, R. L.; Stafford, D. S.; ...

2016-07-26

One of the important roles of cladding in light water reactor fuel rods is to prevent the release of fission products. To that end, it is essential that the cladding maintain its integrity under a variety of thermal and mechanical loading conditions. Local geometric irregularities in fuel pellets caused by manufacturing defects known as missing pellet surfaces (MPS) can in some circumstances lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. The BISON nuclear fuel performance code developed at Idaho National Laboratory can bemore » used to simulate the global thermo-mechanical fuel rod behavior, as well as the local response of regions of interest, in either 2D or 3D. In either case, a full set of models to represent the thermal and mechanical properties of the fuel, cladding and plenum gas is employed. A procedure for coupling 2D full-length fuel rod models to detailed 3D models of the region of the rod containing a MPS defect is detailed in this paper. The global and local model each contain appropriate physics and behavior models for nuclear fuel. This procedure is demonstrated on a simulation of a boiling water reactor (BWR) fuel rod containing a pellet with an MPS defect, subjected to a variety of transient events, including a control blade withdrawal and a ramp to high power. The importance of modeling the local defect using a 3D model is highlighted by comparing 3D and 2D representations of the defective pellet region. Finally, parametric studies demonstrate the effects of the choice of gaseous swelling model and of the depth and geometry of the MPS defect on the response of the cladding adjacent to the defect.« less

Phase 1A Final Report for the AREVA Team Enhanced Accident Tolerant Fuels Concepts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morrell, Mike E.

In response to the Department of Energy (DOE) funded initiative to develop and deploy lead fuel assemblies (LFAs) of Enhanced Accident Tolerant Fuel (EATF) into a US reactor within 10 years, AREVA put together a team to develop promising technologies for improved fuel performance during off normal operations. This team consisted of the University of Florida (UF) and the University of Wisconsin (UW), Savannah River National Laboratory (SRNL), Duke Energy and Tennessee Valley Authority (TVA). This team brought broad experience and expertise to bear on EATF development. AREVA has been designing; manufacturing and testing nuclear fuel for over 50 yearsmore » and is one of the 3 large international companies supplying fuel to the nuclear industry. The university and National Laboratory team members brought expertise in nuclear fuel concepts and materials development. Duke and TVA brought practical utility operating experience. This report documents the results from the initial “discovery phase” where the team explored options for EATF concepts that provide enhanced accident tolerance for both Design Basis (DB) and Beyond Design Basis Events (BDB). The main driver for the concepts under development were that they could be implemented in a 10 year time frame and be economically viable and acceptable to the nuclear fuel marketplace. The economics of fuel design make this DOE funded project very important to the nuclear industry. Even incremental changes to an existing fuel design can cost in the range of $100M to implement through to LFAs. If this money is invested evenly over 10 years then it can take the fuel vendor several decades after the start of the project to recover their initial investment and reach a breakeven point on the initial investment. Step or radical changes to a fuel assembly design can cost upwards of $500M and will take even longer for the fuel vendor to recover their investment. With the projected lifetimes of the current generation of nuclear power plants large scale investment by the fuel vendors is difficult to justify. Specific EATF enhancements considered by the AREVA team were; Improved performance in DB and BDB conditions; Reduced release to the environment in a catastrophic accident; Improved performance during normal operating conditions; Improved performance if US reactors start to load follow; Equal or improved economics of the fuel; and Improvements to the fuel behavior to support future transportation and storage of the used nuclear fuel (UNF). In pursuit of the above enhancements, EATF technology concepts that our team considered were; Additives to the fuel pellets which included; Chromia doping to increase fission gas retention. Chromia doping has the potential to improve load following characteristics, improve performance of the fuel pellet during clad failure, and potentially lock up cesium into the fuel matrix; Silicon Carbide (SiC) Fibers to improve thermal heat transfer in normal operating conditions which also improves margin in accident conditions and the potential to lock up iodine into the fuel matrix; Nano-diamond particles to enhance thermal conductivity; Coatings on the fuel cladding; and Nine coatings on the existing Zircaloy cladding to increase coping time and reduce clad oxidation and hydrogen generation during accident conditions, as well as reduce hydrogen pickup and mitigate hydride reorientation in the cladding. To facilitate the development process AREVA adopted a formal “Gate Review Process” (GR) that was used to review results and focus resources onto promising technologies to reduce costs and identify the technologies that would potentially be carried forward to LFAs within a 10 year period. During the initial discovery phase of the project AREVA took the decision to be relatively hands off and allow our university and National Laboratory partners to be free thinking and consider options that would not be constrained by preconceived ideas from the fuel vendor. To counter this and to keep the partners focused, the GR process was utilized. During this GR process each of the team members presented their findings to a board made up of technical experts from utilities, fuel manufacturing experts, fuel technical experts, and fuel research and development (R&D) experts. During the initial 2 years of the project there were several major accomplishments. These accomplishments, along with the implications for successfully implementing EATF, are; The experimental spark plasma sintering process (SPS) process was successfully used to produce fuel pellets containing either 10% SiC whiskers or nano-diamond particles. The ability to use this process enables the thermal margin enhancements of the fuel additives to be realized. Without the SPS process, the conventional process cannot support adding pellet additives in the required quantities; Coatings of Ti2AlC were successfully applied to Zircaloy-4 cladding. Testing of Ti2AlC coatings at Loss of Cooling Accident (LOCA) conditions showed reduced cladding oxidation compared to present un-coated Zircaloy-4 cladding. This achievement allows the presently used cladding system to be retained so that the 10 year schedule can be met. Having to implement a new cladding material will extend the development schedule beyond 10 years; Several documents were produced to support future development, testing, and licensing of EATF, including a design requirements traceability matrix, a draft business plan, a draft test plan, a draft regulatory plan, and the acceptance criteria for lead fuel assembly insertion into a commercial reactor. This preparatory work lays the foundation for ensuring the future development plans address all the areas required to test, license, and manufacture the new EATF; and In addition, the high velocity oxy-fuel and electrophoretic deposition (EPD) coating application processes were dropped from further consideration due to their inability to meet manufacturing criteria. This allows the resources to be focused on the most promising EATF concepts identified. Future development opportunities that were identified during this work include; The use of SiC or diamond requires that a new pellet production technique (Spark Plasma Sintering), be developed. This entails investment in developing, proving and implementing a new commercial pellet production process. Development of the process to apply thinner coatings is required; Coatings cannot be too “thick” or they will displace a significant volume of water in the core resulting in reduced thermal hydraulic characteristics; Application of the coating at high temperature can affect the Zircaloy substrate. This will require the development and implementation of a new cladding coating manufacturing process; and Replace the Cold Spray (CS) cladding coating application with the Physical Vapor Deposition (PVD) process to eliminate duplication of work and provide greater control over coating thicknesses. This can result in a reduction in the final cycle economic penalty of coatings.« less

Further observations on OCOM MOX fuel: microstructure in the vicinity of the pellet rim and fuel — cladding interaction

NASA Astrophysics Data System (ADS)

Walker, C. T.; Goll, W.; Matsumura, T.

1997-06-01

The fuel investigated was manufactured by Siemens-KWU and irradiated at low rating in the KWO reactor in Germany. The MOX agglomerates in the cold outer region of the fuel shared several common features with the high burn-up structure at the rim of UO 2 fuel. It is proposed that in both cases the mechanism producing the microstructure change is recrystallisation. Further, it is shown that surface MOX agglomerates do not noticeably retard cladding creepdown although they swell into the gap. The contracting cladding appears able to push the agglomerates back into the fuel. The thickness of the oxide layer on the inner cladding surface increased at points where contact with surface MOX agglomerates had occurred. Despite this, the mean thickness of the oxide did not differ significantly from that found in UO 2 fuel rods of like design. It is judged that the high burn-up structure will form in the UO 2 matrix when the local burn-up there reaches 60 to 80 GWd/tM. Limiting the MOX scrap addition in the UO 2 matrix will delay its formation.

Nuclear fuel element with axially aligned fuel pellets and fuel microspheres therein

DOEpatents

Sease, J.D.; Harrington, F.E.

1973-12-11

Elongated single- and multi-region fuel elements are prepared by replacing within a cladding container a coarse fraction of fuel material which includes plutonium and uranium in the appropriate regions of the fuel element and then infiltrating with vibration a fine-sized fraction of uranium-containing microspheres throughout all interstices in the coarse material in a single loading. The fine, rigid material defines a thin annular layer between the coarse fraction and the cladding to reduce adverse mechanical and chemical interactions. (Official Gazette)

Systematic technology evaluation program for SiC/SiC composite-based accident-tolerant LWR fuel cladding and core structures: Revision 2015

DOE Office of Scientific and Technical Information (OSTI.GOV)

Katoh, Yutai; Terrani, Kurt A.

2015-08-01

Fuels and core structures in current light water reactors (LWR’s) are vulnerable to catastrophic failure in severe accidents as unfortunately evidenced by the March 2011 Fukushima Dai-ichi Nuclear Power Plant Accident. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures. Zr alloys are the primary material in LWR cores except for the fuel itself. Therefore, alternative materials with reduced oxidation kinetics as compared to zirconium alloys are sought to enable enhanced accident-tolerant fuels and cores.

Materials technology for an advanced space power nuclear reactor concept: Program summary

NASA Technical Reports Server (NTRS)

Gluyas, R. E.; Watson, G. K.

1975-01-01

The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

Zirconium-based alloys, nuclear fuel rods and nuclear reactors including such alloys, and related methods

DOEpatents

Mariani, Robert Dominick

2014-09-09

Zirconium-based metal alloy compositions comprise zirconium, a first additive in which the permeability of hydrogen decreases with increasing temperatures at least over a temperature range extending from 350.degree. C. to 750.degree. C., and a second additive having a solubility in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. At least one of a solubility of the first additive in the second additive over the temperature range extending from 350.degree. C. to 750.degree. C. and a solubility of the second additive in the first additive over the temperature range extending from 350.degree. C. to 750.degree. C. is higher than the solubility of the second additive in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. Nuclear fuel rods include a cladding material comprising such metal alloy compositions, and nuclear reactors include such fuel rods. Methods are used to fabricate such zirconium-based metal alloy compositions.

Non-destructive evaluation of the cladding thickness in LEU fuel plates by accurate ultrasonic scanning technique

DOE Office of Scientific and Technical Information (OSTI.GOV)

Borring, J.; Gundtoft, H.E.; Borum, K.K.

1997-08-01

In an effort to improve their ultrasonic scanning technique for accurate determination of the cladding thickness in LEU fuel plates, new equipment and modifications to the existing hardware and software have been tested and evaluated. The authors are now able to measure an aluminium thickness down to 0.25 mm instead of the previous 0.35 mm. Furthermore, they have shown how the measuring sensitivity can be improved from 0.03 mm to 0.01 mm. It has now become possible to check their standard fuel plates for DR3 against the minimum cladding thickness requirements non-destructively. Such measurements open the possibility for the acceptancemore » of a thinner nominal cladding than normally used today.« less

Evaluation of tantalum-alloy-clad uranium mononitride fuel specimens from 7500-hour, 1040 C pumped-lithium-loop test

NASA Technical Reports Server (NTRS)

Watson, G. K.

1974-01-01

Simulated nuclear fuel element specimens, consisting of uranium mononitride (UN) fuel cylinders clad with tungsten-lined T-111, were exposed for up to 7500 hr at 1040 C (1900 F) in a pumped-lithium loop. The lithium flow velocity was 1.5 m/sec (5 ft/sec) in the specimen test section. No evidence of any compatibility problems between the specimens and the flowing lithium was found based on appearance, weight change, chemistry, and metallography. Direct exposure of the UN to the lithium through a simulated cladding crack resulted in some erosion of the UN in the area of the defect. The T-111 cladding was ductile after lithium exposure, but it was sensitive to hydrogen embrittlement during post-test handling.

Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

NASA Astrophysics Data System (ADS)

Williamson, R. L.; Capps, N. A.; Liu, W.; Rashid, Y. R.; Wirth, B. D.

2016-11-01

Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial ( R- Z) or plane radial-circumferential ( R- θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. In comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

DOE PAGES

Williamson, R. L.; Capps, N. A.; Liu, W.; ...

2016-09-27

Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial (R-Z) ormore » plane radial-circumferential (R-θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used in this paper to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. Finally, in comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.« less

Creep relaxation of fuel pin bending and ovalling stresses. [BEND code, OVAL code, MARC-CDC code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chan, D.P.; Jackson, R.J.

1981-10-01

Analytical methods for calculating fuel pin cladding bending and ovalling stresses due to pin bundle-duct mechanical interaction taking into account nonlinear creep are presented. Calculated results are in agreement with finite element results by MARC-CDC program. The methods are used to investigate the effect of creep on the FTR fuel cladding bending and ovalling stresses. It is concluded that the cladding of 316 SS 20 percent CW and reference design has high creep rates in the FTR core region to keep the bending and ovalling stresses to acceptable levels. 6 refs.

In-pile and out-of-pile testing of a molybdenum-uranium dioxide cermet fueled themionic diode

NASA Technical Reports Server (NTRS)

Diianni, D. C.

1972-01-01

The behavior of Mo-UO2 cermet fuel in a diode for thermionic reactor application was studied. The diode had a Mo-0.5 Ti emitter and niobium collector. Output power ranged from 1.4 to 2.8 W/cm squared at emitter and collector temperatures of 1500 deg and 540 C. Thermionic performance was stable within the limits of the instrumentation sensitivity. Through 1000 hours of in-pile operation the emitter was dimensionally stable. However, some fission gases (15 percent) leaked through an inner clad imperfection that occurred during fuel fabrication.

BISON Theory Manual The Equations behind Nuclear Fuel Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hales, J. D.; Williamson, R. L.; Novascone, S. R.

2016-09-01

BISON is a finite element-based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO particle fuel, and metallic rod and plate fuel. It solves the fully-coupled equations of thermomechanics and species diffusion, for either 2D axisymmetric or 3D geometries. Fuel models are included to describe temperature and burnup dependent thermal properties, fission product swelling, densification, thermal and irradiation creep, fracture, and fission gas production and release. Plasticity, irradiation growth, and thermal and irradiation creep models are implemented for clad materials. Models are also available to simulate gap heat transfer, mechanical contact,more » and the evolution of the gap/plenum pressure with plenum volume, gas temperature, and fission gas addition. BISON is based on the MOOSE framework and can therefore efficiently solve problems using standard workstations or very large high-performance computers. This document describes the theoretical and numerical foundations of BISON.« less

A Historical Review of Cermet Fuel Development and the Engine Performance Implications

NASA Technical Reports Server (NTRS)

Stewart, Mark E.

2015-01-01

To better understand Cermet engine performance, examined historical material development reports two issues: High vaporization rate of UO2, High temperature chemical stability of UO2. Cladding and chemical stabilizers each result in large, order of magnitude improvements in high temperature performance. Few samples were tested above 2770 K. Results above 2770 K are ambiguous. Contemporary testing may clarify performance. Cermet sample testing during the NERVA Rover era. Important properties, melting temperature, vaporization rate, strength, Brittle-to-Ductile Transition, cermet sample test results, engine performance, location, peak temperature.

Fabrication of fuel pin assemblies, phase 3

NASA Technical Reports Server (NTRS)

Keeton, A. R.; Stemann, L. G.

1972-01-01

Five full size and eight reduced length fuel pins were fabricated for irradiation testing to evaluate design concepts for a fast spectrum lithium cooled compact space power reactor. These assemblies consisted of uranium mononitride fuel pellets encased in a T-111 (Ta-8W-2Hf) clad with a tungsten barrier separating fuel and clad. Fabrication procedures were fully qualified by process development and assembly qualification tests. Detailed specifications and procedures were written for the fabrication and assembly of prototype fuel pins.

Fast reactor safety and related physics. Volume IV. Phenomenology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1976-01-01

Separate abstracts are included for 58 papers concerning single-phase flow and sodium boiling; sodium boiling and subassembly flow blockages; transient-overpower and loss-of-flow experiments; fuel and cladding behavior and relocation; fuel and cladding freezing; molten-fuel-coolant interaction; aerosols and fission product release, and post-accident heat removal. Thirteen papers have been perivously abstracted and included in ERA.

Mercury-free dissolution of aluminum-clad fuel in nitric acid

DOEpatents

Christian, Jerry D.; Anderson, Philip A.

1994-01-01

A mercury-free dissolution process for aluminum involves placing the aluminum in a dissolver vessel in contact with nitric acid-fluoboric acid mixture at an elevated temperature. By maintaining a continuous flow of the acid mixture through the dissolver vessel, an effluent containing aluminum nitrate, nitric acid, fluoboric acid and other dissolved components are removed.

Mercury-free dissolution of aluminum-clad fuel in nitric acid

DOEpatents

Christian, J.D.; Anderson, P.A.

1994-11-15

A mercury-free dissolution process for aluminum involves placing the aluminum in a dissolver vessel in contact with nitric acid-fluoboric acid mixture at an elevated temperature. By maintaining a continuous flow of the acid mixture through the dissolver vessel, an effluent containing aluminum nitrate, nitric acid, fluoboric acid and other dissolved components are removed. 5 figs.

High-temperature compatibility between liquid metal as PWR fuel gap filler and stainless steel and high-density concrete

NASA Astrophysics Data System (ADS)

Wongsawaeng, Doonyapong; Jumpee, Chayanit; Jitpukdee, Manit

2014-08-01

In conventional nuclear fuel rods for light-water reactors, a helium-filled as-fabricated gap between the fuel and the cladding inner surface accommodates fuel swelling and cladding creep down. Because helium exhibits a very low thermal conductivity, it results in a large temperature rise in the gap. Liquid metal (LM; 1/3 weight portion each of lead, tin, and bismuth) has been proposed to be a gap filler because of its high thermal conductivity (∼100 times that of He), low melting point (∼100 °C), and lack of chemical reactivity with UO2 and water. With the presence of LM, the temperature drop across the gap is virtually eliminated and the fuel is operated at a lower temperature at the same power output, resulting in safer fuel, delayed fission gas release and prevention of massive secondary hydriding. During normal reactor operation, should an LM-bonded fuel rod failure occurs resulting in a discharge of liquid metal into the bottom of the reactor pressure vessel, it should not corrode stainless steel. An experiment was conducted to confirm that at 315 °C, LM in contact with 304 stainless steel in the PWR water chemistry environment for up to 30 days resulted in no observable corrosion. Moreover, during a hypothetical core-melt accident assuming that the liquid metal with elevated temperature between 1000 and 1600 °C is spread on a high-density concrete basement of the power plant, a small-scale experiment was performed to demonstrate that the LM-concrete interaction at 1000 °C for as long as 12 h resulted in no penetration. At 1200 °C for 5 h, the LM penetrated a distance of ∼1.3 cm, but the penetration appeared to stop. At 1400 °C the penetration rate was ∼0.7 cm/h. At 1600 °C, the penetration rate was ∼17 cm/h. No corrosion based on chemical reactions with high-density concrete occurred, and, hence, the only physical interaction between high-temperature LM and high-density concrete was from tiny cracks generated from thermal stress. Moreover, for as high as 1600 °C, the non-reactive LM was experimentally confirmed not to show any chemical reaction with air or moisture in the air. This experimental work confirmed the excellent compatibility behaviors between the LM as a PWR fuel gap filler and stainless steel and high-density concrete in the high-temperature regime.

Annual progress report on the NSRR experiments, (21)

NASA Astrophysics Data System (ADS)

1992-05-01

Fuel behavior studies under simulated reactivity-initiated accident (RIA) conditions have been performed in the Nuclear Safety Research Reactor (NSRR) since 1975. This report gives the results of experiments performed from April, 1989 through March, 1990 and discussions of them. A total of 41 tests were carried out during this period. The tests are distinguished into pre-irradiated fuel tests and fresh fuel tests; the former includes 2 JMTR pre-irradiated fuel tests, 2 PWR pre-irradiated fuel tests, and 2 BWR pre-irradiated fuel tests, and the latter includes 6 standard fuel tests (6 SP(center dot)CP scoping tests), 7 power and cooling condition parameter tests (4 flow shrouded fuel tests, 1 bundle simulation test, 1 fully water-filled vessel test, 1 high pressure/high temperature loop test), 12 special fuel tests (3 stainless steel clad fuel tests, 3 improved PWR fuel tests, 6 improved BWR fuel tests), 3 severe fuel damage tests (1 high temperature flooding test, 1 flooding behavior observation test, 1 debris coolability test), 3 fast breeder reactor fuel tests (2 moderator material characteristic measurement tests, 1 fuel behavior observation test), and 2 miscellaneous tests (2 preliminary tests for pre-irradiated fuel tests).

Characterization of LWRS Hybrid SiC-CMC-Zircaloy-4 Fuel Cladding after Gamma Irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Isabella J van Rooyen

2012-09-01

The purpose of the gamma irradiation tests conducted at the Idaho National Laboratory (INL) was to obtain a better understanding of chemical interactions and potential changes in microstructural properties of a mock-up hybrid nuclear fuel cladding rodlet design (unfueled) in a simulated PWR water environment under irradiation conditions. The hybrid fuel rodlet design is being investigated under the Light Water Reactor Sustainability (LWRS) program for further development and testing of one of the possible advanced LWR nuclear fuel cladding designs. The gamma irradiation tests were performed in preparation for neutron irradiation tests planned for a silicon carbide (SiC) ceramic matrixmore » composite (CMC) zircaloy-4 (Zr-4) hybrid fuel rodlet that may be tested in the INL Advanced Test Reactor (ATR) if the design is selected for further development and testing« less

A comparative study of the mechanical and thermal properties of defective ZrC, TiC and SiC.

PubMed

Jiang, M; Zheng, J W; Xiao, H Y; Liu, Z J; Zu, X T

2017-08-24

ZrC and TiC have been proposed to be alternatives to SiC as fuel-cladding and structural materials in nuclear reactors due to their strong radiation tolerance and high thermal conductivity at high temperatures. To unravel how the presence of defects affects the thermo-physical properties under irradiation, first-principles calculations based on density function theory were carried out to investigate the mechanical and thermal properties of defective ZrC, TiC and SiC. As compared with the defective SiC, the ZrC and TiC always exhibit larger bulk modulus, smaller changes in the Young's and shear moduli, as well as better ductility. The total thermal conductivity of ZrC and TiC are much larger than that of SiC, implying that under radiation environment the ZrC and TiC will exhibit superior heat conduction ability than the SiC. One disadvantage for ZrC and TiC is that their Debye temperatures are generally lower than that of SiC. These results suggest that further improving the Debye temperature of ZrC and TiC will be more beneficial for their applications as fuel-cladding and structural materials in nuclear reactors.

Diffusion Barrier Selection from Refractory Metals (Zr, Mo and Nb) via Interdiffusion Investigation for U-Mo RERTR Fuel Alloy

DOE Office of Scientific and Technical Information (OSTI.GOV)

K. Huang; C. Kammerer; D. D. Keiser, Jr.

2014-04-01

U-Mo alloys are being developed as low enrichment monolithic fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) Program. Diffusional interactions between the U-Mo fuel alloy and Al-alloy cladding within the monolithic fuel plate construct necessitate incorporation of a barrier layer. Fundamentally, a diffusion barrier candidate must have good thermal conductivity, high melting point, minimal metallurgical interaction, and good irradiation performance. Refractory metals, Zr, Mo, and Nb are considered based on their physical properties, and the diffusion behavior must be carefully examined first with U-Mo fuel alloy. Solid-to-solid U-10wt.%Mo vs. Mo, Zr, or Nb diffusion couples were assembledmore » and annealed at 600, 700, 800, 900 and 1000 degrees C for various times. The interdiffusion microstructures and chemical composition were examined via scanning electron microscopy and electron probe microanalysis, respectively. For all three systems, the growth rate of interdiffusion zone were calculated at 1000, 900 and 800 degrees C under the assumption of parabolic growth, and calculated for lower temperature of 700, 600 and 500 degrees C according to Arrhenius relationship. The growth rate was determined to be about 10 3 times slower for Zr, 10 5 times slower for Mo and 10 6 times slower for Nb, than the growth rates reported for the interaction between the U-Mo fuel alloy and pure Al or Al-Si cladding alloys. Zr, however was selected as the barrier metal due to a concern for thermo- mechanical behavior of UMo/Nb interface observed from diffusion couples, and for ductile-to-brittle transition of Mo near room temperature.« less

FOIL ELEMENT FOR NUCLEAR REACTOR

DOEpatents

Noland, R.A.; Walker, D.E.; Spinrad, B.I.

1963-07-16

A method of making a foil-type fuel element is described. A foil of fuel metal is perforated in; regular design and sheets of cladding metal are placed on both sides. The cladding metal sheets are then spot-welded to each other through the perforations, and the edges sealed. (AEC)

DISSOLUTION OF ZIRCONIUM AND ALLOYS THEREFOR

DOEpatents

Swanson, J.L.

1961-07-11

The dissolution of zirconium cladding in a water solution of ammonium fluoride and ammonium nitrate is described. The method finds particular utility in processing spent fuel elements for nuclear reactors. The zirconium cladding is first dissolved in a water solution of ammonium fluoride and ammonium nitrate; insoluble uranium and plutonium fiuorides formed by attack of the solvent on the fuel materiai of the fuel element are then separated from the solution, and the fuel materiai is dissolved in another solution.

Methodology for Mechanical Property Testing of Fuel Cladding Using a Expanded Plug Wedge Test

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiang, Hao; Wang, Jy-An John

2014-01-01

An expanded plug method was developed earlier for determining the tensile properties of irradiated fuel cladding. This method tests fuel rod cladding ductility by utilizing an expandable plug to radially stretch a small ring of irradiated cladding material. The circumferential or hoop strain is determined from the measured diametrical expansion of the ring. A developed procedure is used to convert the load circumferential strain data from the ring tests into material pseudo-stress-strain curves, from which material properties of the cladding can be extracted. However, several deficiencies existed in this expanded-plug test that can impact the accuracy of test results, suchmore » as that the large axial compressive stress resulted from the expansion plug test can potentially induce the shear failure mode of the tested specimen. Moreover, highly nonuniform stress and strain distribution in the deformed clad gage section and significant compressive stresses, induced by bending deformation due to clad bulging effect, will further result in highly nonconservative estimates of the mechanical properties for both strength and ductility of the tested clad. To overcome the aforementioned deficiencies associated with the current expansion plug test, systematic studies have been conducted. By optimizing the specific geometry designs, selecting the appropriate material for the expansion plug, and adding new components into the testing system, a modified expansion plug testing protocol has been developed. A general procedure was also developed to determine the hoop stress in the tested ring specimen. A scaling factor, -factor, was used to convert the ring load Fring into hoop stress , and is written as _ = F_ring/tl , where t is the clad thickness and l is the clad length. The generated stress-strain curve agrees well with the associated tensile test data in both elastic and plastic deformation regions.« less

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Montgomery, Robert; Tomé, Carlos; Liu, Wenfeng

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. CASL has endeavored to improve upon this approach by incorporating a microstructurally-based, atomistically-informed, zirconium alloy mechanical deformation analysis capability into the BISON-CASL engineering scale fuel performance code. Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed bymore » Lebensohn and Tome´ [2], has been coupled with BISON-CASL to represent the mechanistic material processes controlling the deformation behavior of the cladding. A critical component of VPSC is the representation of the crystallographic orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON-CASL and provides initial results utilizing the coupled functionality.« less

76 FR 2243 - List of Approved Spent Fuel Storage Casks: NUHOMS ® HD System Revision 1

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-13

... the requirements of reconstituted fuel assemblies; add requirements to qualify metal matrix composite... requirements to qualify metal matrix composite neutron absorbers with integral aluminum cladding; clarify the... requirements to qualify metal matrix composite neutron absorbers with integral aluminum cladding; clarify the...

Chemical Dissolution of Simulant FCA Cladding and Plates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daniel, G.; Pierce, R.; O'Rourke, P.

The Savannah River Site (SRS) has received some fast critical assembly (FCA) fuel from the Japan Atomic Energy Agency (JAEA) for disposition. Among the JAEA FCA fuel are approximately 7090 rectangular Stainless Steel clad fuel elements. Each element has an internal Pu-10.6Al alloy metal wafer. The thickness of each element is either 1/16 inch or 1/32 inch. The dimensions of each element ranges from 2 inches x 1 inch to 2 inches x 4 inches. This report discusses the potential chemical dissolution of the FCA clad material or stainless steel. This technology uses nitric acid-potassium fluoride (HNO 3-KF) flowsheets ofmore » H-Canyon to dissolve the FCA elements from a rack of materials. Historically, dissolution flowsheets have aimed to maximize Pu dissolution rates while minimizing stainless steel dissolution (corrosion) rates. Because the FCA cladding is made of stainless steel, this work sought to accelerate stainless steel dissolution.« less

Fuels irradiation testing for the SP-100 program

NASA Technical Reports Server (NTRS)

Makenas, Bruce J.; Hales, Janell W.; Ward, Alva L.

1991-01-01

An SP-100 fuel pin irradiation testing program is well on the way to providing data for performance correlations and demonstrating the lifetime and safety of the fuel system of the compact lithium-cooled reactor. Key SP-100 fuel performance issues addressed are the need for low fuel swelling and low fission gas release to minimize cladding strain, and the need for barrier integrity to prevent fuel/cladding chemical interaction. This paper provides a description of the irradiation test program that addresses these key issues and summarizes recent results of posttest examinations including data obtained at 6 atom percent goal burnup.

