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High density dispersion fuel

Description: A fuel development campaign that results in an aluminum plate-type fuel of unlimited LEU burnup capability with an uranium loading of 9 grams per cm{sup 3} of meat should be considered an unqualified success. The current worldwide approved and accepted highest loading is 4.8 g cm{sup {minus}3} with U{sub 3}Si{sub 2} as fuel. High-density uranium compounds offer no real density advantage over U{sub 3}Si{sub 2} and have less desirable fabrication and performance characteristics as well. Of the higher-density compounds, U{sub 3}Si has approximately a 30% higher uranium density but the density of the U{sub 6}X compounds would yield the factor 1.5 needed to achieve 9 g cm{sup {minus}3} uranium loading. Unfortunately, irradiation tests proved these peritectic compounds have poor swelling behavior. It is for this reason that the authors are turning to uranium alloys. The reason pure uranium was not seriously considered as a dispersion fuel is mainly due to its high rate of growth and swelling at low temperatures. This problem was solved at least for relatively low burnup application in non-dispersion fuel elements with small additions of Si, Fe, and Al. This so called adjusted uranium has nearly the same density as pure {alpha}-uranium and it seems prudent to reconsider this alloy as a dispersant. Further modifications of uranium metal to achieve higher burnup swelling stability involve stabilization of the cubic {gamma} phase at low temperatures where normally {alpha} phase exists. Several low neutron capture cross section elements such as Zr, Nb, Ti and Mo accomplish this in various degrees. The challenge is to produce a suitable form of fuel powder and develop a plate fabrication procedure, as well as obtain high burnup capability through irradiation testing.
Date: September 1, 1996
Creator: Hofman, G.L.
Partner: UNT Libraries Government Documents Department

Improved performance of U-Mo dispersion fuel by Si addition in Al matrix.

Description: The purpose of this report is to collect in one publication and fit together work fragments presented in many conferences in the multi-year time span starting 2002 to the present dealing with the problem of large pore formation in U-Mo/Al dispersion fuel plates first observed in 2002. Hence, this report summarizes the excerpts from papers and reports on how we interpreted the relevant results from out-of-pile and in-pile tests and how this problem was dealt with. This report also provides a refined view to explain in detail and in a quantitative manner the underlying mechanism of the role of silicon in improving the irradiation performance of U-Mo/Al.
Date: June 1, 2011
Creator: Kim, Y. S. & Hofman, G. L.
Partner: UNT Libraries Government Documents Department

Design of high density gamma-phase uranium alloys for LEU dispersion fuel applications.

Description: Uranium alloys are candidates for the fuel phase in aluminum matrix dispersion fuels requiring high uranium loading. Certain uranium alloys have been shown to have good irradiation performance at intermediate burnup. Previous studies have shown that acceptable fission gas swelling behavior and fuel-aluminum interaction is possible only if the fuel alloy can be maintained in the high temperature body-centered-cubic {gamma}-phase during fabrication and irradiation, i.e., at temperatures at which {alpha}-U is the equilibrium phase. Transition metals in Groups V through VIII are known to allow metastable retention of the gamma phase below the equilibrium isotherm. These metals have varying degrees of effectiveness in stabilizing the gamma phase. Certain alloys are metastable for very long times at the relatively low fuel temperatures seen in research reactor operation. In this paper, the existing data on the gamma stability of binary and ternary uranium alloys is analyzed. The mechanism and kinetics of decomposition of the gamma phase are assessed with the help of metal alloy theory. Alloys with the highest possible uranium content, good gamma-phase stability, and good neutronic performance are identified for further metallurgical studies and irradiation tests. Results from theory will be compared with experimentally generated data.
Date: October 19, 1998
Creator: Hofman, G. L.
Partner: UNT Libraries Government Documents Department

AAA fuels handbook.

