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Mathematical models for predicting tritium transport in lithium ceramics

Description: Tritium retention and release data for lithium ceramics (Li/sub 2/O, LiAlO/sub 2/, and Li/sub 4/SiO/sub 4/) are available from in-reactor and post-irradiation anneal tests on single crystals, powders, and sintered products. With the exception of the single-crystal tests in which bulk diffusion is the rate-limiting release mechanism, it is very difficult to interpret the results of these tests and extrapolate the results to design conditions for a fusion solid-breeder blanket. Mathematical models are presented for various bulk, grain-boundary, and free-surface phenomena to aid in the interpretation and extrapolation of the data. 24 refs., 6 figs., 4 tabs.
Date: April 1, 1987
Creator: Billone, M.C.
Partner: UNT Libraries Government Documents Department

Modeling tritium behavior in Li{sub 2}ZrO{sub 3}.

Description: Lithium metazirconate (Li{sub 2}ZrO{sub 3}) is a promising tritium breeder material for fusion reactors because of its excellent tritium release characteristics. In particular, for water-cooled breeding blankets (e.g., ITER), Li{sub 2}ZrO{sub 3} is appealing from a design perspective because of its good tritium release at low operating temperatures. The steady-state and transient tritium release/retention database for Li{sub 2}ZrO{sub 3} is reviewed, along with conventional diffusion and first-order surface resorption models which have been used to match the database. A first-order surface resorption model is recommended in the current work both for best-estimate and conservative (i.e., inventory upper-bound) predictions. Model parameters we determined and validated for both types of predictions, although emphasis is placed on conservative design predictions. The effects on tritium retention of ceramic microstructure, protium partial pressure in the purge gas and purge gas flow rate are discussed, along with other mechanisms for tritium retention which may not be dominant in the experiments, but may be important in blanket design analyses. The proposed tritium retention/release model can be incorporated into a transient thermal performance code to enable whole-blanket predictions of tritium retention/release during cyclic reactor operation. Parameters for the ITER driver breeding blanket are used to generate a numerical set of model predictions for steady-state operation.
Date: January 13, 1998
Creator: Billone, M. C.
Partner: UNT Libraries Government Documents Department

Revision of the tensile database for V-Ti and V-Cr-Ti alloys tested at ANL.

Description: The published database for the tensile properties of unirradiated and irradiated vanadium-based alloys tested at Argonne National Laboratory (ANL) has been reviewed. The alloys tested are in the ranges of V-(0-18)wt.%Ti and V-(4-15)wt.%Cr-(3-15)wt.%Ti. A consistent methodology, based on ASTM terminology and standards, has been used to re-analyze the unpublished load vs. displacement curves for 162 unirradiated samples and 91 irradiated samples to determine revised values for yield strength (YS), ultimate tensile strength (UTS), uniform elongation (UE) and total elongation (TE). The revised data set contains lower values for UE ({minus}5{+-}2% strain) and TE ({minus}4{+-}2% strain) than previously reported. Revised values for YS and UTS are consistent with the previously-published values in that they are within the scatter usually associated with these properties.
Date: January 13, 1998
Creator: Billone, M. C.
Partner: UNT Libraries Government Documents Department

Modeling of tritium transport in lithium aluminate fusion solid breeders

Description: Lithium aluminate is a candidate tritium-breeding material for fusion reactor blankets. One of the concerns with using LiAlO/sub 2/ is tritium recovery from this material, particularly at low operating temperatures and high fluences. The data from various tritium release experiments with ..gamma..-LiAlO/sub 2/ and related materials are reviewed and analyzed to determine under what conditions bulk diffusion is the rate-limiting mechanism for tritium transport and what the effective bulk diffusion coefficient should be. Steady-state and transient models based on bulk diffusion are developed and used to interpret the data. Design calculations are then performed with the verified models to determine the steady-state inventory and time to reach equilibrium for a full-scale fusion blanket.
Date: February 1, 1985
Creator: Billone, M.C. & Clemmer, R.G.
Partner: UNT Libraries Government Documents Department

