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Debris-bed cooling following an HCDA in a fast reactor. Annual progress report

Description: Natural convection within simulated core debris beds has been investigated experimentally and numerically. The objectives of this research have been: (1) to develop a three-dimensional transient model for analyzing single-phase cooling of debris beds; (2) to validate the model using out-of-pile simulant experiments which measure the detailed structure of the convection cells within the bed as well as integral heat transfer rates; and (3) to apply the model to typical core debris beds over a wide range of parameters in order to determine the relative importance of conduction and convection in sodium-cooled debris beds. The bed has been simulated using directly heated two-dimensional packed tube bundles of different particle diameters, bed loadings, porosities, heat generation rates, and overlaying fluid layer heights. Detailed velocity and temperature profiles within the bed and overlaying fluid have been measured using particle tracing techniques and Mach-Zehnder interferometry. Measurements of the downward and upward power fractions and Nusselt numbers over a wide range of experimental variables have been made. The data have been compared with predictions of the transient three-dimensional natural convection computer code COMMIX-1A. 14 figures.
Date: October 1, 1982
Creator: Abdel-Khalik, S.I.
Partner: UNT Libraries Government Documents Department

Neutronic Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint U.S./Russian Progress Report for Fiscal Year 1997 - Volume 4, Part 2--Saxton Plutonium Program Critical Experiments

Description: Critical experiments with water-moderated, single-region PuO{sub 2}-UO{sub 2} or UO{sub 2}, and multiple-region PuO{sub 2}-UO{sub 2}- and UO{sub 2}-fueled cores were performed at the CRX reactor critical facility at the Westinghouse Reactor Evaluation Center (WREC) at Waltz Mill, Pennsylvania in 1965 [1]. These critical experiments were part of the Saxton Plutonium Program. The mixed oxide (MOX) fuel used in these critical experiments and then loaded in the Saxton reactor contained 6.6 wt% PuO{sub 2} in a mixture of PuO{sub 2} and natural UO{sub 2}. The Pu metal had the following isotopic mass percentages: 90.50% {sup 239}Pu; 8.57% {sup 239}Pu; 0.89% {sup 240}Pu; and 0.04% {sup 241}Pu. The purpose of these critical experiments was to verify the nuclear design of Saxton partial plutonium cores while obtaining parameters of fundamental significance such as buckling, control rod worth, soluble poison worth, flux, power peaking, relative pin power, and power sharing factors of MOX and UO{sub 2} lattices. For comparison purposes, the core was also loaded with uranium dioxide fuel rods only. This series is covered by experiments beginning with the designation SX.
Date: October 12, 2000
Creator: Abdurrahman, NM
Partner: UNT Libraries Government Documents Department

Speciation studies of radionuclides in low level wastes and process waters from pressurized water reactor

Description: Physicochemical characterization studies of aqueous process streams at San Onofre Generating Station Unit No. 1 are providing important source term information concerning the forms of radionuclides being released into the marine environment. The primary coolant, secondary steam condensate, processed low level wastes and tertiary coolant were sampled at several different times in the nuclear fuel cycle. Rdionuclides were partitioned into particulate, cationic, anionic, and nonionic species in the reactor process streams, and into particulate and soluble species in the tertiary seawater coolant. Characterization of the particulate species has included a detailed size distribution. The purpose of this research was to provide information concerning chemical and physical forms of the radionuclides being released to the coastal zone from a nuclear generating station in order to facilitate design of radiobiological studies necessary for assessment of their environmental significance.
Date: October 1, 1977
Creator: Abel, K. H.; Robertson, D. E.; Crecelius, E. A. & Silker, W. B.
Partner: UNT Libraries Government Documents Department

Power Burst Facility/Boron Neutron Capture Therapy Program for cancer treatment

Description: This bulletin discusses activities during this reporting period in the areas of: supporting technology development; large animal model studies; melanoma project; human studies; stability, pharmacology, and toxicology of drugs; and PBF technical support. (FL)
Date: October 1, 1990
Creator: Ackermann, A.L. (ed.)
Partner: UNT Libraries Government Documents Department