Bayesian network representing system dynamics in risk analysis of nuclear systems

NASA Astrophysics Data System (ADS)

Varuttamaseni, Athi

2011-12-01

A dynamic Bayesian network (DBN) model is used in conjunction with the alternating conditional expectation (ACE) regression method to analyze the risk associated with the loss of feedwater accident coupled with a subsequent initiation of the feed and bleed operation in the Zion-1 nuclear power plant. The use of the DBN allows the joint probability distribution to be factorized, enabling the analysis to be done on many simpler network structures rather than on one complicated structure. The construction of the DBN model assumes conditional independence relations among certain key reactor parameters. The choice of parameter to model is based on considerations of the macroscopic balance statements governing the behavior of the reactor under a quasi-static assumption. The DBN is used to relate the peak clad temperature to a set of independent variables that are known to be important in determining the success of the feed and bleed operation. A simple linear relationship is then used to relate the clad temperature to the core damage probability. To obtain a quantitative relationship among different nodes in the DBN, surrogates of the RELAP5 reactor transient analysis code are used. These surrogates are generated by applying the ACE algorithm to output data obtained from about 50 RELAP5 cases covering a wide range of the selected independent variables. These surrogates allow important safety parameters such as the fuel clad temperature to be expressed as a function of key reactor parameters such as the coolant temperature and pressure together with important independent variables such as the scram delay time. The time-dependent core damage probability is calculated by sampling the independent variables from their probability distributions and propagate the information up through the Bayesian network to give the clad temperature. With the knowledge of the clad temperature and the assumption that the core damage probability has a one-to-one relationship to it, we have calculated the core damage probably as a function of transient time. The use of the DBN model in combination with ACE allows risk analysis to be performed with much less effort than if the analysis were done using the standard techniques.

Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors

DOE PAGES

Yamamoto, Yukinori; Pint, Bruce A.; Terrani, Kurt A.; ...

2015-10-19

Development of nuclear grade, iron-based wrought FeCrAl alloys has been initiated for light water reactor (LWR) fuel cladding to serve as a substitute for zirconium-based alloys with enhanced accident tolerance. Ferritic alloys with sufficient chromium and aluminum additions can exhibit significantly improved oxidation kinetics in high-temperature steam environments when compared to zirconium-based alloys. In the first phase, a set of model FeCrAl alloys containing 10–20Cr, 3–5Al, and 0–0.12Y in weight percent, were prepared by conventional arc-melting and hot-working processes to explore the effect of composition on the properties of FeCrAlY alloys. It was found that the tensile properties were insensitivemore » to the alloy compositions studied; however, the steam oxidation resistance strongly depended on both the chromium and the aluminum contents. The second phase development focused on strengthening Fe-13Cr-5Al with minor alloying additions of molybdenum, niobium, and silicon. Combined with an optimized thermo-mechanical treatment, a thermally stable microstructure was produced with improved tensile properties at temperatures up to 741°C.« less

Advanced materials for space nuclear power systems

NASA Technical Reports Server (NTRS)

Titran, Robert H.; Grobstein, Toni L.; Ellis, David L.

1991-01-01

The overall philosophy of the research was to develop and characterize new high temperature power conversion and radiator materials and to provide spacecraft designers with material selection options and design information. Research on three candidate materials (carbide strengthened niobium alloy PWC-11 for fuel cladding, graphite fiber reinforced copper matrix composites for heat rejection fins, and tungsten fiber reinforced niobium matrix composites for fuel containment and structural supports considered for space power system applications is discussed. Each of these types of materials offers unique advantages for space power applications.

Measurement station for interim inspections of Lightbridge metallic fuel rods at the Halden Boiling Water Reactor

NASA Astrophysics Data System (ADS)

Hartmann, C.; Totemeier, A.; Holcombe, S.; Liverud, J.; Limi, M.; Hansen, J. E.; Navestad, E. AB(; )

2018-01-01

Lightbridge Corporation has developed a new Uranium-Zirconium based metallic fuel. The fuel rods aremanufactured via a co-extrusion process, and are characterized by their multi-lobed (cruciform-shaped) cross section. The fuel rods are also helically-twisted in the axial direction. Two experimental fuel assemblies, each containing four Lightbridge fuel rods, are scheduled to be irradiated in the Halden Boiling Water Reactor (HBWR) starting in 2018. In addition to on-line monitoring of fuel rod elongation and critical assembly conditions (e.g. power, flow rates, coolant temperatures, etc.) during the irradiation, several key parameters of the fuel will be measured out-of-core during interim inspections. An inspection measurement station for use in the irradiated fuel handling compartment at the HBWR has therefore been developed for this purpose. The multi-lobed cladding cross section combined with the spiral shape of the Lightbridge metallic fuel rods requires a high-precision guiding system to ensure good position repeatability combined with low-friction guiding. The measurement station is equipped with a combination of instruments and equipment supplied from third-party vendors and instruments and equipment developed at Institute for Energy Technology (IFE). Two sets of floating linear voltage differential transformer (LVDT) pairs are used to measure swelling and diameter changes between the lobes and the valleys over the length of the fuel rods. Eddy current probes are used to measure the thickness of oxide layers in the valleys and on the lobe tips and also to detect possible surface cracks/pores. The measurement station also accommodates gamma scans. Additionally, an eddy-current probe has been developed at IFE specifically to detect potential gaps or discontinuities in the bonding layer between the metallic fuel and the Zirconium alloy cladding. Potential gaps in the bonding layer will be hidden behind a 0.5-1.0 mm thick cladding wall. It has therefore been necessary to perform a careful design study of the probe geometry. For this, finite element analysis (FEA) has been performed in combination with practical validation tests on representative fuel dummies with machined flaws to find the probe geometry that best detects a hidden flaw. Tests performed thus far show that gaps down to 25 μm thickness can be detected with good repeatability and good discrimination from lift-off signals.

Performance evaluation and post-irradiation examination of a novel LWR fuel composed of U0.17ZrH1.6 fuel pellets bonded to Zircaloy-2 cladding by lead bismuth eutectic

NASA Astrophysics Data System (ADS)

Balooch, Mehdi; Olander, Donald R.; Terrani, Kurt A.; Hosemann, Peter; Casella, Andrew M.; Senor, David J.; Buck, Edgar C.

2017-04-01

A novel light water reactor fuel has been designed and fabricated at the University of California, Berkeley; irradiated at the Massachusetts Institute of Technology Reactor; and examined within the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. This fuel consists of U0.17ZrH1.6 fuel pellets core-drilled from TRIGA reactor fuel elements that are clad in Zircaloy-2 and bonded with lead-bismuth eutectic. The performance evaluation and post irradiation examination of this fuel are presented here.

Elastic Modulus Measurement of ORNL ATF FeCrAl Alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Thompson, Zachary T.; Terrani, Kurt A.; Yamamoto, Yukinori

2015-10-01

Elastic modulus and Poisson’s ratio for a number of wrought FeCrAl alloys, intended for accident tolerant fuel cladding application, are determined via resonant ultrasonic spectroscopy. The results are reported as a function of temperature from room temperature to 850°C. The wrought alloys were in the fully annealed and unirradiated state. The elastic modulus for the wrought FeCrAl alloys is at least twice that of Zr-based alloys over the temperature range of this study. The Poisson’s ratio of the alloys was 0.28 on average and increased very slightly with increasing temperature.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas; Burns, Joseph R.

The aftermath of the Tōhoku earthquake and the Fukushima accident has led to a global push to improve the safety of existing light water reactors. A key component of this initiative is the development of nuclear fuel and cladding materials with potentially enhanced accident tolerance, also known as accident-tolerant fuels (ATF). These materials are intended to improve core fuel and cladding integrity under beyond design basis accident conditions while maintaining or enhancing reactor performance and safety characteristics during normal operation. To complement research that has already been carried out to characterize ATF neutronics, the present study provides an initial investigationmore » of the sensitivity and uncertainty of ATF systems responses to nuclear cross section data. ATF concepts incorporate novel materials, including SiC and FeCrAl cladding and high density uranium silicide composite fuels, in turn introducing new cross section sensitivities and uncertainties which may behave differently from traditional fuel and cladding materials. In this paper, we conducted sensitivity and uncertainty analysis using the TSUNAMI-2D sequence of SCALE with infinite lattice models of ATF assemblies. Of all the ATF materials considered, it is found that radiative capture in 56Fe in FeCrAl cladding is the most significant contributor to eigenvalue uncertainty. 56Fe yields significant potential eigenvalue uncertainty associated with its radiative capture cross section; this is by far the largest ATF-specific uncertainty found in these cases, exceeding even those of uranium. We found that while significant new sensitivities indeed arise, the general sensitivity behavior of ATF assemblies does not markedly differ from traditional UO2/zirconium-based fuel/cladding systems, especially with regard to uncertainties associated with uranium. We assessed the similarity of the IPEN/MB-01 reactor benchmark model to application models with FeCrAl cladding. We used TSUNAMI-IP to calculate similarity indices of the application model and IPEN/MB-01 reactor benchmark model. This benchmark was selected for its use of SS304 as a cladding and structural material, with significant 56Fe content. The similarity indices suggest that while many differences in reactor physics arise from differences in design, sensitivity to and behavior of 56Fe absorption is comparable between systems, thus indicating the potential for this benchmark to reduce uncertainties in 56Fe radiative capture cross sections.« less

RERTR-9 Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. M. Perez

2011-05-01

The RERTR-9 experiment was designed to test the effect of modified fuel/clad interfaces in monolithic fuel plates and to demonstrate that the addition of Si to the matrix material in dispersion plates continued to be effective at high loading (~8.5 g U/cc). Several monolithic fuel plates were fabricated by Hot Isostatic Pressing (HIP) and Friction Bonding (FB) with thin layers of Si inserted and by HIP with a Zr diffusion barrier between the fuel and cladding. Si was applied to the interface by thermal spray of Al Si mixtures and by the insertion of thin Si-rich Al alloy foil betweenmore » the fuel/clad interface. The dispersion fuel plates were fabricated by semi-standard rolling techniques (the reduction by rolling was lowered to limit fabrication defects). Matrix materials consisted of Al-Si alloys and mixtures with various levels of Si. The following report summarizes the life of the RERTR-9A/B experiment through end of irradiation, including as-run neutronic analysis, thermal analysis and hydraulic testing results.« less

SELECTIVE SEPARATION OF URANIUM FROM FERRITIC STAINLESS STEELS

DOEpatents

Beaver, R.J.; Cherubini, J.H.

1963-05-14

A process is described for separating uranium from a nuclear fuel element comprising a uranium-containing core and a ferritic stainless steel clad by heating said element in a non-carburizing atmosphere at a temperature in the range 850-1050 un. Concent 85% C, rapidly cooling the heated element through the temperature range 815 un. Concent 85% to 650 EC to avoid annealing said steel, and then contacting the cooled element with an aqueous solution of nitric acid to selectively dissolve the uranium. (AEC)

Inert matrix fuel in dispersion type fuel elements

NASA Astrophysics Data System (ADS)

Savchenko, A. M.; Vatulin, A. V.; Morozov, A. V.; Sirotin, V. L.; Dobrikova, I. V.; Kulakov, G. V.; Ershov, S. A.; Kostomarov, V. P.; Stelyuk, Y. I.

2006-06-01

The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg-1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

Development of LWR Fuels with Enhanced Accident Tolerance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lahoda, Edward J.; Boylan, Frank A.

2015-10-30

Significant progress was made on the technical, licensing, and business aspects of the Westinghouse Electric Company’s Enhanced Accident Tolerant Fuel (ATF) by the Westinghouse ATF team. The fuel pellet options included waterproofed U 15N and U 3Si 2 and the cladding options SiC composites and zirconium alloys with surface treatments. Technology was developed that resulted in U 3Si 2 pellets with densities of >94% being achieved at the Idaho National Laboratory (INL). The use of U 3Si 2 will represent a 15% increase in U235 loadings over those in UO₂ fuel pellets. This technology was then applied to manufacture pelletsmore » for 6 test rodlets which were inserted in the Advanced Test Reactor (ATR) in early 2015 in zirconium alloy cladding. The first of these rodlets are expected to be removed in about 2017. Key characteristics to be determined include verification of the centerline temperature calculations, thermal conductivity, fission gas release, swelling and degree of amorphization. Waterproofed UN pellets have achieved >94% density for a 32% U 3Si 2/68% UN composite pellet at Texas A&M University. This represents a U235 increase of about 31% over current UO 2 pellets. Pellets and powders of UO 2, UN, and U 3Si 2the were tested by Westinghouse and Los Alamos National Laboratory (LANL) using differential scanning calorimetry to determine what their steam and 20% oxygen corrosion temperatures were as compared to UO 2. Cold spray application of either the amorphous steel or the Ti 2AlC was successful in forming an adherent ~20 micron coating that remained after testing at 420°C in a steam autoclave. Tests at 1200°C in 100% steam on coatings for Zr alloy have not been successful, possibly due to the low density of the coatings which allowed steam transport to the base zirconium metal. Significant modeling and testing has been carried out for the SiC/SiC composite/SiC monolith structures. A structure with the monolith on the outside and composite on the inside was developed which is the current baseline structure and a SiC to SiC tube closure approach. Permeability tests and mechanical tests were developed to verify the operation of the SiC cladding. Steam autoclave (420°C), high temperature (1200°C) flowing steam tests and quench tests were carried out with minimal corrosion, mechanical or hermeticity degradation effect on the SiC cladding or end plug closure. However, in-reactor loop tests carried out in the MIT reactor indicated an unacceptable degree of corrosion, likely due to the corrosive effect of radiolysis products which attacked the SiC.« less

Development of Metallic Fuels for Actinide Transmutation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hayes, Steven Lowe; Fielding, Randall Sidney; Benson, Michael Timothy

Research and development activities on metallic fuels are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is also a need for a near zero-loss fabrication process and a desire to demonstrate a multifold increase in burnup potential. The incorporation of Am and Np into the traditional U-20Pu-10Zr metallic fuel alloy was demonstrated in the US during the Integral Fast Reactor Program of the 1980’s and early 1990’s. However, the conventional counter gravity injection casting method performed under vacuum, previously used to fabricate these metallic fuel alloys, was not optimizedmore » for mitigating loss of the volatile Am constituent in the casting charge; as a result, approximately 40% of the Am casting charge failed to be incorporated into the as-cast fuel alloys. Fabrication development efforts of the past few years have pursued an optimized bottom-pour casting method to increase utilization of the melted charge to near 100%, and a differential pressure casting approach, performed under an argon overpressure, has been demonstrated to result in essentially no loss of Am due to volatilization during fabrication. In short, a path toward zero-loss fabrication of metallic fuels including minor actinides has been shown to be feasible. Irradiation testing of advanced metallic fuel alloys in the Advanced Test Reactor (ATR) has been underway since 2003. Testing in the ATR is performed inside of cadmium-shrouded positions to remove >99% of the thermal flux incident on the test fuels, resulting in an epi-thermal driven fuel test that is free from gross flux depression and producing an essentially prototypic radial temperature profile inside the fuel rodlets. To date, three irradiation test series (AFC-1,2,3) have been completed. Over 20 different metallic fuel alloys have been tested to burnups as high as 30% with constituent compositions of Pu up to 30%, Am up to 12%, Np up to 10%, and Zr between 10 and 60%. In general, the performance of all of these substantially disparate metallic fuel alloys has been observed to be excellent, and their irradiation behaviors are generally consistent with historic norms for metallic fuels without minor actinide additions and having lower Pu or Zr contents. Future work is being undertaken with a view toward increasing the burnup potential of metallic fuels even more. Design innovations under investigation include: 1) lowering the fuel smear density in order to accommodate more swelling, 2) annular fuel geometry to eliminate the need for a sodium bond, 3) minor alloy additions to stabilize lanthanide fission products inside the fuel and prevent their transport to the cladding where they can participate in fuel-cladding chemical interaction (FCCI), and 4) coatings/liners on the cladding inner surface to mitigate FCCI and enable higher temperature operation. This paper will present the current state of development of metallic fuels for actinide transmutation in the US. Highlights will include recent results from metallic fuel casting experiments, experiments to identify alloy additions to immobilize lanthanide fission products, and postirradiation examinations of annular metallic fuels at low burnup.« less

Steam Oxidation Testing in the Severe Accident Test Station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pint, Bruce A.

After the March 2011 accident at Fukushima Daiichi, Oak Ridge National Laboratory (ORNL) began conducting high temperature steam oxidation testing of candidate materials for accident tolerant fuel (ATF) cladding in August 2011 [1-11]. The ATF concept is to enhance safety margins in light water reactors (LWR) during severe accident scenarios by identifying materials with 100× slower steam oxidation rates compared to current Zr-based alloys. In 2012, the ORNL laboratory equipment was expanded and made available to the entire ATF community as the Severe Accident Test Station (SATS) [4,12]. Compared to the current UO2/Zr-based alloy fuel system, an ATF alternative wouldmore » significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident [13-14]. The steam oxidation behavior of candidate materials is a key metric in the evaluation of ATF concepts and also an important input into models [15-17]. However, initial modeling work of FeCrAl cladding has used incomplete information on the physical properties of FeCrAl. Also, the steam oxidation data being collected at 1200°-1700°C is unique as no prior work has considered steam oxidation of alloys at such high temperatures. Also, because many accident scenarios include steadily increasing temperatures, the required data are not traditional isothermal exposures but exposures with varying “ramp” rates. In some cases, the steam oxidation behavior has been surprising and difficult to interpret. Thus, more fundamental information continues to be collected. In addition, more work continues to focus on commercially-manufactured tube material. This report summarizes recent work to characterize the behavior of candidate alloys exposed to high temperature steam, evaluate steam oxidation behavior in various ramp scenarios and continue to collect integral data on FeCrAl compared to conventional Zr-based cladding.« less

Nuclear fuel element

DOEpatents

Meadowcroft, Ronald Ross; Bain, Alastair Stewart

1977-01-01

A nuclear fuel element wherein a tubular cladding of zirconium or a zirconium alloy has a fission gas plenum chamber which is held against collapse by the loops of a spacer in the form of a tube which has been deformed inwardly at three equally spaced, circumferential positions to provide three loops. A heat resistant disc of, say, graphite separates nuclear fuel pellets within the cladding from the plenum chamber. The spacer is of zirconium or a zirconium alloy.

Deformation behavior of laser welds in high temperature oxidation resistant Fe-Cr-Al alloys for fuel cladding applications

NASA Astrophysics Data System (ADS)

Field, Kevin G.; Gussev, Maxim N.; Yamamoto, Yukinori; Snead, Lance L.

2014-11-01

Ferritic-structured Fe-Cr-Al alloys are being developed and show promise as oxidation resistant accident tolerant light water reactor fuel cladding. This study focuses on investigating the weldability and post-weld mechanical behavior of three model alloys in a range of Fe-(13-17.5)Cr-(3-4.4)Al (wt.%) with a minor addition of yttrium using modern laser-welding techniques. A detailed study on the mechanical performance of bead-on-plate welds using sub-sized, flat dog-bone tensile specimens and digital image correlation (DIC) has been carried out to determine the performance of welds as a function of alloy composition. Results indicated a reduction in the yield strength within the fusion zone compared to the base metal. Yield strength reduction was found to be primarily constrained to the fusion zone due to grain coarsening with a less severe reduction in the heat affected zone. For all proposed alloys, laser welding resulted in a defect free weld devoid of cracking or inclusions.

Modeling of complex wear behavior associated with grid-to-rod fretting in light water nuclear reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blau, P. J.; Qu, J.; Lu, R.

One significant concern in the operation of light water nuclear reactors is the fretting wear damage to fuel cladding from flow-induced vibrations. For years, research on the grid-to-rod fretting (GTRF) phenomena has been underway in countries where nuclear power production is a significant industry. Under the auspices of the U.S. Department of Energy Consortium for Advanced Simulation of Light Water Reactors, an effort has been underway to develop and test an engineering wear model for zirconium alloy fuel rod cladding against a supporting grid. Furthermore, the multi-stage model accounts for oxide layers and wear rate transitions. Our paper describes themore » basis for a GTRF engineering wear model, the physical significance of the wear factor it contains, and recent progress toward model validation based on a fretting wear testing apparatus that accounts for coolant temperature, pressure, and the presence of periodic impacts (gaps) in grid/rod contact.« less

HRTEM and chemical study of an ion-irradiated chromium/zircaloy-4 interface

NASA Astrophysics Data System (ADS)

Wu, A.; Ribis, J.; Brachet, J.-C.; Clouet, E.; Leprêtre, F.; Bordas, E.; Arnal, B.

2018-06-01

Chromium-coated zirconium alloys are being studied as Enhanced Accident Tolerant Fuel Cladding for Light Water Reactors (LWRs). Those materials are especially studied to improve the oxidation resistance of LWRs current fuel claddings in nominal and at High Temperature (HT) for hypothetical accidental conditions such as LOss of Coolant Accident. Beyond their HT behavior, it is essential to assess the materials behavior under irradiation. A first generation chromium/Zircaloy-4 interface was thus irradiated with 20 MeV Kr8+ ions at 400 °C up to 10 dpa. High-Resolution Transmission Electron Microscopy and chemical analysis (EDS) were conducted at the Cr/Zr interface. The atomic structure of the interface reveals the presence of Zr(Fe, Cr)2 Laves phase, displaying both C14 and C15 structure. After irradiation, only the C14 structure was observed and atomic row matching was preserved across the different interfaces, thus ensuring a good adhesion of the coating after irradiation.

Modeling of complex wear behavior associated with grid-to-rod fretting in light water nuclear reactors

DOE PAGES

Blau, P. J.; Qu, J.; Lu, R.

2016-09-21

One significant concern in the operation of light water nuclear reactors is the fretting wear damage to fuel cladding from flow-induced vibrations. For years, research on the grid-to-rod fretting (GTRF) phenomena has been underway in countries where nuclear power production is a significant industry. Under the auspices of the U.S. Department of Energy Consortium for Advanced Simulation of Light Water Reactors, an effort has been underway to develop and test an engineering wear model for zirconium alloy fuel rod cladding against a supporting grid. Furthermore, the multi-stage model accounts for oxide layers and wear rate transitions. Our paper describes themore » basis for a GTRF engineering wear model, the physical significance of the wear factor it contains, and recent progress toward model validation based on a fretting wear testing apparatus that accounts for coolant temperature, pressure, and the presence of periodic impacts (gaps) in grid/rod contact.« less

Fuel development for gas-cooled fast reactors

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Fielding, R.; Gan, J.

2007-09-01

The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

Development of ODS FeCrAl alloys for accident-tolerant fuel cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dryepondt, Sebastien N.; Hoelzer, David T.; Pint, Bruce A.

2015-09-18

FeCrAl alloys are prime candidates for accident-tolerant fuel cladding due to their excellent oxidation resistance up to 1400 C and good mechanical properties at intermediate temperature. Former commercial oxide dispersion strengthened (ODS) FeCrAl alloys such as PM2000 exhibit significantly better tensile strength than wrought FeCrAl alloys, which would alloy for the fabrication of a very thin (~250 m) ODS FeCrAl cladding and limit the neutronic penalty from the replacement of Zr-based alloys by Fe-based alloys. Several Fe-12-Cr-5Al ODS alloys where therefore fabricated by ball milling FeCrAl powders with Y2O3 and additional oxides such as TiO 2 or ZrO 2. Themore » new Fe-12Cr-5Al ODS alloys showed excellent tensile strength up to 800 C but limited ductility. Good oxidation resistance in steam at 1200 and 1400 C was observed except for one ODS FeCrAl alloy containing Ti. Rolling trials were conducted at 300, 600 C and 800 C to simulate the fabrication of thin tube cladding and a plate thickness of ~0.6mm was reached before the formation of multiple edge cracks. Hardness measurements at different stages of the rolling process, before and after annealing for 1h at 1000 C, showed that a thinner plate thickness could likely be achieved by using a multi-step approach combining warm rolling and high temperature annealing. Finally, new Fe-10-12Cr-5.5-6Al-Z gas atomized powders have been purchased to fabricate the second generation of low-Cr ODS FeCrAl alloys. The main goals are to assess the effect of O, C, N and Zr contents on the ODS FeCrAl microstructure and mechanical properties, and to optimize the fabrication process to improve the ductility of the 2nd gen ODS FeCrAl while maintaining good mechanical strength and oxidation resistance.« less

High temperature gradient cobalt based clad developed using microwave hybrid heating

NASA Astrophysics Data System (ADS)

Prasad, C. Durga; Joladarashi, Sharnappa; Ramesh, M. R.; Sarkar, Anunoy

2018-04-01

The development of cobalt based cladding on a titanium substrate using microwave cladding technique is benchmark in coating area. The developed cladding would serve the function of a corrosion resistant coating under high temperatures. Clads of thickness 500 µm have been developed by microwave hybrid heating. A microwave furnace of 2.45GHz frequency was used at a 900W power level for processing. Impact of processing time on melting and adhesion of clad has been discussed. The study also extended to static thermal analysis of simple parts with cladding using commercial Finite Element analysis (FEA) software. A comparative study is explored between four variants of the clad being developed. The analysis has been conducted using a square sample. Similar temperature gradient is also shown for a proposed multi-layer coating, which includes a thermal barrier coating yttria stabilized zirconia (YSZ) on top of the corrosion resistant clad. The YSZ coating would protect the corrosion resistant cladding and substrate from high temperatures.

Overview of lower length scale model development for accident tolerant fuels regarding U3Si2 fuel and FeCrAl cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, Yongfeng

2016-09-01

U3Si2 and FeCrAl have been proposed as fuel and cladding concepts, respectively, for accident tolerance fuels with higher tolerance to accident scenarios compared to UO2. However, a lot of key physics and material properties regarding their in-pile performance are yet to be explored. To accelerate the understanding and reduce the cost of experimental studies, multiscale modeling and simulation are used to develop physics-based materials models to assist engineering scale fuel performance modeling. In this report, the lower-length-scale efforts in method and material model development supported by the Accident Tolerance Fuel (ATF) high-impact-problem (HIP) under the NEAMS program are summarized. Significantmore » progresses have been made regarding interatomic potential, phase field models for phase decomposition and gas bubble formation, and thermal conductivity for U3Si2 fuel, and precipitation in FeCrAl cladding. The accomplishments are very useful by providing atomistic and mesoscale tools, improving the current understanding, and delivering engineering scale models for these two ATF concepts.« less

Application of the SHARP Toolkit to Sodium-Cooled Fast Reactor Challenge Problems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shemon, E. R.; Yu, Y.; Kim, T. K.

The Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) toolkit is under development by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign of the U.S. Department of Energy, Office of Nuclear Energy. To better understand and exploit the benefits of advanced modeling simulations, the NEAMS Campaign initiated the “Sodium-Cooled Fast Reactor (SFR) Challenge Problems” task, which include the assessment of hot channel factors (HCFs) and the demonstration of zooming capability using the SHARP toolkit. If both challenge problems are resolved through advanced modeling and simulation using the SHARP toolkit, the economic competitiveness of a SFR can be significantly improved. The effortsmore » in the first year of this project focused on the development of computational models, meshes, and coupling procedures for multi-physics calculations using the neutronics (PROTEUS) and thermal-hydraulic (Nek5000) components of the SHARP toolkit, as well as demonstration of the HCF calculation capability for the 100 MWe Advanced Fast Reactor (AFR-100) design. Testing the feasibility of the SHARP zooming capability is planned in FY 2018. The HCFs developed for the earlier SFRs (FFTF, CRBR, and EBR-II) were reviewed, and a subset of these were identified as potential candidates for reduction or elimination through high-fidelity simulations. A one-way offline coupling method was used to evaluate the HCFs where the neutronics solver PROTEUS computes the power profile based on an assumed temperature, and the computational fluid dynamics solver Nek5000 evaluates the peak temperatures using the neutronics power profile. If the initial temperature profile used in the neutronics calculation is reasonably accurate, the one-way offline method is valid because the neutronics power profile has weak dependence on small temperature variation. In order to get more precise results, the proper temperature profile for initial neutronics calculations was obtained from the STAR-CCM+ calculations. The HCFs of the peak temperatures at cladding outer surface, cladding inner wall surface, and fuel centerline induced by cladding manufacturing tolerance and uncertainties on the cladding, coolant, and fuel properties were evaluated for the AFR-100. Some assessment on the effect of wire wrap configuration and size of the bundle shows that it is practical to use the 7-pin bare rod bundle to calculate the HCFs. The resulting HCFs obtained from the high-fidelity SHARP calculations are generally smaller than those developed for the earlier SFRs because the most uncertainties involved in the modeling and simulations were disappeared. For completeness, additional investigations are planned in FY 2018, which will use random sampling techniques.« less

Design of a fuel element for a lead-cooled fast reactor

NASA Astrophysics Data System (ADS)

Sobolev, V.; Malambu, E.; Abderrahim, H. Aït

2009-03-01

The options of a lead-cooled fast reactor (LFR) of the fourth generation (GEN-IV) reactor with the electric power of 600 MW are investigated in the ELSY Project. The fuel selection, design and optimization are important steps of the project. Three types of fuel are considered as candidates: highly enriched Pu-U mixed oxide (MOX) fuel for the first core, the MOX containing between 2.5% and 5.0% of the minor actinides (MA) for next core and Pu-U-MA nitride fuel as an advanced option. Reference fuel rods with claddings made of T91 ferrite-martensitic steel and two alternative fuel assembly designs (one uses a closed hexagonal wrapper and the other is an open square variant without wrapper) have been assessed. This study focuses on the core variant with the closed hexagonal fuel assemblies. Based on the neutronic parameters provided by Monte-Carlo modeling with MCNP5 and ALEPH codes, simulations have been carried out to assess the long-term thermal-mechanical behaviour of the hottest fuel rods. A modified version of the fuel performance code FEMAXI-SCK-1, adapted for fast neutron spectrum, new fuels, cladding materials and coolant, was utilized for these calculations. The obtained results show that the fuel rods can withstand more than four effective full power years under the normal operation conditions without pellet-cladding mechanical interaction (PCMI). In a variant with solid fuel pellets, a mild PCMI can appear during the fifth year, however, it remains at an acceptable level up to the end of operation when the peak fuel pellet burnup ∼80 MW d kg-1 of heavy metal (HM) and the maximum clad damage of about 82 displacements per atom (dpa) are reached. Annular pellets permit to delay PCMI for about 1 year. Based on the results of this simulation, further steps are envisioned for the optimization of the fuel rod design, aiming at achieving the fuel burnup of 100 MW d kg-1 of HM.

