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Preliminary analysis of fission gas behavior and fuel response during an LMFBR operational transient

Description: This summary presents results obtained from a preliminary analysis of gas behavior and oxide fuel response during an LMFBR operational transient. The DiMelfi and Deitrich model is extrapolated to operational transient regimes to delineate brittle versus ductile fuel response modes. All pertinent parameters necessary for application of the DiMelfi and Deitrich model were obtained from the LIFE-3 code.
Date: January 1, 1983
Creator: Liu, Y.Y.
Partner: UNT Libraries Government Documents Department

Benchmark and physics testing of LIFE-4C. Summary

Description: LIFE-4C is a steady-state/transient analysis code developed for performance evaluation of carbide ((U,Pu)C and UC) fuel elements in advanced LMFBRs. This paper summarizes selected results obtained during a crucial step in the development of LIFE-4C - benchmark and physics testing.
Date: June 1, 1984
Creator: Liu, Y.Y.
Partner: UNT Libraries Government Documents Department

Scoping analysis of fission gas behavior in UO/sub 2/ fuel for an in-core thermionic reactor

Description: This paper gives results from a preliminary evaluation of swelling, and, in particular, swelling caused by the retained fission gases (Xe, Kr) in the fuel. Design parameters of the thermionic unit cell and its operating conditions were provided by Space Power, Inc. The analysis tool used is a mechanistic fission-gas behavior code, FASTGRASS, developed by J. Rest at ANL (Rest, 1984). In the FASTGRASS analysis, the UO/sub 2/ fuel column is taken as a solid cylinder. No external restraint is imposed on the UO/sub 2/ fuel other than the ambient pressure. The maximum fuel burnup is assumed to be 5 at%. The operating conditions specified for the UO/sub 2/ fuel fall into two categories: (a) low linear power (q' = 5-8 kW/ft) and high fuel surface temperature (T/sub s/ = 1700, 1800, 1900 K) and (b) high linear power (q' = 8-12 kW/ft) and low fuel surface temperature (T/sub s/ = 1300, 1400, 1500, 1600 K). Taking the upper and lower linear powers in each category in combination with the specified fuel surface temperatures resulted in a total of 14 cases for the scoping analysis. The total irradiation time (based on linear power and 5 at% burnup) for these cases varies from 2.5 to 6 years.
Date: October 1, 1984
Creator: Liu, Y.Y. & Rest, J.
Partner: UNT Libraries Government Documents Department

Accuracy of nodal transport and simplified-P/sub 3/ fluxes in benchmark tests

Description: Here we summarize recent work exploring the accuracy of fluxes computed, both by nodal transport methods, and by the simplified-spherical harmonics (SP/sub l/) method. Apparently, significant errors in nodal transport fluxes were first noted by Wagner et al., and attributed to the isotopic-transverse-leakage (ITL) approximation. Later Lawrence detected substantial errors, due to the ITL approximation, in his nodal transport (NTT) solution of the IAEA Stepanek benchmark problem. Gelbard concluded on theoretical grounds that nodal transport fluxes, computed in XY geometry using ITL, should be much more accurate on the coordinate axes than halfway between them and that, at 45/sup 0/ from the axes, nodal transport methods using ITL should give only about half of the true transport correction.
Date: January 1, 1986
Creator: Liu, Y.W.H. & Gelbard, E.M.
Partner: UNT Libraries Government Documents Department

Collisional Cooling of Negative Ion Beams

Description: Investigations have been conducted to determine the feasibility of using collisional cooling for reducing the energy spreads and, consequently, the emittances of negative-ion beams. We have designed a gas-filled RF-quadrupole ion cooler equipped with provisions for retarding energetic negative ion beams to energies below thresholds for electron detachment at injection and for re-acceleration to high energies after the cooling process. The device has been used to cool O{sup -} and F{sup -} ion beams with initial energy spreads, {Delta}E > 10 eV to final energy spreads, {Delta}E {approx} 2 eV FWHM. Overall transmission efficiencies of {approx}14% for F{sup -} beams have been obtained. Experimental results show that electron detachment is the major loss mechanism for negative ions.
Date: June 29, 2001
Creator: Liu, Y.
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

Comparison of various sink strengths for analyzing radiation creep, growth and swelling

Description: The essential physics involved in the reaction-rate-theory analysis of radiation effects at temperatures where both vacancies and self interstitials are mobile is contained in the expressions used for the strengths of distributed point-defect sinks such as dislocations, cavities and grain boundaries. These sink strengths have been obtained by various authors in distinctly different ways, thus giving rise to some possible confusion in comparing the various results. This is even more true with respect to the effect of interaction fields on these sink strengths and the so-called bias factors or sink efficiencies have been defined in entirely different ways, thus rendering quantitative comparisons difficult. We present here a comparison of several procedures in the literature, and attempt to make reasonable quantitative comparisons.
Date: February 1, 1986
Creator: Nichols, F.A. & Liu, Y.Y.
Partner: UNT Libraries Government Documents Department