Description: PART A of this handbook is for metal alloy fuels. The metal alloy of transuranic elements (i.e., Pu, Np, Am, Cm, etc) in Zr, designated as TRU-Zr, is one of the primary candidate fuel types for the AAA system. The data found in the literature were critically reviewed and assessed to provide the recommended ones. For the convenience of the user, most of the materials properties are given in model correlations; performance models are also provided in mathematical formulas. Tabulations were made in case where these were judged to allow more flexibility for the user. The information for the materials properties of the TRU-Zr alloy, however, is extremely scarce in general. Therefore, where no data exists, the values and models based on theoretical estimations and extrapolations from the U-Zr and U-Pu-Zr data are inevitably recommended. The justifications for this will be possible when sufficient measured data are available in the future. In this respect, this part is subject to modification whenever new data or better methods of deduction become available. The purpose of PART B is to provide the best available fuel materials properties and performance models of the (Pu,Zr)N and (TRU,Zr)N solid-solution fuels for fuel design and safety calculation of the Advanced Accelerator Assisted (AAA) system. PART B parallels PART A, Metal Alloy Fuels, in form and topics. The solid solution of the mononitrides of transuranic elements (i.e., Pu, Np, Am, Cm, etc) with zirconium mononitride, designated as (TRU,Zr)N, is one of the primary candidate fuel types for the AAA system, together with metallic TRU-Zr alloy fuels studied in Argonne National Laboratory. The data found in the literature were critically reviewed and assessed to provide the recommended ones. For the convenience of the user, most of the materials properties are given in model correlations; performance models are also provided ...
Date: March 27, 2003
Creator: Kim, Y. S. & Hofman, G. L.
Partner: UNT Libraries Government Documents Department

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

Description: The Dispersion Analysis Research Tool (DART) contains models for fission-gas induced fuel swelling, interaction of fuel with the matrix aluminum, resultant reaction-product swelling, and calculation of the stress gradient within the fuel particle. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by a comparison of DART calculations of fuel swelling of U{sub 3}SiAl-Al and U{sub 3}Si{sub 2}-Al for various dispersion fuel element designs with the data. DART results are compared with data for fuel swelling Of U{sub 3}SiAl-Al in plate, tube, and rod configurations as a function of fission density. Plate and tube calculations were performed at a constant fuel temperature of 373 K and 518 K, respectively. An irradiation temperature of 518 K results in a calculated aluminide layer thickness for the Russian tube that is in the center of the measured range (16 {mu}m). Rod calculations were performed with a temperature gradient across the rod characterized by surface and central temperatures of 373 K and 423 K, respectively. The effective yield stress of irradiated Al matrix material and the aluminide was determined by comparing the results of DART calculations with postirradiation immersion volume measurement of U{sub 3}SiAl plates. The values for the effective yield stress were used in all subsequent simulations. The lower calculated fuel swelling in the rod-type element is due to an assumed biaxial stress state. Fuel swelling in plates results in plate thickness increase only. Likewise, in tubes, only the wall thickness increases. Irradiation experiments have shown that plate-type dispersion fuel elements can develop blisters or pillows at high U-235 burnup when fuel compounds exhibiting breakaway swelling are used at moderate to high fuel volume fractions. DART-calculated interaction layer thickness and fuel swelling follows the trends of the observations. 3 refs., 2 figs.
Date: June 1, 1997
Creator: Rest, J. & Hofman, G.L.
Partner: UNT Libraries Government Documents Department

Postirradiation evaluations of capsules HANS-1 and HANS-2 irradiated in the HFIR target region in support of fuel development for the advanced neutron source