Tritium percolation, convection, and permeation in fusion solid breeder blankets

Description: Models are developed to describe the percolation of released tritium through the breeder interconnected porosity to the purge stream, convection of tritium by the helium purge stream, and leakage or permeation of tritium through the structural material to the primary coolant system. Important parameters in the models are tritium generation rate, breeder microstructure, tritium species in the gas phase, temperatures, tritium diffusivities and permeabilities, and effectiveness of oxide barriers.
Date: January 1, 1985
Creator: Billone, M.C. & Liu, Y.Y.
Partner: UNT Libraries Government Documents Department

Solid breeder/structure mechanical interaction and thermal stability

Description: Solid breeder/structure mechanical interaction (BSMI) during fusion reactor blanket operation is a potential failure mode which could limit the lifetime of the blanket. The severity of BSMI will generally depend on the materials, specific blanket designs, and blanket operating conditions. Thermomechanical analyses performed for a helium-cooled blanket employing Li/sub 2/O/HT-9 plates indicate that BSMI could be a serious concern for this blanket.
Date: April 1, 1985
Creator: Liu, Y.Y.; Billone, M.C. & Taghavi, K.
Partner: UNT Libraries Government Documents Department

Thermal conductivity and tritium retention in Li{sub 2}O and Li{sub 2}ZrO{sub 3}

Description: Lithium oxide (Li{sub 2}O) and lithium zirconate (Li{sub 2}ZrO{sub 3}) are promising ceramic breeder materials for fusion reactor blankets. The thermal and tritium transport databases for these materials are reviewed. Algorithms are presented for predicting both the temperature distribution and the retained tritium profile across sintered-product and pebble-bed regions. Sample design calculations are also performed to demonstrate the relative advantages of each breeder ceramic. For Li{sub 2}O, the thermal conductivity of sintered-product material has been measured over a wide range of temperatures and densities. Data are also available for the effective thermal conductivity of a pebble bed (in atmospheric helium) with 55% packing fraction for the 5-mm-diameter/75%-dense pebbles. Similar results are available for sintered-product and pebble-bed (60% packing fraction for 1.2-mm-diameter/80%-dense pebbles in atmospheric He) Li{sub 2}ZrO{sub 3}. Hall and Martin model predictions are in reasonable agreement with both sets of pebble bed data. Thus, the databases and calculational algorithms are well established for performing thermal analyses. 15 refs., 5 figs.
Date: August 1, 1997
Creator: Billone, M.C.
Partner: UNT Libraries Government Documents Department

Recommended design correlations for S-65 beryllium

Description: The properties of tritium and helium behavior in irradiated beryllium are reviewed, along with the thermal-mechanical properties needed for ITER design analysis. Correlations are developed to describe the performance of beryllium in a fusion reactor environment. While this paper focuses on the use of beryllium as a plasma-facing component (PFC) material, the correlations presented here can also be used to describe the performance of beryllium as a neutron multiplier for a tritium breeding blanket. The performance properties for beryllium are subdivided into two categories: properties which do not change with irradiation damage to the bulk of the material; and properties which are degraded by neutron irradiation. The approach taken in developing properties correlations is to describe the behavior of dense, pressed S-65 beryllium as a function of temperature. As there are essentially no data on the performance of porous and/or irradiated S-65 beryllium, the degradation of properties with as-fabricated porosity and irradiation are determined form the broad data base on S-200F, as well as other types and grades, and applied to S-65 beryllium by scaling factors. The resulting correlations can be used for Be produced by vacuum hot pressing (VHP) and cold-pressing (CP)/sintering(S)/hot-isostatic-pressing(HIP). The performance of plasma-sprayed beryllium is discussed but not quantified.
Date: December 31, 1995
Creator: Billone, M.C.
Partner: UNT Libraries Government Documents Department

Materials for breeding blankets

Description: There are several candidate concepts for tritium breeding blankets that make use of a number of special materials. These materials can be classified as Primary Blanket Materials, which have the greatest influence in determining the overall design and performance, and Secondary Blanket Materials, which have key functions in the operation of the blanket but are less important in establishing the overall design and performance. The issues associated with the blanket materials are specified and several examples of materials performance are given. Critical data needs are identified.
Date: September 1, 1995
Creator: Mattas, R.F. & Billone, M.C.
Partner: UNT Libraries Government Documents Department