Steady-state and transient fission gas release and swelling model for LIFE-4. [LMFBR]

Description: The fuel-pin modeling code LIFE-4 and the mechanistic fission gas behavior model FASTGRASS have been coupled and verified against gas release data from mixed-oxide fuels which were transient tested in the TREAT reactor. Design of the interface between LIFE-4 and FASTGRASS is based on an earlier coupling between an LWR version of LIFE and the GRASS-SST code. Fission gas behavior can significantly affect steady-state and transient fuel performance. FASTGRASS treats fission gas release and swelling in an internally consistent manner and simultaneously includes all major mechanisms thought to influence fission gas behavior. The FASTGRASS steady-state and transient analysis has evolved through comparisons of code predictions with fission-gas release and swelling data from both in- and ex-reactor experiments. FASTGRASS was chosen over other fission-gas behavior models because of its availability, its compatibility with the LIFE-4 calculational framework, and its predictive capability.
Date: June 1, 1984
Creator: Villalobos, A.; Liu, Y.Y. & Rest, J.
Partner: UNT Libraries Government Documents Department

Thermal conductivities for sintered and sphere-pac Li/sub 2/O and. gamma. /sup -/LiAlO/sub 2/ solid breeders with and without irradiation effects

Description: Thermal conductivities (k, k/sub eff/) have been estimated for sintered and sphere-pac Li/sub 2/O and ..gamma..-LiAlO/sub 2/ with and without neutron irradiation effects. The estimation is based on (1) data from unirradiated UO/sub 2/, Li/sub 2/O, and ..gamma..-LiAlO/sub 2/; (2) data from irradiated dielectric insulator materials; and (3) relatively simple physical models. Comparison of model predictions with limited ex- and in-reactor data found reasonable agreement, thus lending credence for their use in design applications. The impact of thermal conductivities on tritium breeding and power generation in fusion solid-breeder blankets is briefly highlighted.
Date: July 1, 1984
Creator: Liu, Y.Y. & Tam, S.W.
Partner: UNT Libraries Government Documents Department

Thermal conductivity of fusion solid breeder materials

Description: Several simple and useful formulae for estimating the thermal conductivity of lithium-containing ceramic tritium breeder materials for fusion reactor blankets are given. These formulae account for the effects of irradiation, as well as solid breeder configuration, i.e., monolith or a packed bed. In the latter case, a coated-sphere concept is found more attractive in incorporating beryllia (a neutron multiplier) into the blanket than a random mixture of solid breeder and beryllia spheres.
Date: June 1, 1986
Creator: Liu, Y.Y. & Tam, S.W.
Partner: UNT Libraries Government Documents Department

FASTER: A new DOE effort to bridge ESM and ASR sciences

Description: In order to better use the long-term ARM measurements to evaluate parameterizations of fast processes used in global climate models --- mainly those related to clouds, precipitation and aerosols, the DOE Earth System Modeling (ESM) program funds a new multi-institution project led by the Brookhaven National Laboratory, FAst -physics System Testbed and Research (FASTER). This poster will present an overview of this new project and its scientific relationships to the ASR sciences and ARM measurements.
Date: March 15, 2010
Creator: Liu, 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

Criticality control in shipments of fissile materials

Description: This paper describes a procedure for finite-array criticality analysis to ensure criticality safety of shipments of fissile materials in US DOE-certified packages. After the procedure has been performed, one can obtain the minimum transport index and determine the maximum number of fissile packages allowable in a shipment that meets the 10 CFR 71 criticality safety requirements.
Date: March 14, 2000
Creator: Liaw, J. R. & Liu, Y. Y.
Partner: UNT Libraries Government Documents Department

Analysis of radiation measurement data of the BUSS cask

Description: The Beneficial Uses Shipping System (BUSS) is a Type-B packaging developed for shipping nonfissile, special-form radioactive materials to facilities such as sewage, food, and medical-product irradiators. The primary purpose of the BUSS cask is to provide shielding and confinement, as well as impact, puncture, and thermal protection for its certified special-form contents under both normal transport and hypothetical accident conditions. A BUSS cask that contained 16 CsCl capsules (2.723 {times} 10{sup 4} TBq total activity) was recently subjected to radiation survey measurements at a Westinghouse Hanford facility, which provided data that could be used to validate computer codes. Two shielding analysis codes, MICROSHIELD (User`s Manual 1988) and SAS4 (Tan 1993), that are used at Argonne National Laboratory to evaluate the safety of packaging of radioactive materials during transportation, have been selected for analysis of radiation data obtained from the BUSS cask. MICROSHIELD, which performs only gamma radiation shielding calculation, is based on a point-kernel model with idealized geometry, whereas SAS4 is a control module in the SCALE code system (1995) that can perform three-dimensional Monte Carlo shielding calculation for photons and neutrons, with built-in procedures for cross-section data processing and automated variance reduction. The two codes differ in how they model the details of the physics of gamma photon attenuation in materials, and this difference is reflected in the associated engineering cost of the analysis. One purpose of the analysis presented in this paper, therefore, is to examine the effects of the major modeling assumptions in the two codes on calculated dose rates, and to use the measured dose rates for comparison. The focus in this paper is on analysis of radiation dose rates measured on the general body of the cask and away from penetrations.
Date: December 31, 1995
Creator: Liu, Y.Y. & Tang, J.S.
Partner: UNT Libraries Government Documents Department