TEST SYSTEM FOR EVALUATING SPENT NUCLEAR FUEL BENDING STIFFNESS AND VIBRATION INTEGRITY

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong; Bevard, Bruce Balkcom

2013-01-01

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements specified by federal regulations. For normal conditions of transport, vibration loads incident to transport must be considered. This is particularly relevant for high-burnup fuel (>45 GWd/MTU). As the burnup of the fuel increases, a number of changes occur that may affect the performance of the fuel and cladding in storage and during transportation. The mechanical properties of high-burnup de-fueled cladding have been previously studied by subjecting defueled cladding tubes to longitudinal (axial) tensile tests, ring-stretch tests, ring-compression tests, and biaxial tube burst tests. The objective of this study ismore » to investigate the mechanical properties and behavior of both the cladding and the fuel in it under vibration/cyclic loads similar to the sustained vibration loads experienced during normal transport. The vibration loads to SNF rods during transportation can be characterized by dynamic, cyclic, bending loads. The transient vibration signals in a specified transport environment can be analyzed, and frequency, amplitude and phase components can be identified. The methodology being implemented is a novel approach to study the vibration integrity of actual SNF rod segments through testing and evaluating the fatigue performance of SNF rods at defined frequencies. Oak Ridge National Laboratory (ORNL) has developed a bending fatigue system to evaluate the response of the SNF rods to vibration loads. A three-point deflection measurement technique using linear variable differential transformers is used to characterize the bending rod curvature, and electromagnetic force linear motors are used as the driving system for mechanical loading. ORNL plans to use the test system in a hot cell for SNF vibration testing on high burnup, irradiated fuel to evaluate the pellet-clad interaction and bonding on the effective lifetime of fuel-clad structure bending fatigue performance. Technical challenges include pure bending implementation, remote installation and detachment of the SNF test specimen, test specimen deformation measurement, and identification of a driving system suitable for use in a hot cell. Surrogate test specimens have been used to calibrate the test setup and conduct systematic cyclic tests. The calibration and systematic cyclic tests have been used to identify test protocol issues prior to implementation in the hot cell. In addition, cyclic hardening in unidirectional bending and softening in reverse bending were observed in the surrogate test specimens. The interface bonding between the surrogate clad and pellets was found to impact the bending response of the surrogate rods; confirming this behavior in the actual spent fuel segments will be an important aspect of the hot cell test implementation,« less

The Development of Expansion Plug Wedge Test for Clad Tubing Structure Mechanical Property Evaluation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Jiang, Hao

2016-01-12

To determine the tensile properties of irradiated fuel cladding in a hot cell, a simple test was developed at the Oak Ridge National Laboratory (ORNL) and is described fully in US Patent Application 20060070455, “Expanded plug method for developing circumferential mechanical properties of tubular materials.” This method is designed for testing fuel rod cladding ductility in a hot cell using an expandable plug to stretch a small ring of irradiated cladding material. The specimen strain is determined using the measured diametrical expansion of the ring. This method removes many complexities associated with specimen preparation and testing. The advantages are themore » simplicity of measuring the test component assembly in the hot cell and the direct measurement of the specimen’s strain. It was also found that cladding strength could be determined from the test results.« less

METHOD OF MANUFACTURE OF METAL ENCASED CORE MATERIAL

DOEpatents

Peters, E.J.

1963-06-01

A method of making reactor fuel elements in which the fissionable material is encased and sealed in steel or aluminum cladding having an enclosed channel from the fissionable material to the surface of the cladding at one end is described. Heat and pressure sufficient to bond the assembled fuel element are applied in a nonoxidizing atmosphere. (AEC)

High-temperature steam oxidation and oxide crack effects of Zr-1Nb-1Sn-0.1Fe fuel cladding

NASA Astrophysics Data System (ADS)

Lee, Cheol Min; Mok, Yong-Kyoon; Sohn, Dong-Seong

2017-12-01

In this study, high-temperature steam oxidation experiments were performed at 1012-1207 °C on Zr-1Nb-1Sn-0.1Fe fuel cladding tubes to study their weight gains and microstructural characteristics. Many specimens were tested at each test temperature, and the results were reproducible and reliable. It is often debated whether the Zr-1Nb-1Sn-0.1Fe alloy follows the weight gain correlation developed by Cathcart and Pawel (C-P correlation) at around 1000 °C. According to our results, the C-P correlation overpredicts the weight gain at around 1000 °C, and this observation agrees well with the data reported by Westinghouse. In addition, the microstructures of the specimens were analyzed using scanning electron microscopy, and it was found that circumferential cracks are formed at the oxide-metal interface only at around 1000 °C. In previous studies, it has been postulated that cracks in the oxide promote the oxidation process, but it appears that the circumferential cracks at the oxide-metal interface decrease the oxidation rate before the breakaway oxidation occurs by disturbing the diffusion of oxygen. The oxidation rate reduction due to the circumferential cracks appears to be the reason for the overprediction of the C-P correlation at around 1000 °C.

FPIN2 posttest analysis of cylindrical canisters in SLSF Experiment P4

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hughes, T H; Kramer, J M

Results demonstrate that the clad deformation is dominated by the expansion of the fuel when it melts. In our analysis we moved the end space volume and some of the fuel-clad radial gap volume to an artificial central hole. This approximation may affect the details in the early parts of the transient, but clearly did not affect the major cladding deformation. It is also clear that the accuracy of the value of the fuel expansion upon melting is significant as is the dimensional accuracy of the fuel and canisters. The major conclusions from the FPIN2 posttest analysis of the cylindricalmore » canisters in SLSF Experiment P4 are: The maximum melt fractions in the two canisters were about 75%. Both canisters experienced about the same diametral strains of 12% prior to failure. These strains were almost entirely due to the additional volume that must be created inside the canisters to accommodate the expansion of fuel on melting. The mode of cladding failure was plastic instability by necking of the canister walls. The failure time of the 20% CW canister and the nonmechanical failure of the 10% CW canister are consistent with the FPIN2 calculations using the plastic instability failure criteria.« less

An investigation of FeCrAl cladding behavior under normal operating and loss of coolant conditions

DOE PAGES

Gamble, Kyle A.; Barani, Tommaso; Pizzocri, David; ...

2017-04-30

Iron-chromium-aluminum (FeCrAl) alloys are candidates to be used as nuclear fuel cladding for increased accident tolerance. An analysis of the response of FeCrAl under normal operating and loss of coolant conditions has been performed using fuel performance modeling. In particular, recent information on FeCrAl material properties and phenomena from separate effects tests has been implemented in the BISON fuel performance code and analyses of integral fuel rod behavior with FeCrAl cladding have been performed. BISON simulations included both light water reactor normal operation and loss-of-coolant accidental transients. In order to model fuel rod behavior during accidents, a cladding failure criterionmore » is desirable. For FeCrAl alloys, a failure criterion is developed using recent burst experiments under loss of coolant like conditions. The added material models are utilized to perform comparative studies with Zircaloy-4 under normal operating conditions and oxidizing and non-oxidizing out-of-pile loss of coolant conditions. The results indicate that for all conditions studied, FeCrAl behaves similarly to Zircaloy-4 with the exception of improved oxidation performance. Here, further experiments are required to confirm these observations.« less

An investigation of FeCrAl cladding behavior under normal operating and loss of coolant conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gamble, Kyle A.; Barani, Tommaso; Pizzocri, David

Iron-chromium-aluminum (FeCrAl) alloys are candidates to be used as nuclear fuel cladding for increased accident tolerance. An analysis of the response of FeCrAl under normal operating and loss of coolant conditions has been performed using fuel performance modeling. In particular, recent information on FeCrAl material properties and phenomena from separate effects tests has been implemented in the BISON fuel performance code and analyses of integral fuel rod behavior with FeCrAl cladding have been performed. BISON simulations included both light water reactor normal operation and loss-of-coolant accidental transients. In order to model fuel rod behavior during accidents, a cladding failure criterionmore » is desirable. For FeCrAl alloys, a failure criterion is developed using recent burst experiments under loss of coolant like conditions. The added material models are utilized to perform comparative studies with Zircaloy-4 under normal operating conditions and oxidizing and non-oxidizing out-of-pile loss of coolant conditions. The results indicate that for all conditions studied, FeCrAl behaves similarly to Zircaloy-4 with the exception of improved oxidation performance. Here, further experiments are required to confirm these observations.« less

Critical Safe Disposal of Spent Fuel: Behavior of Neutron Poisons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kienzler, Bernhard; Gmal, Bernhard

2007-07-01

In contrast to Yucca Mountain, European repository concepts rely on deep underground conditions which guarantee permanently a reducing geochemical environment. As long as no water comes into contact with the disposed nuclear fuel, criticality is excluded by compliance with the disposal conditions (limitation of U/Pu in the canisters). Penetration of water into the canister may also be considered as a scenario. However, water in a disposal results in geochemical reactions proceeding over very long periods of time: (1) Presence of water allows the corrosion of the steel of the canister material forming hydrogen and iron corrosion products. (2) Hydrogen pressuresmore » affect the zircaloy cladding even at low temperatures. Failure of fuel cladding and spacers leads to changes in the geometrical configuration. (3) UO{sub 2} matrix corrosion results in geochemically controlled reformation of secondary phase. (4) Even if the dissolution rate of UO{sub 2} is low, elements accounting for burnup credit do not behave similar as uranium. Geochemical reactions are analyzed in detail and compositions are presented which have a high probability to be formed in the long-term needing to be analyzed with respect to K{sub eff}. (authors)« less

Effect of the oxidation front penetration on in-clad hydrogen migration

NASA Astrophysics Data System (ADS)

Feria, F.; Herranz, L. E.

2018-03-01

In LWR fuel claddings the embrittlement due to hydrogen precipitates (i.e., hydrides) is a degrading mechanism that concerns in nuclear safety, particularly in dry storage. A relevant factor is the radial distribution of the hydrogen absorbed, especially the hydride rim formed. Thus, a reliable assessment of fuel performance should account for hydrogen migration. Based on the current state of modelling of hydrogen dynamics in the cladding, a 1D radial model has been derived and coupled with the FRAPCON code. The model includes the effect of the oxidation front progression on in-clad hydrogen migration, based on experimental observations found (i.e., dissolution/diffusion/re-precipitation of the hydrogen in the matrix ahead of the oxidation front). A remarkable quantitative impact of this new contribution has been shown by analyzing the hydrogen profile across the cladding of several high burnup fuel scenarios (>60 GW d/tU); other potential contributions like thermodiffusion and diffusion in the hydride phase hardly make any difference. Comparisons against PIE measurements allow concluding that the model accuracy notably increases when the effect of the oxidation front is accounted for in the hydride rim formation. In spite of the promising results, further validation would be needed.

Occurence and prediction of sigma phase in fuel cladding alloys for breeder reactors. [LMFBR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Anantatmula, R.P.

1982-01-01

In sodium-cooled fast reactor systems, fuel cladding materials will be exposed for several thousand hours to liquid sodium. Satisfactory performance of the materials depends in part on the sodium compatibility and phase stability of the materials. This paper mainly deals with the phase stability aspect, with particular emphasis on sigma phase formation of the cladding materials upon extended exposures to liquid sodium. A new method of predicting sigma phase formation is proposed for austenitic stainless steels and predictions are compared with the experimental results on fuel cladding materials. Excellent agreement is obtained between theory and experiment. The new method ismore » different from the empirical methods suggested for superalloys and does not suffer from the same drawbacks. The present method uses the Fe-Cr-Ni ternary phase diagram for predicting the sigma-forming tendencies and exhibits a wide range of applicability to austenitic stainless steels and heat-resistant Fe-Cr-Ni alloys.« less

Kr ion irradiation study of the depleted-uranium alloys

NASA Astrophysics Data System (ADS)

Gan, J.; Keiser, D. D.; Miller, B. D.; Kirk, M. A.; Rest, J.; Allen, T. R.; Wachs, D. M.

2010-12-01

Fuel development for the reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium nuclear fuels that can be employed to replace existing high enrichment uranium fuels currently used in some research reactors throughout the world. For dispersion type fuels, radiation stability of the fuel-cladding interaction product has a strong impact on fuel performance. Three depleted-uranium alloys are cast for the radiation stability studies of the fuel-cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Al, Si) 3, (U, Mo)(Al, Si) 3, UMo 2Al 20, U 6Mo 4Al 43 and UAl 4. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 °C to ion doses up to 2.5 × 10 19 ions/m 2 (˜10 dpa) with an Kr ion flux of 10 16 ions/m 2/s (˜4.0 × 10 -3 dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

High burnup fuel behavior related to fission gas effects under reactivity initiated accidents (RIA) conditions

NASA Astrophysics Data System (ADS)

Lemoine, F.

1997-09-01

Specific aspects of irradiated fuel result from the increasing retention of gaseous and volatile fission products with burnup, which, under overpower conditions, can lead to solid fuel pressurization and swelling causing severe PCMI (pellet clad mechanical interaction). In order to assess the reliability of high burnup fuel under RIAs, experimental programs have been initiated which have provided important data concerning the transient fission gas behavior and the clad loading mechanisms. The importance of the rim zone is demonstrated based on three experiments resulting in clad failure at low enthalpy, which are explained by energetic considerations. High gas release in non-failure tests with low energy deposition underlines the importance of grain boundary and porosity gas. Measured final releases are strongly correlated to the microstructure evolution, depending on energy deposition, pulse width, initial and refabricated fuel rod design. Observed helium release can also increase internal pressure and gives hints to the gas behavior understanding.

Design and testing a high fuel volume fraction, externally finned, thermionic emitter.

NASA Technical Reports Server (NTRS)

Peelgren, M. L.; Ernst, D. M.

1971-01-01

A prototypical, high fuel volume fraction, thermionic emitter body was designed and tested. The emitter body is all tungsten, with a 1.40-cm ID, a 3.23-cm OD, and eight full-length axial fins. The emitter thickness is 0.15 cm while the fins and outer clad are 0.075 cm thick. Different methods of fabrication were used in making the test samples. Stress analysis was performed with a three-dimensional elastic code. Thermal testing of the samples, duplicating calculated radial temperature gradients, heatup and cooldown rates, and emitter body temperatures in operation, was performed with no structural failures noted (six heatup and cooldown cycles per sample). Further emitter analysis and testing is planned.

Supercritical Brayton Cycle Nuclear Power System Concepts

NASA Astrophysics Data System (ADS)

Wright, Steven A.

2007-01-01

Both the NASA and DOE have programs that are investigating advanced power conversion cycles for planetary surface power on the moon or Mars, and for next generation nuclear power plants on earth. The gas Brayton cycle offers many practical solutions for space nuclear power systems and was selected as the nuclear power system of choice for the NASA Prometheus project. An alternative Brayton cycle that offers high efficiency at a lower reactor coolant outlet temperature is the supercritical Brayton cycle (SCBC). The supercritical cycle is a true Brayton cycle because it uses a single phase fluid with a compressor inlet temperature that is just above the critical point of the fluid. This paper describes the use of a supercritical Brayton cycle that achieves a cycle efficiency of 26.6% with a peak coolant temperature of 750 K and for a compressor inlet temperature of 390 K. The working fluid uses a clear odorless, nontoxic refrigerant C318 perflurocarbon (C4F8) that always operates in the gas phase. This coolant was selected because it has a critical temperature and pressure of 388.38 K and 2.777 MPa. The relatively high critical temperature allows for efficient thermal radiation that keeps the radiator mass small. The SCBC achieves high efficiency because the loop design takes advantage of the non-ideal nature of the coolant equation of state just above the critical point. The lower coolant temperature means that metal fuels, uranium oxide fuels, and uranium zirconium hydride fuels with stainless steel, ferretic steel, or superalloy cladding can be used with little mass penalty or reduction in cycle efficiency. The reactor can use liquid-metal coolants and no high temperature heat exchangers need to be developed. Indirect gas cooling or perhaps even direct gas cooling can be used if the C4F8 coolant is found to be sufficiently radiation tolerant. Other fluids can also be used in the supercritical Brayton cycle including Propane (C3H8, Tcritical = 369 K) and Hexane (C6H14, Tcritical = 506.1 K) provided they have adequate chemical compatibility and stability. Overall the use of supercritical Brayton cycles may offer ``break through'' operating capabilities for space nuclear power plants because high efficiencies can be achieved a very low reactor operating temperatures which in turn allows for the use of available fuels, cladding, and structural materials.

Analysis of Loss-of-Coolant Accidents in the NIST Research Reactor - Early Phase

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baek, Joo S.; Diamond, David

A study of the fuel temperature during the early phase of a loss-of-coolant accident (LOCA) in the NIST research reactor (NBSR) was completed. Previous studies had been reported in the preliminary safety analysis report for the conversion of the NBSR from high-enriched uranium (HEU) fuel to low-enriched (LEU) fuel. Those studies had focused on the most vulnerable LOCA situation, namely, a double-ended guillotine break in the time period after reactor trip when water is drained from either the coolant channels inside the fuel elements or the region outside the fuel elements. The current study fills in a gap in themore » analysis which is the early phase of the event when there may still be water present but the reactor is at power or immediately after reactor trip and pumps have tripped. The calculations were done, for both the current HEU-fueled core and the proposed LEU core, with the TRACE thermal-hydraulic systems code. Several break locations and different break sizes were considered. In all cases the increase in the clad (or fuel meat) temperature was relatively small so that a large margin to the temperature threshold for blistering (the Safety Limit for the NBSR) remained.« less

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

NASA Astrophysics Data System (ADS)

Montgomery, Robert; Tomé, Carlos; Liu, Wenfeng; Alankar, Alankar; Subramanian, Gopinath; Stanek, Christopher

2017-01-01

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical constitutive models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. To improve upon this approach, a microstructurally-based zirconium alloy mechanical deformation analysis capability is being developed within the United States Department of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL). Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed by Lebensohn and Tomé [1], has been coupled with the BISON engineering scale fuel performance code to represent the mechanistic material processes controlling the deformation behavior of light water reactor (LWR) cladding. A critical component of VPSC is the representation of the crystallographic nature (defect and dislocation movement) and orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. A future goal is for VPSC to obtain information on reaction rate kinetics from atomistic calculations to inform the defect and dislocation behavior models described in VPSC. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON and provides initial results utilizing the coupled functionality.

Sensitivity analysis for best-estimate thermal models of vertical dry cask storage systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

DeVoe, Remy R.; Robb, Kevin R.; Skutnik, Steven E.

Loading requirements for dry cask storage of spent nuclear fuel are driven primarily by decay heat capacity limitations, which themselves are determined through recommended limits on peak cladding temperature within the cask. This study examines the relative sensitivity of peak material temperatures within the cask to parameters that influence both the stored fuel residual decay heat as well as heat removal mechanisms. Here, these parameters include the detailed reactor operating history parameters (e.g., soluble boron concentrations and the presence of burnable poisons) as well as factors that influence heat removal, including non-dominant processes (such as conduction from the fuel basketmore » to the canister and radiation within the canister) and ambient environmental conditions. By examining the factors that drive heat removal from the cask alongside well-understood factors that drive decay heat, it is therefore possible to make a contextual analysis of the most important parameters to evaluation of peak material temperatures within the cask.« less

Sensitivity analysis for best-estimate thermal models of vertical dry cask storage systems

DOE PAGES

DeVoe, Remy R.; Robb, Kevin R.; Skutnik, Steven E.

2017-07-08

Loading requirements for dry cask storage of spent nuclear fuel are driven primarily by decay heat capacity limitations, which themselves are determined through recommended limits on peak cladding temperature within the cask. This study examines the relative sensitivity of peak material temperatures within the cask to parameters that influence both the stored fuel residual decay heat as well as heat removal mechanisms. Here, these parameters include the detailed reactor operating history parameters (e.g., soluble boron concentrations and the presence of burnable poisons) as well as factors that influence heat removal, including non-dominant processes (such as conduction from the fuel basketmore » to the canister and radiation within the canister) and ambient environmental conditions. By examining the factors that drive heat removal from the cask alongside well-understood factors that drive decay heat, it is therefore possible to make a contextual analysis of the most important parameters to evaluation of peak material temperatures within the cask.« less

Verification and Validation of the BISON Fuel Performance Code for PCMI Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gamble, Kyle Allan Lawrence; Novascone, Stephen Rhead; Gardner, Russell James

2016-06-01

BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. A brief overview of BISON’s computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described. Validation for application to light water reactor (LWR) PCMI problems is assessed by comparing predicted and measured rod diameter following base irradiation andmore » power ramps. Results indicate a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. Initial rod diameter comparisons have led to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.« less

RERTR-12 Post-irradiation Examination Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rice, Francine; Williams, Walter; Robinson, Adam

2015-02-01

The following report contains the results and conclusions for the post irradiation examinations performed on RERTR-12 Insertion 2 experiment plates. These exams include eddy-current testing to measure oxide growth; neutron radiography for evaluating the condition of the fuel prior to sectioning and determination of fuel relocation and geometry changes; gamma scanning to provide relative measurements for burnup and indication of fuel- and fission-product relocation; profilometry to measure dimensional changes of the fuel plate; analytical chemistry to benchmark the physics burnup calculations; metallography to examine the microstructural changes in the fuel, interlayer and cladding; and microhardness testing to determine the material-propertymore » changes of the fuel and cladding.« less

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-05-01

A nuclear reactor fuel element comprising high density ceramic fissionable material enclosed in a tubular cladding of corrosion-resistant material is described. The fissionable material is in the form of segments of a tube which have cooperating tapered interfaces which produce outward radial displacement when the segments are urged axially together. A resilient means is provided within the tubular housing to constantly urge the fuel segments axially. This design maintains the fuel material in tight contacting engagement against the inner surface of the outer cladding tube to eliminate any gap therebetween which may be caused by differential thermal expansion between the fuel material and the material of the tube.

Protected Nuclear Fuel Element

DOEpatents

Kittel, J. H.; Schumar, J. F.

1962-12-01

A stainless steel-clad actinide metal fuel rod for use in fast reactors is reported. In order to prevert cladding failures due to alloy formation between the actinide metal and the stainless steel, a mesh-like sleeve of expanded metal is interposed between them, the sleeve metal being of niobium, tantalum, molybdenum, tungsten, zirconium, or vanadium. Liquid alkali metal is added as a heat transfer agent. (AEC)

Hydrogen motion in Zircaloy-4 cladding during a LOCA transient

NASA Astrophysics Data System (ADS)

Elodie, T.; Jean, D.; Séverine, G.; M-Christine, B.; Michel, C.; Berger, P.; Martine, B.; Antoine, A.

2016-04-01

Hydrogen and oxygen are key elements influencing the embrittlement of zirconium-based nuclear fuel cladding during the quench phase following a Loss Of Coolant Accident (LOCA). The understanding of the mechanisms influencing the motion of these two chemical elements in the metal is required to fully describe the material embrittlement. High temperature steam oxidation tests were performed on pre-hydrided Zircaloy-4 samples with hydrogen contents ranging between 11 and 400 wppm prior to LOCA transient. Thanks to the use of both Electron Probe Micro-Analysis (EPMA) and Elastic Recoil Detection Analysis (μ-ERDA), the chemical elements partitioning has been systematically quantified inside the prior-β phase. Image analysis and metallographic examinations were combined to provide an average oxygen profile as well as hydrogen profile within the cladding thickness after LOCA transient. The measured hydrogen profile is far from homogeneous. Experimental distributions are compared to those predicted numerically using calculations derived from a finite difference thermo-diffusion code (DIFFOX) developed at IRSN.

General-Purpose Heat Source Safety Verification Test Program: Edge-on flyer plate tests

NASA Astrophysics Data System (ADS)

George, T. G.

1987-03-01

The radioisotope thermoelectric generator (RTG) that will supply power for the Galileo and Ulysses space missions contains 18 General-Purpose Heat Source (GPHS) modules. The GPHS modules provide power by transmitting the heat of Pu-238 alpha-decay to an array of thermoelectric elements. Each module contains four Pu-238O2-fueled clads and generates 250 W(t). Because the possibility of a launch vehicle explosion always exists, and because such an explosion could generate a field of high-energy fragments, the fueled clads within each GPHS module must survive fragment impact. The edge-on flyer plate tests were included in the Safety Verification Test series to provide information on the module/clad response to the impact of high-energy plate fragments. The test results indicate that the edge-on impact of a 3.2-mm-thick, aluminum-alloy (2219-T87) plate traveling at 915 m/s causes the complete release of fuel from capsules contained within a bare GPHS module, and that the threshold velocity sufficient to cause the breach of a bare, simulant-fueled clad impacted by a 3.5-mm-thick, aluminum-alloy (5052-TO) plate is approximately 140 m/s.

SiC-CMC-Zircaloy-4 Nuclear Fuel Cladding Performance during 4-Point Tubular Bend Testing

DOE Office of Scientific and Technical Information (OSTI.GOV)

IJ van Rooyen; WR Lloyd; TL Trowbridge

2013-09-01

The U.S. Department of Energy Office of Nuclear Energy (DOE NE) established the Light Water Reactor Sustainability (LWRS) program to develop technologies and other solutions to improve the reliability, sustain the safety, and extend the life of current reactors. The Advanced LWR Nuclear Fuel Development Pathway in the LWRS program encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. Recent investigations of potential options for “accident tolerant” nuclear fuel systems point to the potential benefits of silicon carbide (SiC) cladding. One of the proposed SiC-based fuel cladding designsmore » being investigated incorporates a SiC ceramic matrix composite (CMC) as a structural material supplementing an internal Zircaloy-4 (Zr-4) liner tube, referred to as the hybrid clad design. Characterization of the advanced cladding designs will include a number of out-of-pile (nonnuclear) tests, followed by in-pile irradiation testing of the most promising designs. One of the out-of-pile characterization tests provides measurement of the mechanical properties of the cladding tube using four point bend testing. Although the material properties of the different subsystems (materials) will be determined separately, in this paper we present results of 4-point bending tests performed on fully assembled hybrid cladding tube mock-ups, an assembled Zr-4 cladding tube mock-up as a standard and initial testing results on bare SiC-CMC sleeves to assist in defining design parameters. The hybrid mock-up samples incorporated SiC-CMC sleeves fabricated with 7 polymer impregnation and pyrolysis (PIP) cycles. To provide comparative information; both 1- and 2-ply braided SiC-CMC sleeves were used in this development study. Preliminary stress simulations were performed using the BISON nuclear fuel performance code to show the stress distribution differences for varying lengths between loading points and clad configurations. The 2-ply sleeve samples show a higher bend momentum compared to those of the 1-ply sleeve samples. This is applicable to both the hybrid mock-up and bare SiC-CMC sleeve samples. Comparatively both the 1- and 2-ply hybrid mock-up samples showed a higher bend stiffness and strength compared with the standard Zr-4 mock-up sample. The characterization of the hybrid mock-up samples showed signs of distress and preliminary signs of fraying at the protective Zr-4 sleeve areas for the 1-ply SiC-CMC sleeve. In addition, the microstructure of the SiC matrix near the cracks at the region of highest compressive bending strain shows significant cracking and flaking. The 2-ply SiC-CMC sleeve samples showed a more bonded, cohesive SiC matrix structure. This cracking and fraying causes concern for increased fretting during the actual use of the design. Tomography was proven as a successful tool to identify open porosity during pre-test characterization. Although there is currently insufficient data to make conclusive statements regarding the overall merit of the hybrid cladding design, preliminary characterization of this novel design has been demonstrated.« less

NATCRCTR: One-dimensional thermal-hydraulics analysis code for natural-circulation TRIGA reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Rubinaccio, G.

1996-12-31

The Pennsylvania State University nuclear engineering department is evaluating the upgrade of the Reed College (Portland, Oregon) TRIGA reactor from 250 kW to 1 MW in two areas: thermal-hydraulics and steady-state neutronics analysis. This analysis was initiated as a cooperative effort between Penn State and Reed College as a training project for two International Atomic Energy Agency (IAEA) fellows from Ghana. The two Ghanaian IAEA fellows were assisted by G. Rubinaccio, an undergraduate, who undertook the task of writing the new computer programs for the thermal-hydraulic and physics evaluation as a three-credit special design project course. The Reed College TRIGA,more » which has a fixed graphite radial reflector, is cooled by natural circulation, without external cross-flow; whereas, the Penn State Breazeale Reactor has significant crossflow into its sides. To model the Reed TRIGA, the NATCRCTR program has been developed from first principles using the following assumptions: 1. The core is surrounded by the fixed reflector structure, which acts as a one-dimensional channel. 2. The core inlet temperature distribution is constant at the core bottom. 3. The axial heat flux distribution is a chopped cosine shape. 4. The heat transfer in the fuel is primarily in the radial directions. 5. A small gap between the fuel and cladding exists. The NATCRCTR code is used to find the peak centerline fuel, gap, and cladding surface temperatures, based on assumed flux and engineering peaking factors.« less

Design Evolutuion of Hot Isotatic Press Cans for NTP Cermet Fuel Fabrication

NASA Technical Reports Server (NTRS)

Mireles, O. R.; Broadway, J.; Hickman, R.

2014-01-01

Nuclear Thermal Propulsion (NTP) is under consideration for potential use in deep space exploration missions due to desirable performance properties such as a high specific impulse (> 850 seconds). Tungsten (W)-60vol%UO2 cermet fuel elements are under development, with efforts emphasizing fabrication, performance testing and process optimization to meet NTP service life requirements [1]. Fuel elements incorporate design features that provide redundant protection from crack initiation, crack propagation potentially resulting in hot hydrogen (H2) reduction of UO2 kernels. Fuel erosion and fission product retention barriers include W coated UO2 fuel kernels, W clad internal flow channels and fuel element external W clad resulting in a fully encapsulated fuel element design as shown.

Model calculation of Cr dissolution behavior of ODS ferritic steel in high-temperature flowing sodium environment

NASA Astrophysics Data System (ADS)

Ohtsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Kato, Shoichi; Furukawa, Tomohiro; Kaito, Takeji

2018-07-01

A calculation model was constructed to systematically study the effects of environmental conditions (i.e. Cr concentration in sodium, test temperature, axial temperature gradient of fuel pin, and sodium flow velocity) on Cr dissolution behavior. Chromium dissolution was largely influenced by small changes in Cr concentration (i.e. chemical potential of Cr) in liquid sodium in the model calculation. Chromium concentration in sodium coolant, therefore, should be recognized as a critical parameter for the prediction and management of Cr dissolution behavior in the sodium-cooled fast reactor (SFR) core. Because the fuel column length showed no impact on dissolution behavior in the model calculation, no significant downstream effects possibly take place in the SFR fuel cladding tube due to the much shorter length compared with sodium loops in the SFR plant and the large axial temperature gradient. The calculated profile of Cr concentration along the wall-thickness direction was consistent with that measured in BOR-60 irradiation test where Cr concentration in inlet sodium bulk flow was set at 0.07 wt ppm in the calculation.

