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UPDATE: nuclear power program information and data, July-September 1981

Description: UPDATE is published by the Office of Coordination and Special Projects, Office of Nuclear Reactor Programs, to provide a quick reference source on the current status of nuclear powerplant construction and operation in the United States and for information on the fuel cycle, economics, and performance of nuclear generating units. Similar information on other means of electric generation as related to nuclear power is included when appropriate. The subject matter of the reports and analyses presented in UPDATE will vary from issue to issue, reflecting changes in foci of interest and new developments in the field of commercial nuclear power generation. UPDATA is intended to provide a timely source of current statistics, results of analyses, and programmatic information proceeding from the activities of the Office of Nuclear Reactor Programs and other components of the Department of Energy, as well as condensations of topical articles from other sources of interest to the nuclear community. It also facilitates quick responses to requests for data and information of the type often solicited from this office.
Date: January 1, 1981
Creator: /NBM--6011986, DOE
Partner: UNT Libraries Government Documents Department

Computer simulation of LMFBR piping systems. [Accident conditions]

Description: Integrity of piping systems is one of the main concerns of the safety issues of Liquid Metal Fast Breeder Reactors (LMFBR). Hypothetical core disruptive accidents (HCDA) and water-sodium interaction are two examples of sources of high pressure pulses that endanger the integrity of the heat transport piping systems of LMFBRs. Although plastic wall deformation attenuates pressure peaks so that only pressures slightly higher than the pipe yield pressure propagate along the system, the interaction of these pulses with the different components of the system, such as elbows, valves, heat exchangers, etc.; and with one another produce a complex system of pressure pulses that cause more plastic deformation and perhaps damage to components. A generalized piping component and a tee branching model are described. An optional tube bundle and interior rigid wall simulation model makes such a generalized component model suited for modelling of valves, reducers, expansions, and heat exchangers. The generalized component and the tee branching junction models are combined with the pipe-elbow loop model so that a more general piping system can be analyzed both hydrodynamically and structurally under the effect of simultaneous pressure pulses.
Date: January 1, 1977
Creator: A-Moneim, M.T.; Chang, Y.W. & Fistedis, S.H.
Partner: UNT Libraries Government Documents Department

Confinement of airborne radioactivity. Final progress report, January-December 1978

Description: A new test method has been developed at the Savannah River Laboratory for evaluating the iodine retention capabilities of carbon used in the airborne-activity confinement system. Methyl iodide tagged with I-131 is injected into a test gas stream continuously for 5 hours with test conditions of 80/sup 0/C temperature, 95% relative humidity, and 55 feet per minute linear flow velocity. Results show that the CH/sub 3/I retention efficiency is independent of the inlet CH/sub 3/I concentration over the range of at least 0.9 to 200 ..mu..g/m/sup 3/ in the test gas stream. The method was also used to evaluate the effects of paint fumes on in-service carbons and showed that solvent exposure reduced carbon service life by 5 to 7 months. Experimental carbons both before and after service exposure in the SRP carbon test facility were also evaluated.
Date: February 1, 1980
Creator: A.G., Evans
Partner: UNT Libraries Government Documents Department

Interim assessment of the denatured /sup 233/U fuel cycle: feasibility and nonproliferation characteristics

Description: A fuel cycle that employs /sup 233/U denatured with /sup 238/U and mixed with thorium fertile material is examined with respect to its proliferation-resistance characteristics and its technical and economic feasibility. The rationale for considering the denatured /sup 233/U fuel cycle is presented, and the impact of the denatured fuel on the performance of Light-Water Reactors, Spectral-Shift-Controlled Reactors, Gas-Cooled Reactors, Heavy-Water Reactors, and Fast Breeder Reactors is discussed. The scope of the R, D and D programs to commercialize these reactors and their associated fuel cycles is also summarized and the resource requirements and economics of denatured /sup 233/U cycles are compared to those of the conventional Pu/U cycle. In addition, several nuclear power systems that employ denatured /sup 233/U fuel and are based on the energy center concept are evaluated.
Date: December 1, 1979
Creator: Abbott, L.S.; Bartine, D.E. & Burns, T.J. (eds.)
Partner: UNT Libraries Government Documents Department

