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Behavior of fission product gases in HTGR fuel material

Description: This document is a status report on experimental data relating to the behavior of fission product gases in HTGR fuel material. It contains (1) a review of published experimental data on the behavior of gaseous fission products (noble gases, halogens, and related elements) in HTGR fuel materials and (2) new data on fission gas behavior. From the experimental data, the effects of fission product element, fuel configuration, nuclide half-life, temperature, pressure, neutron flux, neutron fluence, and fuel hydrolysis on fission gas release are deduced. Also on the basis of the experimental data, a model is formulated for the release of fission gases from fuel rods. The data presented in this report are being confirmed and extended through continuing work at General Atomic and other laboratories.
Date: October 1, 1977
Creator: Myers, B.F.; Baldwin, N.L.; Bell, W.E. & Burnette, R.D.
Partner: UNT Libraries Government Documents Department

Bulk Shielding Facility quarterly report, April, May, and June of 1977

Description: The BSR operated at an average power level of 1,821 kW for 39.82 percent of the time during April, May, and June. Water-quality control in both the reactor primary and secondary cooling systems was satisfactory. The BSR was operated at low and variable power during this quarter for 91.517 hours as part of the training programs for nuclear reactor operator trainees from Memphis State University and nuclear engineering students from the University of Tennessee. The PCA was also used in the above-mentioned training programs and was operated on twelve occasions when Central Florida Community College, the University of Kentucky, and Memphis State University students actively participated in training laboratories.
Date: October 1, 1977
Creator: Hurt, III, S. S.; Lance, E. D. & Thomas, J. R.
Partner: UNT Libraries Government Documents Department

Compendium of computer codes for the safety analysis of fast breeder reactors

Description: The objective of the compendium is to provide the reader with a guide which briefly describes many of the computer codes used for liquid metal fast breeder reactor safety analyses, since it is for this system that most of the codes have been developed. The compendium is designed to address the following frequently asked questions from individuals in licensing and research and development activities: (1) What does the code do. (2) To what safety problems has it been applied. (3) What are the code's limitations. (4) What is being done to remove these limitations. (5) How does the code compare with experimental observations and other code predictions. (6) What reference documents are available.
Date: October 1, 1977
Partner: UNT Libraries Government Documents Department

Conversion on an operating system oriented towards transaction processing from two to three modes of logical address space

Description: A computer system to control and acquire data from a set of ten neutron and x-ray scattering and diffraction experiments located at the High Flux Beam Reactor at Brookhaven National Laboratory has operated in a routine manner for over two years. The system was constructed according to a functionally distributed architecture, and thus consists of a set of functional nodes. Ten of these nodes, the private or application nodes, perform the function ''execute programs to control and acquire data from experiment number x.'' An additional function node, the common or shared service node, performs the function ''provide a set of shared services to the application nodes.'' The shared service node was successfully implemented in software with in-house code oriented towards transaction processing and in hardware with a Digital Equipment Corporation PDP-11/40 computer. However,recent demands that this node provide an expanded set of services have required that its implementation elements be modified and extended. In particular, the node hardware has been changed to a PDP-11/45 processor, and the software present at the node has been extended from operation in two modes of logical address space to three modes. A discussion of the systems analysis principles which influenced the manner in which these modifications and extensions were carried out is given. The structure of the old two-mode software is briefly reviewed in order to provide a basis for an examination of its three-mode replacement. Finally, the possibility of extending the technique to operation in many modes of logical address space is indicated. 1 figure, 4 tables.
Date: October 1, 1977
Creator: Stubblefield, F.W.
Partner: UNT Libraries Government Documents Department

Experiment data report for Semiscale Mod-1 Test S-28-1 (steam generator tube rupture test series)

Description: Recorded test data are presented for Test S-28-1 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-1 was conducted from initial conditions of 15 767 kPa and 557 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Sixty steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg.
Date: October 1, 1977
Creator: Collins, B.L.; Coppin, C.E. & Sackett, K.E.
Partner: UNT Libraries Government Documents Department

Experiment data report for Semiscale Mod-1 test S-28-3 (steam generator tube rupture test). [PWR]

Description: Recorded test data are presented for Test S-28-3 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-3 was conducted from initial conditions of 15621 kPa and 555 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Twelve steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg.
Date: October 1, 1977
Creator: Gillins, R.L. & Sackett, K.E.
Partner: UNT Libraries Government Documents Department

Experiment data report for semiscale Mod-1 test S-28-4 (steam generator tube rupture test)

Description: Recorded test data are presented for Test S-28-4 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-4 was conducted from initial conditions of 15 646 kPa and 557 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Thirty steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg.
Date: October 1, 1977
Creator: Esparza, V. & Sackett, K.E.
Partner: UNT Libraries Government Documents Department

In-pile intragranular densification of oxide fuels (AWBA Development Program)

