Simplified analysis of nuclear fuel pin swelling
Effect of fuel swelling on strains in cladding of cylindrical nuclear fuel elements
Lietzke, A. F.↗
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Theoretical analysis of swelling characteristics of cylindrical uranium dioxide fuel pins with a niobium - 1-percent-zirconium clad
The relations between clad creep strain and fuel volume swelling are shown for cylindrical UO2 fuel pins with a Nb-1Zr clad. These relations were obtained by using the computer code CYGRO-2. These clad-strain - fuel-volume-swelling relations may be used with any fuel-volume-swelling model, provided the fuel volume swelling is isotropic and independent of the clad restraints. The effects of clad temperature (over a range from 118 to 1642 K (2010 to 2960 R)), pin diameter, clad thickness and central hole size in the fuel have been investigated. In all calculations the irradiation time was 500 hours. The burnup rate was varied.
Saltsman, J. F.↗
Testing of uranium nitride fuel in T-111 cladding at 1200 K cladding temperature
Two groups of six fuel pins each were assembled, encapsulated, and irradiated in the Plum Brook Reactor. The fuel pins employed uranium mononitride (UN) in a tantalum alloy clad. The first group of fuel pins was irradiated for 1500 hours to a maximum burnup of 0.7-atom-percent uranium. The second group of fuel pins was irradiated for about 3000 hours to a maximum burnup of 1.0-atom-percent uranium. The average clad surface temperature during irradiation of both groups of fuel pins was approximately 1200 K. The postirradiation examination revealed the following: no clad failures or fuel swelling occurred; less than 1 percent of the fission gases escaped from the fuel; and the clad of the first group of fuel pins experienced clad embrittlement whereas the second group, which had modified assembly and fabrication procedures to minimize contamination, had a ductile clad after irradiation.
Rohal, R. G.↗
EXFILE: A program for compiling irradiation data on UN and UC fuel pins
A FORTRAN-4 computer program for handling fuel pin data is described. Its main features include standardized output, easy access for data manipulation, and tabulation of important material property data. An additional feature allows simplified preparation of input decks for a fuel swelling computer code (CYGRO-2). Data from over 300 high temperature nitride and carbide based fuel pin irradiations are listed.
Mayer, J. T.↗
Examination of T-111 clad uranium nitride fuel pins irradiated up to 13,000 hours at a clad temperature of 990 C
The examination of 27 fuel pins irradiated for up to 13,000 hours at 990 C is described. The fuel pin clad was a tantalum alloy with uranium nitride as the nuclear fuel. Two nominal fuel pin diameters were tested with a maximum burnup of 2.34 atom percent. Twenty-two fuel pins were tested for fission gas leaks; thirteen pins leaked. Clad ductility tests indicated clad embrittlement. The embrittlement is attributed to hydrogen from an n,p reaction in the fuel. Fuel swelling was burnup dependent, and the amount of fission gas release was low, generally less than 0.5 percent. No incompatibilities between fuel, liner, and clad were in evidence.
Slaby, J. G.↗
Experimental evaluation of thermal ratcheting behavior in UO2 fuel elements
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.
Phillips, W. M.↗
Goals of thermionic program for space power
The thermionic and Brayton reactor concepts were compared for application to space power. For a turbine inlet temperature of 15000 K the Brayton powerplant weighted 5 to 40% less than the thermionic concept. The out of core concept separates the thermionic converters from their reactor. Technical risks are diminished by: (1) moving the insolator out of the reactor; (2) allowing a higher thermal flux for the thermionic converters than is required of the reactor fuel; and (3) eliminating fuel swelling's threat against lifetime of the thermionic converters. Overall performance can be improved by including power processing in system optimization for design and technology on more efficient, higher temperature power processors. The thermionic reactors will be larger than those for competitive systems with higher conversion efficiency and lower reactor operating temperatures. It is concluded that although the effect of reactor size on shield weight will be modest for unmanned spacecraft, the penalty in shield weight will be large for manned or man-tended spacecraft.
English, R. E.↗
Fuels irradiation testing for the SP-100 program
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.