Interim assessment of the denatured /sup 233/U fuel cycle: feasibility and nonproliferation characteristics

Description: A fuel cycle that employs /sup 233/U denatured with /sup 238/U and mixed with thorium fertile material is examined with respect to its proliferation-resistance characteristics and its technical and economic feasibility. The rationale for considering the denatured /sup 233/U fuel cycle is presented, and the impact of the denatured fuel on the performance of Light-Water Reactors, Spectral-Shift-Controlled Reactors, Gas-Cooled Reactors, Heavy-Water Reactors, and Fast Breeder Reactors is discussed. The scope of the R, D and D programs to commercialize these reactors and their associated fuel cycles is also summarized and the resource requirements and economics of denatured /sup 233/U cycles are compared to those of the conventional Pu/U cycle. In addition, several nuclear power systems that employ denatured /sup 233/U fuel and are based on the energy center concept are evaluated. Under this concept, dispersed power reactors fueled with denatured or low-enriched uranium fuel are supported by secure energy centers in which sensitive activities of the nuclear cycle are performed. These activities include /sup 233/U production by Pu-fueled transmuters (thermal or fast reactors) and reprocessing. A summary chapter presents the most significant conclusions from the study and recommends areas for future work.
Date: December 1, 1978
Creator: Abbott, L.S.; Bartine, D.E. & Burns, T.J. (eds.)
Partner: UNT Libraries Government Documents Department

Fast reactor analytical shielding progress report for July and August 1973. 189a No. 10318, activity No. HN 04 01 01 1

Description: Analytical shielding work performed for the fast reactor program during the months of July and August included calculations for both the FFTF and the Demo plant, computer code development, analyses of TSF experiments, and a cross- section sensitivity study of a TSF experiment. Most of the FFTF studies were aimed at determining the effect of various design changes on the dose rates at the maintenance deck during reactor operation, in particular, the effects of penetrations through the radial cavity shield and the effect of a new design for the head temperature control system shield. Gamma-ray dose rates due to the activated sodium coolant were also calculated for two locations in the system: around one of the branch-arm piping shields and near the entrances of the 28-in. and 16-in. coolant lines into the IHX cell. The main Demo plant study was for the lower axial shield region; another study has been initiated to calculate the Demo stored-fuel power and the ex-vessel detector response. The code development consisted in setting forth a set of specifications for the fourth version of the discrete ordinates code DOT. The TSF experiments that were analyzed were an experiment in which secondary gammaray production in iron-polyethylene configurations was investigated and another in which the transmission of the Westinghouse spectrum modifier and 5-in. radial blanket was measured. The sensitivtty study was performed to demonstrate that estimates of the accuracy of calculations for the FFTF and Demo plant can be deduced from a sensitivity study of an experimental configuration that only roughly mocks up the real system. (auth)
Date: November 1, 1973
Creator: Abbott, L.S.; Childs, R.L.; Engle, W.W. & Maerker, R.E.
Partner: UNT Libraries Government Documents Department

Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

Description: During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor.
Date: January 1, 1982
Creator: Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J. & Slater, C.O.
Partner: UNT Libraries Government Documents Department

A nonlinear dynamic model of a once-through, helical-coil steam generator

Description: A dynamic model of a once-through, helical-coil steam generator is presented. The model simulates the advanced liquid metal reactor superheated cycle steam generator with a four-region, moving-boundary, drift-flux model. The model is described by a set of nonlinear differential equations derived from the fundamental equations of conversation of mass, energy, and momentum. Sample results of steady-state and transient calculations are presented.
Date: July 1, 1993
Creator: Abdalla, M. A.
Partner: UNT Libraries Government Documents Department

Growth of molten core debris pools in concrete. Part II. A. Pool growth in composite beds; B. Effect of overlaying steel layers. Final report, March 1, 1978-September 30, 1979. [LMFBR]