Dry Storage of Research Reactor Spent Nuclear Fuel - 13321

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adams, T.M.; Dunsmuir, M.D.; Leduc, D.R.

2013-07-01

Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. Themore » initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry storage requires integration with current facility operations, and selection of equipment that will allow safe operation within the constraints of existing facility conditions. Examples of such constraints that are evaluated and addressed by the dry storage program include limited basin depth, varying fuel lengths up to 4 m, (13 ft), fissile loading limits, canister closure design, post-load drying and closure of the canisters, instrument selection and installation, and movement of the canisters to storage casks. The initial pilot phase restricts the fuels to shorter length fuels that can be loaded to the canister directly underwater; subsequent phases will require use of a shielded transfer system. Removal of the canister from the basin, followed by drying, inerting, closure of the canister, and transfer of the canister to the storage cask are completed with remotely operated equipment and appropriate shielding to reduce personnel radiation exposure. (authors)« less

A model for recovery of scrap monolithic uranium molybdenum fuel by electrorefining

NASA Astrophysics Data System (ADS)

Van Kleeck, Melissa A.

The goal of the Reduced Enrichment for Research and Test Reactors program (RERTR) is toreduce enrichment at research and test reactors, thereby decreasing proliferation risk at these facilities. A new fuel to accomplish this goal is being manufactured experimentally at the Y12 National Security Complex. This new fuel will require its own waste management procedure,namely for the recovery of scrap from its manufacture. The new fuel is a monolithic uraniummolybdenum alloy clad in zirconium. Feasibility tests were conducted in the Planar Electrode Electrorefiner using scrap U-8Mo fuel alloy. These tests proved that a uranium product could be recovered free of molybdenum from this scrap fuel by electrorefining. Tests were also conducted using U-10Mo Zr clad fuel, which confirmed that product could be recovered from a clad version of this scrap fuel at an engineering scale, though analytical results are pending for the behavior of Zr in the electrorefiner. A model was constructed for the simulation of electrorefining the scrap material produced in the manufacture of this fuel. The model was implemented on two platforms, Microsoft Excel and MatLab. Correlations, used in the model, were developed experimentally, describing area specific resistance behavior at each electrode. Experiments validating the model were conducted using scrap of U-10Mo Zr clad fuel in the Planar Electrode Electrorefiner. The results of model simulations on both platforms were compared to experimental results for the same fuel, salt and electrorefiner compositions and dimensions for two trials. In general, the model demonstrated behavior similar to experimental data but additional refinements are needed to improve its accuracy. These refinements consist of a function for surface area at anode and cathode based on charge passed. Several approximations were made in the model concerning areas of electrodes which should be replaced by a more accurate function describing these areas.

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Montgomery, Robert, E-mail: robert.montgomery@pnnl.gov; Tomé, Carlos, E-mail: tome@lanl.gov; Liu, Wenfeng, E-mail: wenfeng.liu@anatech.com

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical constitutive models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. To improve upon this approach, a microstructurally-based zirconium alloy mechanical deformation analysis capability is being developed within the United States Department of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL). Specifically, the viscoplastic self-consistent (VPSC)more » polycrystal plasticity modeling approach, developed by Lebensohn and Tomé [1], has been coupled with the BISON engineering scale fuel performance code to represent the mechanistic material processes controlling the deformation behavior of light water reactor (LWR) cladding. A critical component of VPSC is the representation of the crystallographic nature (defect and dislocation movement) and orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. A future goal is for VPSC to obtain information on reaction rate kinetics from atomistic calculations to inform the defect and dislocation behavior models described in VPSC. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON and provides initial results utilizing the coupled functionality.« less

Modeling and Simulation of Used Nuclear Fuel During Transportation with Consideration of Hydride Effects and Cyclic Fatigue

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chakraborty, Pritam; Sabharwall, Piyush; Spears, Robert Edward

2015-09-30

The objective of this work is to understand the integrity of Used Nuclear Fuel (UNF) during transportation. Previous analysis work has been performed to look at the integrity of UNF during transportation but these analyses have neglected to analyze the effect of hydrides and flaws (fracture mechanics models to capture radial cracking in the cladding). In this study, the clad regions of interest are near the pellet-pellet interfaces. These regions can experience more complex stress-states than the rest of the clad during cooling and have a greater possibility to develop radially reoriented hydrides during vacuum drying.

Summary of LCRE fuel element design including supporting experimental data

DOE Office of Scientific and Technical Information (OSTI.GOV)

None, None

Declassified 18 Sep 1973. The design basis of the LCRE fuel pin is presented. The fuel pin consists of a Cb-1 Zr alloy cladding tube 0.305 inch diameter, 0.015 inch wall thickness and 35.96 inches long. The active fuel section is 13.5 inches long, with top and bottom reflector rods each 6.9 inches long and with a 4 inch gas accumulation space at each end. The cladding is designed as a pressure vessel to contain the gases released from the fuel and end refiector materials, which results in an internal gas pressure buildup in the pins during reactor operation. (23more » referencea) (auth)« less

Effects of thermal treatment on the co-rolled U-Mo fuel foils

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dennis D. Keiser, Jr.; Tammy L. Trowbridge; Cynthia R. Breckenridge

2014-11-01

A monolithic fuel type is being developed to convert US high performance research and test reactors such as Advanced Test Reactor (ATR) at Idaho National Laboratory from highly enriched uranium (HEU) to low-enriched uranium (LEU). The interaction between the cladding and the U-Mo fuel meat during fuel fabrication and irradiation is known to have negative impacts on fuel performance, such as mechanical integrity and dimensional stability. In order to eliminate/minimize the direct interaction between cladding and fuel meat, a thin zirconium diffusion barrier was introduced between the cladding and U-Mo fuel meat through a co-rolling process. A complex interface betweenmore » the zirconium and U-Mo was developed during the co-rolling process. A predictable interface between zirconium and U-Mo is critical to achieve good fuel performance since the interfaces can be the weakest link in the monolithic fuel system. A post co-rolling annealing treatment is expected to create a well-controlled interface between zirconium and U-Mo. A systematic study utilizing post co-rolling annealing treatment has been carried out. Based on microscopy results, the impacts of the annealing treatment on the interface between zirconium and U-Mo will be presented and an optima annealing treatment schedule will be suggested. The effects of the annealing treatment on the fuel performance will also be discussed.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Montgomery, Rose; Scaglione, John M; Bevard, Bruce Balkcom

The High Burnup Spent Fuel Data project pulled 25 sister rods (9 from the project assemblies and 16 from similar HBU assemblies) for characterization. The 25 sister rods are all high burnup and cover the range of modern domestic cladding alloys. The 25 sister rods were shipped to Oak Ridge National Laboratory (ORNL) in early 2016 for detailed non-destructive and destructive examination. Examinations are intended to provide baseline data on the initial physical state of the cladding and fuel prior to the loading, drying, and long-term dry storage process. Further examinations are focused on determining the effects of temperatures encounteredmore » during and following drying. Similar tests will be performed on rods taken from the project assemblies at the end of their long-term storage in a TN-32 dry storage cask (the cask rods ) to identify any significant changes in the fuel rods that may have occurred during the dry storage period. Additionally, some of the sister rods will be used for separate effects testing to expand the applicability of the project data to the fleet, and to address some of the data-related gaps associated with extended storage and subsequent transportation of high burnup fuel. A draft test plan is being developed that describes the experimental work to be conducted on the sister rods. This paper summarizes the draft test plan and necessary coordination activities for the multi-year experimental program to supply data relevant to the assessment of the safety of long-term storage followed by transportation of high burnup spent fuel.« less

SPERT I DESTRUCTIVE TEST PROGRAM SAFETY ANALYSIS REPORT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spano, A.H.; Miller, R.W.

1962-06-15

The water-moderated core used for destructive experiments is mounted in the Spent I open-type reactor vessel, which has no provision for pressurization or forced coolant flow. The core is an array of highly enriched aluminum clad, plate-type fuel assemblies, using four bladetype, gang-operated control rods. Reactor transients are initiated at ambient temperature by step-insentions of reactivity, using a control rod which can be quickly ejected from the core. Following an initial series of static measurements to determine the basic- reactor properties of the test core, a series of nondestructive, self-limiting power excursion tests was performed, which covered a reactor periodmore » range down to the point where minor fuel plate damage first occurred -approximately for a 10- msec period test. These tests provided power, temperature, and pressure data. Additional kinetic teste in the period region between 10 and 5 msec were completed to explore the region of limited core damage. Fuel plate damage results included plate distortion, cladding cracking, and fuel melting. These exploratory tests were valuable in revealing unexpected changes in the dependence of pressure, temperature, burst energy, and burst shape parameters on reactor period, although the dependence of peak power on reactor period was not significantly changed. An evaluation of hazards involved in conducting the 2- msec test, based on pessimistic assumptions regarding fission product release and weather conditions, indicates that with the procedural controls normally exercised in the conduct of any transient test at Spent and the special controls to be in effect during the destructive test series, no significant hazard to personnel or to the general public will be obtained. All nuclear operation is conducted remotely approximately 1/2 mile from the reactor building. Discussion is also given of the supervision and control of personnel during and after each destructive test, and of the plans for re-entry, cleanup, and restoration of the facility. (auth)« less

Bending testing and characterization of surrogate nuclear fuel rods made of Zircaloy-4 cladding and aluminum oxide pellets

NASA Astrophysics Data System (ADS)

Wang, Hong; Wang, Jy-An John

2016-10-01

Behavior of surrogate nuclear fuel rods made of Zircaloy-4 (Zry-4) cladding with alumina pellets under reversed cyclic bending was studied. Tests were performed under load or moment control at 5 Hz. The surrogate rods fractured under moment amplitudes greater than 10.16 Nm with fatigue lives between 2.4 × 103 and 2.2 × 106 cycles. Fatigue response of Zry-4 cladding was characterized by using flexural rigidity. Degradation of flexural rigidity was shown to depend on the moment and the prefatigue condition of specimens. Pellet-to-pellet interface (PPI), pellet-to-cladding interface (PCI), and pellet condition affect surrogate rod failure. Both debonding of PPI/PCI and pellet fracturing contribute to surrogate rod bending fatigue. The effect of sensor spacing on curvature measurement using three-point deflections was studied; the method based on effective gauge length is effective in sensor spacing correction. The database developed and the understanding gained in this study can serve as input to analysis of SNF (spent nuclear fuel) vibration integrity.

The COSIMA experiments and their verification, a data base for the validation of two phase flow computer codes

NASA Astrophysics Data System (ADS)

Class, G.; Meyder, R.; Stratmanns, E.

1985-12-01

The large data base for validation and development of computer codes for two-phase flow, generated at the COSIMA facility, is reviewed. The aim of COSIMA is to simulate the hydraulic, thermal, and mechanical conditions in the subchannel and the cladding of fuel rods in pressurized water reactors during the blowout phase of a loss of coolant accident. In terms of fuel rod behavior, it is found that during blowout under realistic conditions only small strains are reached. For cladding rupture extremely high rod internal pressures are necessary. The behavior of fuel rod simulators and the effect of thermocouples attached to the cladding outer surface are clarified. Calculations performed with the codes RELAP and DRUFAN show satisfactory agreement with experiments. This can be improved by updating the phase separation models in the codes.

Initial experimental evaluation of crud-resistant materials for light water reactors

NASA Astrophysics Data System (ADS)

Dumnernchanvanit, I.; Zhang, N. Q.; Robertson, S.; Delmore, A.; Carlson, M. B.; Hussey, D.; Short, M. P.

2018-01-01

The buildup of fouling deposits on nuclear fuel rods, known as crud, continues to challenge the worldwide fleet of light water reactors (LWRs). Crud causes serious operational problems for LWRs, including axial power shifts, accelerated fuel clad corrosion, increased primary circuit radiation dose rates, and in some instances has led directly to fuel failure. Numerous studies continue to attempt to model and predict the effects of crud, but each assumes that it will always be present. In this study, we report on the development of crud-resistant materials as fuel cladding coatings, to reduce or eliminate these problems altogether. Integrated loop testing experiments at flowing LWR conditions show significantly reduced crud adhesion and surface crud coverage, respectively, for TiC and ZrN coatings compared to ZrO2. The loop testing results roughly agree with the London dispersion component of van der Waals force predictions, suggesting that they contribute most significantly to the adhesion of crud to fuel cladding in out-of-pile conditions. These results motivate a new look at ways of reducing crud, thus avoiding many expensive LWR operational issues.

Texture evolution in Oxide Dispersion Strengthened (ODS) steel tubes during pilgering process

NASA Astrophysics Data System (ADS)

Vakhitova, E.; Sornin, D.; Barcelo, F.; François, M.

2017-10-01

Oxide Dispersion Strengthened (ODS) steels are foreseen as fuel cladding material in the coming generation of Sodium Fast Reactors (SFR). Cladding tubes are manufactured by hot extrusion and subsequent cold forming steps. In this study, a 9 wt% Cr ODS steel exhibiting α-γ phase transformation at high temperature is cold formed under industrial conditions with a large section reduction in two pilgering steps. The influence of pilgering process parameters and intermediate heat treatment on the microstructure evolution is studied experimentally using Electron Backscattering Diffraction (EBSD) and X-ray Diffraction (XRD) methods. Pilgered samples show elongated grains and a high texture formation with a preferential orientation along the rolling direction. During the heat treatment, grain morphology is recovered from elongated grains to almost equiaxed ones, while the well-known α-fiber texture presents an unexpected increase in intensity. The remarkable temperature stability of this fiber is attributed to a crystallographic structure memory effect during phase transformations.

(Project 14-6770) An Investigation to Establish Multiphysical Property Dataset of Nuclear Materials Based on in-situ Observations and Measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tomar, Vikas; Haque, Aman; Hattar, Khalid

In-core nuclear materials including fuel pins and cladding materials fail due to issues including corrosion, mechanical wear, and pellet cladding interaction. In most such scenario microstructure dependent and corrosioninduced chemistry dependent property changes significantly affect performance of cladding, pellet, and housing. Emphasis of this work was on replace conventional pellet-cladding material models with a new straingradient viscoplasticity model that is informed by transmission electron microscopy (TEM) based measurements and by nanomechanical Raman spectroscopy (NMRS) based measurements. The TEM measurements are quantitative in nature and therefore reveal stress-strain relations with simultaneous insights into mechanisms of deformation at nanoscale. The NMRS measurementsmore » reveal the similar information at mesoscale along with additional information on relating local microstructural stresses with applied stresses. The resulting information is used to fit constants in the strain gradient viscoplasticity model as well as to validate one. During TEM measurements, a micro-electro-mechanical system based setup was developed with mechanical actuation, sensing, heating, and electrical loading. Contrary to post-mortem analysis or qualitative visualization, this setup combines direct visualization of the mechanisms behind deformation with measurement of stress, strain, thermal and electrical properties. The unique research philosophy of visualizing the microstructure at high resolution while measuring the properties led to fundamental understanding in grain size and temperature effects on measured mechanical properties such as fracture toughness. A key contribution is the role of mechanical loading boundary conditions to deconvolute the insitu TEM based nanoscale and NMRS based mesoscale data to bulk behavior. First the literature based pellet cladding mechanical interaction model based on the work of Retel’s and Williamson’s in literature work to predict tempurature and stress distribution in cladding and pellet at normal operating condition was analyzed. Later the data was fitted to find constants for a viscoplastic strain gradient model. The developed model still needs to be refined and calibrated using various experimental results. That remains the focus of future work. Overall, a major thrust of the work was therefore on active control of the microstructure (grain size, defect density and types) exploiting the multi-physics coupling in materials. In particular, using experiments the synergy of current density, mechanical stress and temperature were studied to annihilate defects and recrystallize grains. The developed model is being examined for implementation in BISON. Multiple invited talks, international journal publications, and conference publications were performed by students supported on this work. Another output is support multiple PhD and masters thesis students who will be an important asset for future basic nuclear research. Future Work Recommendations: A nuclear reactor operates under significant variations of thermal loads due to energy cycling and mechanical loads due to constraint effects. Significant thermal and chemical diffusion takes place at the pallet-cladding level. While the proposed work established new experimental approach and new dataset for Zircaloy-4, the irradiation level was in the range of 1-2 dpa. Samples with higher dpa need to be examined. Therefore, a continual of support of the performed work is essential. Currently, these are the only experiments that can measure the produced data. The work also needs to be extended to different fuel types and cladding types such as SiC and FeCrAl based claddings. A combination of datasets for these materials can then be used to analyze accurately predict behavior of critical pellet cladding systems in accident scenario with high heat flux and high thermal loads. This is a BIG unknown as if now.« less

Deterministic Local Sensitivity Analysis of Augmented Systems - II: Applications to the QUENCH-04 Experiment Using the RELAP5/MOD3.2 Code System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ionescu-Bujor, Mihaela; Jin Xuezhou; Cacuci, Dan G.

2005-09-15

The adjoint sensitivity analysis procedure for augmented systems for application to the RELAP5/MOD3.2 code system is illustrated. Specifically, the adjoint sensitivity model corresponding to the heat structure models in RELAP5/MOD3.2 is derived and subsequently augmented to the two-fluid adjoint sensitivity model (ASM-REL/TF). The end product, called ASM-REL/TFH, comprises the complete adjoint sensitivity model for the coupled fluid dynamics/heat structure packages of the large-scale simulation code RELAP5/MOD3.2. The ASM-REL/TFH model is validated by computing sensitivities to the initial conditions for various time-dependent temperatures in the test bundle of the Quench-04 reactor safety experiment. This experiment simulates the reflooding with water ofmore » uncovered, degraded fuel rods, clad with material (Zircaloy-4) that has the same composition and size as that used in typical pressurized water reactors. The most important response for the Quench-04 experiment is the time evolution of the cladding temperature of heated fuel rods. The ASM-REL/TFH model is subsequently used to perform an illustrative sensitivity analysis of this and other time-dependent temperatures within the bundle. The results computed by using the augmented adjoint sensitivity system, ASM-REL/TFH, highlight the reliability, efficiency, and usefulness of the adjoint sensitivity analysis procedure for computing time-dependent sensitivities.« less

PRELIMINARY EVALUATION OF FeCrAl CLADDING AND U-Si FUEL FOR ACCIDENT TOLERANT FUEL CONCEPTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hales, J. D.; Gamble, K. A.

2015-09-01

Since the accident at the Fukushima Daiichi Nuclear Power Station, enhancing the accident tolerance of light water reactors (LWRs) has become an important research topic. In particular, the community is actively developing enhanced fuels and cladding for LWRs to improve safety in the event of accidents in the reactor or spent fuel pools. Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system, can tolerate loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normalmore » operations and operational transients. This paper presents early work in developing thermal and mechanical models for two materials that may have promise: U-Si for fuel, and FeCrAl for cladding. These materials would not necessarily be used together in the same fuel system, but individually have promising characteristics. BISON, the finite element-based fuel performance code in development at Idaho National Laboratory, was used to compare results from normal operation conditions with Zr-4/UO2 behavior. In addition, sensitivity studies are presented for evaluating the relative importance of material parameters such as ductility and thermal conductivity in FeCrAl and U-Si in order to provide guidance on future experiments for these materials.« less

Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

NASA Astrophysics Data System (ADS)

Govers, K.; Verwerft, M.

2016-09-01

The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive.

Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robb, Kevin R

2015-01-01

Iron-chromium-aluminum (FeCrAl) alloys are being considered for fuel concepts with enhanced accident tolerance. FeCrAl alloys have very slow oxidation kinetics and good strength at high temperatures. FeCrAl could be used for fuel cladding in light water reactors and/or as channel box material in boiling water reactors (BWRs). To estimate the potential safety gains afforded by the FeCrAl concept, the MELCOR code was used to analyze a range of postulated station blackout severe accident scenarios in a BWR/4 reactor employing FeCrAl. The simulations utilize the most recently known thermophysical properties and oxidation kinetics for FeCrAl. Overall, when compared to the traditionalmore » Zircaloy-based cladding and channel box, the FeCrAl concept provides a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. Finally, due to the slower oxidation kinetics, substantially less hydrogen is generated, and the generation is delayed in time. This decreases the amount of non-condensable gases in containment and the potential for deflagrations to inhibit the accident response.« less

Stress corrosion cracking of Zircaloys in unirradiated and irradiated CsI

NASA Astrophysics Data System (ADS)

Cox, B.; Surette, B. A.; Wood, J. C.

1986-03-01

Unirradiated split-ring specimens of Zircaloy fuel cladding, coated with CsI, cracked when stressed at elevated temperatures. The specimens have been reexamined fractographically and metallographically in order to confirm that the cause of cracking was stress corrosion (SCC) and not delayed hydride cracking (DHC). Further specimens have been cracked at 350°C by a solution of CsI in a fused mixture of nitrates of rubidium, cesium, strontium and barium, by a similar mechanism. CsI dissolved in a fused molybdate melt was not stable at 400°C, and rapidly evolved iodine, leaving a melt that was incapable of causing SCC. Irradiation of stressed split-ring specimens of Zircaloy fuel cladding in a γ-irradiator of 10 6 R/h and in the U-5 loop in the NRU reactor at an estimated 10 9 R/h caused SCC when the specimens were packed in dry CsI powder. Care had to be taken to dry the CsI, otherwise cracking occurred by a DHC mechanism from hydrogen absorbed from residual moisture in the CsI. Fractography showed that the crack surfaces obtained with dry CsI were typical of iodine-induced SCC rather than cesium-induced metal vapour embrittlement. Thus, if a transport process is provided for the iodide to obtain access to the zirconium surface, CsI is capable of causing SCC of Zircaloy. This transport process might be ionic diffusion in a fission product oxide melt in the fuel-clad gap, however, radiolysis of CsI to form a volatile iodine species in a radiation field is the more probable explanation of PCI failures.

Demonstration of optimum fuel-to-moderator ratio in a PWR unit fuel cell

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Pozsgai, C.

1992-01-01

Nuclear engineering students at The Pennsylvania State University develop scaled-down [[approx]350 MW(thermal)] pressurized water reactors (PWRs) using actual plants as references. The design criteria include maintaining the clad temperature below 2200[degree]F, fuel temperature below melting point, sufficient departure from nucleate boiling ratio (DNBR) margin, a beginning-of-life boron concentration that yields a negative moderator temperature coefficient, an adequate cycle power production (330 effective full-power days), and a batch loading scheme that is economical. The design project allows for many degrees of freedom (e.g., assembly number, pitch and height and batch enrichments) so that each student's result is unique. The iterative naturemore » of the design process is stressed in the course. The LEOPARD code is used for the unit cell depletion, critical boron, and equilibrium xenon calculations. Radial two-group diffusion equations are solved with the TWIDDLE-DEE code. The steady-state ZEBRA thermal-hydraulics program is used for calculating DNBR. The unit fuel cell pin radius and pitch (fuel-to-moerator ratio) for the scaled-down design, however, was set equal to the already optimized ratio for the reference PWR. This paper describes an honors project that shows how the optimum fuel-to-moderator ratio is found for a unit fuel cell shown in terms of neutron economics. This exercise illustrates the impact of fuel-to-moderator variations on fuel utilization factor and the effect of assuming space and energy separability.« less

SOLID SOLUTION CARBIDES ARE THE KEY FUELS FOR FUTURE NUCLEAR THERMAL PROPULSION

NASA Technical Reports Server (NTRS)

Panda, Binayak; Hickman, Robert R.; Shah, Sandeep

2005-01-01

Nuclear thermal propulsion uses nuclear energy to directly heat a propellant (such as liquid hydrogen) to generate thrust for space transportation. In the 1960 s, the early Rover/Nuclear Engine for Rocket Propulsion Application (NERVA) program showed very encouraging test results for space nuclear propulsion but, in recent years, fuel research has been dismal. With NASA s renewed interest in long-term space exploration, fuel researchers are now revisiting the RoverMERVA findings, which indicated several problems with such fuels (such as erosion, chemical reaction of the fuel with propellant, fuel cracking, and cladding issues) that must be addressed. It is also well known that the higher the temperature reached by a propellant, the larger the thrust generated from the same weight of propellant. Better use of fuel and propellant requires development of fuels capable of reaching very high temperatures. Carbides have the highest melting points of any known material. Efforts are underway to develop carbide mixtures and solid solutions that contain uranium carbide, in order to achieve very high fuel temperatures. Binary solid solution carbides (U, Zr)C have proven to be very effective in this regard. Ternary carbides such as (U, Zr, X) carbides (where X represents Nb, Ta, W, and Hf) also hold great promise as fuel material, since the carbide mixtures in solid solution generate a very hard and tough compact material. This paper highlights past experience with early fuel materials and bi-carbides, technical problems associated with consolidation of the ingredients, and current techniques being developed to consolidate ternary carbides as fuel materials.

Phase decomposition of γ-U (bcc) in U-10 wt% Mo fuel alloy during hot isostatic pressing of monolithic fuel plate

NASA Astrophysics Data System (ADS)

Park, Y.; Eriksson, N.; Newell, R.; Keiser, D. D.; Sohn, Y. H.

2016-11-01

Eutectoid decomposition of γ-phase (cI2) into α-phase (oC4) and γ‧-phase (tI6) during the hot isostatic pressing (HIP) of the U-10 wt% Mo (U10Mo) alloy was investigated using monolithic fuel plate samples consisting of U10Mo fuel alloy, Zr diffusion barrier and AA6061 cladding. The decomposition of the γ-phase was observed because the HIP process is carried out near the eutectoid temperature, 555 °C. Initially, a cellular structure, consisting of γ‧-phase surrounded by α-phase, developed from the destabilization of the γ-phase. The cellular structure further developed into an alternating lamellar structure of α- and γ‧-phases. Using scanning electron microscopy and transmission electron microscopy, qualitative and quantitative microstructural analyses were carried out to identify the phase constituents, and elucidate the microstructural development based on time-temperature-transformation diagram of the U10Mo alloy. The destabilization of γ -phase into α- and γ‧-phases would be minimized when HIP process was carried out with rapid ramping/cooling rate and dwell temperature higher than 560 °C.

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Accident Tolerant Fuels High Impact Problem: Coordinate Multiscale U 3Si 2 Modeling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gamble, K. A.; Hales, J. D.; Miao, Y.

Since the events at the Fukushima-Daiichi nuclear power plant in March 2011 significant research has unfolded at national laboratories, universities and other institutions into alternative materials that have potential enhanced accident tolerance when compared to traditional \\uo~fuel zircaloy clad fuel rods. One of the potential replacement fuels is uranium silicide (\\usi) for its higher thermal conductivity and uranium density. The lower melting temperature is of potential concern during postulated accident conditions. Another disadvantage for \\usi~ is the lack of experimental data under power reactor conditions. Due to the aggressive development schedule for inserting some of the potential materials into leadmore » test assemblies or rods by 2022~\\cite{bragg-sitton_2014} multiscale multiphysics modeling approaches have been used to provide insight into these materials. \\\\ \

Effect of indium addition in U-Zr metallic fuel on lanthanide migration

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Yeon Soo; Wiencek, T.; O'Hare, E.

Advanced fast reactor concepts to achieve ultra-high burnup (~50%) require prevention of fuel-cladding chemical interaction (FCCI). Fission product lanthanide accumulation at high burnup is substantial and significantly contributes to FCCI upon migration to the cladding interface. Diffusion barriers are typically used to prevent interaction of the lanthanides with the cladding. A more active method has been proposed which immobilizes the lanthanides through formation of stable compounds with an additive. Theoretical analysis showed that indium, thallium, and antimony are good candidates. Indium was the strongest candidate because of its low reactivity with iron-based cladding alloys. Characterization of the as-fabricated alloys wasmore » performed to determine the effectiveness of the indium addition in forming compounds with lanthanides, represented by cerium. Tests to examine how effectively the dopant prevents lanthanide migration under a thermal gradient were also performed. The results showed that indium effectively prevented cerium migration.« less

Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident conditions

NASA Astrophysics Data System (ADS)

Ott, L. J.; Robb, K. R.; Wang, D.

2014-05-01

Following the severe accidents at the Japanese Fukushima Daiichi Nuclear Power Station in 2011, the US Department of Energy initiated research and development on the enhancement of the accident tolerance of light water reactors by the development of fuels/cladding that, in comparison with the standard UO2/Zircaloy (Zr) system, can tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations. Analyses are presented that illustrate the impact of these new candidate fuel/cladding materials on the fuel performance at normal operating conditions and on the reactor system under DB and BDB accident conditions.