Development of LIFE4-CN: a combined code for steady-state and transient analyses of advanced LMFBR fuels

Description: The methodology used to develop the LMFBR carbide/nitride fuels code, LIFE4-CN, is described in detail along with some subtleties encountered in code development. Fuel primary and steady-state thermal creep have been used as an example to illustrate the need for physical modeling and the need to recognize the importance of the materials characteristics. A self-consistent strategy for LIFE4-CN verification against irradiation data has been outlined with emphasis on the establishment of the gross uncertainty bands. These gross uncertainty bands can be used as an objective measure to gauge the overall success of the code predictions. Preliminary code predictions for sample steady-state and transient cases are given.
Date: January 1, 1979
Creator: Liu, Y.Y.; Zawadzki, S.; Billone, M.C.; Nayak, U.P. & Roth, T.
Partner: UNT Libraries Government Documents Department

Examination of spent PWR fuel rods after 15 years in dry storage.

Description: Virginia Power Surry Nuclear Station Pressurized Water Reactor (PWR) fuel was stored in a dry inert atmosphere Castor V/21 cask at the Idaho National Environmental and Engineering Laboratory (INEEL) for 15 years at peak cladding temperatures decreasing from about 350 to 150 C. Prior to the storage, the loaded cask was subjected to extensive thermal benchmark tests. The cask was opened to examine the fuel for degradation and to determine if it was suitable for extended storage. No rod breaches had occurred and no visible degradation or crud/oxide spallation were observed. Twelve rods were removed from the center of the T11 assembly and shipped from INEEL to the Argonne-West HFEF for profilometric scans. Four of these rods were punctured to determine the fission gas release from the fuel matrix and internal pressure in the rods. Three of the four rods were cut into five segments each, then shipped to the Argonne-East AGHCF for detailed examination. The test plan calls for metallographic examination of six samples from two of the rods, microhardness and hydrogen content measurements at or near the six metallographic sample locations, tensile testing of six samples from the two rods, and thermal creep testing of eight samples from the two rods to determine the extent of residual creep life. The results from the profilometry (12 rods), gas release measurements (4 rods), metallographic examinations (2 samples from 1 rod), and microhardness and hydrogen content characterization (2 samples from 1 rod) are reported here. The tensile and creep studies are just starting and will be reported at a later date, along with the additional characterization work to be performed. Although only limited prestorage characterization is available, a number of preliminary conclusions can be drawn based on comparison with characterization of Florida Power Turkey Point rods of a similar vintage. Based ...
Date: February 11, 2002
Creator: Einziger, R. E.; Tsai, H. C.; Billone, M. C. & Hilton, B. A.
Partner: UNT Libraries Government Documents Department

The ARIES-RS power core -- Recent development in Li/V designs

Description: The ARIES-RS fusion power plant design study is based on reversed-shear (RS) physics with a Li/V (lithium breeder and vanadium structure) blanket. The reversed-shear discharge has been documented in many large tokamak experiments. The plasma in the RS mode has a high beta, low current, and low current drive requirements. Therefore, it is an attractive physics regime for a fusion power plant. The blanket system based on a Li/V has high temperature operating capability, good tritium breeding, excellent high heat flux removal capability, long structural life time, low activation, low after heat and good safety characteristics. For these reasons, the ARIES-RS reactor study selected Li/V as the reference blanket. The combination of attractive physics and attractive blanket engineering is expected to result in a superior power plant design. This paper summarizes the power core design of the ARIES-RS power plant study.
Date: April 1, 1997
Creator: Sze, D.K.; Billone, M.C. & Hua, T.Q.
Partner: UNT Libraries Government Documents Department

Characteristics of hydride precipitation and reorientation in spent-fuel cladding.