Description: The heat and mass transfer processes taking place in molten core debris/concrete systems have been experimentally investigated. Two types of experiments have been conducted. The first experiment simulates the growth of a molten debris pool in a composite sacrificial bed. This experiment models debris pool growth in an inner, low-melting point, sacrificial material zone followed by a melting attack on the concrete bed. The purpose of the inner zone is to quickly melt and dilute the debris pool so that its subsequent downward growth in the concrete may be slowed. In the second experiment a two-layer immiscible liquid system is volumetrically heated and allowed to melt into a low-density gas releasing solid bed which is miscible in the initially-higher-density bottom liquid. The solid melts, mixes with, and dilutes the bottom liquid pool until its density is lower than that of the top liquid.
Date: December 26, 1979
Creator: Abdel-Khalik, S I
Partner: UNT Libraries Government Documents Department

Growth of molten core debris pools in concrete. Progress report, April 1--June 30, 1979. [LMFBR]

Description: The heat and mass transfer processes taking place in molten core debris/concrete systems have been experimentally investigated. Two types of experiments have been conducted. The first experiment simulates the growth of a molten debris pool in a composite sacrificial bed. This experiment models debris pool growth in an inner, low-melting point, sacrificial material zone followed by a melting attack on the concrete bed. The purpose of the inner zone is to quickly melt and dilute the debris pool so that its subsequent downward growth in the concrete may be slowed. In the second experiment a two-layer immiscible liquid system is volumetrically heated and allowed to melt into a low-density gas-releasing solid bed which is miscible in the initially-higher-density bottom liquid. The solid melts, mixes with, and dilutes the bottom liquid pool until its density is lower than that of the top liquid. At this time pool inversion occurs and the immiscible liquid sinks to the bottom of the pool displacing the now lighter fuel-concrete simulant.
Date: July 1, 1979
Creator: Abdel-Khalik, S I
Partner: UNT Libraries Government Documents Department

Growth of molten core debris pools in concrete. Progress report, March 1, 1977--November 30, 1977. [LMFBR]

Description: Experiments have been conducted using a volumetrically-heated pool with noncondensable gas injection at the boundaries to simulate the heat transfer processes taking place in molten core debris/concrete systems. Measurements of the upward, downward, and sideward heat transfer rates have been made over wide ranges of internal Rayleigh number (6.0 x 10/sup 5/ < Ra < 2.7 x 10/sup 9/), superficial gas velocity (0 < V/sub g/ < 2.5 cm/min), and pool aspect ratio (1.32 < W/L < 5.16). Pools with either an isothermal solid upper boundary or a free surface have been examined. Detailed measurements of the temperature distribution within the pools have also been made. The results indicate that the downward heat transfer rates are significantly increased while the sideward heat transfer rates are slightly affected by gas injection. The heat transfer rates at the pool boundaries are sensitive to the upper surface boundary condition. Efforts are continuing on the second part of the present study where the phase-change and gas evolution processes are combined.
Date: November 1, 1977
Creator: Abdel-Khalik, S I
Partner: UNT Libraries Government Documents Department

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

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

Radionuclide characterization at US commercial light-water reactors for decommissioning assessment: Distributions, inventories, and waste disposal considerations

Description: A continuing research program, conducted by Pacific Northwest Laboratory for the US Nuclear Regulatory Commission, characterizing radionuclide concentrations associated with US light-water reactors has been conducted for more than a decade. The research initially focused upon sampling and analytical measurements for the purpose of establishing radionuclide distributions and inventories for decommissioning assessment, since very little empirical data existed. The initial phase of the research program examined radionuclide concentrations and distributions external to the reactor vessel at seven US light water reactors. Later stages of the research program have examined the radionuclide distributions in the highly radioactive reactor internals and fuel assembly. Most recently, the research program is determining radionuclide concentrations in these highly radioactive components and comparing empirical results with those derived from the several nonempirical methodologies employed to estimate radionuclide inventories for disposal classification. The results of the research program to date are summarized, and their implications and significance for the decommissioning process are noted.
Date: September 1, 1992
Creator: Abel, K.H.; Robertson, D.E. & Thomas, C.W.
Partner: UNT Libraries Government Documents Department