Rate Theory Modeling and Simulations of Silicide Fuel at LWR Conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miao, Yinbin; Ye, Bei; Mei, Zhigang

Uranium silicide (U 3Si 2) fuel has higher thermal conductivity and higher uranium density, making it a promising candidate for the accident-tolerant fuel (ATF) used in light water reactors (LWRs). However, previous studies on the fuel performance of U 3Si 2, including both experimental and computational approaches, have been focusing on the irradiation conditions in research reactors, which usually involve low operation temperatures and high fuel burnups. Thus, it is important to examine the fuel performance of U 3Si 2 at typical LWR conditions so as to evaluate the feasibility of replacing conventional uranium dioxide fuel with this silicide fuelmore » material. As in-reactor irradiation experiments involve significant time and financial cost, it is appropriate to utilize modeling tools to estimate the behavior of U 3Si 2 in LWRs based on all those available research reactor experimental references and state-of-the-art density functional theory (DFT) calculation capabilities at the early development stage. Hence, in this report, a comprehensive investigation of the fission gas swelling behavior of U 3Si 2 at LWR conditions is introduced. The modeling efforts mentioned in this report was based on the rate theory (RT) model of fission gas bubble evolution that has been successfully applied for a variety of fuel materials at devious reactor conditions. Both existing experimental data and DFT-calculated results were used for the optimization of the parameters adopted by the RT model. Meanwhile, the fuel-cladding interaction was captured by the coupling of the RT model with simplified mechanical correlations. Therefore, the swelling behavior of U 3Si 2 fuel and its consequent interaction with cladding in LWRs was predicted by the rate theory modeling, providing valuable information for the development of U 3Si 2 fuel as an accident-tolerant alternative for uranium dioxide.« less

A Jacobian-free Newton Krylov method for mortar-discretized thermomechanical contact problems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hansen, Glen, E-mail: Glen.Hansen@inl.gov

2011-07-20

Multibody contact problems are common within the field of multiphysics simulation. Applications involving thermomechanical contact scenarios are also quite prevalent. Such problems can be challenging to solve due to the likelihood of thermal expansion affecting contact geometry which, in turn, can change the thermal behavior of the components being analyzed. This paper explores a simple model of a light water reactor nuclear fuel rod, which consists of cylindrical pellets of uranium dioxide (UO{sub 2}) fuel sealed within a Zircalloy cladding tube. The tube is initially filled with helium gas, which fills the gap between the pellets and cladding tube. Themore » accurate modeling of heat transfer across the gap between fuel pellets and the protective cladding is essential to understanding fuel performance, including cladding stress and behavior under irradiated conditions, which are factors that affect the lifetime of the fuel. The thermomechanical contact approach developed here is based on the mortar finite element method, where Lagrange multipliers are used to enforce weak continuity constraints at participating interfaces. In this formulation, the heat equation couples to linear mechanics through a thermal expansion term. Lagrange multipliers are used to formulate the continuity constraints for both heat flux and interface traction at contact interfaces. The resulting system of nonlinear algebraic equations are cast in residual form for solution of the transient problem. A Jacobian-free Newton Krylov method is used to provide for fully-coupled solution of the coupled thermal contact and heat equations.« less

A Jacobian-Free Newton Krylov Method for Mortar-Discretized Thermomechanical Contact Problems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Glen Hansen

2011-07-01

Multibody contact problems are common within the field of multiphysics simulation. Applications involving thermomechanical contact scenarios are also quite prevalent. Such problems can be challenging to solve due to the likelihood of thermal expansion affecting contact geometry which, in turn, can change the thermal behavior of the components being analyzed. This paper explores a simple model of a light water reactor nuclear reactor fuel rod, which consists of cylindrical pellets of uranium dioxide (UO2) fuel sealed within a Zircalloy cladding tube. The tube is initially filled with helium gas, which fills the gap between the pellets and cladding tube. Themore » accurate modeling of heat transfer across the gap between fuel pellets and the protective cladding is essential to understanding fuel performance, including cladding stress and behavior under irradiated conditions, which are factors that affect the lifetime of the fuel. The thermomechanical contact approach developed here is based on the mortar finite element method, where Lagrange multipliers are used to enforce weak continuity constraints at participating interfaces. In this formulation, the heat equation couples to linear mechanics through a thermal expansion term. Lagrange multipliers are used to formulate the continuity constraints for both heat flux and interface traction at contact interfaces. The resulting system of nonlinear algebraic equations are cast in residual form for solution of the transient problem. A Jacobian-free Newton Krylov method is used to provide for fully-coupled solution of the coupled thermal contact and heat equations.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

James A. Smith; Jeffrey M. Lacy; Barry H. Rabin

12. Other advances in QNDE and related topics: Preferred Session Laser-ultrasonics Developing A Laser Shockwave Model For Characterizing Diffusion Bonded Interfaces 41st Annual Review of Progress in Quantitative Nondestructive Evaluation Conference QNDE Conference July 20-25, 2014 Boise Centre 850 West Front Street Boise, Idaho 83702 James A. Smith, Jeffrey M. Lacy, Barry H. Rabin, Idaho National Laboratory, Idaho Falls, ID ABSTRACT: The US National Nuclear Security Agency has a Global Threat Reduction Initiative (GTRI) which is assigned with reducing the worldwide use of high-enriched uranium (HEU). A salient component of that initiative is the conversion of research reactors from HEUmore » to low enriched uranium (LEU) fuels. An innovative fuel is being developed to replace HEU. The new LEU fuel is based on a monolithic fuel made from a U-Mo alloy foil encapsulated in Al-6061 cladding. In order to complete the fuel qualification process, the laser shock technique is being developed to characterize the clad-clad and fuel-clad interface strengths in fresh and irradiated fuel plates. The Laser Shockwave Technique (LST) is being investigated to characterize interface strength in fuel plates. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves to characterize interfaces in nuclear fuel plates. However the deposition of laser energy into the containment layer on specimen’s surface is intractably complex. The shock wave energy is inferred from the velocity on the backside and the depth of the impression left on the surface from the high pressure plasma pulse created by the shock laser. To help quantify the stresses and strengths at the interface, a finite element model is being developed and validated by comparing numerical and experimental results for back face velocities and front face depressions with experimental results. This paper will report on initial efforts to develop a finite element model for laser shock.« less

Core thermal response and hydrogen generation of the N Reactor hydrogen mitigation design basis accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

White, M.D.; Lombardo, N.J.; Heard, F.J.

1988-04-01

Calculations were performed to determine core heatup, core damage, and subsequent hydrogen production of a hypothetical loss-of-cooling accident at the Department of Energy's N Reactor. The thermal transient response of the reactor core was solved using the TRUMP-BD computer program. Estimates of whole-core thermal damage and hydrogen production were made by weighting the results of multiple half-length pressure tube simulations at various power levels. The Baker-Just and Wilson parabolic rate equations for the metal-water chemical reactions modeled the key phenomena of chemical energy and hydrogen evolution. Unlimited steam was assumed available for continuous oxidation of exposed Zircaloy-2 surfaces and formore » uranium metal with fuel cladding beyond the failure temperature (1038 C). Intact fuel geometry was modeled. Maximum fuel temperatures (1181 C) in the cooled central regions of the core were predicted to occur one-half hour into the accident scenario. Maximum fuel temperatures of 1447 C occurred in the core GSCS-regions at the end of the 10-h transient. After 10-h 26% of the fuel inventory was predicted to have failed. Peak hydrogen evolution equaled 42 g/s, while 10-h integrated hydrogen evolution equaled 167 kg. 12 refs., 12 figs., 2 tabs.« less

Microstructure and properties of laser-clad high-temperature wear-resistant alloys

NASA Astrophysics Data System (ADS)

Yang, Yongqiang

1999-02-01

A 2-kW CO 2 laser with a powder feeder was used to produce alloy coatings with high temperature-wear resistance on the surface of steel substrates. To analyze the microstructure and microchemical composition of the laser-clad layers, a scanning electron microscope (SEM) equipped with an energy dispersive X-ray microanalysis system was employed. X-ray diffraction techniques were applied to characterize the phases formed during the cladding process. The results show that the microstructure of the cladding alloy consists mainly of many dispersed particles (W 2C, (W,Ti)C 1- x, WC), a lamellar eutectic carbide M 12C, and an (f.c.c) matrix. Hardness tested at room and high temperature showed that the laser-clad zone has a moderate room temperature hardness and relatively higher elevated temperature hardness. The application of the laser-clad layer to a hot tool was very successful, and its operational life span was prolonged 1 to 4 times.

Effects of heat treatment on U-Mo fuel foils with a zirconium diffusion barrier

NASA Astrophysics Data System (ADS)

Jue, Jan-Fong; Trowbridge, Tammy L.; Breckenridge, Cynthia R.; Moore, Glenn A.; Meyer, Mitchell K.; Keiser, Dennis D.

2015-05-01

A monolith fuel design based on U-Mo alloy has been selected as the fuel type for conversion of the United States' high performance research reactors (HPRRs) from highly enriched uranium (HEU) to low-enriched uranium (LEU). In this fuel design, a thin layer of zirconium is used to eliminate the direct interaction between the U-Mo fuel meat and the aluminum-alloy cladding during irradiation. The co-rolling process used to bond the Zr barrier layer to the U-Mo foil during fabrication alters the microstructure of both the U-10Mo fuel meat and the U-Mo/Zr interface. This work studied the effects of post-rolling annealing treatment on the microstructure of the co-rolled U-Mo fuel meat and the U-Mo/Zr interaction layer. Microscopic characterization shows that the grain size of U-Mo fuel meat increases with the annealing temperature, as expected. The grain sizes were ∼9, ∼13, and ∼20 μm for annealing temperature of 650, 750, and 850 °C, respectively. No abnormal grain growth was observed. The U-Mo/Zr interaction-layer thickness increased with the annealing temperature with an Arrhenius constant for growth of 184 kJ/mole, consistent with a previous diffusion-couple study. The interaction layer thickness was 3.2 ± 0.5 μm, 11.1 ± 2.1 μm, 27.1 ± 0.9 μm for annealing temperature of 650, 750, to 850 °C, respectively. The homogeneity of Mo improves with post rolling annealing temperature and with U-Mo coupon homogenization. The phases in the Zr/U-Mo interaction layer produced by co-rolling, however, differ from those reported in the previous diffusion couple studies.

Effects of heat treatment on U–Mo fuel foils with a zirconium diffusion barrier

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jue, Jan-Fong; Trowbridge, Tammy L.; Breckenridge, Cynthia R.

A monolith fuel design based on U–Mo alloy has been selected as the fuel type for conversion of the United States’ high performance research reactors (HPRRs) from highly enriched uranium (HEU) to low-enriched uranium (LEU). In this fuel design, a thin layer of zirconium is used to eliminate the direct interaction between the U–Mo fuel meat and the aluminum-alloy cladding during irradiation. The co-rolling process used to bond the Zr barrier layer to the U–Mo foil during fabrication alters the microstructure of both the U–10Mo fuel meat and the U–Mo/Zr interface. This work studied the effects of post-rolling annealing treatmentmore » on the microstructure of the co-rolled U–Mo fuel meat and the U–Mo/Zr interaction layer. Microscopic characterization shows that the grain size of U–Mo fuel meat increases with the annealing temperature, as expected. The grain sizes were ~9, ~13, and ~20 μm for annealing temperature of 650, 750, and 850 °C, respectively. No abnormal grain growth was observed. The U–Mo/Zr interaction-layer thickness increased with the annealing temperature with an Arrhenius constant for growth of 184 kJ/mole, consistent with a previous diffusion-couple study. The interaction layer thickness was 3.2 ± 0.5 μm, 11.1 ± 2.1 μm, 27.1 ± 0.9 μm for annealing temperature of 650, 750, to 850 °C, respectively. The homogeneity of Mo improves with post rolling annealing temperature and with U–Mo coupon homogenization. The phases in the Zr/U–Mo interaction layer produced by co-rolling, however, differ from those reported in the previous diffusion couple studies.« less

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

NASA Astrophysics Data System (ADS)

Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

2008-12-01

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

General corrosion properties of modified PNC1520 austenitic stainless steel in supercritical water as a fuel cladding candidate material for supercritical water reactor

NASA Astrophysics Data System (ADS)

Nakazono, Y.; Iwai, T.; Abe, H.

2010-03-01

The Super-Critical Water-cooled Reactor (SCWR) has been designed and investigated because of its high thermal efficiency and plant simplification. There are some advantages including the use of a single phase coolant with high enthalpy but there are numerous potential problems, particularly with materials. As the operating temperature of supercritical water reactor will be between 280°C and 620°C with a pressure of 25MPa, the selection of materials is difficult and important. Austenitic stainless steels were selected for possible use in supercritical water systems because of their corrosion resistance and radiation resistance. The PNC1520 austenitic stainless steel developed by Japan Atomic Energy Agency (JAEA) as a nuclear fuel cladding material for a Na-cooled fast breeder reactor. The corrosion data of PNC1520 in supercritical water (SCW) is required but does not exist. The purpose of the present study is to research the corrosion properties for PNC1520 austenitic stainless steel in supercritical water. The supercritical water corrosion test was performed for the standard PNC1520 (1520S) and the Ti-additional type of PNC1520 (1520Ti) by using a supercritical water autoclave. Corrosion tests on the austenitic 1520S and 1520Ti steels in supercritical water were performed at 400, 500 and 600°C with exposures up to 1000h. The amount of weight gain, weight loss and weight of scale were evaluated after the corrosion test in supercritical water for both austenitic steels. After 1000h corrosion test performed, the weight gains of both austenitic stainless steels were less than 2 g/m2 at 400°C and 500°C . But both weight gain and weight loss of 1520Ti were larger than those of 1520S at 600°C . By increasing the temperature to 600°C, the surface of 1520Ti was covered with magnetite formed in supercritical water and dissolution of the steel alloying elements has been observed. In view of corrosion, 1520S may have larger possibility than 1520Ti to adopt a supercritical water reactor core fuel cladding.

High temperature sensor properties of a specialty double cladding fiber

NASA Astrophysics Data System (ADS)

Zhou, Ting; Pang, Fufei; Wang, Tingyun

2011-12-01

A simple high temperature fiber sensor is proposed and demonstrated. The sensor head is made of a short section of specialty double cladding fiber (DCF). The DCF consists of a depressed inner cladding which is boron (B)-doped silica. Through an evanescent wave, the cladding mode can be excited, and thus the transmission presents a resonant spectral dip. The high temperature sensing properties was studied according to the shift of the transmission spectrum shifts. With increasing the temperature from 28 °C to 850 °C, the resonant spectrum shifts to longer wavelengths. The sensitivity is 0.112 nm / °C.

Cyclic furnace oxidation of clad WI-52 systems at 1040 C and 1090 C

NASA Technical Reports Server (NTRS)

Gedwill, M. A.

1972-01-01

Cyclic furnace oxidation studies were conducted on the cobalt alloy WI-52 clad with Ni-30Cr, Fe-25Cr-4A1, and Ni-20Cr-4A1 foils (0.051 to 0.254 mm thick). Tests as long as 400 hours using 1- and 20-hour cycles showed that the Ni-Cr- and Fe-Cr-A1 claddings were about equally protective at both temperatures. The protective ability of these alloys was influenced by exposure temperature and cladding thickness. At both temperatures, they protected WI-52 about as well as, or better than, a widely used commercial aluminide coating. The Ni-Cr-Al claddings did not protect WI-52 nearly as well. Interdiffusion generally influenced the oxidation behavior of all clad WI-52 systems.

The honey insertion cladding to improve the sensitivity of temperature polymer optical fiber sensor

NASA Astrophysics Data System (ADS)

Arwani, M.; Kuswanto, H.

2018-04-01

The sensitivity of temperature polymer optical fiber (POF) sensor has been studied. Part of cladding (9 cm) was substituted with honey. Polymer cladding was stripped mechanically and the honey inserted into the tube. Plastic gel closed the two end sides of the tubes. The optical power output was detected by Optical Power Meter (OPM). Honey cladding and temperature changing effect to the internal reflection and optical fiber output intensity. Highest output intensity changing at 20°C was shown by optical fiber coated by longan honey as cladding. The range of 10-50° C, as the rise of surroundings temperature, the attenuation was getting smaller. Best sensitivity was fiber with sensing part coated by Longan honey. Best linearity was sensing fiber with sensing part coated by Pracimantoro honey.

Funnel for fuel pin loading system

DOEpatents

Christiansen, D.W.; Steffen, J.M.; Brown, W.F.

1984-01-01

An enlarged funnel is described which is releasably mounted at the open end of a length of cladding by an encircling length of shrink tubing which securely engages outer surfaces of both the funnel and cladding. The shrink tubing overlaps an annular shoulder against which pulling force can be exerted to remove the tubing from the cladding. The shoulder can be provided on a separate collar or ring, or on the funnel itself.

Funnel for fuel pin loading system

DOEpatents

Christiansen, David W.; Steffen, Jim M.; Brown, William F.

1985-01-01

An enlarged funnel is releasably mounted at the open end of a length of cladding by an encircling length of shrink tubing which securely engages outer surfaces of both the funnel and cladding. The shrink tubing overlaps an annular shoulder against which pulling force can be exerted to remove the tubing from the cladding. The shoulder can be provided on a separate collar or ring, or on the funnel itself.

Capture of Tritium Released from Cladding in the Zirconium Recycle Process

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spencer, Barry B.; Walker, T. B.; Bruffey, S. H.

2016-08-31

Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when themore » solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using nonradioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.« less

Capture of Tritium Released from Cladding in the Zirconium Recycle Process

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spencer, Barry B.; Walker, T. B.; Bruffey, Stephanie H.

2016-08-31

This report is issued as the first revision to FCRD-MRWFD-2016-000297. Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-basedmore » cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using non-radioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.« less

Ultra-high temperature tensile properties of ODS steel claddings under severe accident conditions

NASA Astrophysics Data System (ADS)

Yano, Y.; Tanno, T.; Oka, H.; Ohtsuka, S.; Inoue, T.; Kato, S.; Furukawa, T.; Uwaba, T.; Kaito, T.; Ukai, S.; Oono, N.; Kimura, A.; Hayashi, S.; Torimaru, T.

2017-04-01

Ultra-high temperature ring tensile tests were performed to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions with temperatures ranging from room temperature to 1400 °C which is close to the melting point of core materials. The experimental results showed that the tensile strength of 9Cr-ODS steel claddings was highest in the core materials at ultra-high temperatures of 900-1200 °C, but there was significant degradation in the tensile strength of 9Cr-ODS steel claddings above 1200 °C. This degradation was attributed to grain boundary sliding deformation with γ/δ transformation, which is associated with reduced ductility. By contrast, the tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 °C, unlike the other tested materials.

Development and Validation of Capabilities to Measure Thermal Properties of Layered Monolithic U-Mo Alloy Plate-Type Fuel

NASA Astrophysics Data System (ADS)

Burkes, Douglas E.; Casella, Andrew M.; Buck, Edgar C.; Casella, Amanda J.; Edwards, Matthew K.; MacFarlan, Paul J.; Pool, Karl N.; Smith, Frances N.; Steen, Franciska H.

2014-07-01

The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium to low enriched uranium. One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the thermal-conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify functionality of equipment installed in hot cells for eventual measurements on irradiated uranium-molybdenum (U-Mo) monolithic fuel specimens, refine procedures to operate the equipment, and validate models to extract the desired thermal properties. The results presented here demonstrate the adequacy of the equipment, procedures, and models that have been developed for this purpose based on measurements conducted on surrogate depleted uranium-molybdenum (DU-Mo) alloy samples containing a Zr diffusion barrier and clad in aluminum alloy 6061 (AA6061). The results are in excellent agreement with thermal property data reported in the literature for similar U-Mo alloys as a function of temperature.

Nanocrystalline SiC and Ti 3SiC 2 Alloys for Reactor Materials: Diffusion of Fission Product Surrogates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Henager, Charles H.; Jiang, Weilin

2014-11-01

MAX phases, such as titanium silicon carbide (Ti 3SiC 2), have a unique combination of both metallic and ceramic properties, which make them attractive for potential nuclear applications. Ti 3SiC 2 has been suggested in the literature as a possible fuel cladding material. Prior to the application, it is necessary to investigate diffusivities of fission products in the ternary compound at elevated temperatures. This study attempts to obtain relevant data and make an initial assessment for Ti 3SiC 2. Ion implantation was used to introduce fission product surrogates (Ag and Cs) and a noble metal (Au) in Ti 3SiC 2,more » SiC, and a dual-phase nanocomposite of Ti 3SiC 2/SiC synthesized at PNNL. Thermal annealing and in-situ Rutherford backscattering spectrometry (RBS) were employed to study the diffusivity of the various implanted species in the materials. In-situ RBS study of Ti 3SiC 2 implanted with Au ions at various temperatures was also performed. The experimental results indicate that the implanted Ag in SiC is immobile up to the highest temperature (1273 K) applied in this study; in contrast, significant out-diffusion of both Ag and Au in MAX phase Ti 3SiC 2 occurs during ion implantation at 873 K. Cs in Ti 3SiC 2 is found to diffuse during post-irradiation annealing at 973 K, and noticeable Cs release from the sample is observed. This study may suggest caution in using Ti 3SiC 2 as a fuel cladding material for advanced nuclear reactors operating at very high temperatures. Further studies of the related materials are recommended.« less

JOYO-1 Irradiation Test Campaign Technical Close-out, For Information

DOE Office of Scientific and Technical Information (OSTI.GOV)

G. Borges

2006-01-31

The JOYO-1 irradiation testing was designed to screen the irradiation performance of candidate cladding, structural and reflector materials in support of space reactor development. The JOYO-1 designation refers to the first of four planned irradiation tests in the JOYO reactor. Limited irradiated material performance data for the candidate materials exists for the expected Prometheus-1 duration, fluences and temperatures. Materials of interest include fuel element cladding and core materials (refractory metal alloys and silicon carbide (Sic)), vessel and plant structural materials (refractory metal alloys and nickel-base superalloys), and control and reflector materials (BeO). Key issues to be evaluated were long termmore » microstructure and material property stability. The JOYO-1 test campaign was initiated to irradiate a matrix of specimens at prototypical temperatures and fluences anticipated for the Prometheus-1 reactor [Reference (1)]. Enclosures 1 through 9 describe the specimen and temperature monitors/dosimetry fabrication efforts, capsule design, disposition of structural material irradiation rigs, and plans for post-irradiation examination. These enclosures provide a detailed overview of Naval Reactors Prime Contractor Team (NRPCT) progress in specific areas; however, efforts were in various states of completion at the termination of NRPCT involvement with and restructuring of Project Prometheus.« less

Development of a full-length external-fuel thermionic converter for in-pile testing.

NASA Technical Reports Server (NTRS)

Schock, A.; Raab, B.

1971-01-01

Description of an external-fuel thermionic converter which utilizes a thoriated-tungsten fuel-emitter body. Performance in out-of-pile tests was comparable to that of an arc-cast tungsten emitter body, with 400-eW output power (about 5 W/sq cm) at 10.8% efficiency. Maximum fuel clad temperature averaged from 1650 to 1700 C during the 300-hour test. This converter has been processed for in-pile testing. The various processing steps, including the installation of six emitter thermocouples, encapsulation in the secondary container, and joining to the fission-gas collection system, are described in detail. In addition to the converter assembly, a doubly contained fission gas collection assembly with radiation-hardened differential pressure transducers was fabricated. The experiment support plate required for the in-pile test, containing electrically insulated instrumentation feedthroughs and coolant line feedthroughs to the vacuum test chamber, was also fabricated.

Fast-spectrum space-power-reactor concepts using boron control devices

NASA Technical Reports Server (NTRS)

Mayo, W.

1973-01-01

Several fast-spectrum space power reactor concepts that use boron carbide control devices were examined to determine the neutronic feasibility of the designs. The designs considered were (1) a 199-fuel-pin, 12-poison-reflector-control-drum reactor; (2) a 232-fuel-pin reactor with 12 reflector drums and three in-core control rods; (3) a 337-fuel-pin design with 12 incore control rods; and a 181-fuel-pin design with six drums closely coupled to the core to increase reactivity per drum. Adequate reactivity control and excess reactivity could be obtained for each concept, and the goals of 50,000 hours at 2.17 thermal megawatts with a lithium-7 coolant outlet temperature of 1222 K could be met without exceeding the 1-percent-clad-creep criterion. Heating rates in the boron carbide were calculated, but a heat transfer analysis was not done.

Fission products and nuclear fuel behaviour under severe accident conditions part 2: Fuel behaviour in the VERDON-1 sample

NASA Astrophysics Data System (ADS)

Geiger, E.; Le Gall, C.; Gallais-During, A.; Pontillon, Y.; Lamontagne, J.; Hanus, E.; Ducros, G.

2017-11-01

Within the framework of the International Source Term Programme (ISTP), the VERDON programme aims at quantifying the source term of radioactive materials in case of a hypothetical severe accident in a light water reactor (LWR). Tests were performed in a new experimental laboratory (VERDON) built in the LECA-STAR facility (CEA Cadarache). The VERDON-1 test was devoted to the study of a high burn-up UO2 fuel and FP releases at very high temperature (≈2873 K) in a reducing atmosphere. Post-test qualitative and quantitative characterisations of the VERDON-1 sample led to the proposal of a scenario explaining the phenomena occurring during the experimental sequence. Hence, the fuel and the cladding may have interacted which led to the melting of UO2-ZrO2 alloy. Although no relocation was observed during the test, it may have been imminent.

Automated closure system for nuclear reactor fuel assemblies

DOEpatents

Christiansen, David W.; Brown, William F.

1985-01-01

A welder for automated closure of fuel pins by a pulsed magnetic process in which the open end of a length of cladding is positioned within a complementary tube surrounded by a pulsed magnetic welder. Seals are provided at each end of the tube, which can be evacuated or can receive tag gas for direct introduction to the cladding interior. Loading of magnetic rings and end caps is accomplished automatically in conjunction with the welding steps carried out within the tube.

Final report of fuel dynamics Test E7

DOE Office of Scientific and Technical Information (OSTI.GOV)

Doerner, R.C.; Murphy, W.F.; Stanford, G.S.

1977-04-01

Test data from an in-pile failure experiment of high-power LMFBR-type fuel pins in a simulated $3/s transient-overpower (TOP) accident are reported and analyzed. Major conclusions are that (1) a series of cladding ruptures during the 100-ms period preceding fuel release injected small bursts of fission gas into the flow stream; (2) gas release influenced subsequent cladding melting and fuel release (there were no measurable FCI's (fuel-coolant interactions), and all fuel motion observed by the hodoscope was very slow); (3) the predominant postfailure fuel motion appears to be radial swelling that left a spongy fuel crust on the holder wall; (4)more » less than 4 to 6 percent of the fuel moved axially out of the original fuel zone, and most of this froze within a 10-cm region above the original top of the fuel zone to form the outlet blockage. An inlet blockage approximately 1 cm long was formed and consisted of large interconnected void regions. Both blockages began just beyond the ends of the fuel pellets.« less

Tensile properties from room temperature to 1315 C of tungsten-lined tantalum-alloy (T-111) tubing fabricated by hot isostatic pressing

NASA Technical Reports Server (NTRS)

Buzzard, R. J.; Metroka, R. R.

1974-01-01

The effects were studied of a thin tungsten liner on the tensile properties of T-111 tubing considered for fuel cladding in a space power nuclear reactor concept. The results indicate that the metallurgically bonded liner had no appreciable effects on the properties of the T-111 tubing. A hot isostatic pressing method used to apply the liners is described along with a means for overcoming the possible embrittling effects of hydrogen contamination.

A novel plant protection strategy for transient reactors

NASA Astrophysics Data System (ADS)

Bhattacharyya, Samit K.; Lipinski, Walter C.; Hanan, Nelson A.

The present plant protection system (PPS) has been defined for use in the TREAT-upgrade (TU) reactor for controlled transient operation of reactor-fuel behavior testing under simulated reactor-accident conditions. A PPS with energy-dependent trip set points lowered worst-case clad temperatures by as much as 180 K, relative to the use of conventional fixed-level trip set points. The multilayered multilevel protection strategy represents the state-of-the-art in terrestrial transient reactor protection systems, and should be applicable to multi-MW space reactors.

An analysis of the sliding pressure start-up of SCWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, F.; Yang, J.; Li, H.

In this paper, the preliminary sliding pressure start-up system and scheme of supercritical water-cooled reactor in CGNPC (CGN-SCWR) were proposed. Thermal-hydraulic behavior in start-up procedures was analyzed in detail by employing advanced reactor subchannel analysis software ATHAS. The maximum cladding temperature (MCT for short) and core power of fuel assembly during the whole start-up process were investigated comparatively. The results show that the recommended start-up scheme meets the design requirements from the perspective of thermal-hydraulic. (authors)

Study of clad ballooning and rupture behaviour of Indian PHWR fuel pins under transient heating condition in steam environment

NASA Astrophysics Data System (ADS)

Sawarn, Tapan K.; Banerjee, Suparna; Sheelvantra, Smita S.; Singh, J. L.; Bhasin, Vivek

2017-11-01

This paper presents the results of the investigation on the deformation and rupture characteristics of Indian pressurized heavy water reactor (IPHWR) fuel pins under simulated loss of coolant accident (LOCA) condition in steam environment. Transient heating experiments were carried out on single fuel pin internally pressurized with argon gas in the range 3-70 bar. Effect of internal pressure on burst temperature, influence of burst temperature on the circumferential strain and rupture opening area were also studied. Two circumferential strain maxima at the burst temperatures of 740 & ∼979 °C and a minimum at the burst temperature of ∼868 °C were observed. It was found that oxidation had considerable effect on the burst behavior. Test data were used to derive a direct empirical correlation for burst stress exclusively as a function of temperature. The ballooning and rupture behaviours in steam and argon environments have been compared. Experimental data were examined against various correlations using Erbacher equation and author's previous correlation in argon. A second burst correlation has also been developed combining the equation in argon from the previous work of the authors and an exponential factor with oxygen content as a parameter assuming the burst stress to be a function of both temperature and oxygen concentration. The burst temperatures predicted by this empirical correlation are in good agreement with the test data.

Chemical vapor deposition of Mo tubes for fuel cladding applications

DOE PAGES

Beaux, Miles F.; Vodnik, Douglas R.; Peterson, Reuben J.; ...

2018-01-31

In this study, chemical vapor deposition (CVD) techniques have been evaluated for fabrication of free-standing 0.25 mm thick molybdenum tubes with the end goal of nuclear fuel cladding applications. In order to produce tubes with the wall thickness and microstructures desirable for this application, long deposition durations on the order of 50 h with slow deposition rates were employed. A standard CVD method, involving molybdenum pentachloride reduction by hydrogen, as well as a fluidized-bed CVD (FBCVD) method was applied towards these objectives. Characterization of the tubes produced in this manner revealed regions of material with fine grain microstructure and wallmore » thickness suitable for fuel cladding applications, but lacking necessary uniformity across the length of the tubes. Finally, a path forward for the production of freestanding molybdenum tubes that possess the desired properties across their entire length has been identified and can be accomplished by future optimization of the deposition system.« less

Chemical vapor deposition of Mo tubes for fuel cladding applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beaux, Miles F.; Vodnik, Douglas R.; Peterson, Reuben J.