Description: This report describes the design, fabrication, irradiation, and evaluation of two capsule tests containing U{sub 3}Si{sub 2} fuel particles in contact with aluminum. The tests were in support of fuel qualification for the Advanced Neutron Source (ANS) reactor, a high-powered research reactor that was planned for the Oak Ridge National Laboratory. At the time of these tests, the fuel consisted of U{sub 3}Si{sub 2}, containing highly enriched uranium dispersed in aluminum at a volume fraction of {approximately}0.15. The extremely high thermal flux in the target region of the High Flux Isotope Reactor provided up to 90% burnup in one 23-d cycle. Temperatures up to 450{degrees}C were maintained by gamma heating. Passive SiC temperature monitors were employed. The very small specimen size allowed only microstructural examination of the fuel particles but also allowed many specimens to be tested at a range of temperatures. The determination of fission gas bubble morphology by microstructural examination has been beneficial in developing a fuel performance model that allows prediction of fuel performance under these extreme conditions. The results indicate that performance of the reference fuel would be satisfactory under the ANS conditions. In addition to U{sub 3}Si{sub 2}, particles of U{sub 3}Si, UAl{sub 2}, UAl{sub x}, and U{sub 3}O{sub 8} were tested.
Date: August 1, 1995
Creator: Hofman, G.L.; Snelgrove, J.L. & Copeland, G.L.
Partner: UNT Libraries Government Documents Department

Progress in development of low-enriched U-Mo dispersion fuels.

Description: Results from postirradiation examinations and analyses of U-Mo/Al dispersion miniplates are presented. Irradiation test RERTR-5 contained mini-fuel plates with fuel loadings of 6 and 8 gU cm{sup -3}. The fuel material consisted of 6, 7 and 10 wt.% Mo-uranium-alloy powders in atomized and machined form. The swelling behavior of the various fuel types is analyzed, indicating athermal swelling of the U-Mo alloy and temperature-dependent swelling owing to U-Mo/Al interdiffusion.
Date: March 4, 2002
Creator: Hofman, G. L.; Snelgrove, J. L.; Hayes, S. L. & Meyer, M. K.
Partner: UNT Libraries Government Documents Department

Calculation of the evolution of the fuel microstructure in UMo alloys and implications for fuel swelling.

Description: The evolution of a cellular dislocation structure and subsequent recrystallization have been identified as important aspects of the irradiated UMo alloy microstructure that can have a strong impact on dispersion fuel swelling. Dislocation kinetics depends on the preferential bias of dislocations for interstitial compared to vacancies. This paper presents theoretical calculations for the evolution of a cellular dislocation structure, and recrystallization in U-10Mo. Implications for fuel swelling are discussed.
Date: October 1, 1999
Creator: Rest, J.; Hofman, G. L.; Konovalov, I. & Maslov, A.
Partner: UNT Libraries Government Documents Department

Thermal properties for the thermal-hydraulics analyses of the BR2 maximum nominal heat flux.

Description: This memo describes the assumptions and references used in determining the thermal properties for the various materials used in the BR2 HEU (93% enriched in {sup 235}U) to LEU (19.75% enriched in {sup 235}U) conversion feasibility analysis. More specifically, this memo focuses on the materials contained within the pressure vessel (PV), i.e., the materials that are most relevant to the study of impact of the change of fuel from HEU to LEU. This section is regrouping all of the thermal property tables. Section 2 provides a summary of the thermal properties in form of tables while the following sections present the justification of these values. Section 3 presents a brief background on the approach used to evaluate the thermal properties of the dispersion fuel meat and specific heat capacity. Sections 4 to 7 discuss the material properties for the following materials: (i) aluminum, (ii) dispersion fuel meat (UAlx-Al and U-7Mo-Al), (iii) beryllium, and (iv) stainless steel. Section 8 discusses the impact of irradiation on material properties. Section 9 summarizes the material properties for typical operating temperatures. Appendix A elaborates on how to calculate dispersed phase's volume fraction. Appendix B shows the evolution of the BR2 maximum heat flux with burnup.
Date: May 23, 2011
Creator: Dionne, B.; Kim, Y. S. & Hofman, G. L. (Nuclear Engineering Division)
Partner: UNT Libraries Government Documents Department

Effect of recrystallization in high-burnup UO{sub 2} on gas release during RIA-type transients