International nuclear waste management fact book

Description: The International Nuclear Waste Management Fact Book has been compiled to provide current data on fuel cycle and waste management facilities, R and D programs, and key personnel in 24 countries, including the US; four multinational agencies; and 20 nuclear societies. This document, which is in its second year of publication supersedes the previously issued International Nuclear Fuel Cycle Fact Book (PNL-3594), which appeared annually for 12 years. The content has been updated to reflect current information. The Fact Book is organized as follows: National summaries--a section for each country that summarizes nuclear policy, describes organizational relationships, and provides addresses and names of key personnel and information on facilities. International agencies--a section for each of the international agencies that has significant fuel cycle involvement and a list of nuclear societies. Glossary--a list of abbreviations/acronyms of organizations, facilities, and technical and other terms. The national summaries, in addition to the data described above, feature a small map for each country and some general information that is presented from the perspective of the Fact Book user in the US.
Date: November 1995
Creator: Abrahms, C. W.; Patridge, M. D. & Widrig, J. E.
Partner: UNT Libraries Government Documents Department

Radioactive waste management at a Liquid Metal Fast Breeder Reactor

Description: This paper presents the radioactive waste production and management at a Liquid Metal Fast Breeder Reactor-II (EBR-II), which is operated for the US Department of Energy by the Argonne National Laboratory at the Idaho National Engineering Laboratory (INEL). Since this facility, in addition to supplying power has been used to demonstrate the breeder, fuel cycling, and recently operations with defective fuel elements, various categories of waste have been handled safely over some 14 years of operation. Liquid wastes are processed such that the resulting effluent can be discharged to an uncontrolled area. Solid wastes up to 10,000 R/hr are packaged and shipped contamination-free to a disposal site or interim storage with exposures to personnel approximately 10 mrem. Gaseous waste discharges are low such as 143 Ci of noble gases in 1978 and do not have a significant effect on the environment even with operations with breached fuel.
Date: January 1, 1979
Creator: Abrams, C.S.; Fryer, R.H. & Witbeck, L.C.
Partner: UNT Libraries Government Documents Department

Calculational limitations in PWR system simulation

Description: Engineering transient analysis codes, which are in general more accurate than the present generation of simulator software, can be expected to yield reasonably accurate results (+-20% or so on system pressure) if carefully utilized and if the two-phase and transient flow conditions are not severe. As the severity of the transient increases, the confidence that one may have in the results decreases. None of the existing engineering analysis codes is well assessed or verified for transient analysis, but all give qualitatively the same results lending credence to their results. Recent comparisons to transients in LOFT and SEMISCALE are encouraging as are various comparisons to actual plant data.
Date: January 1, 1982
Creator: Abramson, P.B.; Kennedy, M.F. & Speis, T.P.
Partner: UNT Libraries Government Documents Department

Optical probe for local void fraction and interface velocity measurements. [BWR; PWR]

Description: In view of the importance of obtaining unsteady local void fraction and interface velocities in liquid-vapor two-phase flows, an optical probe with a controlled tip geometry was developed and is described. In order to minimize the disturbances caused to the flow field by the presence of the probe, its dimensions have been miniaturized. The electronic and hydrodynamic response of the probe were investigated experimentally. The probe was found to be sensitive to both the interface velocities and the phase present at the probe tip. A possible explanation for the behavior of the probe is presented. Within the velocity range checked and with proper calibration, the optical probe developed can be used to determine both local void fractions and interface velocities.
Date: March 1, 1978
Creator: Abuaf, N.; Jones, O.C. Jr. & Zimmer, G.A.
Partner: UNT Libraries Government Documents Department