Uranium oxide and sodium oxide aerosol experiments: NSPP mixed-oxide tests 303-307, data record report. [LMFBR]

Description: This data record report summarizes five tests, involving mixtures of uranium oxide and sodium oxide aerosols, conducted in the Nuclear Safety Pilot Plant project at Oak Ridge National Laboratory. The goal of this project is to establish the validity (or level of conservatism) of the aerosol behavioral code, HAARM-3, and follow-on codes under development at Battelle Columbus Laboratories for the US Nuclear Regulatory Commission. Descriptions of the five tests with tables and graphs summarizing the results are included.
Date: October 1, 1982
Creator: Adams, R.E.; Kress, T.S. & Tobias, M.L.
Partner: UNT Libraries Government Documents Department

Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics

Description: A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences.
Date: October 1, 1985
Creator: Adams, S.R.
Partner: UNT Libraries Government Documents Department

Natural circulation in FFTF

Description: This paper provides a review of the evaluation of natural circulation in the Fast Flux Test Facility (FFTF), a loop type reactor plant. The emphasis is on the role of natural circulation in decay heat removal; however, free convection during either operation at power or normal shutdown does influence some aspects of the design and these are mentioned. In treating decay heat removal the topics discussed include steady state loop performance and transient dynamics for conditions immediately after scram and for the longer term which involves different considerations. The review summarizes complex dynamics, specific to the FFTF design evaluation, which particularly illustrate large available margins for decay heat removal (possibly involving coolant boiling). Also addressed are modelling considerations and the status of experimental programs. To date, design adequacy has been assessed largely by a combination of component tests and system analyses, with design margins increased to accommodate uncertainties. Presently planned tests, when completed, will provide an empirical base for characterizing system dynamic behavior in the range most relevant to current LMFBR designs and will demonstrate FFTF decay heat removal capability directly.
Date: October 1, 1979
Creator: Additon, S.L. & Bouchey, G.D.
Partner: UNT Libraries Government Documents Department

Decay Heat Calculations for PWR and BWR Assemblies Fueled with Uranium and Plutonium Mixed Oxide Fuel using SCALE

Description: In currently operating commercial nuclear power plants (NPP), there are two main types of nuclear fuel, low enriched uranium (LEU) fuel, and mixed-oxide uranium-plutonium (MOX) fuel. The LEU fuel is made of pure uranium dioxide (UO{sub 2} or UOX) and has been the fuel of choice in commercial light water reactors (LWRs) for a number of years. Naturally occurring uranium contains a mixture of different uranium isotopes, primarily, {sup 235}U and {sup 238}U. {sup 235}U is a fissile isotope, and will readily undergo a fission reaction upon interaction with a thermal neutron. {sup 235}U has an isotopic concentration of 0.71% in naturally occurring uranium. For most reactors to maintain a fission chain reaction, the natural isotopic concentration of {sup 235}U must be increased (enriched) to a level greater than 0.71%. Modern nuclear reactor fuel assemblies contain a number of fuel pins potentially having different {sup 235}U enrichments varying from {approx}2.0% to {approx}5% enriched in {sup 235}U. Currently in the United States (US), all commercial nuclear power plants use UO{sub 2} fuel. In the rest of the world, UO{sub 2} fuel is still commonly used, but MOX fuel is also used in a number of reactors. MOX fuel contains a mixture of both UO{sub 2} and PuO{sub 2}. Because the plutonium provides the fissile content of the fuel, the uranium used in MOX is either natural or depleted uranium. PuO{sub 2} is added to effectively replace the fissile content of {sup 235}U so that the level of fissile content is sufficiently high to maintain the chain reaction in an LWR. Both reactor-grade and weapons-grade plutonium contains a number of fissile and non-fissile plutonium isotopes, with the fraction of fissile and non-fissile plutonium isotopes being dependent on the source of the plutonium. While only RG plutonium is currently used in MOX, there ...
Date: October 2011
Creator: Ade, Brian J. & Gauld, Ian C.
Partner: UNT Libraries Government Documents Department