Description: The morphology, number density, orientation, distribution, and crystallographic aspects of Zr hydrides in Zircaloy fuel cladding play important roles in fuel performance during all phases before and after discharge from the reactor, i.e., during normal operation, transient and accident situations in the reactor, temporary storage in a dry cask, and permanent storage in a waste repository. In the past, partly because of experimental difficulties, hydriding behavior in irradiated fuel cladding has been investigated mostly by optical microscopy (OM). In the present study, fundamental metallurgical and crystallographic characteristics of hydride precipitation and reorientation were investigated on the microscopic level by combined techniques of OM and transmission electron and scanning electron microscopy (TEM and SEM) of spent-fuel claddings discharged from several boiling and pressurized water reactors (BWRs and PWRs). Defueled sections of standard and Zr-lined Zircaloy-2 fuel claddings, irradiated to fluences of {approx}3.3 x 10{sup 21} n cm{sup {minus}2} and {approx}9.2 x 10{sup 21} n cm{sup {minus}2} (E > 1 MeV), respectively, were obtained from spent fuel rods discharged from two BWRs. Sections of standard and low-tin Zircaloy-4 claddings, irradiated to fluences of {approx}4.4 x 10{sup 21} n cm{sup {minus}2}, {approx}5.9 x 10{sup 21} n cm{sup {minus}2}, and {approx}9.6 x 10{sup 21} n cm{sup {minus}2} (E > 1 MeV) in three PWRs, were also obtained. Microstructural characteristics of hydrides were analyzed in as-irradiated condition and after gas-pressurization-burst or expanding-mandrel tests at 292-325 C in Ar for some of the spent-fuel claddings. Analyses were also conducted of hydride habit plane, morphology, and reorientation characteristics on unirradiated Zircaloy-4 cladding that contained dense radial hydrides. Reoriented hydrides in the slowly cooled unirradiated cladding were produced by expanding-mandrel loading.
Date: November 14, 2000
Creator: Chung, H. M.; Strain, R. V. & Billone, M. C.
Partner: UNT Libraries Government Documents Department

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

Description: This report discusses a fuel system utilizing metallic U-Pu-Zr alloys which has been developed for advanced liquid metal-cooled reactors (LMRs). Result`s from extensive irradiation testing conducted in EBR-II show a design having the following key features can achieve both high reliability and high burnup capability: a cast nominally U-20wt %Pu-10wt %Zr slug with the diameter sized to yield a fuel smear density of {approx}75% theoretical density, low-swelling tempered martensitic stainless steel cladding, sodium bond filling the initial fuel/cladding gap, and an as-built plenum/fuel volume ratio of {approx}1.5. The robust performance capability of this design stems primarily from the negligible loading on the cladding from either fuel/cladding mechanical interaction or fission-gas pressure during the irradiation. The effects of these individual design parameters, e.g., fuel smear density, zirconium content in fuel, plenum volume, and cladding types, on fuel element performance were investigated in a systematic irradiation experiment in EBR-II. The results show that, at the discharge burnup of {approx}11 at. %, variations on zirconium content or plenum volume in the ranges tested have no substantial effects on performance. Fuel smear density, on the other hand, has pronounced but countervailing effects: increased density results in greater cladding strain, but lesser cladding wastage from fuel/cladding chemical interaction.
Date: October 1, 1994
Creator: Tsai, H.; Cohen, A.B.; Billone, M.C. & Neimark, L.A.
Partner: UNT Libraries Government Documents Department

Relationship between fabrication parameters and structural characteristics of sintered lithium orthosilicate

Description: Lithium orthosilicate (Li/sub 4/SiO/sub 4/) powder was synthesized by the solid-state reaction of lithium oxide with amorphous silica, and the effects of fabrication parameters on the structural characteristics of the product were investigated. Processing considerations such as milling media, drying technique, calcination time and temperature, pressing behavior, sintering time and temperatures, and impurity concentration were addressed. The initial powder particle size was observed to be important in achieving high sintered density, with densities as high as 98% TD achieved with a particle size of approximately 1 ..mu..m. 9 refs., 6 figs.
Date: February 1, 1988
Creator: Chu, C.Y.; Bar, K.; Singh, J.B.; Poeppel, R.B. & Billone, M.C.
Partner: UNT Libraries Government Documents Department