Fission product plateout/liftoff/washoff test plan. Revision 1

Description: A test program is planned in the COMEDIE loop of the Commissariat a l`Energy Atomique (CEA), Grenoble, France, to generate integral test data for the validation of computer codes used to predict fission product transport and core corrosion in the Modular High Temperature Gas-Cooled Reactor (MHTGR). The inpile testing will be performed by the CEA under contract from the US Department of Energy (DOE); the contract will be administered by Oak Ridge National Laboratory (ORNL). The primary purpose of this test plan is to provide an overview of the proposed program in terms of the overall scope and schedule. 8 refs, 3 figs.
Date: May 1, 1988
Creator: Acharya, R. & Hanson, D.
Partner: UNT Libraries Government Documents Department

Oxidation of 316 stainless steel and other alloys in prototypic GCFR environments

Description: The oxidation behavior of type 316 stainless steel and candidate advanced alloys for the gas-cooled fast reactor (GCFR) is being investigated at General Atomic Company. The test program consists of oxidation tests in prototypic GCFR environments. Two tests have been completed to date and a third test is under way. The first test was performed in an environment containing a hydrogen/water ratio of 10. The oxidation behavior of all the alloys was good to excellent in this environment. Preferential oxidation of chromium was responsible for this behavior. The second test was performed in an environment containing a hydrogen/water ratio of 0.25, where both chromium and iron oxides are thermodynamically stable. Some of the alloys and some of the ribbed type 316 stainless steel test specimens showed unacceptable oxidation resistance in this environment. In the third test, presently under way, two different pretreatment procedures are being used to control the poor oxidation behavior observed in the second test. Early results show some degree of success.
Date: May 1, 1980
Creator: Acharya, R.T.
Partner: UNT Libraries Government Documents Department

Gas-cooled fast breeder reactor steady-state irradiation testing program

Description: The requirements for the gas-cooled fast breeder reactor irradiation program are specified, and an irradiation program plan which satisfies these requirements is presented. The irradiation program plan consists of three parts and includes a schedule and a preliminary cost estimate: (1) a steady-state irradiation program, (2) irradiations in support of the design basis transient test program, and (3) irradiations in support of the GRIST-2 safety test program. Data from the liquid metal fast breeder reactor program are considered, and available irradiation facilities are examined.
Date: August 1, 1980
Creator: Acharya, R.T.; Campana, R.J. & Langer, S.
Partner: UNT Libraries Government Documents Department

Process to remove rare earth from IFR electrolyte

Description: The invention is a process for the removal of rare earths from molten chloride electrolyte salts used in the reprocessing of integrated fast reactor fuel (IFR). The process can be used either continuously during normal operation of the electrorefiner or as a batch process. The process consists of first separating the actinide values from the salt before purification by removal of the rare earths. After replacement of the actinides removed in the first step, the now-purified salt electrolyte has the same uranium and plutonium concentration and ratio as when the salt was removed from the electrorefiner.
Date: January 1, 1992
Creator: Ackerman, J. P. & Johnson, T. R.
Partner: UNT Libraries Government Documents Department

Waste removal in pyrochemical fuel processing for the Integral Fast Reactor

Description: Electrorefining in a molten salt electrolyte is used in the Integral Fast Reactor fuel cycle to recover actinides from spent fuel. Processes that are being developed for removing the waste constituents from the electrorefiner and incorporating them into the waste forms are described in this paper. During processing, halogen, chalcogen, alkali, alkaline earth, and rare earth fission products build up in the molten salt as metal halides and anions, and fuel cladding hulls and noble metal fission products remain as metals of various particle sizes. Essentially all transuranic actinides are collected as metals on cathodes, and are converted to new metal fuel. After processing, fission products and other waste are removed to a metal and a mineral waste form. The metal waste form contains the cladding hulls, noble metal fission products, and (optionally) most rare earths in a copper or stainless steel matrix. The mineral waste form contains fission products that have been removed from the salt into a zeolite or zeolite-derived matrix.
Date: January 1, 1994
Creator: Ackerman, J. P.; Johnson, T. R. & Laidler, J. J.
Partner: UNT Libraries Government Documents Department