Description: The authors recently proposed a model for irradiation-induced recrystallization (grain subdivision) and swelling in UO{sub 2} fuels wherein the stored energy in the material is concentrated in a network of sink-like nuclei that diminish with dose due to interaction with radiation-produced defects. It is of considerable interest to explore the consequences of recrystallization on gas release during a reactivity initiated accident (RIA). In the absence of recrystallization, gas release during RIA-type transients is generally limited to gas available on grain boundaries and edges due to the very short heatup times (milliseconds), short cooldown periods (seconds), and relatively long intragranular diffusion distances (on the order of micrometers). However, recrystallization provides grain-boundary surfaces that are substantially closer to the gas retained in the bulk material, and thus the potential for much higher gas release. The authors show the calculated burnup at which grain subdivision will occur as a function of fractional radius and fuel temperature for a generic pressurized water reactor irradiation. The FASTGRASS code was used to calculate fission gas behavior during in-reactor irradiation and during the RIA-type transient. Results are given. It is clear from these results that recrystallization of high-burnup UO{sub 2} has implications for the potential consequences of severe accident scenarios such as the RIA type.
Date: October 1, 1994
Creator: Rest, J. & Hofman, G.L.
Partner: UNT Libraries Government Documents Department

Method for fabricating {sup 99}Mo production targets using low enriched uranium, {sup 99}Mo production targets comprising low enriched uranium

Description: A radioisotope production target and a method for fabricating a radioisotope production target is provided, wherein the target comprises an inner cylinder, a foil of fissionable material (low enriched U) circumferentially contacting the outer surface of the inner cylinder, and an outer hollow cylinder adapted to receive the substantially foil-covered inner cylinder and compress tightly against the foil to provide good mechanical contact therewith. The method for fabricating a primary target for the production of fission products comprises preparing a first substrate to receive a foil of fissionable material so as to allow for later removal of the foil from the first substrate, preparing a second substrate to receive the foil so as to allow for later removal of the foil from the second substrate; attaching the first substrate to the second substrate such that the foil is sandwiched between the first substrate and second substrate to prevent foil exposure to ambient atmosphere, and compressing the exposed surfaces of the first and second substrate to assure snug mechanical contact between the foil, the first substrate and the second substrate.
Date: December 31, 1993
Creator: Wiencek, T.C.; Matos, J.E. & Hofman, G.L.
Partner: UNT Libraries Government Documents Department

Development of very-high-density fuels by the RERTR program

Description: The RERTR program has recently begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm{sup 3}, based on the use of {gamma}-stabilized uranium alloys. Fabrication development teams and facilities are being put into place, and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and the first results should be available by the end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun.
Date: December 1, 1996
Creator: Snelgrove, J.L.; Hofman, G.L.; Trybus, C.L. & Wiencek, T.C.
Partner: UNT Libraries Government Documents Department

Development and processing of LEU targets for {sup 99}Mo production

Description: Substituting LEU for HEU in targets for producing fission-product {sup 99}Mo requires changes in target design and chemical processing. We have made significant progress in developing targets and chemical processes for this purpose. Target development was concentrated on a U- metal foil target as a replacement for the coated-UO{sub 2} Cintichem- type target. Although the first designs were not successful because of ion mixing-induced bonding of the U foil to the target tubes, recent irradiations of modified targets have proven successful. It was shown that only minor modifications of the Cintichem chemical process are required for the U-metal foil targets. A demonstration using prototypically irradiated targets is anticipated by the end of 1996. Progress was also made in basic dissolution of both U-metal foil and Al-clad U{sub 3}Si{sub 2} dispersion fuel targets, and work in this area is also continuing.
Date: February 1, 1997
Creator: Snelgrove, J.L.; Vandergrift, G.F. & Hofman, G.L.
Partner: UNT Libraries Government Documents Department

Irradiation behavior of uranium oxide-aluminum dispersion fuel

Description: An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO{sub 2}-Al dispersion fuel. The aluminum-fuel interaction models were developed based on U{sub 3}O{sub 8}-Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products, as well as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show, that with the assumption that the correlations derived from U{sub 3}O{sub 8} are valid for UO{sub 2}, the LEU UO{sub 2}-Al with a 42% fuel volume loading (4 gm/cc) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10{sup 27} fissions m{sup {minus}3} ({approximately} 63% {sup 235}U burnup).
Date: December 1, 1996
Creator: Hofman, G.L.; Rest, J. & Snelgrove, J.L.
Partner: UNT Libraries Government Documents Department