Cost update technology, safety, and costs of decommissioning a reference uranium hexafluoride conversion plant

Description: The purpose of this study is to update the cost estimates developed in a previous report, NUREG/CR-1757 (Elder 1980) for decommissioning a reference uranium hexafluoride conversion plant from the original mid-1981 dollars to values representative of January 1993. The cost updates were performed by using escalation factors derived from cost index trends over the past 11.5 years. Contemporary price quotes wee used for costs that have increased drastically or for which is is difficult to find a cost trend. No changes were made in the decommissioning procedures or cost element requirements assumed in NUREG/CR-1757. This report includes only information that was changed from NUREG/CR-1757. Thus, for those interested in detailed descriptions and associated information for the reference uranium hexafluoride conversion plant, a copy of NUREG/CR-1757 will be needed.
Date: August 1, 1995
Creator: Miles, T.L. & Liu, Y.
Partner: UNT Libraries Government Documents Department

Evaluation of accident frequencies at the canister storage building

Description: By using the fault tree logic, an evaluation of the design basis accident frequencies at the Canister Storage Building has been performed. The evaluation demonstrates that due to low frequency of occurrences, the following design basis accidents are considered not credible (annual frequency of less than 10{sup -6}): Rearrangement of multi-canister overpack (MCO) internals; Gaseous release from the MCO; MCO internal hydrogen explosion; MCO external hydrogen explosion; Thermal runaway reactions inside the MCO; and Violation of design temperature criteria.
Date: May 13, 1999
Creator: LIU, Y.J.
Partner: UNT Libraries Government Documents Department

Gas breakdown limits for inverse Cherenkov laser accelerators

Description: The probability of avalanche, tunneling and multiphoton ionization induced by a CO{sub 2} laser in H{sub 2} gas has been calculated. Laser light screening by a self-induced plasma density gradient is considered as the limiting factor for upscaling a CO{sub 2} laser-driven Inverse Cherenkov Laser Accelerator beyond 650 MeV/m. However, in near-resonance inverse Cherenkov acceleration where a shorter wavelength laser is used at a wavelength near the resonance of the gas (e.g. 248nm in H{sub 2}), the formation of a plasma is not a problem because the plasma density is below the critical density. In that case, the laser beam propagates unaffected through the plasma and the acceleration gradient is not limited by gas breakdown. Gradients > 1 GeV/m are possible.
Date: July 1, 1995
Creator: Liu, Y. & Pogorelsky, I.V.
Partner: UNT Libraries Government Documents Department

Generic aging management programs for license renewal of BWR reactor coolant systems components.

Description: The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of the AMPs, which do not require further evaluation, do need enhancements to allow for an extended period of ...
Date: February 15, 2002
Creator: Shah, V.N. & Liu, Y.Y.
Partner: UNT Libraries Government Documents Department

Evaluation of Accident Frequencies at the Canister Storage Bldg (CSB)

Description: By using the fault tree logic, an evaluation of the design basis accident frequencies at the Canister Storage Building has been performed. The evaluation demonstrates that due to low frequency of occurrences, the following design basis accidents are considered not credible (annual frequency of less than 10{sup -6}): Rearrangement of multidster overpack (MCO) internals; Gaseous release from the MCO; MCO internal hydrogen explosion; MCO external hydrogen explosion; Thermal runaway reactions inside the MCO; and Violation of design temperature criteria.
Date: September 2, 1999
Creator: Liu, Y. J.
Partner: UNT Libraries Government Documents Department