ADVANCED COMPUTATIONAL MODEL FOR THREE-PHASE SLURRY REACTORS

Description: In this project, an Eulerian-Lagrangian formulation for analyzing three-phase slurry flows in a bubble column was developed. The approach used an Eulerian analysis of liquid flows in the bubble column, and made use of the Lagrangian trajectory analysis for the bubbles and particle motions. The bubble-bubble and particle-particle collisions are included the model. The model predictions are compared with the experimental data and good agreement was found An experimental setup for studying two-dimensional bubble columns was developed. The multiphase flow conditions in the bubble column were measured using optical image processing and Particle Image Velocimetry techniques (PIV). A simple shear flow device for bubble motion in a constant shear flow field was also developed. The flow conditions in simple shear flow device were studied using PIV method. Concentration and velocity of particles of different sizes near a wall in a duct flow was also measured. The technique of Phase-Doppler anemometry was used in these studies. An Eulerian volume of fluid (VOF) computational model for the flow condition in the two-dimensional bubble column was also developed. The liquid and bubble motions were analyzed and the results were compared with observed flow patterns in the experimental setup. Solid-fluid mixture flows in ducts and passages at different angle of orientations were also analyzed. The model predictions were compared with the experimental data and good agreement was found. Gravity chute flows of solid-liquid mixtures were also studied. The simulation results were compared with the experimental data and discussed A thermodynamically consistent model for multiphase slurry flows with and without chemical reaction in a state of turbulent motion was developed. The balance laws were obtained and the constitutive laws established.
Date: October 1, 2004
Creator: Ahmadi, Goodarz
Partner: UNT Libraries Government Documents Department

ADVANCED COMPUTATIONAL MODEL FOR THREE-PHASE SLURRY REACTORS

Description: In the second year of the project, the Eulerian-Lagrangian formulation for analyzing three-phase slurry flows in a bubble column is further developed. The approach uses an Eulerian analysis of liquid flows in the bubble column, and makes use of the Lagrangian trajectory analysis for the bubbles and particle motions. An experimental set for studying a two-dimensional bubble column is also developed. The operation of the bubble column is being tested and diagnostic methodology for quantitative measurements is being developed. An Eulerian computational model for the flow condition in the two-dimensional bubble column is also being developed. The liquid and bubble motions are being analyzed and the results are being compared with the experimental setup. Solid-fluid mixture flows in ducts and passages at different angle of orientations were analyzed. The model predictions were compared with the experimental data and good agreement was found. Gravity chute flows of solid-liquid mixtures is also being studied. Further progress was also made in developing a thermodynamically consistent model for multiphase slurry flows with and without chemical reaction in a state of turbulent motion. The balance laws are obtained and the constitutive laws are being developed. Progress was also made in measuring concentration and velocity of particles of different sizes near a wall in a duct flow. The technique of Phase-Doppler anemometry was used in these studies. The general objective of this project is to provide the needed fundamental understanding of three-phase slurry reactors in Fischer-Tropsch (F-T) liquid fuel synthesis. The other main goal is to develop a computational capability for predicting the transport and processing of three-phase coal slurries. The specific objectives are: (1) To develop a thermodynamically consistent rate-dependent anisotropic model for multiphase slurry flows with and without chemical reaction for application to coal liquefaction. Also establish the material parameters of the model. (2) ...
Date: October 1, 2001
Creator: Ahmadi, Goodarz
Partner: UNT Libraries Government Documents Department

SLSF loop handling system. Volume I. Structural analysis. [LMFBR]

Description: SLSF loop handling system was analyzed for deadweight and postulated dynamic loading conditions, identified in Chapters II and III in Volume I of this report, using a linear elastic static equivalent method of stress analysis. Stress analysis of the loop handling machine is presented in Volume I of this report. Chapter VII in Volume I of this report is a contribution by EG and G Co., who performed the work under ANL supervision.
Date: October 1, 1978
Creator: Ahmed, H.; Cowie, A. & Ma, D.
Partner: UNT Libraries Government Documents Department