Development and processing of LEU targets for {sup 99}Mo production-overview of the ANL program

Description: Most of the world`s supply of {sup 99m}Tc for medical purposes is currently produced from the decay of {sup 99}Mo derived from the fissioning of high-enriched uranium (HEU). Substitution of low-enriched uranium (LEU) silicide fuel for the HEU alloy and aluminide fuels used in most current target designs will allow equivalent {sup 99}Mo yields with little change in target geometries. Substitution of uranium metal for uranium oxide films in other target designs will also allow the substitution of LEU for HEU. During 1995, we have continued to study the modification of current targets and processes to allow the conversion from HEU to LEU. A uranium-metal-foil target was fabricated at ANL and irradiated to prototypic burnup in the Indonesian RSG-GAS reactor. Postirradiation examination indicated that minor design modifications will be required to allow the irradiated foil to be removed for chemical processing. Means to dissolve and process LEU foil have been developed, and a mock LEU foil target was processed in Indonesia. We have also developed means to dissolve the LEU foil in alkaline peroxide, where it can be used to replace HEU targets that are currently dissolved in base before recovering and purifying the {sup 99}Mo. We have also continued work on the dissolution of U{sub 3}Si{sub 2} and have a firm foundation on dissolving these targets in alkaline peroxide. The technology-exchange agreement with Indonesia is well underway, and we hope to expand our international cooperations in 1996.
Date: September 1, 1995
Creator: Snelgrove, J.L.; Hofman, G.L. & Wiencek, T.C.
Partner: UNT Libraries Government Documents Department

Analysis of the swelling behavior of U-alloys

Description: Available data on two alloys from the EBR-II driver fuel development program have been utilized in the construction and validation of mechanistic models aimed at elucidating swelling mechanisms in high density uranium alloys. Swelling predictions are made under ATR conditions for U-10Mo fuels, currently under irradiation in the ATR, and for U-10Zr.
Date: October 1, 1997
Creator: Rest, J.; Hofman, G.L.; Coffey, K.L.; Konovalov, I. & Maslov, A.
Partner: UNT Libraries Government Documents Department

DART model for thermal conductivity of U{sub 3}Si{sub 2} aluminum dispersion fuel

Description: This paper describes the primary physical models that form the basis of the DART model for calculating irradiation-induced changes in the thermal conductivity of aluminium dispersion fuel. DART calculations of fuel swelling, pore closure, and thermal conductivity are compared with measured values.
Date: September 1, 1995
Creator: Rest, J.; Snelgrove, J.L. & Hofman, G.L.
Partner: UNT Libraries Government Documents Department

Development and processing of LEU targets for {sup 99}Mo production

Description: Most of the world`s supply of {sup 99m}Tc for medical purposes is currently produced from the decay of {sup 99}Mo derived from the fissioning of high-enriched uranium (HEU). Substantial progress has been made in developing targets and chemical processes for producing {sup 99}Mo using low-enriched uranium (LEU). Target development has been focused on a uranium-metal foil target as a replacement for the coated-UO{sub 2} Cintichem-type target. Although the first designs were not successful because of ion mixing-induced bonding of the uranium foil to the target tubes, recent irradiations of modified targets have proven successful. Only minor modifications of the Cintichem chemical process are required for the uranium-metal foil targets. A demonstration using prototypically irradiated targets is anticipated in February 1997. Progress has also been made in basic dissolution of both uranium-metal foil and aluminum-clad U{sub 3}Si{sub 2} dispersion fuel targets.
Date: April 1, 1997
Creator: Snelgrove, J.L.; Vandegrift, G.F. & Hofman, G.L.
Partner: UNT Libraries Government Documents Department