In this study, chemical vapor deposition (CVD) techniques have been evaluated for fabrication of free-standing 0.25 mm thick molybdenum tubes with the end goal of nuclear fuel cladding applications. In order to produce tubes with the wall thickness and microstructures desirable for this application, long deposition durations on the order of 50 h with slow deposition rates were employed. A standard CVD method, involving molybdenum pentachloride reduction by hydrogen, as well as a fluidized-bed CVD (FBCVD) method was applied towards these objectives. Characterization of the tubes produced in this manner revealed regions of material with fine grain microstructure and wallmore » thickness suitable for fuel cladding applications, but lacking necessary uniformity across the length of the tubes. Finally, a path forward for the production of freestanding molybdenum tubes that possess the desired properties across their entire length has been identified and can be accomplished by future optimization of the deposition system.« less

Pulsed Magnetic Welding for Advanced Core and Cladding Steel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cao, Guoping; Yang, Yong

2013-12-19

To investigate a solid-state joining method, pulsed magnetic welding (PMW), for welding the advanced core and cladding steels to be used in Generation IV systems, with a specific application for fuel pin end-plug welding. As another alternative solid state welding technique, pulsed magnetic welding (PMW) has not been extensively explored on the advanced steels. The resultant weld can be free from microstructure defects (pores, non-metallic inclusions, segregation of alloying elements). More specifically, the following objectives are to be achieved: 1. To design a suitable welding apparatus fixture, and optimize welding parameters for repeatable and acceptable joining of the fuel pinmore » end-plug. The welding will be evaluated using tensile tests for lap joint weldments and helium leak tests for the fuel pin end-plug; 2 Investigate the microstructural and mechanical properties changes in PMW weldments of proposed advanced core and cladding alloys; 3. Simulate the irradiation effects on the PWM weldments using ion irradiation.« less

Effects of the foil flatness on irradiation performance of U10Mo monolithic mini-plates

DOE PAGES

Ozaltun, Hakan; Medvedev, Pavel G.; Rabin, Barry H.

2015-09-03

Monolithic plate-type fuels comprise of a high density, low enrichment, U10Mo fuel foil encapsulated in a cladding material. This concept generates several fabrication challenges such as flatness, centering or thickness variation. There are concerns, if these parameters have implications on overall performance. To investigate these inquiries, the effects of the foil flatness were studied. For this, a representative plate was simulated for an ideal case. The simulations were repeated for additional cases with various foil curvatures to evaluate the effects on the irradiation performance. The results revealed that the stresses and strains induced by fabrication process are not affected bymore » the flatness of the foil. Furthermore, fabrication stresses in the foil are relieved relatively fast in the reactor. The effects of the foil flatness on peak irradiation stressstrains are minimal. There is a slight increase in temperature for the case with maximum curvature. The major impact is on the displacement characteristics. Furthermore, while the case with a flat foil produces a symmetrical swelling, if the foil is curved, more swelling occurs on the thin-cladding side and the plate bows during irradiation.« less

Development of low-Cr ODS FeCrAl alloys for accident-tolerant fuel cladding

NASA Astrophysics Data System (ADS)

Dryepondt, Sebastien; Unocic, Kinga A.; Hoelzer, David T.; Massey, Caleb P.; Pint, Bruce A.

2018-04-01

Low-Cr oxide dispersion strengthened (ODS) FeCrAl alloys were developed as accident tolerant fuel cladding because of their excellent oxidation resistance at very high temperature, high strength and improved radiation tolerance. Fe-12Cr-5Al wt.% gas atomized powder was ball milled with Y2O3+FeO, Y2O3+ZrO2 or Y2O3+TiO2, and the resulting powders were extruded at 950 °C. The resulting fine grain structure, particularly for the Ti and Zr containing alloys, led to very high strength but limited ductility. Comparison with variants of commercial PM2000 (Fe-20Cr-5Al) highlighted the significant impact of the powder consolidation step on the alloy grain size and, therefore, on the alloy mechanical properties at T < 500 °C. These low-Cr compositions exhibited good oxidation resistance at 1400 °C in air and steam for 4 h but could not form a protective alumina scale at 1450 °C, similar to observations for fine grained PM2000 alloys. The effect of alloy grain size, Zr and Ti additions, and impurities on the alloy mechanical and oxidation behaviors are discussed.

A Multi-Stage Wear Model for Grid-to-Rod Fretting of Nuclear Fuel Rods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blau, Peter Julian

The wear of fuel rod cladding against the supporting structures in the cores of pressurized water nuclear reactors (PWRs) is an important and potentially costly tribological issue. Grid-to-rod fretting (GTRF), as it is known, involves not only time-varying contact conditions, but also elevated temperatures, flowing hot water, aqueous tribo-corrosion, and the embrittling effects of neutron fluences. The multi-stage, closed-form analytical model described in this paper relies on published out-of-reactor wear and corrosion data and a set of simplifying assumptions to portray the conversion of frictional work into wear depth. The cladding material of interest is a zirconium-based alloy called Zircaloy-4,more » and the grid support is made of a harder and more wear-resistant material. Focus is on the wear of the cladding. The model involves an incubation stage, a surface oxide wear stage, and a base alloy wear stage. The wear coefficient, which is a measure of the efficiency of conversion of frictional work into wear damage, can change to reflect the evolving metallurgical condition of the alloy. Wear coefficients for Zircaloy-4 and for a polyphase zirconia layer were back-calculated for a range of times required to wear to a critical depth. Inputs for the model, like the friction coefficient, are taken from the tribology literature in lieu of in-reactor tribological data. Concepts of classical fretting were used as a basis, but are modified to enable the model to accommodate the complexities of the PWR environment. Factors like grid spring relaxation, pre-oxidation of the cladding, multiple oxide phases, gap formation, impact, and hydrogen embrittlement are part of the problem definition but uncertainties in their relative roles limits the ability to validate the model. Sample calculations of wear depth versus time in the cladding illustrate how GTRF wear might occur in a discontinuous fashion during months-long reactor operating cycles. A means to account for grid/rod gaps and repetitive impact effects on GTRF wear is proposed« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ballinger, Ronald G

A material that resists lead-bismuth eutectic (LBE) attack and retains its strength at 700°C would be an enabling technology for LBE-cooled reactors. No single alloy currently exists that can economically meet the required performance criteria of high strength and corrosion resistance. A Functionally Graded Composite (FGC) was developed with layers engineered to perform these functions. F91 was chosen as the structural layer of the composite for its strength and radiation resistance. Fe-12Cr-2Si, an alloy developed from previous work in the Fe-Cr-Si system, was chosen as the corrosion-resistant cladding layer because of its chemical similarity to F91 and its superior corrosionmore » resistance in both oxidizing and reducing environments. Fe-12Cr-2Si experienced minimal corrosion due to its self-passivation in oxidizing and reducing environments. Extrapolated corrosion rates are below one micron per year at 700ï°C. Corrosion of F91 was faster, but predictable and manageable. Diffusion studies showed that 17 microns of the cladding layer will be diffusionally diluted during the three year life of fuel cladding. 33 microns must be accounted for during the sixty year life of coolant piping. 5 cm coolant piping and 6.35 mm fuel cladding preforms were produced on a commercial scale by weld-overlaying Fe-12Cr-2Si onto F91 billets and co-extruding them. An ASME certified weld was performed followed by the prescribed quench-and-tempering heat treatment for F91. A minimal heat affected zone was observed, demonstrating field weldability. Finally, corrosion tests were performed on the fabricated FGC at 700ï°C after completely breaching the cladding in a small area to induce galvanic corrosion at the interface. None was observed. This FGC has significant impacts on LBE reactor design. The increases in outlet temperature and coolant velocity allow a large increase in power density, leading to either a smaller core for the same power rating or more power output for the same size core. This FGC represents an enabling technology for LBE cooled fast reactors.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kohnert, Aaron A.; Dasgupta, Dwaipayan; Wirth, Brian

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, the material response must be demonstrated to provide suitable radiation stability, in order to ensuremore » that there will not be significant dimensional changes (e.g., swelling), as well as quantifying the radiation hardening and radiation creep behavior. In this report, we describe the use of cluster dynamics modeling to evaluate the defect physics and damage accumulation behavior of FeCrAl alloys subjected to neutron irradiation, with a particular focus on irradiation-induced swelling and defect fluxes to dislocations that are required to model irradiation creep behavior.« less

Gas emission from the UO2 samples, containing fission products and burnable absorber

NASA Astrophysics Data System (ADS)

Kopytin, V. P.; Baranov, V. G.; Burlakova, M. A.; Tenishev, A. V.; Kuzmin, R. S.; Pokrovskiy, S. A.; Mikhalchik, V. V.

2016-04-01

The process gas released from the fuel pellets of uranium fuel during fuel burn-up reduces the thermal conductivity of the rod-shell gap, enhances hydrogen embrittlement of the cladding material, causes it's carbonization, as well as transport processes in the fuel. In this study a technique of investigating the thermal desorption of gases from the UO2 fuel material were perfected in the temperature range 300-2000 K for uniform sample heating rate of 15 K/min in vacuum. The characteristic kinetic dependences are acquired for the gas emission from UO2 samples, containing simulators of fission products (SFP) and the burnable neutron absorber (BNA). Depending on the amount of SFP and BNA contained in the sample thermal desorption gas spectra (TDGS) vary. The composition of emitted gas varies, as well as the number of peaks in the TDGS and the peaks shift to higher temperatures. This indicates that introduction of SFPs and BNA alters the sample material structure and cause the creation of so- called traps which have different bonding energies to the gases. The traps can be a grid of dislocations, voids, and contained in the UO2 matrix SFP and BNA. Similar processes will occur in the fuel pellets in the real conditions of the Nuclear Power Plant as well.

Loss-of-flow-without-scram tests in Experimental Breeder Reactor-II and comparison with pretest predictions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chang, L.K.; Mohr, D.; Planchon, H.P.

This article discusses a series of successful loss-of-flow-without-scram tests conducted in Experimental Breeder Reactor-II (EBR-II), a metal-fueled, sodium-cooled fast reactor. These May 1985 tests demonstrated the capability of the EBR to reduce reactor power passively during a loss of flow and to maintain reactor temperatures within bounds without any reliance on an active safety system. The tests were run from reduced power to ensure that temperatures could be maintained well below the fuel-clad eutectic temperature. Good agreement was found between selected test data and pretest predictions made with the EBR-II system analysis code NATDEMO and the hot channel analysis codemore » HOTCHAN. The article also discusses safety assessments of the tests as well as modifications required on the EBR-II reactor safety system for conducting required on the EBR-II reactor safety system for the conducting the tests.« less

FABRICATION DEVELOPMENT OF UO$sub 2$-STAINLESS STEEL COMPOSITE FUEL PLATES FOR CORE B OF THE ENRICO FERMI FAST BREEDER REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cherubini, J.H.; Beaver, R.J.; Leitten, C.F. Jr.

1961-04-18

The development of an inexpensive composite fuel plate with a high burnup potential for application in a 500 deg C sodium environment as Core B of the Enrico Fermi Fast Breeder Reactor is described. The dispersion fuel product consists of 35 wt.% spheroidal UO/sub 2/ dispersed in type 347B stainless steel powder and clad with wrought type 347 stainless steel. Nominal over-all dimensions of Type II design fuel plates are 18.97 in. long x 2.406 in. wide x 0.112 in. thick with 0.005-in. cladding. Reliable processing methods for achieving a uniform distribution of spheroidal UO/sub 2/ in the matrix powdermore » and cladding the sintered powder compact by roll bonding are described. Examination of experimental plates reveals that the degree of UO/sub 2/ fragmentation and stringering encountered during processing is primarily a function of the degree of cold work employed in the finishing operation snd the starting quality of the UO/sub 2/ powder. Cladding studies indicate that a sound metallurgical bond can be achieved with an 87.5% reduction in thickness at 1200 deg C and that close processing control is required to meet the stringent tolerances specified. The developed process meets all criteria except possibly the surface finish requirement; occasionally, pitting occurs due to scale embedded during hot working. Detailed procedures covering composite plate manufacture are presented. (auth)« less

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Accident Tolerant Fuels High Impact Problem: Coordinate Multiscale FeCrAl Modeling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gamble, K. A.; Hales, J. D.; Zhang, Y.

Since the events at the Fukushima-Daiichi nuclear power plant in March 2011 significant research has unfolded at national laboratories, universities and other institutions into alternative materials that have potential enhanced ac- cident tolerance when compared to traditional UO2 fuel zircaloy clad fuel rods. One of the potential replacement claddings are iron-chromium-alunimum (FeCrAl) alloys due to their increased oxidation resistance [1–4] and higher strength [1, 2]. While the oxidation characteristics of FeCrAl are a benefit for accident tolerance, the thermal neu- tron absorption cross section of FeCrAl is about ten times that of Zircaloy. This neutronic penalty necessitates thinner cladding. Thismore » allows for slightly larger pellets to give the same cold gap width in the rod. However, the slight increase in pellet diameter is not sufficient to compensate for the neutronic penalty and enriching the fuel beyond the current 5% limit appears to be necessary [5]. Current estimates indicate that this neutronic penalty will impose an increase in fuel cost of 15-35% [1, 2]. In addition to the neutronic disadvantage, it is anticipated that tritium release to the coolant will be larger because the permeability of hydrogen in FeCrAl is about 100 times higher than in Zircaloy [6]. Also, radiation-induced hardening and embrittlement of FeCrAl need to be fully characterized experimentally [7]. Due to the aggressive development schedule for inserting some of the potential materials into lead test assemblies or rods by 2022 [8] multiscale multiphysics modeling approaches have been used to provide insight into these the use of FeCrAl as a cladding material. The purpose of this letter report is to highlight the multiscale modeling effort for iron-chromium-alunimum (FeCrAl) cladding alloys as part of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program through its Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The approach taken throughout the HIP is to utilize lower length scale approaches (e.g., density functional theory, cluster dynamics, rate theory, phase field, and Visco-Plastic- Self-Consistent (VPSC)) to develop more physically informed models at the engineering scale for use in the BISON [9] fuel performance code.« less

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Picklesimer, M.L.; Thurber, W.C.

1961-01-01

A chemically nonreactive fuel composition for incorporation in aluminum- clad, plate type fuel elements for neutronic reactors is described. The composition comprises a mixture of aluminum and uranium carbide particles, the uranium carbide particles containing at least 80 wt.% UC/sub 2/.

Iron-chrome-aluminum alloy cladding for increasing safety in nuclear power plants

NASA Astrophysics Data System (ADS)

Rebak, Raul B.

2017-12-01

After a tsunami caused plant black out at Fukushima, followed by hydrogen explosions, the US Department of Energy partnered with fuel vendors to study safer alternatives to the current UO2-zirconium alloy system. This accident tolerant fuel alternative should better tolerate loss of cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. General electric, Oak ridge national laboratory, and their partners are proposing to replace zirconium alloy cladding in current commercial light water power reactors with an iron-chromium-aluminum (FeCrAl) cladding such as APMT or C26M. Extensive testing and evaluation is being conducted to determine the suitability of FeCrAl under normal operation conditions and under severe accident conditions. Results show that FeCrAl has excellent corrosion resistance under normal operation conditions and FeCrAl is several orders of magnitude more resistant than zirconium alloys to degradation by superheated steam under accident conditions, generating less heat of oxidation and lower amount of combustible hydrogen gas. Higher neutron absorption and tritium release effects can be minimized by design changes. The implementation of FeCrAl cladding is a near term solution to enhance the safety of the current fleet of commercial light water power reactors.

Use of ion beams to simulate reaction of reactor fuels with their cladding

NASA Astrophysics Data System (ADS)

Birtcher, R. C.; Baldo, P.

2006-01-01

Processes occurring within reactor cores are not amenable to direct experimental observation. Among major concerns are damage, fission gas accumulation and reaction between the fuel and its cladding all of which lead to swelling. These questions can be investigated through simulation with ion beams. As an example, we discuss the irradiation driven interaction of uranium-molybdenum alloys, intended for use as low-enrichment reactor fuels, with aluminum, which is used as fuel cladding. Uranium-molybdenum coated with a 100 nm thin film of aluminum was irradiated with 3 MeV Kr ions to simulate fission fragment damage. Mixing and diffusion of aluminum was followed as a function of irradiation with RBS and nuclear reaction analysis using the 27Al(p,γ)28Si reaction which occurs at a proton energy of 991.9 keV. During irradiation at 150 °C, aluminum diffused into the uranium alloy at a irradiation driven diffusion rate of 30 nm2/dpa. At a dose of 90 dpa, uranium diffusion into the aluminum layer resulted in formation of an aluminide phase at the initial interface. The thickness of this phase grew until it consumed the aluminum layer. The rapid diffusion of Al into these reactor fuels may offer explanation of the observation that porosity is not observed in the fuel particles but on their periphery.

Finite element simulation of gap opening between cladding tube and spacer grid in a fuel rod assembly using crystallographic models of irradiation growth and creep

DOE Office of Scientific and Technical Information (OSTI.GOV)

Patra, Anirban; Tomé, Carlos N.

A physically-based crystal plasticity framework for modeling irradiation growth and creep is interfaced with the finite element code ABAQUS in order to study the contact forces and the gap evolution between the spacer grid and the cladding tube as a function of irradiation in a representative section of a fuel rod assembly. Deformation mechanisms governing the gap opening are identified and correlated to the texture-dependent material response. Thus, in the absence of coolant flow-induced vibrations, these simulations predict the contribution of irradiation growth and creep to the gap opening between the cladding tube and the springs and dimples on themore » spacer grid. The simulated contact forces on the springs and dimples are compared to available experimental and modeling data. Various combinations of external loads are applied on the springs and dimples to simulate fuel rods in the interior and at the periphery of the fuel rod assembly. Furthermore, we found that loading conditions representative (to a first order approximation) of fuel rods at the periphery show higher gap opening. This is in agreement with in-reactor data, where rod leakages due to the synergistic effects of gap opening and coolant flow-induced vibrations were generally found to occur at the periphery of the fuel rod assembly.« less

Finite element simulation of gap opening between cladding tube and spacer grid in a fuel rod assembly using crystallographic models of irradiation growth and creep

DOE PAGES

Patra, Anirban; Tomé, Carlos N.

2017-03-06

A physically-based crystal plasticity framework for modeling irradiation growth and creep is interfaced with the finite element code ABAQUS in order to study the contact forces and the gap evolution between the spacer grid and the cladding tube as a function of irradiation in a representative section of a fuel rod assembly. Deformation mechanisms governing the gap opening are identified and correlated to the texture-dependent material response. Thus, in the absence of coolant flow-induced vibrations, these simulations predict the contribution of irradiation growth and creep to the gap opening between the cladding tube and the springs and dimples on themore » spacer grid. The simulated contact forces on the springs and dimples are compared to available experimental and modeling data. Various combinations of external loads are applied on the springs and dimples to simulate fuel rods in the interior and at the periphery of the fuel rod assembly. Furthermore, we found that loading conditions representative (to a first order approximation) of fuel rods at the periphery show higher gap opening. This is in agreement with in-reactor data, where rod leakages due to the synergistic effects of gap opening and coolant flow-induced vibrations were generally found to occur at the periphery of the fuel rod assembly.« less

Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Su'ud, Zaki, E-mail: szaki@fi.itba.c.id; Sekimoto, H., E-mail: hsekimot@gmail.com

2014-09-30

Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature canmore » be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.« less

System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

NASA Astrophysics Data System (ADS)

Porter, Ian Edward

A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several additional fuels will also be analyzed, including uranium nitride (UN), uranium carbide (UC) and uranium silicide (U3Si2). Focusing on the system response in an accident scenario, an emphasis is placed on the fracture mechanics of the ceramic cladding by design the fuel rods to eliminate pellet cladding mechanical interaction (PCMI). The time to failure and how much of the fuel in the reactor fails with an advanced fuel design will be analyzed and compared to the current UO2/Zircaloy design using a full scale reactor model.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kraus, A.; Garner, P.; Hanan, N.

Thermal-hydraulic simulations have been performed using computational fluid dynamics (CFD) for the highly-enriched uranium (HEU) design of the IVG.1M reactor at the Institute of Atomic Energy (IAE) at the National Nuclear Center (NNC) in the Republic of Kazakhstan. Steady-state simulations were performed for both types of fuel assembly (FA), i.e. the FA in rows 1 & 2 and the FA in row 3, as well as for single pins in those FA (600 mm and 800 mm pins). Both single pin calculations and bundle sectors have been simulated for the most conservative operating conditions corresponding to the 10 MW outputmore » power, which corresponds to a pin unit cell Reynolds number of only about 7500. Simulations were performed using the commercial code STAR-CCM+ for the actual twisted pin geometry as well as a straight-pin approximation. Various Reynolds-Averaged Navier-Stokes (RANS) turbulence models gave different results, and so some validation runs with a higher-fidelity Large Eddy Simulation (LES) code were performed given the lack of experimental data. These singled out the Realizable Two-Layer k-ε as the most accurate turbulence model for estimating surface temperature. Single-pin results for the twisted case, based on the average flow rate per pin and peak pin power, were conservative for peak clad surface temperature compared to the bundle results. Also the straight-pin calculations were conservative as compared to the twisted pin simulations, as expected, but the single-pin straight case was not always conservative with regard to the straight-pin bundle. This was due to the straight-pin temperature distribution being strongly influenced by the pin orientation, particularly near the outer boundary. The straight-pin case also predicted the peak temperature to be in a different location than the twisted-pin case. This is a limitation of the straight-pin approach. The peak temperature pin was in a different location from the peak power pin in every case simulated, and occurred at an inner pin just before the enrichment change. The 600 mm case demonstrated a peak clad surface temperature of 370.4 K, while the 800 mm case had a temperature of 391.6 K. These temperatures are well below the necessary temperatures for boiling to occur at the rated pressure. Fuel temperatures are also well below the melting point. Future bundle work will include simulations of the proposed low-enriched uranium (LEU) design. Two transient scenarios were also investigated for the single-pin geometries. Both were “model” problems that were focused on pure thermal-hydraulic behavior, and as such were simple power changes that did not incorporate neutron kinetics modeling. The first scenario was a high-power, ramp increase, while the second scenario was a low-power, step increase. A cylindrical RELAP model was also constructed to investigate its accuracy as compared to the higher-fidelity CFD. Comparisons between the two codes showed good agreement for peak temperatures in the fuel and at the cladding surface for both cases. In the step transient, temperatures at four axial levels were also computed. These showed greater but reasonable discrepancy, with RELAP outputting higher temperatures. These results provide some evidence that RELAP can be used with confidence in modeling transients for IVG.« less

Performance of U3Si2 Fuel in a Reactivity Insertion Accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng, Lap Y.; Cuadra, Arantxa; Todosow, Michael

In this study we examined the performance of the U3Si2 fuel cladded with Zircaloy (Zr) in a reactivity insertion accident (RIA) in a PWR core. The power excursion as a result of a $1 reactivity insertion was calculated by a TRACE PWR plant model using point-kinetics, for alternative cores with UO2 and U3Si2 fuel assemblies. The point-kinetics parameters (feedback coefficients, prompt-neutron lifetime and group constants for six delayed-neutron groups) were obtained from beginning-of-cycle equilibrium full core calculations with PARCS. In the PARCS core calculations, the few-group parameters were developed utilizing the TRITON/NEWT tools in the SCALE package. In order tomore » assess the fuel response in finer detail (e.g. the maximum fuel temperature) the power shape and thermal boundary conditions from the TRACE/PARCS calculations were used to drive a BISON model of a fuel pin with U3Si2 and UO2 respectively. For a $1 reactivity transient both TRACE and BISON predicted a higher maximum fuel temperature for the UO2 fuel than the U3Si2 fuel. Furthermore, BISON is noted to calculate a narrower gap and a higher gap heat transfer coefficient than TRACE. This resulted in BISON predicting consistently lower fuel temperatures than TRACE. This study also provides a systematic comparison between TRACE and BISON using consistent transient boundary conditions. The TRACE analysis of the RIA only reflects the core-wide response in power. A refinement to the analysis would be to predict the local peaking in a three-dimensional core as a result of control rod ejection.« less

Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

NASA Astrophysics Data System (ADS)

Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin; Park, Jong Man; Sohn, Dong-Seong

2018-04-01

A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature- and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS). The code was validated using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code.

Laser Cladding of Ni, Nb, and Mg Alloys for Improved Environmental Resistance at High Temperature

DTIC Science & Technology

1989-01-01

v*LASER CLADDING OF NI , Nb AND Mg ALLOYS < FOR 7IMPR-OVED ENVIIONM ENTAL I RESISTANCE AT HIGH TEMPERATURE Final Report for Research Conducted through...resistance at high temperature. Major emphasis has been on Ni -Cr-Al-Hf system. Microstructural evolution and oxidation properties of Ni and Nb alloys ...metastable crystalline and amorphous structure on a) the high temperature oxidation properties of laser clad Ni and Nb alloys , and b) the corrosion

Cladding and duct materials for advanced nuclear recycle reactors

NASA Astrophysics Data System (ADS)

Allen, T. R.; Busby, J. T.; Klueh, R. L.; Maloy, S. A.; Toloczko, M. B.

2008-01-01

The expanded use of nuclear energy without risk of nuclear weapons proliferation and with safe nuclear waste disposal is a primary goal of the Global Nuclear Energy Partnership (GNEP). To achieve that goal the GNEP is exploring advanced technologies for recycling spent nuclear fuel that do not separate pure plutonium, and advanced reactors that consume transuranic elements from recycled spent fuel. The GNEP’s objectives will place high demands on reactor clad and structural materials. This article discusses the materials requirements of the GNEP’s advanced nuclear recycle reactors program.

PRESSURIZED WATER REACTOR PROGRAM TECHNICAL PROGRESS REPORT FOR THE PERIOD MAY 5, 1955 TO JUNE 16, 1955

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

The current PWR plant and core parameters are listed. Resign requirements are briefly summarized for a radiation monitoring system, a fuel handling water system, a coolant purification system, an electrical power distribution system, and component shielding. Results of studies on thermal bowing and stressing of UO/sub 2/ are reported. A graph is presented of reactor power vs. reactor flow for various hot channel conditions. Development of U-- Mo and U-Nb alloys has been stopped because of the recent selection of UO/sub 2/ fuel material for the PWR core and blanket. The fabrication characteristics of UO/sub 2/ powders are being studied.more » Seamless Zircaloy-2 tubing has been tested to determine elastic limits, bursting pressures, and corrosion resistance. Fabrication techniques and tests for corrosion and defects in Zircaloy-clad U-Mo and UO/sub 2/ fuel rods are described. The preparation of UO/sub 2/ by various methods is being studied to determine which method produces a material most suitable for PWR fuel elements. The stability of UO/sub 2/ compacts in high temperature water and steam is being determined. Surface area and density measurements have been performed on samples of UO/sub 2/ powder prepared by various methods. Revelopment work on U-- Mo and U--Nb alloys has included studies of the effect on corrosion behavior of additions to the test water, additions to the alloys, homogenization of the alloys, annealing times, cladding, and fabrication techniques. Data are presented on relaxation in spring materials after exposure to a corrosive environment. Results are reported from loop and autoclave tests on fission product and crud deposition. Results of irradiation and corrosion testing of clad and unclad U--Mo and U-Nh alloys are described. The UO/sub 2/ irradiation program has included studies of dimensional changes, release of fission gases, and activity in the water surrounding the samples. A review of the methods of calculating reactor physics parameters has been completed, and the established procedures have been applied to determination of PWR reference design parameters. Critical experiments and primary loop shielding analyses are described. (D.E.B.)« less

Status of advanced fuel candidates for Sodium Fast Reactor within the Generation IV International Forum

DOE Office of Scientific and Technical Information (OSTI.GOV)

F. Delage; J. Carmack; C. B. Lee

2013-10-01

The main challenge for fuels for future Sodium Fast Reactor systems is the development and qualification of a nuclear fuel sub-assembly which meets the Generation IV International Forum goals. The Advanced Fuel project investigates high burn-up minor actinide bearing fuels as well as claddings and wrappers to withstand high neutron doses and temperatures. The R&D outcome of national and collaborative programs has been collected and shared between the AF project members in order to review the capability of sub-assembly material and fuel candidates, to identify the issues and select the viable options. Based on historical experience and knowledge, both oxidemore » and metal fuels emerge as primary options to meet the performance and the reliability goals of Generation IV SFR systems. There is a significant positive experience on carbide fuels but major issues remain to be overcome: strong in-pile swelling, atmosphere required for fabrication as well as Pu and Am losses. The irradiation performance database for nitride fuels is limited with longer term R&D activities still required. The promising core material candidates are Ferritic/Martensitic (F/M) and Oxide Dispersed Strengthened (ODS) steels.« less

Measurement of thermal diffusivity of depleted uranium metal microspheres

NASA Astrophysics Data System (ADS)

Humrickhouse-Helmreich, Carissa J.; Corbin, Rob; McDeavitt, Sean M.

2014-03-01

The high void space of nuclear fuels composed of homogeneous uranium metal microspheres may allow them to achieve ultra-high burnup by accommodating fuel swelling and reducing fuel/cladding interactions; however, the relatively low thermal conductivity of microsphere nuclear fuels may limit their application. To support the development of microsphere nuclear fuels, an apparatus was designed in a glovebox and used to measure the apparent thermal diffusivity of a packed bed of depleted uranium (DU) microspheres with argon fill in the void spaces. The developed Crucible Heater Test Assembly (CHTA) recorded radial temperature changes due to an initial heat pulse from a central thin-diameter cartridge heater. Using thermocouple positions and time-temperature data, the apparent thermal diffusivity was calculated. The thermal conductivity of the DU microspheres was calculated based on the thermal diffusivity from the CHTA, known material densities and specific heat capacities, and an assumed 70% packing density based on prior measurements. Results indicate that DU metal microspheres have very low thermal conductivity, relative to solid uranium metal, and rapidly form an oxidation layer even in a low oxygen environment. At 500 °C, the thermal conductivity of the DU metal microsphere bed was 0.431 ± 0.0560 W/m-K compared to the literature value of approximately 32 W/m-K for solid uranium metal.