Makenas, Bruce J.↗
Calculation of radiation induced swelling of uranium mononitride using the digital computer program CYGRO 2
Fuel volume swelling and clad diametral creep strains were calculated for five fuel pins, clad with either T-111 (Ta-8W-2.4Hf) or PWC-11 (Nb-1Zr-0.1C). The fuel pins were irradiated to burnups between 2.7 and 4.6%. Clad temperatures were between 1750 and 2400 F (1228 and 1589 K). The maximum percentage difference between calculated and experimentally measured values of volumetric fuel swelling is 60%.
Davison, H. W.↗
Unrestrained swelling of uranium-nitride fuel irradiated at temperatures ranging from 1100 to 1400 K (1980 to 2520 R)
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.
Rohal, R. G.↗
Irradiation of TZM: Uranium dioxide fuel pin at 1700 K
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.
Mcdonald, G. E.↗
Mathematical Modeling Of A Nuclear/Thermionic Power Source
Report discusses mathematical modeling to predict performance and lifetime of spacecraft power source that is integrated combination of nuclear-fission reactor and thermionic converters. Details of nuclear reaction, thermal conditions in core, and thermionic performance combined with model of swelling of fuel.
Vandersande, Jan W.↗
Molybdenum-UO2 cerment irradiation at 1145 K
Two molybdenum-UO2 cermet fuel pins were fission heated in a helium-cooled loop at a temperature of 1145 K and to a total burnup of 5.3 % of the U-235. After irradiation the fuel pins were measured to check dimensional stability, punctured at the plenums to determine fission gas release, and examined metallographically to determine the effect of irradiation. Burnup was determined in several sections of the fuel pin. The results of the postirradiation examination indicated: (1) There was no visible change in the fuel pins on irradiation under the above conditions. (2) The maximum swelling of the fuel pins was less than 1%. (3) There was no migration of UO2 and no visible interaction between the molybdenum and the UO2. (4) Approximately 12% of the fission gas formed was released from the cermet cone into the gas plenum.
Mcdonald, G.↗
Irradiation of three T-111 clad uranium nitride fuel pins for 8070 hours at 990 C (1815 F)
The design and successful operation of three tantalum alloy (Ta-8W-2Hf) clad uranium mononitride (UN) fuel pins irradiated for 8070 hr at 990 C (1815 F) is described. Two pin diameters having measured burnups of 0.47 and 0.90 uranium atom percent were tested. No clad failures or swelling was detected; however, postirradiation clad samples tested failed with 1 percent strain. The fuel density decrease was 2 percent, and the fission gas release was less than 0.05 percent. Isotropic fuel swelling, which averaged about 0.5 percent, was less than fuel pin assembly clearances. Thus the clad was not strained. Thermocouples with a modified hot zone operated at average temperatures to 1100 C (2012 F) without failure. Factors that influence the ability to maintain uniform clad temperature as well as the results of the heat transfer calculations are discussed.
Slaby, J. G.↗
A Rapid Survey of the Compatibility of Selected Seal Materials with Conventional and Semi-Synthetic JP-8
Since the synthesis of a liquid hydrocarbon fuel from coal by Franz Fischer and Hans Tropsch in 1923, there has been cyclic interest in developing this fuel for military and commercial applications. In recent years the U.S. Department of Defense has taken interest in producing a unified battlespace fuel using the Fischer Tropsch (FT) process for a variety of reasons including cost, quality, and logistics. In the past year there has been a particular emphasis on moving quickly to demonstrate that an FT fuel can be used in the form of a blend with conventional petroleum-derived jet fuel. The initial objective is to employ this semi-synthetic fuel with blend ratios as high as 50 percent FT with longer range goals to use even high blend ratios and ultimately a fully synthetic jet fuel. A significant concern associated with the use of a semi-synthetic jet fuel with high FT blend ratios is the effect these low aromatic fuels will have on fuel-wetted polymeric materials, most notably seals and sealants. These materials typically swell and soften to some degree when exposed to jet fuel and the aromatic content of these fuels contribute to this effect. Semi-synthetic jet fuels with very low aromatic contents may cause seals and sealants to shrink and harden leading to acute or chronic failure. Unfortunately, most of the material qualification tests are more concerned with excessive swelling than shrinkage and there is little guidance offered as to an acceptable level of shrinkage or other changes in physical properties related to low aromatic content. Given the pressing need for guidance data, a program was developed to rapidly survey the volume swell of selected fuel-wetted materials in a range of conventional and semi-synthetic jet fuels and through a statistical analysis to make a determination as to whether there was a basis to be concerned about using fuels with FT blend ratios as high as 50 percent. Concurrent with this analysis data was obtained as to the composition of the fuel absorbed in fuel-wetted materials through the use of GC-MS analysis of swollen samples as well as other supporting data. In this presentation the authors will present a summary of the results of the volume swell and fuel absorbed by selected O-rings and sealants as well as a description of the measurement protocols developed for this program.