Aluminum-U{sub 3}Si{sub 2} interdiffusion and its implications for the performance of highly loaded fuel operating at higher temperatures and fission rates

Description: Recent irradiation tests of U{sub 3}Si-Al dispersion fuel have shown performance limitations of this fuel when high volume fractions of U{sub 3}Si{sub 2} operate at high temperatures and high fission rates. This potential problem is associated with high rates of Al-U{sub 3}Si{sub 2} interdiffusion that may lead to complete consumption of matrix aluminum and the formation of excessive porosity.
Date: December 1, 1996
Creator: Hofman, G.L.; Rest, J.; Snelgrove, J.L.; Wiencek, T. & Koster van Groos, S.
Partner: UNT Libraries Government Documents Department

Metallographic analysis of irradiated RERTR-3 fuel test specimens.

Description: The RERTR-3 irradiation test was designed to investigate the irradiation behavior of aluminum matrix U-MO alloy dispersion fuels under high-temperature, high-fission-rate conditions. Initial postirradiation examination of RERTR-3 fuel specimens has concentrated on binary U-MO atomized fuels. The rate of matrix aluminum depletion was found to be higher than predictions based on low temperature irradiation data. Wavelength Dispersive X-ray Spectroscopy (WDS) indicates that aluminum is present in the interior of the fuel particles. WDS data is supported by a mass and volume balance calculation performed on the basis of image analysis results. The depletion of matrix aluminum seems to have no detrimental effects on fuel performance under the conditions tested to date.
Date: November 8, 2000
Creator: Meyer, M. K.; Hofman, G. L.; Strain, R. V.; Clark, C. R. & Stuart, J. R.
Partner: UNT Libraries Government Documents Department

Progress in converting {sup 99}Mo production from high- to low-enriched uranium--1999.

Description: Over this past year, extraordinary progress has been made in executing our charter to assist in converting Mo-99 production worldwide from HEU to LEU. Building on the successful development of the experimental LEU-foil target, we have designed a new, economical irradiation target. We have also successfully demonstrated, in collaboration with BATAN in Indonesia, that LEU can be substituted for HEU in the Cintichem target without loss of product yield or purity; in fact, conversion may make economic sense. We are interacting with a number of commercial producers--we have begun active collaborations with the CNEA and ANSTO; we are working to define the scope of collaborations with MDS Nordion and Mallinckrodt; and IRE has offered its services to irradiate and test a target at the appropriate time. Conversion of the CNEA process is on schedule. Other papers presented at this meeting will present specific results on the demonstration of the LEU-modified Cintichem process, the development of the new target, and progress in converting the CNEA process.
Date: September 29, 1999
Creator: Snelgrove, J. L.; Vandegrift, G. F.; Conner, C.; Wiencek, T. C. & Hofman, G. L.
Partner: UNT Libraries Government Documents Department

Irradiation tests of {sup 99}Mo isotope production targets employing uranium metal foils

Description: Most of the world`s supply of {sup 99m}Tc for medical purposes is currently produced form the decay of {sup 99}Mo derived from the fissioning of high-enriched uranium (HEU). Substitution of low-enriched uranium (LEU) metal foils for the HEU UO{sub 2} used in current target designs will allow equivalent {sup 99}Mo yields with little change in target geometries. Substitution of uranium metal for uranium alloy and aluminide in other target designs will also allow the conversion of HEU to LEU. Several uranium-metal-foil targets have been fabricated at ANL and irradiated to prototypic burnup in the Indonesian RSG-GAS reactor. Postirradiation examination of the initial test indicated that design modifications were required to allow the irradiated foil to be removed for chemical processing. The latest test has shown good irradiation behavior, satisfactory dismantling and foil removal when the U-foil is separated from its containment by metallic, fission-recoil absorbing barriers.
Date: December 1, 1996
Creator: Hofman, G.L.; Wiencek, T.C.; Wood, E.L.; Snelgrove, J.L.; Suripto, , A.; Nasution, H. et al.
Partner: UNT Libraries Government Documents Department