Review of SiCf/SiCm corrosion, erosion and erosion-corrosion in high temperature helium relevant to GFR conditions

NASA Astrophysics Data System (ADS)

Fitzgerald, Kerr; Shepherd, Daniel

2018-01-01

The good thermal, mechanical and nuclear properties of silicon carbide composites (SiCf/SiCm) has led to their proposal as a gas-cooled fast reactor (GFR) fuel cladding material. Accordingly, previous studies on the corrosion, erosion and erosion-corrosion of SiCf/SiCm have been reviewed. A significant amount of corrosion data was compared for various SiCf/SiCm types tested under different conditions, which suggested that it may be suitable for use in a GFR. However some issues remain, specifically the potential for removal by oxidation of the pyrolytic carbon (PyC) interphase below 900 °C and highly damaging active oxidation at temperatures above 1200 °C. Few relevant papers were found on the erosion and erosion-corrosion behaviour of SiCf/SiCm, though those that were did indicate improved erosion resistance compared to metals. However, most data concerned particulate rather than gas erosion, and at relatively low temperatures. Exacerbation of erosion by corrosion (and vice versa) is hypothesised due to both phenomena potentially causing and being worsened by increased composite porosity. The possibility for this to be further exacerbated mechanically was also identified e.g. by surface damage or by crack formation and growth. The potential impact of irradiation in fast reactor conditions also needs to be better understood. Overall, limited data was found that is representative of anticipated GFR conditions and only in the form of separate effect corrosion tests, thus highlighting the need for erosion and combined effect experiments to progress SiCf/SiCm towards qualification as GFR fuel cladding. Finally, some potential means for improving performance have been identified including environmental adjustment and/or alternative composite design.

Bending testing and characterization of surrogate nuclear fuel rods made of Zircaloy-4 cladding and aluminum oxide pellets

DOE PAGES

Wang, Hong; Wang, Jy-An John

2016-07-20

We studied behavior of surrogate nuclear fuel rods made of Zircaloy-4 (Zry-4) cladding with alumina pellets under reversed cyclic bending. Tests were performed under load or moment control at 5 Hz, and an empirical correlation was established between rod fatigue life and amplitude of the applied moment. Fatigue response of Zry-4 cladding was further characterized by using flexural rigidity. Degradation of flexural rigidity was shown to depend on the moment applied and the prefatigue condition of specimens. Pellet-to-pellet interface (PPI), pellet-to-cladding interface (PCI), and pellet condition all affect surrogate rod failure. Bonding/debonding of PPI/PCI and pellet fracturing contribute to surrogatemore » rod bending fatigue. Also, the effect of sensor spacing on curvature measurement using three-point deflections was studied; the method based on effective specimen gauge length is effective in sensor spacing correction. Finally, we developed the database and gained understanding in this study such that it will serve as input to analysis of SNF vibration integrity.« less

Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

DOE Office of Scientific and Technical Information (OSTI.GOV)

M. Chen; CM Regan; D. Noe

2006-01-09

Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas releasemore » and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.« less

Effect of Temperature and Sheet Temper on Isothermal Solidification Kinetics in Clad Aluminum Brazing Sheet

NASA Astrophysics Data System (ADS)

Benoit, Michael J.; Whitney, Mark A.; Wells, Mary A.; Winkler, Sooky

2016-09-01

Isothermal solidification (IS) is a phenomenon observed in clad aluminum brazing sheets, wherein the amount of liquid clad metal is reduced by penetration of the liquid clad into the core. The objective of the current investigation is to quantify the rate of IS through the use of a previously derived parameter, the Interface Rate Constant (IRC). The effect of peak temperature and initial sheet temper on IS kinetics were investigated. The results demonstrated that IS is due to the diffusion of silicon (Si) from the liquid clad layer into the solid core. Reduced amounts of liquid clad at long liquid duration times, a roughened sheet surface, and differences in resolidified clad layer morphology between sheet tempers were observed. Increased IS kinetics were predicted at higher temperatures by an IRC model as well as by experimentally determined IRC values; however, the magnitudes of these values are not in good agreement due to deficiencies in the model when applied to alloys. IS kinetics were found to be higher for sheets in the fully annealed condition when compared with work-hardened sheets, due to the influence of core grain boundaries providing high diffusivity pathways for Si diffusion, resulting in more rapid liquid clad penetration.

Temperature influence on the cladding mode distribution in highly localized point-by-point fibre Bragg gratings

NASA Astrophysics Data System (ADS)

Caucheteur, C.; Gonzalez-Vila, A.; Chikh-Bled, H.; Lasri, B.; Kinet, D.; Chah, K.

2016-05-01

An infrared femtosecond pulses laser is used to manufacture point-by-point gratings in telecommunication-grade optical fibres. The refractive index modulations are localized close to the core-cladding interface, yielding a strong coupling to cladding mode resonances together with an important photo-induced birefringence. Such gratings have been recently used for refractrometric measurements. In this work, their transmitted amplitude spectrum is measured with polarized light while they are exposed to temperature changes up to 900 °C. Despite an overall good thermal stability of the gratings that confirms their robustness for high-temperature refractometry, we report an interesting polarization effect depending on both the cladding mode resonance family (radially- and azimuthally-polarized modes) and mode order. While the birefringence of the core mode resonance decreases with the temperature, certain cladding mode resonances show an increase of the wavelength splitting between their orthogonally-polarized components. This differential behaviour can be of high interest to develop high-resolution multiparametric sensing platforms.

Perform Tests and Document Results and Analysis of Oxide Layer Effects and Comparisons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collins, E. D.; DelCul, G. D.; Spencer, B. B.

2014-08-30

During the initial feasibility test using actual used nuclear fuel (UNF) cladding in FY 2012, an incubation period of 30–45 minutes was observed in the initial dry chlorination. The cladding hull used in the test had been previously oxidized in a dry air oxidation pretreatment prior to removal of the fuel. The cause of this incubation period was attributed to the resistance to chlorination of an oxide layer imparted by the dry oxidation pretreatment on the cladding. Subsequently in 2013, researchers at the Korea Atomic Energy Institute (KAERI) reported on their chlorination study [R1] on ~9-gram samples of unirradiated ZirloTMmore » cladding tubes that had been previously oxidized in air at 500oC for various time periods to impart oxide layers of varying thickness. In early 2014, discussions with Indefinite Delivery, Indefinite Quantity (IDIQ) contracted technical consultants from Westinghouse described their previous development (and patents) [R2] on methods of chemical washing to remove some or all of the hydrous oxide layer imparted on UNF cladding during irradiation in light water reactors (LWRs) . Thus, the Oak Ridge National Laboratory (ORNL) study, described herein, was planned to extend the KAERI study on the effects of anhydrous oxide layers, but on larger ~100-gram samples of unirradiated zirconium alloy cladding tubes, and to investigate the effects of various methods of chemical pretreatment prior to chlorination with 100% chlorine on the average reaction rates and Cl2 usage efficiencies.« less

Oxygen potentials of mixed oxide fuels for fast reactors

NASA Astrophysics Data System (ADS)

Kato, M.; Tamura, T.; Konashi, K.

2009-03-01

Oxygen potentials of homogenous (Pu0.2U0.8)O2-x and (Am0.02Pu0.30Np0.02U0.66)O2-x which have been developed as fuels for fast breeder reactors were measured at temperatures of 1473-1623 K by a gas equilibrium method using an (Ar, H2, H2O) gas mixture. The measured oxygen potentials of (Pu0.2U0.8)O2-x were about 25 kJ mol-1 lower than those of (Pu0.3U0.7)O2-x measured previously and were consistent with the values calculated by Besmann and Lindemer's model. The measured oxygen potentials of (Am0.02Pu0.30Np0.02U0.66)O2-x were slightly higher than those of MOX without minor actinides. No fuel-cladding chemical interaction is affected significantly by adding their minor actinides.

75 FR 25120 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-07

...-235, clarify the requirements of reconstituted fuel assemblies, add requirements to qualify metal matrix composite neutron absorbers with integral aluminum cladding, delete use of nitrogen for draining...

EPRI-NASA Cooperative Project on Stress Corrosion Cracking of Zircaloys. [nuclear fuel failures

NASA Technical Reports Server (NTRS)

Cubicciotti, D.; Jones, R. L.

1978-01-01

Examinations of the inside surface of irradiated fuel cladding from two reactors show the Zircaloy cladding is exposed to a number of aggressive substances, among them iodine, cadmium, and iron-contaminated cesium. Iodine-induced stress corrosion cracking (SCC) of well characterized samples of Zircaloy sheet and tubing was studied. Results indicate that a threshold stress must be exceeded for iodine SCC to occur. The existence of a threshold stress indicates that crack formation probably is the key step in iodine SCC. Investigation of the crack formation process showed that the cracks responsible for SCC failure nucleated at locations in the metal surface that contained higher than average concentrations of alloying elements and impurities. A four-stage model of iodine SCC is proposed based on the experimental results and the relevance of the observations to pellet cladding interaction failures is discussed.

UFD Storage and Transportation - Transportation Working Group Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maheras, Steven J.; Ross, Steven B.

2011-08-01

The Used Fuel Disposition (UFD) Transportation Task commenced in October 2010. As its first task, Pacific Northwest National Laboratory (PNNL) compiled a list of structures, systems, and components (SSCs) of transportation systems and their possible degradation mechanisms during extended storage. The list of SSCs and the associated degradation mechanisms [known as features, events, and processes (FEPs)] were based on the list of used nuclear fuel (UNF) storage system SSCs and degradation mechanisms developed by the UFD Storage Task (Hanson et al. 2011). Other sources of information surveyed to develop the list of SSCs and their degradation mechanisms included references suchmore » as Evaluation of the Technical Basis for Extended Dry Storage and Transportation of Used Nuclear Fuel (NWTRB 2010), Transportation, Aging and Disposal Canister System Performance Specification, Revision 1 (OCRWM 2008), Data Needs for Long-Term Storage of LWR Fuel (EPRI 1998), Technical Bases for Extended Dry Storage of Spent Nuclear Fuel (EPRI 2002), Used Fuel and High-Level Radioactive Waste Extended Storage Collaboration Program (EPRI 2010a), Industry Spent Fuel Storage Handbook (EPRI 2010b), and Transportation of Commercial Spent Nuclear Fuel, Issues Resolution (EPRI 2010c). SSCs include items such as the fuel, cladding, fuel baskets, neutron poisons, metal canisters, etc. Potential degradation mechanisms (FEPs) included mechanical, thermal, radiation and chemical stressors, such as fuel fragmentation, embrittlement of cladding by hydrogen, oxidation of cladding, metal fatigue, corrosion, etc. These degradation mechanisms are discussed in Section 2 of this report. The degradation mechanisms have been evaluated to determine if they would be influenced by extended storage or high burnup, the need for additional data, and their importance to transportation. These categories were used to identify the most significant transportation degradation mechanisms. As expected, for the most part, the transportation importance was mirrored by the importance assigned by the UFD Storage Task. A few of the more significant differences are described in Section 3 of this report« less

Fuel Thermo-physical Characterization Project. Fiscal Year 2014 Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burkes, Douglas; Casella, Andrew M.; Buck, Edgar C.

2015-03-15

The Office of Material Management and Minimization (M3) Reactor Conversion Fuel Thermo-Physical Characterization Project at Pacific Northwest National Laboratory (PNNL) was tasked with using PNNL facilities and processes to receive irradiated low enriched uranium–molybdenum (LEU-Mo) fuel plate samples and perform analysis in support of the M3 Reactor Conversion Program. This work is in support of the M3 Reactor Conversion Fuel Development Pillar that is managed by Idaho National Laboratory. The primary research scope was to determine the thermo-physical properties as a function of temperature and burnup. Work conducted in Fiscal Year (FY) 2014 complemented measurements performed in FY 2013 onmore » four additional irradiated LEU-Mo fuel plate samples. Specifically, the work in FY 2014 investigated the influence of different processing methods on thermal property behavior, the absence of aluminum alloy cladding on thermal property behavior for additional model validation, and the influence of higher operating surface heat flux / more aggressive irradiation conditions on thermal property behavior. The model developed in FY 2013 and refined in FY 2014 to extract thermal properties of the U-Mo alloy from the measurements conducted on an integral fuel plate sample (i.e., U-Mo alloy with a thin Zr coating and clad in AA6061) continues to perform very well. Measurements conducted in FY 2014 on samples irradiated under similar conditions compare well to measurements performed in FY 2013. In general, there is no gross influence of fabrication method on thermal property behavior, although the difference in LEU-Mo foil microstructure does have a noticeable influence on recrystallization of grains during irradiation. Samples irradiated under more aggressive irradiation conditions, e.g., higher surface heat flux, revealed lower thermal conductivity when compared to samples irradiated at moderate surface heat fluxes, with the exception of one sample. This report documents thermal property measurements conducted in FY 2014 and compares results to values obtained from literature and measurements performed in FY 2013, where applicable, along with appropriate discussion.« less

Technique development for modulus, microcracking, hermeticity, and coating evaluation capability characterization of SiC/SiC tubes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Xunxiang; Ang, Caen K.; Singh, Gyanender P.

Driven by the need to enlarge the safety margins of nuclear fission reactors in accident scenarios, research and development of accident-tolerant fuel has become an important topic in the nuclear engineering and materials community. A continuous-fiber SiC/SiC composite is under consideration as a replacement for traditional zirconium alloy cladding owing to its high-temperature stability, chemical inertness, and exceptional irradiation resistance. An important task is the development of characterization techniques for SiC/SiC cladding, since traditional work using rectangular bars or disks cannot directly provide useful information on the properties of SiC/SiC composite tubes for fuel cladding applications. At Oak Ridge Nationalmore » Laboratory, experimental capabilities are under development to characterize the modulus, microcracking, and hermeticity of as-fabricated, as-irradiated SiC/SiC composite tubes. Resonant ultrasound spectroscopy has been validated as a promising technique to evaluate the elastic properties of SiC/SiC composite tubes and microcracking within the material. A similar technique, impulse excitation, is efficient in determining the basic mechanical properties of SiC bars prepared by chemical vapor deposition; it also has potential for application in studying the mechanical properties of SiC/SiC composite tubes. Complete evaluation of the quality of the developed coatings, a major mitigation strategy against gas permeation and hydrothermal corrosion, requires the deployment of various experimental techniques, such as scratch indentation, tensile pulling-off tests, and scanning electron microscopy. In addition, a comprehensive permeation test station is being established to assess the hermeticity of SiC/SiC composite tubes and to determine the H/D/He permeability of SiC/SiC composites. This report summarizes the current status of the development of these experimental capabilities.« less

Development of U-frame bending system for studying the vibration integrity of spent nuclear fuel

NASA Astrophysics Data System (ADS)

Wang, Hong; Wang, Jy-An John; Tan, Ting; Jiang, Hao; Cox, Thomas S.; Howard, Rob L.; Bevard, Bruce B.; Flanagan, Michelle

2013-09-01

A bending fatigue system developed to evaluate the response of spent nuclear fuel rods to vibration loads is presented. A U-frame testing setup is used for imposing bending loads on the fuel rod specimen. The U-frame setup consists of two rigid arms, side connecting plates to the rigid arms, and linkages to a universal testing machine. The test specimen's curvature is obtained through a three-point deflection measurement method. The tests using surrogate specimens with stainless steel cladding revealed increased flexural rigidity under unidirectional cyclic bending, significant effect of cladding-pellets bonding on the response of surrogate rods, and substantial cyclic softening in reverse bending mode. These phenomena may cast light on the expected response of a spent nuclear fuel rod. The developed U-frame system is thus verified and demonstrated to be ready for further pursuit in hot-cell tests.

Diffusion Bonding Technology of Tungsten and SiC/SiC Composites for Nuclear Applications

NASA Astrophysics Data System (ADS)

Kishimoto, Hirotatsu; Shibayama, Tamaki; Abe, Takahiro; Shimoda, Kazuya; Kawamura, Satoshi; Kohyama, Akira

2011-10-01

Silicon carbide (SiC) is a candidate for the structural material in the next generation nuclear plants. Use of SiC/SiC composites is expected to increase the operation temperature of system over 1000 °C. For the high temperature system, refractory metals are planned to be used for several components. Tungsten is a candidate of armor on the divertor component in fusion, and is planned to be used for an upper-end plug of SiC/SiC fuel pin in a Gas cooled Fast Reactor (GFR). Joining technique of the SiC/SiC composites and tungsten is an important issue for nuclear systems in future. Nano-Infiltration and Transient Eutectoid (NITE) method is able to provide dense stable and high strength SiC/SiC composites having high resistance against pressure at elevated temperature, a diffusion bonding technique is usable to join the materials. Present research produces a NITE-SiC/SiC composite and tungsten as the similar dimension as a projected cladding tube of fuel pin for GFR using diffusion bonding, and investigated microstructure and mechanical properties.

RIA simulation tests using driver tube for ATF cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cinbiz, Mahmut N.; Brown, N. R.; Lowden, R. R.

Pellet-cladding mechanical interaction (PCMI) is a potential failure mechanism for accident-tolerant fuel (ATF) cladding candidates during a reactivity-initiated accident (RIA). This report summarizes Fiscal Year (FY) 2017 research activities that were undertaken to evaluate the PCMI-like hoop-strain-driven mechanical response of ATF cladding candidates. To achieve various RIA-like conditions, a modified-burst test (MBT) device was developed to produce different mechanical pulses. The calibration of the MBT instrument was accomplished by performing mechanical tests on unirradiated Generation-I iron-chromium-aluminum (FeCrAl) alloy samples. Shakedown tests were also conducted in both FY 2016 and FY 2017 using unirradiated hydrided ZIRLO™ tube samples. This milestone reportmore » focuses on testing of ATF materials, but the benchmark tests with hydrided ZIRLO™ tube samples are documented in a recent journal article.a For the calibration and benchmark tests, the hoop strain was monitored using strain gauges attached to the sample surface in the hoop direction. A novel digital image correlation (DIC) system composed of a single high-speed camera and an array of six mirrors was developed for the MBT instrument to better resolve the failure behavior of samples and to provide useful data for validation of high-fidelity modeling and simulation tools. The DIC system enable a 360° view of a sample’s outer surface. This feature was added to the instrument to determine the precise failure location on a sample’s surface for strain predictions. The DIC system was tested on several silicon carbide fiber/silicon carbide matrix (SiC/SiC) composite tube samples at various pressurization rates of the driver tube (which correspond to the strain rates for the samples). The hoop strains for various loading conditions were determined for the SiC/SiC composite tube samples. Future work is planned to enhance understanding of the failure behavior of the ATF cladding candidates of age-hardened FeCrAl alloys and SiC/SiC composites in detail during RIA conditions informed by the computational studies performed under the US Department of Energy Office of Nuclear Energy Advanced Fuels Campaign. The testing instrument and the new DIC system will be further developed to reach different stress-state conditions and to perform tests at elevated temperatures.« less

Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin

A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature-and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS).The code was validatedmore » using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code. (c) 2018 Elsevier B.V. All rights reserved.« less

Research on the interfacial behaviors of plate-type dispersion nuclear fuel elements

NASA Astrophysics Data System (ADS)

Wang, Qiming; Yan, Xiaoqing; Ding, Shurong; Huo, Yongzhong

2010-04-01

The three-dimensional constitutive relations are constructed, respectively, for the fuel particles, the metal matrix and the cladding of dispersion nuclear fuel elements, allowing for the effects of large deformation and thermal-elastoplasticity. According to the constitutive relations, the method of modeling their irradiation behaviors in ABAQUS is developed and validated. Numerical simulations of the interfacial performances between the fuel meat and the cladding are implemented with the developed finite element models for different micro-structures of the fuel meat. The research results indicate that: (1) the interfacial tensile stresses and shear stresses for some cases will increase with burnup, but the relative stresses will decrease with burnup for some micro-structures; (2) at the lower burnups, the interfacial stresses increase with the particle sizes and the particle volume fractions; however, it is not the case at the higher burnups; (3) the particle distribution characteristics distinctly affect the interfacial stresses, and the face-centered cubic case has the best interfacial performance of the three considered cases.

Thermal Analysis of the Mound One Kilowatt Package

DOE Office of Scientific and Technical Information (OSTI.GOV)

Or, Chuen T.

The Mound One Kilowatt (1 KW) package was designed for the shipment of plutonium (Pu-238) with not more than 1 kW total heat dissipation. To comply with regulations, the Mound 1 kW package has to pass all the requirements under Normal Conditions of Transport (NCT; 38 degrees C ambient temperature) and Hypothetical Accident Conditions (HAC; package engulfed in fire for 30 minutes). Analytical and test results were presented in the Safety Analysis Report for Packaging (SARP) for the Mound 1 kW package, revision 1, April 1991. Some issues remained unresolved in that revision. In March 1992, Fairchild Space and Defensemore » Corporation was commissioned by the Department of Energy to perform the thermal analyses. 3-D thermal models were created to perform the NCT and HAC analyses. Four shipping configurations in the SARP revision 3 were analyzed. They were: (1) The GPHS graphite impact shell (GIS) in the threaded product can (1000 W total heat generation); (2) The fueled clads in the welded product can (1000 W total heat generation); (3) The General Purpose Heat Source (GPHS) module (750 W total heat generation); and (4) The Multi-Hundred Watt (MHW) spheres (810 W total heat generation). Results from the four cases show that the GIS or fuel clad in the product can is the worse case. The temperatures predicted under NCT and HAC in all four cases are within the design limits. The use of helium instead of argon as cover gas provides a bigger safety margin. There is a duplicate copy.« less

Advanced Pellet-Cladding Interaction Modeling using the US DOE CASL Fuel Performance Code: Peregrine

DOE Office of Scientific and Technical Information (OSTI.GOV)

Montgomery, Robert O.; Capps, Nathan A.; Sunderland, Dion J.

The US DOE’s Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermo-mechanical-chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale code thatmore » is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.« less

Advanced Pellet Cladding Interaction Modeling Using the US DOE CASL Fuel Performance Code: Peregrine

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jason Hales; Various

The US DOE’s Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermomechanical- chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale codemore » that is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.« less

Effects of the foil flatness on the stress-strain characteristics of U10Mo alloy based monolithic mini-plates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hakan Ozaltun; Pavel Medvedev

The effects of the foil flatness on stress-strain behavior of monolithic fuel mini-plates during fabrication and irradiation were studied. Monolithic plate-type fuels are a new fuel form being developed for research and test reactors to achieve higher uranium densities. This concept facilitates the use of low-enriched uranium fuel in the reactor. These fuel elements are comprised of a high density, low enrichment, U–Mo alloy based fuel foil encapsulated in a cladding material made of Aluminum. To evaluate the effects of the foil flatness on the stress-strain behavior of the plates during fabrication, irradiation and shutdown stages, a representative plate frommore » RERTR-12 experiments (Plate L1P756) was considered. Both fabrication and irradiation processes of the plate were simulated by using actual irradiation parameters. The simulations were repeated for various foil curvatures to observe the effects of the foil flatness on the peak stress and strain magnitudes of the fuel elements. Results of fabrication simulations revealed that the flatness of the foil does not have a considerable impact on the post fabrication stress-strain fields. Furthermore, the irradiation simulations indicated that any post-fabrication stresses in the foil would be relieved relatively fast in the reactor. While, the perfectly flat foil provided the slightly better mechanical performance, overall difference between the flat-foil case and curved-foil case was not significant. Even though the peak stresses are less affected, the foil curvature has several implications on the strain magnitudes in the cladding. It was observed that with an increasing foil curvature, there is a slight increase in the cladding strains.« less

Atomistic modeling and simulation of the role of Be and Bi in Al diffusion in U-Mo fuel

NASA Astrophysics Data System (ADS)

Hofman, G. L.; Bozzolo, G.; Mosca, H. O.; Yacout, A. M.

2011-07-01

Within the RERTR program, previous experimental and modeling studies identified Si as the alloying addition to the Al cladding responsible for inhibiting Al interdiffusion in the UMo fuel. However, difficulties with reprocessing have rendered this choice inappropriate, leading to the need to study alternative elements. In this work, we discuss the results of an atomistic modeling effort which allows for the systematic study of several possible alloying additions. Based on the behavior observed in the phase diagrams, beryllium or bismuth additions suggest themselves as possible options to replace Si. The results of temperature-dependent simulations using the Bozzolo-Ferrante-Smith (BFS) method for the energetics for varying concentrations of either element are shown, indicating that Be could have a substantial effect in stopping Al interdiffusion, while Bi does not. Details of the calculations and the dependence of the role of each alloying addition as a function of temperature and concentration (of beryllium or bismuth in Al) are shown.

Potentialities of TEC topping: A simplified view of parametric effects

NASA Technical Reports Server (NTRS)

Morris, J. F.

1980-01-01

An examination of the benefits of thermionic-energy-conversion (TEC)-topped power plants and methods of increasing conversion efficiency are discussed. Reductions in the cost of TEC modules yield direct decreases in the cost of electricity (COE) from TEC-topped central station power plants. Simplified COE, overall-efficiency charts presented illustrate this trend. Additional capital-cost diminution results from designing more compact furnaces with considerably increased heat transfer rates allowable and desirable for high temperature TEC and heat pipes. Such improvements can evolve of the protection from hot corrosion and slag as well as the thermal expansion compatibilities offered by silicon-carbide clads on TEC-heating surfaces. Greater efficiencies and far fewer modules are possible with high-temperature, high-power-density TEC: This decreases capital and fuel costs much more and substantially increases electric power outputs for fixed fuel inputs. In addition to more electricity, less pollution, and lower costs, TEC topping used directly in coal-combustion products contributes balance-of-payment gains.

Science based integrated approach to advanced nuclear fuel development - integrated multi-scale multi-physics hierarchical modeling and simulation framework Part III: cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tome, Carlos N; Caro, J A; Lebensohn, R A

2010-01-01

Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Reactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems to develop predictive tools is critical. Not only are fabrication and performance models needed to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating themore » phase and microstructural behavior of the nuclear fuel system materials and matrices. In this paper we review the current status of the advanced modeling and simulation of nuclear reactor cladding, with emphasis on what is available and what is to be developed in each scale of the project, how we propose to pass information from one scale to the next, and what experimental information is required for benchmarking and advancing the modeling at each scale level.« less

Comments on ""Contact Diffusion Interaction of Materials with Cladding''

NASA Technical Reports Server (NTRS)

Morris, J. F.

1972-01-01

A Russian paper by A. A. Babad-Zakhryapina contributes much to the understanding of fuel, clad interactions, and thus to nuclear thermionic technology. In that publication the basic diffusion expression is a simple one. A more general but complicated equation for this mass transport results from the present work. With appropriate assumptions, however, the new relation reduces to Babad-Zakhryapina's version.

Fundamental metallurgical aspects of axial splitting in zircaloy cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chung, H. M.

Fundamental metallurgical aspects of axial splitting in irradiated Zircaloy cladding have been investigated by microstructural characterization and analytical modeling, with emphasis on application of the results to understand high-burnup fuel failure under RIA situations. Optical microscopy, SEM, and TEM were conducted on BWR and PWR fuel cladding tubes that were irradiated to fluence levels of 3.3 x 10{sup 21} n cm{sup {minus}2} to 5.9 x 10{sup 21} n cm{sup {minus}2} (E > 1 MeV) and tested in hot cell at 292--325 C in Ar. The morphology, distribution, and habit planes of macroscopic and microscopic hydrides in as-irradiated and posttest claddingmore » were determined by stereo-TEM. The type and magnitude of the residual stress produced in association with oxide-layer growth and dense hydride precipitation, and several synergistic factors that strongly influence axial-splitting behavior were analyzed. The results of the microstructural characterization and stress analyses were then correlated with axial-splitting behavior of high-burnup PWR cladding reported for simulated-RIA conditions. The effects of key test procedures and their implications for the interpretation of RIA test results are discussed.« less

Status of Wrought FeCrAl-UO 2 Capsules Irradiated in the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G.; Harp, J.; Core, G.

2017-07-01

Candidate cladding materials for accident tolerant fuel applications require extensive testing and validation prior to commercial deployment within the nuclear power industry. One class of cladding materials, FeCrAl alloys, is currently undergoing such effort. Within these activities is a series of irradiation programs within the Advanced Test Reactor. These programs are developed to aid in commercial maturation and understand the fundamental mechanisms controlling the cladding performance during normal operation of a typical light water reactor. Three different irradiation programs are on-going; one designed as a simple proof-of-principle concept, the other to evaluate the susceptibility of FeCrAl to fuel-cladding chemical interaction,more » and the last to fully simulate the conditions of a pressurized water reactor experimentally. To date, nondestructive post-irradiation examination has been completed on the rodlet deemed FCA-L3 from the simple proof-of-concept irradiation program. Initial results show possible breach of the rodlet under irradiation but further studies are needed to conclusively determine whether breach has occurred and the underlying reasons for such a possible failure. Further work includes characterizing additional rodlets following irradiation.« less

DISSOLUTION OF ZIRCONIUM-CONTAINING FUEL ELEMENTS

DOEpatents

Horn, F.L.

1961-12-12

Uranium is recovered from spent uranium fuel elements containing or clad with zirconium. These fuel elements are placed in an anhydrous solution of hydrogen fluoride and nitrogen dioxide. Within this system uranium forms a soluble complex and zirconium forms an insoluble complex. The uranium can then be separated, treated, and removed from solution as uranium hexafluoride. (AEC)

Analysis and Implementation of Accident Tolerant Nuclear Fuels

NASA Astrophysics Data System (ADS)

Prewitt, Benjamin Joseph

To improve the reliability and robustness of LWR, accident tolerant nuclear fuels and cladding materials are being developed to possibly replace the current UO2/zirconium system. This research highlights UN and U3Si 2, two of the most favorable accident tolerant fuels being developed. To evaluate the commercial feasiblilty of these fuels, two areas of research were conducted. Chemical fabrication routes for both fuels were investigated in detail, considering UO2 and UF6 as potential starting materials. Potential pathways for industrial scale fabrication using these methods were discussed. Neutronic performance of 70%UN-30%U3Si2 composite was evaluated in MNCP using PWR assembly and core models. The results showed comparable performance to an identical UO2 fueled simulation with the same configuration. The parameters simulated for composite and oxide fuel include the following: fuel to moderator ratio curves; energy dependent flux spectra; temperature coefficients for fuel and moderator; delayed neutron fractions; power peaking factors; axial and radial flux profiles in 2D and 3D; burnup; critical boron concentration; and shutdown margin. Overall, the neutronic parameters suggest that the transition from UO2 to composite in existing nuclear systems will not require significant changes in operating procedures or modifications to standards and regulations.