Steady-state deformation of some lithium ceramics

Description: The stress-strain behavior of Li/sub 2/O, LiAlO/sub 2/ and Li/sub 2/ZrO/sub 3/ polycrystals, with densities varying from 0.70 to 0.95 of the theoretical, has been measured in constant-crosshead-speed compression tests at temperatures of 700 to 1000/sup 0/C with strain rates ranging from about 10/sup -6/ to 10/sup -4/ s/sup -1/. A steady-state stress, sigma/sub s/, for which the work-hardening rate becomes zero, was achieved. These results, therefore, yield information equivalent to that obtained from creep experiments. Limited data on LiAlO/sub 2/ and Li/sub 2/ZrO/sub 3/ were obtained. Nevertheless, under comparable conditions the lithium aluminate and zirconate were considerably stronger than the Li/sub 2/O. This finding may be related to differences in crystal structure. It is, however, likely that in operation as a function breeder blanket material, the oxide will swell whereas the aluminate and the zirconate will crack. 5 refs., 6 figs., 1 tab.
Date: May 1, 1987
Creator: Poeppel, R.B.; Routbort, J.L.; Billone, M.C.; Applegate, D.S.; Buchmann, E. & Londschien, B.
Partner: UNT Libraries Government Documents Department

Tritium retention and release analysis for US-ITER blanket

Description: The US design for the ITER tritium-breeding blanket consists of layers of Be multiplier, stainless steel cladding, and Li{sub 2}O ceramic breeder. Tritium is recovered from the ceramic breeder by purging it with He + 0.2% H{sub 2}. Models have been developed to describe the purge-flow thermal-hydraulics and gas reactions and the tritium retention/release due to lattice diffusion, desorption/adsorption, solubility/precipitation, and percolation through interconnected porosity. These have been incorporated into the steady-state code TIARA for the purpose of performing design calculations for Tritium Inventory and Release Analysis. Transient calculations for pulsed operation are done with a modified version of the DISPL code. The results of both steady-state and transient analyses for tritium retention and releases are given for anticipated ITER operating conditions. 13 refs., 6 figs., 3 tabs.
Date: November 1, 1990
Creator: Billone, M.C.; Lin, C.C.; Attaya, H. & Gohar, Y.
Partner: UNT Libraries Government Documents Department

Irradiation creep of vanadium-base alloys.

Description: A study of irradiation creep in vanadium-base alloys is underway with experiments in the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR) in the US. Test specimens are thin-wall sealed tubes with internal pressure loading. The results from the initial ATR irradiation at low temperature (200-300 C) to a neutron damage level of 4.7 dpa show creep rates ranging from {approx}0 to 1.2 x 10{sup {minus}5}/dpa/MPa for a 500-kg heat of V-4Cr-4Ti alloy. These rates were generally lower than reported from a previous experiment in BR-10. Because both the attained neutron damage levels and the creep strains were low in the present study, however, these creep rates should be regarded as only preliminary. Substantially more testing is required before a data base on irradiation creep of vanadium alloys can be developed and used with confidence.
Date: May 18, 1998
Creator: Tsai, H.; Matsui, H.; Billone, M. C.; Strain, R. V. & Smith, D. L.
Partner: UNT Libraries Government Documents Department

Mechanical properties and deformation of polycrystalline lithium orthosilicate

Description: Room-temperature strength, fracture toughness, Young's modulus, and thermal-shock resistance were determined for 68--98% dense lithium orthosilicate (Li/sub 4/SiO/sub 4/) specimens. In the low-density regime, both strength and fracture toughness were controlled by the density of the specimen. At high density, the strength depends on grain size. Young's modulus values ranged from 30--103 GPa at densities between 68 and 98% TD. A critical quenching temperature difference in the range of 150--170/degree/C was observed in thermal-shock tests of bar specimens. Steady-state creep tests indicated 90% dense Li/sub 4/SiO/sub 4/ fractures at T less than or equal to 800/degree/C before reaching steady state and deforms plastically at 900/degree/C. It is more creep-resistant at 900/degree/C than Li/sub 2/O, about equal to Li/sub 2/Zr)/sub 3/, and less than LiA10/sub 2/. 13 refs., 4 figs., 1 tab.
Date: February 1, 1988
Creator: Bar, K.; Chu, C.Y.; Singh, J.P.; Goretta, K.C.; Routbort, J.L.; Billone, M.C. et al.
Partner: UNT Libraries Government Documents Department