Graham, John L.↗
New RTM/RI Resins for the HSCT
In the first portion of this work, 1,2,3,3,4,4-hexafluoro-1,2-bis[4-(dimethylhydroxysilyl)phenoxy]cyclobutane and 1,2,3,3,4,4-hexafluoro-1,2-bis[3-(dimethylhydroxysilyl)phenoxy]cyclobutane were prepared and homopolymerized to afford polymers with excellent thermal stability and Tgs of 27 C and -12 C, respectively. Despite the moderately high wt% of fluorin in the polymer structure (23.8%), these polymers had poor fuel resistance. In fact, swelling measurements indicate that these polymers had apparent solubility parameters of about 18.2 J (exp 1/2) m (exp -3/2) (toluene). Copolymerization of the disilanol monomers with fluorosilicone monomers afforded copolymers containing 20-30 wt% of the perfluorocyclobutane-containing structure displayed adequate fuel resistance, enhanced thermal stability, and a Tg low enough to meet the requirements of a High Speed Civil Transport (HSCT) fuel tank sealant. In the second part of this work, trifluorovinylether-terminated oligomers were prepared and polymerized via cyclodimerization. Initially, an alpha, omega-silanol-terminated fluorosilicone was endcapped with trifluorovinylether end groups via a two-step synthetic sequence. The oligomer was thermally cyclodimerized to a polymer that displayed thermal stability similar to that of a fluorosilicone homopolymer. Second, 1,3-bis[4-trifluorovinyl(oxy)phenyl]-1,3-(3,3,3-trifluoropropyl)dimethyldisiloxane and 1,3-bis{3-trifluorovinyl(oxy)phenyl]-1,3-(3,3,3-trifluoropropyl)dimethyldisiloxane were prepared and cyclodimerized to afford polymers that contained pendant trifluoropropyl groups. The pendant trifluoropropyl groups did enhance solvent resistance in aliphatic hydrocarbon solvents, however, no improvement was observed in aromatic hydrocarbon solvents. These polymers also displayed excellent thermal stability. In the last part of this work, a series of monomers was prepared by the DCC-promoted esterification of 4-[trifluorovinyl(oxy)benzoic acid with alpha, omega-functionalized hydrogenated and partially fluorinated alcohols. The monomers were cyclodimerized to the corresponding polymers. The polymers that did not contain beta hydrogens displayed significantly higher thermal stability than the fully hydrogenated polymers. A commercially-available alpha,omega-hydroxy-terminated perfluoropolyether was then functionalized with 4-[trifluorovinyl(oxy)benzoylchloride. An attempt was made to polymerize the resulting oligomer via the cyclodimerization of the terminal trifluorovinylether moieties. Although the viscosity of the oligomer increased significantly during polymerization, Gel Permeation Chromatography (GPC) analysis revealed that the Tetrahydrofuran (THF) soluble portion of the polymer did not have high molecular weight.
Harris, Frank W.↗
Modifications of the CYGRO-2 computer program
Modifications of CYGRO-2 computer program for mechanical analysis of irradiated nuclear fuel pins subjected to swelling due to fission products
Fiero, I. B.↗
Compatibility of elastomers in alternate jet fuels
The compatibility of elastomeric compositions of known resistance to aircraft fuels was tested for potential use in Jet A type fuels obtainable from alternate sources, such as coal. Since such fuels were not available at the time, synthetic alternate fuels were prepared by adding tetralin to a petroleum based Jet A type fuel to simulate coal derived fuels which are expected to contain higher amounts of aromatic and hydroaromatic hydrocarbons. The elastomeric compounds tested were based on butadiene-acrylonitrile rubber, a castable Thiokol polysulfide rubber, and a castable fluorosilicone rubber. Batches of various cross-link densities of these rubbers were made and their chemical stress relaxation behavior in fuel, air, and nitrogen, their swelling properties, and response to mechanical testing were determined.
Kalfayan, S. H.↗