Fluid flow and heat transfer modeling for castings

Description: Casting is fundamental to manufacturing of many types of equipment and products. Although casting is a very old technology that has been in existence for hundreds of years, it remains a highly empirical technology, and production of new castings requires an expensive and time-consuming trial-and-error approach. In recent years, mathematical modeling of casting has received increasing attention; however, a majority of the modeling work has been in the area of heat transfer and solidification. Very little work has been done in modeling fluid flow of the liquid melt. This paper presents a model of fluid flow coupled with heat transfer of a liquid melt for casting processes. The model to be described in this paper is an extension of the COMMIX code and is capable of handling castings with any shape, size, and material. A feature of this model is the ability to track the liquid/gas interface and liquid/solid interface. The flow of liquid melt through the sprue and runners and into the mold cavity is calculated as well as three-dimensional temperature and velocity distributions of the liquid melt throughout the casting process. 14 refs., 13 figs.
Date: January 1, 1986
Creator: Domanus, H.M.; Liu, Y.Y. & Sha, W.T.
Partner: UNT Libraries Government Documents Department

Formulation for the analysis of pellet-cladding mechanical interaction

Description: A formulation has been derived for the thermoelastic analysis of pellet-cladding mechanical interaction. The formulation is based on a combination of finite element analysis and the matrix-displacement method. Boundary conditions at the pellet-cladding and pellet-pellet contacting sites can be treated realistically by considering the force and displacement relationships at the interfaces. As a result, no simulating gap elements need be used and the analysis of the two limiting cases, fuel-cladding bonding (stick) and fuel-cladding slippage (slip), follows directly from imposing the appropriate boundary conditions. From the principle of superposition, this formulation also gives a very compact scheme which can be used to study a variety of power ramp cases with ease. No additional finite element analysis is required after the solutions for a number of individual base-case problems have been obtained. Because of thermoelastic response of a fuel-element during a power ramp corresponds to that for a fast ramp rate, which subjects cladding to the most severe mechanical loading, the results from our thermoelastic analysis can be very useful in evaluating the likelihood of cladding failure when they are combined with knowledge of the cladding failure mode (plastic instability or stress-corrosion cracking). Since the failure mode for the Light Water Reactor fuel elements is predominantly stress-corrosion cracking and there is generally no time-independent cladding plastic deformation during a power ramp, our formulation may be applied directly to provide an assessment of the permissible power ramps. This could lead to a relaxation of the current restrictive and empirically based reactor-operation rules.
Date: January 1, 1979
Creator: Liu, Y.Y.; Meyer, J.E. & Argon, A.S.
Partner: UNT Libraries Government Documents Department

Fine-mesh limit of 1D nodal transport equations

Description: Imbedded in multidimensional nodal transport codes is the solution of transverse integrated ID transport equations. Since, in this solution, fluxes on boundaries are DP/sub 1/, it is generally assumed that the ID solutions, in the small-mesh limit, approach DP/sub 1/ solutions. It will be shown that this is not true. Small-mesh limits of the ID nodal equations will be derived, and it will be shown that these are substantially worse than the DP/sub 1/ equations under certain circumstances. Alternate ID nodal equations (which do have a DP/sub 1/ small-mesh limit) are proposed. 2 refs.
Date: January 1, 1987
Creator: Gelbard, E.M.; Liu, Y.W.H. & Olvey, L.
Partner: UNT Libraries Government Documents Department

Evaluation of selected samples from the TMI-2 core

Description: Core-bore samples from the K9 and N12 locations in the TMI-2 core were examined for microstructural and microchemical features. The purpose of the examinations was twofold, first to determine core temperatures at known elevations in the core, and second to obtain insight into materials interactions that lead to core degradation. The temperature at the /approximately/50-cm elevation in the N12 location, a control rod position, was estimated to have been /approximately/960/degree/C and dropping sharply to less than 800/degree/C a few centimeters below. These temperatures were estimated based on the eutectic reaction between a control rod's Zircaloy guide tube and its stainless steel cladding, the ..beta..-phase transformation temperature in Zircaloy, and the melting temperature of the Ag-In-Cd control material. In the K9 location, local transient temperatures to /approximately/900/degree/C at the 37-cm elevation were estimated from the ..beta..-phase transformation in the Zircaloy cladding. Interactions of note were the guide tube/cladding eutectic interaction in the control rod, the apparent degradation of Zircaloy cladding by molten Cd and In, and the attack of stainless steel by Sn in a melt of Ag-In-Cd. 4 refs., 10 figs., 1 tab.
Date: December 1, 1988
Creator: Liu, Y.Y.; Neimark, L.A. & Jackson, W.D.
Partner: UNT Libraries Government Documents Department

The foaming of U-Al fuel under simulated reactor accident conditions

Description: Postirradiation heating tests were conducted on segments of UAl{sub 4}/Al dispersion fuel plates clad with Al to scope the foaming (rapid swelling) behavior of such fuels during beyond-design-basis accident scenarios. Four tests investigated maximum temperature, ramp rate, and duration with a liquid phase as parameters in foam formation and stability. Real-time fission-gas release was also determined during the foaming process. Ramp-rate had the most noticeable effect of foam formation and collapse.
Date: March 1, 1993
Creator: Neimark, L. A. & Liu, Y. Y.
Partner: UNT Libraries Government Documents Department