SLSF loop handling system. Volume III. AISC code evaluations and analysis of critical attachments. [LMFBR]

Description: SLSF loop handling system was analyzed for deadweight and postulated dynamic loading conditions using a linear elastic static equivalent method of stress analysis. Stress computations of Cradle and critical attachments per AISC Code guidelines are presented. HFEF is credited with in-depth review of initial phase of work.
Date: October 1, 1978
Creator: Ahmed, H.; Cowie, A.; Malek, R.A.; Rafer, A.; Ma, D. & Tebo, F.
Partner: UNT Libraries Government Documents Department

FAST FLUX TEST FACILITY. Design and development quality assurance requirements for the FFTF

Description: The document is presented to provide general management requirements for Pacific Northwest Laboratory (PNL) and contractor design and development quality assurance programs to assure the required quality level of the various items required for the FFTF. The document is applicable as imposed by the contract to FFTF contractors and subcontractors. The document is also applicable to PNL design and development activities related to the FFTF.
Date: October 23, 1968
Creator: Albert, W.G.
Partner: UNT Libraries Government Documents Department

Projections of spent fuel to be discharged by the U. S. nuclear power industry

Description: Calculated properties of spent fuel projected to be discharged and accumulated by the U.S. nuclear power industry through the year 2031 A.D. are presented. The projections are based on installed nuclear capacities of 380 and 543 GW(e) in the year 2000 and 2030, respectively. They include compilations of the grams of the elements, curies of radioactivity, thermal decay power, photon and neutron emission rates, and radiotoxicities of the assemblies that are accumulated at a Spent Unreprocessed Fuel Facility (SURFF), allowing for delays of 5 and 10 years before shipment to SURFF.
Date: October 1, 1977
Creator: Alexander, C.W.; Kee, C.W.; Croff, A.G. & Blomeke, J.O.
Partner: UNT Libraries Government Documents Department

OSMOSE experiment representativity studies.

Description: The OSMOSE program aims at improving the neutronic predictions of advanced nuclear fuels through measurements in the MINERVE facility at the CEA-Cadarache (France) on samples containing the following separated actinides: Th-232, U-233, U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-243, Cm-244 and Cm-245. The goal of the experimental measurements is to produce a database of reactivity-worth measurements in different neutron spectra for the separated heavy nuclides. This database can then be used as a benchmark for integral reactivity-worth measurements to verify and validate reactor analysis codes and integral cross-section values for the isotopes tested. In particular, the OSMOSE experimental program will produce very accurate sample reactivity-worth measurements for a series of actinides in various spectra, from very thermalized to very fast. The objective of the analytical program is to make use of the experimental data to establish deficiencies in the basic nuclear data libraries, identify their origins, and provide guidelines for nuclear data improvements in coordination with international programs. To achieve the proposed goals, seven different neutron spectra can be created in the MINERVE facility: UO2 dissolved in water (representative of over-moderated LWR systems), UO2 matrix in water (representative of LWRs), a mixed oxide fuel matrix, two thermal spectra containing large epithermal components (representative of under-moderated reactors), a moderated fast spectrum (representative of fast reactors which have some slowing down in moderators such as lead-bismuth or sodium), and a very hard spectrum (representative of fast reactors with little moderation from reactor coolant). The different spectra are achieved by changing the experimental lattice within the MINERVE reactor. The experimental lattice is the replaceable central part of MINERVE, which establishes the spectrum at the sample location. This configuration leads to a uniform well-behaved system so that the reactor configuration is in the fundamental mode. In fact, an important ...
Date: October 10, 2007
Creator: Aliberti, G. & Klann, R.
Partner: UNT Libraries Government Documents Department

Impact of spectral transition zone in reference ENIGMA configuration.