Structural and thermodynamic study of dicesium molybdate Cs2Mo2O7: Implications for fast neutron reactors

NASA Astrophysics Data System (ADS)

Smith, A. L.; Kauric, G.; van Eijck, L.; Goubitz, K.; Wallez, G.; Griveau, J.-C.; Colineau, E.; Clavier, N.; Konings, R. J. M.

2017-09-01

The structure of α-Cs2Mo2O7 (monoclinic in space group P21 / c), which can form during irradiation in fast breeder reactors in the space between nuclear fuel and cladding, has been refined in this work at room temperature from neutron diffraction data. Furthermore, the compounds' thermal expansion and polymorphism have been investigated using high temperature X-ray diffraction combined with high temperature Raman spectroscopy. A phase transition has been observed at Ttr(α → β)=(621.9±0.8) K using Differential Scanning Calorimetry, and the structure of the β-Cs2Mo2O7 phase, orthorhombic in space group Pbcm, has been solved ab initio from the high temperature X-ray diffraction data. Furthermore, the low temperature heat capacity of α-Cs2Mo2O7 has been measured in the temperature range T=(1.9-313.2) K using a Quantum Design PPMS (Physical Property Measurement System) calorimeter. The heat capacity and entropy values at T=298.15 K have been derived as Cp,mo (Cs2Mo2O7 , cr , 298.15 K) = (211.9 ± 2.1) J K-1mol-1 and Smo (Cs2Mo2O7 , cr , 298.15 K) = (317.4 ± 4.3) J K-1mol-1 . When combined with the enthalpy of formation reported in the literature, these data yield standard entropy and Gibbs energy of formation as Δf Smo (Cs2Mo2O7 , cr , 298.15 K) = - (628.2 ± 4.4) J K-1mol-1 and Δf Gmo (Cs2Mo2O7 , cr , 298.15 K) = - (2115.1 ± 2.5) kJmol-1 . Finally, the cesium partial pressure expected in the gap between fuel and cladding following the disproportionation reaction 2Cs2MoO4=Cs2Mo2O7+2Cs(g)+ 1/2 O2(g) has been calculated from the newly determined thermodynamic functions.

Molybdenum-UO2 cermet irradiation at 1145 K.

NASA Technical Reports Server (NTRS)

Mcdonald, G.

1971-01-01

Two molybdenum-uranium dioxide cermet fuel pins with molybdenum clad were fission-heated in a forced-convection helium coolant for sufficient time to achieve 5.3% burnup. The cermet core contained 20 wt % of 93.2% enriched uranium dioxide. The results were as follows: there was no visible change in the appearance of the molybdenum clad during irradiation; the maximum increase in diameter of the fuel pins was 0.8%; there was no migration of uranium dioxide along grain boundaries and no evident interaction between molybdenum and uranium dioxide; and, finally, approximately 12% of the fission gas formed was released from the cermet core into the gas plenum.

Improved Accident Tolerance of Austenitic Stainless Steel Cladding through Colossal Supersaturation with Interstitial Solutes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ernst, Frank

We proposed a program-supporting research project in the area of fuel-cycle R&D, specifically on the topic of advanced fuels. Our goal was to investigate whether SECIS (surface engineering by concentrated interstitial solute – carbon, nitrogen) can improve the properties of austenitic stainless steels and related structural alloys such that they can be used for nuclear fuel cladding in LWRs (light-water reactors) and significantly excel currently used alloys with regard to performance, safety, service life, and accident tolerance. We intended to demonstrate that SECIS can be adapted for post-processing of clad tubing to significantly enhance mechanical properties (hardness, wear resistance, andmore » fatigue life), corrosion resistance, resistance to stress–corrosion cracking (hydrogen-induced embrittlement), and – potentially – radiation resistance (against electron-, neutron-, or ion-radiation damage). To test this hypothesis, we measured various relevant properties of the surface-engineered alloys and compared them with corresponding properties of the non–treated, as-received alloys. In particular, we studied the impact of heat exposure corresponding to BWR (boiling-water reactor) working and accident (loss-of-coolant) conditions and the effect of ion irradiation.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hayes, S. L.; Harp, J. M.; Chichester, H. J. M.

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

PROCESS OF DISSOLVING FUEL ELEMENTS OF NUCLEAR REACTORS

DOEpatents

Wall, E.M.V.; Bauer, D.T.; Hahn, H.T.

1963-09-01

A process is described for dissolving stainless-steelor zirconium-clad uranium dioxide fuel elements by immersing the elements in molten lead chloride, adding copper, cuprous chloride, or cupric chloride as a catalyst and passing chlorine through the salt mixture. (AEC)

Critical processes and parameters in the development of accident tolerant fuels drop-in capsule irradiation tests

DOE PAGES

Barrett, K. E.; Ellis, K. D.; Glass, C. R.; ...

2015-12-01

The goal of the Accident Tolerant Fuel (ATF) program is to develop the next generation of Light Water Reactor (LWR) fuels with improved performance, reliability, and safety characteristics during normal operations and accident conditions and with reduced waste generation. An irradiation test series has been defined to assess the performance of proposed ATF concepts under normal LWR operating conditions. The Phase I ATF irradiation test series is planned to be performed as a series of drop-in capsule tests to be irradiated in the Advanced Test Reactor (ATR) operated by the Idaho National Laboratory (INL). Design, analysis, and fabrication processes formore » ATR drop-in capsule experiment preparation are presented in this paper to demonstrate the importance of special design considerations, parameter sensitivity analysis, and precise fabrication and inspection techniques for figure innovative materials used in ATF experiment assemblies. A Taylor Series Method sensitivity analysis approach was used to identify the most critical variables in cladding and rodlet stress, temperature, and pressure calculations for design analyses. The results showed that internal rodlet pressure calculations are most sensitive to the fission gas release rate uncertainty while temperature calculations are most sensitive to cladding I.D. and O.D. dimensional uncertainty. The analysis showed that stress calculations are most sensitive to rodlet internal pressure uncertainties, however the results also indicated that the inside radius, outside radius, and internal pressure were all magnified as they propagate through the stress equation. This study demonstrates the importance for ATF concept development teams to provide the fabricators as much information as possible about the material properties and behavior observed in prototype testing, mock-up fabrication and assembly, and chemical and mechanical testing of the materials that may have been performed in the concept development phase. Special handling, machining, welding, and inspection of materials, if known, should also be communicated to the experiment fabrication and inspection team.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong; Jiang, Hao

The objective of this research is to collect dynamic experimental data on spent nuclear fuel (SNF) under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT), the hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). The collected CIRFT data will be utilized to support ongoing spent fuel modeling activities, and support SNF transportation related licensing issues. Recent testing to understand the effects of hydride reorientation on SNF vibration integrity is also being evaluated. CIRFT results have provided insight into the fuel/clad system response to transportation related loads. The major findings of CIRFT on the HBU SNFmore » are as follows: SNF system interface bonding plays an important role in SNF vibration performance, Fuel structure contributes to the SNF system stiffness, There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interaction, and SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous. Because of the non-homogeneous composite structure of the SNF system, finite element analyses (FEA) are needed to translate the global moment-curvature measurement into local stress-strain profiles. The detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained directly from a CIRFT system measurement. Therefore, detailed FEA is used to understand the global test response, and that data will also be presented.« less

Method of fission heat flux determination from experimental data

DOEpatents

Paxton, Frank A.

1999-01-01

A method is provided for determining the fission heat flux of a prime specimen inserted into a specimen of a test reactor. A pair of thermocouple test specimens are positioned at the same level in the holder and a determination is made of various experimental data including the temperature of the thermocouple test specimens, the temperature of bulk water channels located in the test holder, the gamma scan count ratios for the thermocouple test specimens and the prime specimen, and the thicknesses of the outer clads, the fuel fillers, and the backclad of the thermocouple test specimen. Using this experimental data, the absolute value of the fission heat flux for the thermocouple test specimens and prime specimen can be calculated.

LOW TEMPERATURE PROCESS FOR THE REMOVAL AND RECOVERY OF CHLORIDES AND NITRATES FROM AQUEOUS NITRATE SOLUTIONS

DOEpatents

Savolainen, J.E.

1963-01-29

A method is described for reducing the chloride content of a solution derived from the dissolution of a stainless steel clad nuclear fuel element with an aqua regia dissolution medium. The solutlon is adjusted to a nitric acid concentration in the range 5 to 10 M and is countercurrently contacted at room temperature with a gaseous oxide of nitrogen selected from NO, NO/sub 2/, N/sub 2/ O/sub 3/, and N/sub 2/O/sub 4/. Chlo ride is recovered from the contacted solution as nitrosyl chloride. After reduction of the chloride content, the solution is then contacted with gaseous NO to reduce the nitric acid molarity to a desired level. (AEC)

Effect of nitrogen on tensile properties and structures of T-111 (tantalum, 8 percent tungsten, 2 percent hafnium) tubing

NASA Technical Reports Server (NTRS)

Buzzard, R. J.; Metroka, R. R.

1973-01-01

The effect of controlled nitrogen additions was evaluated on the mechanical properties of T-111 (Ta-8W-2Hf) fuel pin cladding material proposed for use in a lithium-cooled nuclear reactor concept. Additions of 80 to 1125 ppm nitrogen resulted in increased strengthening of T-111 tubular section test specimens at temperatures of 25 to 1200 C. Homogeneous distributions of up to 500 ppm nitrogen did not seriously decrease tensile ductility. Both single and two-phase microstructures, with hafnium nitride as the second phase, were evaluated in this study.

The chemical state of defective uranium-plutonium oxide fuel pins irradiated in sodium cooled reactors

NASA Astrophysics Data System (ADS)

Kleykamp, H.

1997-09-01

Steady-state irradiation experiments were conducted in the sodium loop of the Siloe reactor on artificially failed mixed oxide pins that had been pre-irradiated in fast reactors up to 11.5% burnup. The formation of the predominant reaction product Na 3(U,Pu)O 4 starts on the fuel surface and is terminated when a lower O/(U + Pu) threshold of the fuel is attained. The axial extent of the reaction product depends on the size of the initial cladding defect. The occurrence of secondary cracks is possible. Na(U,Pu)O 3 forms at higher fuel temperatures. The existence of Na 3U 1- xPu xO 4 is shown in pre-irradiated blanket pins after artificial defect formation. Caesium in the oxocompounds is reduced to the metallic state and is dissolved in the coolant. Evidence of a very low chemical potential of oxygen in defective fuel pins is sustained by the occurrence of actinide-platinum metal phases formed by coupled reduction of hypostoichiometric fuel with ɛ-(Mo,Tc,Ru,Rh,Pd) precipitates. Continued operation of defective pins is not hazardous by easy precautions.

TRACE/PARCS Analysis of ATWS with Instability for a MELLLA+BWR/5

DOE PAGES

L. Y. Cheng; Baek, J. S.; Cuadra, A.; ...

2016-06-06

A TRACE/PARCS model has been developed to analyze anticipated transient without SCRAM (ATWS) events for a boiling water reactor (BWR) operating in the maximum extended load line limit analysis-plus (MELLLA+) expanded operating domain. The MELLLA+ domain expands allowable operation in the power/flow map of a BWR to low flow rates at high power conditions. Such operation exacerbates the likelihood of large amplitude power/flow oscillations during certain ATWS scenarios. The analysis shows that large amplitude power/flow oscillations, both core-wide and out-of-phase, arise following the establishment of natural circulation flow in the reactor pressure vessel (RPV) after the trip of the recirculationmore » pumps and an increase in core inlet subcooling. The analysis also indicates a mechanism by which the fuel may experience heat-up that could result in localized fuel damage. TRACE predicts the heat-up to occur when the cladding surface temperature exceeds the minimum stable film boiling temperature after periodic cycles of dryout and rewet; and the fuel becomes “locked” into a film boiling regime. Further, the analysis demonstrates the effectiveness of the simulated manual operator actions to suppress the instability.« less

High Sensitive Temperature Sensor Using a Liquid-core Optical Fiber with Small Refractive Index Difference Between Core and Cladding Materials.

PubMed

Xu, Yonghao; Chen, Xianfeng; Zhu, Yu

2008-03-17

An intensive temperature sensor based on a liquid-core optical fiber has been demonstrated for the measuring the temperature of the environment. The core of fiber is filled with a mixture of toluene and chloroform in order to make the refractive index of the liquid-core and the cladding of the fiber close. The experiment shows that a temperature sensitivity of about 5 dB/K and a tunable temperature range (from 20 o C to 60 o C) can be achieved. Based on the dielectric-clad liquid core fiber model, a simulation was carried out and the calculated results were in good accord with the experimental measurement.

Method and apparatus for determining peak temperature along an optical fiber

DOEpatents

Fox, R.J.

1982-07-29

The invention relates to a new method and new apparatus for determining the hottest temperature or the coldest temperature prevailing along the length of an optical-fiber light guide. The invention is conducted with an optical fiber capable of supporting multidiode propagation of light and comprising a core, a cladding, and a jacket. The core is selected to have (1) a higher refractive index than the core and the cladding and (2) a relatively high negative temperature coefficient of refractive index. A light beam capable of establishing substantially single-mode propagation in the core is launched into an end thereof at an angle to the axis. The angle is increased to effect the onset of light fraction from the core into the cladding. The value of the launch angle corresponding to the onset is determined and then used to establish the refractive index of the core corresponding to the onset angle. The maximum temperature prevailing along the fiber then is determined from the (1) refractive index so determined and (2) the temperature coefficient of refractive index for the core. The invention is based on the finding that the launch angle corresponding to the onset of refraction into the cladding is uniquely determined by the maximum value of the ratio of the core refractive index to the cladding refractive index, which maximum occurs at the hottest point along the fiber.

Method and apparatus for determining peak temperature along an optical fiber

DOEpatents

Fox, Richard J.

1985-01-01

The invention relates to a new method and new apparatus for determining the hottest temperature or the coldest temperature prevailing along the length of an optical-fiber light guide. The invention is conducted with an optical fiber capable of supporting multidiode propagation of light and comprising a core, a cladding, and a jacket. The core is selected to have (1) a higher refractive index than the core and the cladding and (2) a relatively high negative temperature coefficient of refractive index. A light beam capable of establishing substantially single-mode propagation in the core is launched into an end thereof at an angle to the axis. The angle is increased to effect the onset of light refraction from the core into the cladding. The value of the launch angle corresponding to the onset is determined and then used to establish the refractive index of the core corresponding to the onset angle. The maximum temperature prevailing along the fiber then is determined from the (1) refractive index so determined and (2) the temperature coefficient of refractive index for the core. The invention is based on the finding that the launch angle corresponding to the onset of refraction into the cladding is uniquely determined by the maximum value of the ratio of the core refractive index to the cladding refractive index, which maximum occurs at the hottest point along the fiber.

Accident-tolerant oxide fuel and cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mariani, Robert D.

Systems and methods for accident tolerant oxide fuel. One or more disks can be placed between fuel pellets comprising UO.sub.2, wherein such disks possess a higher thermal conductivity material than that of the UO.sub.2 to provide enhanced heat rejection thereof. Additionally, a cladding coating comprising zircaloy coated with a material that provides stability and high melting capability can be provided. The pellets can be configured as annular pellets having an annulus filled with the higher thermal conductivity material. The material coating the zircaloy can be, for example, Zr.sub.5Si.sub.4 or another silicide such as, for example, a Zr-Silicide that limits corrosion.more » The aforementioned higher thermal conductivity material can be, for example, Si, Zr.sub.xSi.sub.y, Zr, or Al.sub.2O.sub.3.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Carmack; L. Braase; F. Goldner

The mission of the Advanced Fuels Campaign (AFC) is to perform Research, Development, and Demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors, enhance proliferation resistance of nuclear fuel, effectively utilize nuclear energy resources, and address the longer-term waste management challenges. This includes development of a state of the art Research and Development (R&D) infrastructure to support the use of a “goal oriented science based approach.” AFC uses a “goal oriented, science based approach” aimed at a fundamental understanding of fuel and cladding fabrication methods and performancemore » under irradiation, enabling the pursuit of multiple fuel forms for future fuel cycle options. This approach includes fundamental experiments, theory, and advanced modeling and simulation. One of the most challenging aspects of AFC is the management, integration, and coordination of major R&D activities across multiple organizations. AFC interfaces and collaborates with Fuel Cycle Technologies (FCT) campaigns, universities, industry, various DOE programs and laboratories, federal agencies (e.g., Nuclear Regulatory Commission [NRC]), and international organizations. Key challenges are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Challenged with the research and development of fuels for two different reactor technology platforms, AFC targeted transmutation fuel development and focused ceramic fuel development for Advanced LWR Fuels.« less

Laser Cladding of TiAl Intermetallic Alloy on Ti6Al4V -Process Optimization and Properties

NASA Astrophysics Data System (ADS)

Cárcel, B.; Serrano, A.; Zambrano, J.; Amigó, V.; Cárcel, A. C.

In order to improve Ti6Al4V high-temperature resistance and its tribological properties, the deposition of TiAl intermetallic (Ti-48Al-2Cr-2Nb) coating on a Ti6Al4V substrate by coaxial laser cladding has been investigated. Laser cladding by powder injection is an emerging laser material processing technique that allows the deposition of thick protective coatings on substrates,using a high power laser beam as heat source. Laser cladding is a multiple-parameter-dependent process. The main process parameters involved (laser power, powder feeding rate, scanning speed and preheating temperature) has been optimized. The microstructure and geometrical quantities (clad area and dilution) of the coating was characterized by optical microscopy and scanning electron microscopy (SEM). In addition the cooling rate of the clad during the process was measured by a dual-color pyrometer. This result has been related to defectology and mechanical coating properties.

An analytical model to predict and minimize the residual stress of laser cladding process

NASA Astrophysics Data System (ADS)

Tamanna, N.; Crouch, R.; Kabir, I. R.; Naher, S.

2018-02-01

Laser cladding is one of the advanced thermal techniques used to repair or modify the surface properties of high-value components such as tools, military and aerospace parts. Unfortunately, tensile residual stresses generate in the thermally treated area of this process. This work focuses on to investigate the key factors for the formation of tensile residual stress and how to minimize it in the clad when using dissimilar substrate and clad materials. To predict the tensile residual stress, a one-dimensional analytical model has been adopted. Four cladding materials (Al2O3, TiC, TiO2, ZrO2) on the H13 tool steel substrate and a range of preheating temperatures of the substrate, from 300 to 1200 K, have been investigated. Thermal strain and Young's modulus are found to be the key factors of formation of tensile residual stresses. Additionally, it is found that using a preheating temperature of the substrate immediately before laser cladding showed the reduction of residual stress.

Industry Application Emergency Core Cooling System Cladding Acceptance Criteria Early Demonstration

DOE Office of Scientific and Technical Information (OSTI.GOV)

Szilard, Ronaldo H.; Youngblood, Robert W.; Zhang, Hongbin

2015-09-01

The U. S. NRC is currently proposing rulemaking designated as “10 CFR 50.46c” to revise the loss-of-coolant-accident (LOCA)/emergency core cooling system (ECCS) acceptance criteria to include the effects of higher burnup on cladding performance as well as to address other technical issues. The NRC is also currently resolving the public comments with the final rule expected to be issued in April 2016. The impact of the final 50.46c rule on the industry may involve updating of fuel vendor LOCA evaluation models, NRC review and approval, and licensee submittal of new LOCA evaluations or re-analyses and associated technical specification revisions formore » NRC review and approval. The rule implementation process, both industry and NRC activities, is expected to take 4-6 years following the rule effective date. As motivated by the new rule, the need to use advanced cladding designs may be a result. A loss of operational margin may result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently, there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin. The proposed rule would apply to a light water reactor and to all cladding types.« less

Fuel rod with annular nuclear fuel pellets having same U-235 enrichment and different annulus sizes for graduated enrichment loading

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mildrum, C.M.

1987-08-18

A fuel rod is described for a nuclear reactor fuel assembly, comprising: (a) a hollow cladding tube; (b) a pair of end plugs connected to and sealing the cladding tube at opposite ends thereof; (c) a plurality of fuel pellets contained on the tube and being composed of fissile material having a single enrichment the value of which is at the level of the maximum enrichment loading of the rod, the pellets having provided in a stack having one end disposed adjacent to one of the end plugs and an opposite end disposed remote from the other of the endmore » plugs; and (d) a plenum spring disposed in the tube between the other end plug and the opposite end of the pellet stack for retaining the pellets in a stack form; (e) at least some of the fuel pellets having an annular configuration and at least other of the fuel pellets having a solid configuration; (f) each of some of the annular fuel pellets having an annulus of a first size; (e) each of other of the annual fuel pellets having an annulus of a second size different from the first size, whereby graduation of axial enrichment loading is provided between the annual fuel pellets of the fuel rod.« less

Results for the Aboveground Configuration of the Boiling Water Reactor Dry Cask Simulator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Durbin, Samuel G.; Lindgren, Eric Richard

The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and also by increasing themore » internal convection through greater canister helium pressure. These same canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and belowground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. The purpose of the current investigation was to produce data sets that can be used to test the validity of the assumptions associated with the calculations used to determine steady-state cladding temperatures in modern dry casks that utilize elevated helium pressure in the sealed canister in an aboveground configuration. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly was deployed inside of a representative storage basket and cylindrical pressure vessel that represents a vertical canister system. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. The arrangement of ducting was used to mimic conditions for an aboveground storage configuration in a vertical, dry cask systems with canisters. Transverse and axial temperature profiles were measured for a wide range of decay power and helium cask pressures. Of particular interest was the evaluation of the effect of increased helium pressure on peak cladding temperatures (PCTs) for identical thermal loads. All steady state peak temperatures and induced flow rates increased with increasing assembly power. Peak cladding temperatures decreased with increasing internal helium pressure for a given assembly power, indicating increased internal convection. In addition, the location of the PCT moved from near the top of the assembly to ~1/3 the height of the assembly for the highest (8 bar absolute) to the lowest (0 bar absolute) pressure studied, respectively. This shift in PCT location is consistent with the varying contribution of convective heat transfer proportional with of internal helium pressure.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aleman, S.E.

The task was to compute the maximum aluminum-oxide and oxide-coolant temperatures of assemblies cladded in 99+ percent aluminum. The assemblies considered were the Mark VB, VE, V5, 15 and 25. These assemblies consist of nested slug columns with individual uranium slugs cladded in aluminum cans. The CREDIT code was modified to calculate the oxide film thickness and the aluminum-oxide temperature at each axial increment. This information in this report will be used to evaluate the potential for cladding corrosion of the Mark 25 assembly.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aleman, S.E.

The task was to compute the maximum aluminum-oxide and oxide-coolant temperatures of assemblies cladded in 99 plus percent aluminum. The assemblies considered were the Mark VB, VE, V5, 15 and 25. These assemblies consist of nested slug columns with individual uranium slugs cladded in aluminum cans. The CREDIT code was modified to calculate the oxide film thickness and the aluminum-oxide temperature at each axial increment. The information in this report will be used to evaluate the potential for cladding corrosion of the Mark 25 assembly.

Status Report on Activities of the Systems Assessment Task Force, OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bragg-Sitton, Shannon Michelle

The Organization for Economic Cooperation and Development /Nuclear Energy Agency (OECD/NEA) Nuclear Science Committee approved the formation of an Expert Group on Accident Tolerant Fuel (ATF) for LWRs (EGATFL) in 2014. Chaired by Kemal Pasamehmetoglu, INL Associate Laboratory Director for Nuclear Science and Technology, the mandate for the EGATFL defines work under three task forces: (1) Systems Assessment, (2) Cladding and Core Materials, and (3) Fuel Concepts. Scope for the Systems Assessment task force includes definition of evaluation metrics for ATF, technology readiness level definition, definition of illustrative scenarios for ATF evaluation, parametric studies, and selection of system codes. Themore » Cladding and Core Materials and Fuel Concepts task forces will identify gaps and needs for modeling and experimental demonstration; define key properties of interest; identify the data necessary to perform concept evaluation under normal conditions and illustrative scenarios; identify available infrastructure (internationally) to support experimental needs; and make recommendations on priorities. Where possible, considering proprietary and other export restrictions (e.g., International Traffic in Arms Regulations), the Expert Group will facilitate the sharing of data and lessons learned across the international group membership. The Systems Assessment Task Force is chaired by Shannon Bragg-Sitton (INL), while the Cladding Task Force will be chaired by a representative from France (Marie Moatti, Electricite de France [EdF]) and the Fuels Task Force will be chaired by a representative from Japan (Masaki Kurata, Japan Atomic Energy Agency [JAEA]). This report provides an overview of the Systems Assessment Task Force charter and status of work accomplishment.« less

Assessment of MARMOT. A Mesoscale Fuel Performance Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tonks, M. R.; Schwen, D.; Zhang, Y.

2015-04-01

MARMOT is the mesoscale fuel performance code under development as part of the US DOE Nuclear Energy Advanced Modeling and Simulation Program. In this report, we provide a high level summary of MARMOT, its capabilities, and its current state of validation. The purpose of MARMOT is to predict the coevolution of microstructure and material properties of nuclear fuel and cladding. It accomplished this using the phase field method coupled to solid mechanics and heat conduction. MARMOT is based on the Multiphysics Object-Oriented Simulation Environment (MOOSE), and much of its basic capability in the areas of the phase field method, mechanics,more » and heat conduction come directly from MOOSE modules. However, additional capability specific to fuel and cladding is available in MARMOT. While some validation of MARMOT has been completed in the areas of fission gas behavior and grain growth, much more validation needs to be conducted. However, new mesoscale data needs to be obtained in order to complete this validation.« less

Preliminary engineering design of sodium-cooled CANDLE core

NASA Astrophysics Data System (ADS)

Takaki, Naoyuki; Namekawa, Azuma; Yoda, Tomoyuki; Mizutani, Akihiko; Sekimoto, Hiroshi

2012-06-01

The CANDLE burning process is characterized by the autonomous shifting of burning region with constant reactivity and constant spacial power distribution. Evaluations of such critical burning process by using widely used neutron diffusion and burning codes under some realistic engineering constraints are valuable to confirm the technical feasibility of the CANDLE concept and to put the idea into concrete core design. In the first part of this paper, it is discussed that whether the sustainable and stable CANDLE burning process can be reproduced even by using conventional core analysis tools such as SLAROM and CITATION-FBR. As a result, it is certainly possible to demonstrate it if the proper core configuration and initial fuel composition required as CANDLE core are applied to the analysis. In the latter part, an example of a concrete image of sodium cooled, metal fuel, 2000MWt rating CANDLE core has been presented by assuming an emerging inevitable technology of recladding. The core satisfies engineering design criteria including cladding temperature, pressure drop, linear heat rate, and cumulative damage fraction (CDF) of cladding, fast neutron fluence and sodium void reactivity which are defined in the Japanese FBR design project. It can be concluded that it is feasible to design CADLE core by using conventional codes while satisfying some realistic engineering design constraints assuming that recladding at certain time interval is technically feasible.

Effect of hydrogenation conditions on the microstructure and mechanical properties of zirconium hydride

NASA Astrophysics Data System (ADS)

Muta, Hiroaki; Nishikane, Ryoji; Ando, Yusuke; Matsunaga, Junji; Sakamoto, Kan; Harjo, Stefanus; Kawasaki, Takuro; Ohishi, Yuji; Kurosaki, Ken; Yamanaka, Shinsuke

2018-03-01

Precipitation of brittle zirconium hydrides deteriorate the fracture toughness of the fuel cladding tubes of light water reactor. Although the hydride embrittlement has been studied extensively, little is known about physical properties of the hydride due to the experimental difficulties. In the present study, to elucidate relationship between mechanical properties and microstructure, two δ-phase zirconium hydrides and one ε-phase zirconium hydride were carefully fabricated considering volume changes at the metal-to-hydride transformation. The δ-hydride that was fabricated from α-zirconium exhibits numerous inner cracks due to the large volume change. Analyses of the neutron diffraction pattern and electron backscatter diffraction (EBSD) data show that the sample displays significant stacking faults in the {111} plane and in the pseudo-layered microstructure. On the other hand, the δ-hydride sample fabricated from β-zirconium at a higher temperature displays equiaxed grains and no cracks. The strong crystal orientation dependence of mechanical properties were confirmed by indentation test and EBSD observation. The δ-hydride hydrogenated from α-zirconium displays a lower Young's modulus than that prepared from β-zirconium. The difference is attributed to stacking faults within the {111} plane, for which the Young's modulus exhibits the highest value in the perpendicular direction. The strong influence of the crystal orientation and dislocation density on the mechanical properties should be considered when evaluating hydride precipitates in nuclear fuel cladding.

Advanced Fuels Campaign FY 2015 Accomplishments Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Braase, Lori Ann; Carmack, William Jonathan

2015-10-29

The mission of the Advanced Fuels Campaign (AFC) is to perform research, development, and demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors; enhance proliferation resistance of nuclear fuel; effectively utilize nuclear energy resources; and address the longer-term waste management challenges. This report is a compilation of technical accomplishment summaries for FY-15. Emphasis is on advanced accident-tolerant LWR fuel systems, advanced transmutation fuels technologies, and capability development.

LWRS ATR Irradiation Testing Readiness Status

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kristine Barrett

2012-09-01

The Light Water Reactor Sustainability (LWRS) Program was established by the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors. The LWRS Program is divided into four R&D Pathways: (1) Materials Aging and Degradation; (2) Advanced Light Water Reactor Nuclear Fuels; (3) Advanced Instrumentation, Information and Control Systems; and (4) Risk-Informed Safety Margin Characterization. This report describes an irradiation testing readiness analysis in preparation of LWRS experiments for irradiation testing at the Idaho National Laboratory (INL) Advanced Testmore » Reactor (ATR) under Pathway (2). The focus of the Advanced LWR Nuclear Fuels Pathway is to improve the scientific knowledge basis for understanding and predicting fundamental performance of advanced nuclear fuel and cladding in nuclear power plants during both nominal and off-nominal conditions. This information will be applied in the design and development of high-performance, high burn-up fuels with improved safety, cladding integrity, and improved nuclear fuel cycle economics« less