Description: The gas-cooled fast reactor (GFR) is one of six advanced nuclear energy systems being studied under the auspices of the Gen IV International Forum (GIF). In a bilateral International Nuclear Energy Research Initiative (I-NERI) project French and U.S. national laboratories, industry, and universities are collaborating on the development of the GFR. This effort is led by the ANL in the U.S. and the CEA in France. Some of the attractions of the GFR include: (1) Hard spectrum and core breeding ratio, BR {approx} 1. These features allow minimal waste production, improved transmutation capability, optimal and flexible use of natural resources, potentially better economy (because of use of higher power density relative to current thermal gas-cooled systems), and improved non-proliferation (no fertile blanket); (2) Temperature resistant fuel and structure elements that are favorable to tight fission product confinement and system operation at high temperature; (3) High temperature and transparent helium (He) gas coolant that allows a high thermodynamic conversion efficiency, other energy applications (e.g., hydrogen production), and ease of in-service inspection and repair; and (4) Possible direct energy conversion cycle leading to a simpler design, increased conversion efficiency, and lower investment costs. The French strategy for advanced systems includes the development of the GFR and sodium-cooled fast reactor (SFR) to levels that allow industries to be able to make an informed choice of the fast spectrum system that would provide a sustainable nuclear energy generation option for the future. Current planning calls for the construction of a small experimental research and technology development reactor (ETDR) around 2009 (first operation in 2015) at CEA-Cadarache, France. This would be followed by the construction of a GFR industrial prototype, around 2025. In support of the GFR development efforts, a new physics experimental program (called ENIGMA, Experimental Neutron Investigation of Gas-cooled reactor at Masurca) is ...
Date: October 5, 2005
Creator: Aliberti, G.; Palmiotti, G.; Taiwo, T. A. & Tommasi, J.
Partner: UNT Libraries Government Documents Department

IRIS Reactor a Suitable Option to Provide Energy and Water Desalination for the Mexican Northwest Region

Description: The Northwest region of Mexico has a deficit of potable water, along this necessity is the region growth, which requires of additional energy capacity. The IRIS reactor offers a very suitable source of energy given its modular size of 300 MWe and it can be coupled with a desalination plant to provide the potable water for human consumption, agriculture and industry. The present paper assess the water and energy requirements for the Northwest region of Mexico and how the deployment of the IRIS reactor can satisfy those necessities. The possible sites for deployment of Nuclear Reactors are considered given the seismic constraints and the closeness of the sea for external cooling. And in the other hand, the size of the desalination plant and the type of desalination process are assessed accordingly with the water deficit of the region.
Date: October 3, 2004
Creator: Alonso, G.; Ramirez, R.; Gomez, C. & Viais, J.
Partner: UNT Libraries Government Documents Department

Mark I 1/5-scale boiling water reactor pressure suppression experiment facility report

Description: An accurate Mark I /sup 1///sub 5/-scale, boiling water reactor (BWR), pressure suppression facility was designed and constructed at Lawrence Livermore Laboratory (LLL) in 11 months. Twenty-seven air tests using the facility are described. Cost was minimized by utilizing equipment borrowed from other LLL programs. The total value of borrowed equipment exceeded the program's budget of $2,020,000. Substantial flexibility in the facility was used to permit independent variation in the drywell pressure-time history, initial pressure in the drywell and toroidal wetwells, initial toroidal wetwell water level and downcomer length, vent line flow resistance, and vent line flow asymmetry. The two- and three-dimensional sectors of the toroidal wetwell provided significant data.
Date: October 11, 1977
Creator: Altes, R.G.; Pitts, J.H.; Ingraham, R.F.; Collins, E.K. & McCauley, E.W.
Partner: UNT Libraries Government Documents Department

3-D THERMAL EVALUATIONS FOR a FUELED EXPERIMENT in the ADVANCED TEST REACTOR

Description: The DOE Advanced Fuel Cycle Initiative and Generation IV reactor programs are developing new fuel types for use in the current Light Water Reactors and future advanced reactor concepts. The Advanced Gas Reactor program is planning to test fuel to be used in the Next Generation Nuclear Plant (NGNP) nuclear reactor. Preliminary information for assessing performance of the fuel will be obtained from irradiations performed in the Advanced Test Reactor large ''B'' experimental facility. A test configuration has been identified for demonstrating fuel types typical of gas cooled reactors or fast reactors that may play a role in closing the fuel cycle or increasing efficiency via high temperature operation Plans are to have 6 capsules, each containing 12 compacts, for the test configuration. Each capsule will have its own temperature control system. Passing a helium-neon gas through the void regions between the fuel compacts and the graphite carrier and between the graphite carrier and the capsule wall will control temperature. This design with three compacts per axial level was evaluated for thermal performance to ascertain the temperature distributions in the capsule and test specimens with heating rates that encompass the range of initial heat generation rates.
Date: October 6, 2004
Creator: Ambrosek, R. G.; Chang, G. S. & Utterbeck, D. J.
Partner: UNT Libraries Government Documents Department

K-TIF: a two-fluid computer program for downcomer flow dynamics. [PWR]

Description: The K-TIF computer program has been developed for numerical solution of the time-varying dynamics of steam and water in a pressurized water reactor downcomer. The current status of physical and mathematical modeling is presented in detail. The report also contains a complete description of the numerical solution technique, a full description and listing of the computer program, instructions for its use, with a sample printout for a specific test problem. A series of calculations, performed with no change in the modeling parameters, shows consistent agreement with the experimental trends over a wide range of conditions, which gives confidence to the calculations as a basis for investigating the complicated physics of steam-water flows in the downcomer.
Date: October 1, 1977
Creator: Amsden, A.A. & Harlow, F.H.
Partner: UNT Libraries Government Documents Department

Downflow heat transfer in a heated ribbed vertical annulus with a cosine power profile

Description: Experiments designed to investigate downflow heat transfer in a heated, ribbed annulus test section simulating one of the annular coolant channels of a Savannah River Plant production reactor Mark 22 fuel assembly have been conducted at the Idaho National Engineering Laboratory. The inner surface of the annulus was constructed of aluminum and was electrically heated to provide an axial cosine power profile and a flat azimuthal power shape. Data presented in this report are from the ECS-2c series, which was a follow on series to the ECS-2b series, conducted specifically to provide additional data on the effect of different powers at the same test conditions, for use in evaluation of possible power effects on the aluminum temperature measurements. Electrical powers at 90%, 100%, and 110% of the power required to result in the maximum aluminum temperature at fluid saturation temperature were used at each set of test conditions previously used in the ECS-2b series. The ECS-2b series was conducted in the same test rig as the previous ECS-2b series. Data and experimental description for the ECS-2b series is provided in a previous report. 18 refs., 25 figs., 3 tabs.
Date: October 1, 1991
Creator: Anderson, J.L.; Condie, K.G. & Larson, T.K.
Partner: UNT Libraries Government Documents Department

Power plant system assessment. Final report. SP-100 Program

Description: The purpose of this assessment was to provide system-level insights into 100-kWe-class space reactor electric systems. Using these insights, Rockwell was to select and perform conceptual design studies on a ''most attractive'' system that met the preliminary design goals and requirements of the SP-100 Program. About 4 of the 6 months were used in the selection process. The remaining 2 months were used for the system conceptual design studies. Rockwell completed these studies at the end of FY 1983. This report summarizes the results of the power plant system assessment and describes our choice for the most attractive system - the Rockwell SR-100G System (Space Reactor, 100 kWe, Growth) - a lithium-cooled UN-fueled fast reactor/Brayton turboelectric converter system.
Date: October 31, 1983
Creator: Anderson, R.V.; Atkins, D.F.; Bost, D.S.; Berman, B.; Clinger, D.A.; Determan, W.R. et al.
Partner: UNT Libraries Government Documents Department