Description: This report proposes a model to describe in-pile densification of oxide fuels, by both vacancy boil-off due to thermal excitation and vacancy knockout by the passage of fission fragments through the pores. The model includes the migration rates of both vacancies and interstitials to pores and the production of vacancy-rich damage cascades by fission fragments. It has been coupled with a previously reported swelling and gas release model so that it can predict the total dimensional changes of the fuel as well as predicting intragranular densification for both ThO/sub 2/ and UO/sub 2/ fuels for advanced water breeder reactor applications development effort.
Date: October 1, 1977
Creator: Dollins, C.C. & Nichols, F.A.
Partner: UNT Libraries Government Documents Department

Irradiation Effects Test Series: Test IE-3. Test results report. [PWR]

Description: The objectives of the test reported were to: (a) determine the behavior of irradiated fuel rods subjected to a rapid power increase during which the possibility of a pellet-cladding mechanical interaction failure is enhanced and (b) determine the behavior of these fuel rods during film boiling following this rapid power increase. Test IE-3 used four 0.97-m long pressurized water reactor type fuel rods fabricated from previously irradiated fuel. The fuel rods were subjected to a preconditioning period, followed by a power ramp to 69 kW/m at a coolant mass flux of 4920 kg/s-m/sup 2/. After a flow reduction to 2120 kg/s-m/sup 2/, film boiling occurred on the fuel rods. One rod failed approximately 45 seconds after the reactor was shut down as a result of cladding embrittlement due to extensive cladding oxidation. Data are presented on the behavior of these irradiated fuel rods during steady-state operation, the power ramp, and film boiling operation. The effects of a power ramp and power ramp rates on pellet-cladding interaction are discussed. Test data are compared with FRAP-T3 computer model calculations and data from a previous Irradiation Effects test in which four irradiated fuel rods of a similar design were tested. Test IE-3 results indicate that the irradiated state of the fuel rods did not significantly affect fuel rod behavior during normal, abnormal (power ramp of 20 kW/m per minute), and accident (film boiling) conditions.
Date: October 1, 1977
Creator: Farrar, L. C.; Allison, C. M.; Croucher, D. W. & Ploger, S. A.
Partner: UNT Libraries Government Documents Department

Irradiation test OF-2: high-temperature irradiation behavior of LASL-made fuel rods and LASL-made coated particles. [ZrC coated particles]

Description: Three LASL-made, substoichiometric ZrC-coated particles with inert kernels, and two high-density molded graphite fuel rods that contained LASL-made, ZrC-coated fissile particles were irradiated in the Oak Ridge Research Reactor test OF-2. The severest test conditions were 8.36 x 10/sup 21/ nvt (E greater than 0.18 MeV) at 1350/sup 0/C. The graphite matrix showed no effect of the irradiation. There was no interaction between the matrix and any of the particle coats. The loose ZrC coated particles with inert kernels showed no irradiation effects. The graded ZrC-C coats on the fissile particles were cracked. It is postulated that the cracking is associated with the low LTI deposition rate and is not related to the ZrC.
Date: October 1, 1977
Creator: Wagner, P.; Reiswig, R.D.; Hollabaugh, C.M.; White, R.W.; O'Rourke, J.A.; Davidson, K.V. et al.
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

LOFT three-beam densitometer data interpretation

Description: A calculation procedure has been devised to fit the best of several density distribution models, or a default model, to chordal average fluid density measurements from the LOFT gamma densitometer to obtain an estimate of the coolant density profile and the average density. The models used are homogeneous, continuous and discontinuous stratified, and continuous and discontinuous annular. The default model is a beam length weighted average for an estimate of the average density. A computer program for applying these calculations to LOFT test data has been written.
Date: October 1, 1977
Creator: Lassahn, G. D.
Partner: UNT Libraries Government Documents Department

Near-isotropic petroleum-coke based graphites for high temperature gas-cooled reactor core components

Description: The standard covers procurement requirements for extruded graphite logs, 15 in. (381 mm) or greater in diameter, manufactured with near-isotropic petroleum cokes and coal-tar pitch binders which are candidates or reference materials for replaceable fuel and reflector blocks for High-Temperature Gas-Cooled Reactors (HTGRs). The requirements are designed to produce the degree of lot-to-lot reproducibility which is required to ensure consistent and predictable properties and irradiation performance for specific graphite grades and to ensure traceability of the graphite logs to production processes and raw materials that affect performance. The standard is intended for use in the procurement of developmental and commercial grades of nuclear graphite which are to be evaluated on Department of Energy (DOE) funded programs for use as core components in HTGRs.
Date: October 1, 1977
Partner: UNT Libraries Government Documents Department

Numerical results obtained from the three dimensional transient single phase version of the COMMIX computer code

Description: Results from the first three applications of the three-dimensional, transient, single-phase version of the thermal-hydraulics computer code, COMMIX, are presented. The applications include an LMFBR outlet plenum, a horizontal pipe, and a 19-pin LMFBR hexagonal fuel assembly. Each case is described and the results presented. All three applications use liquid sodium as the fluid. The plenum and pipe analyses are transient, while the fuel assembly involves a steady-state analysis.
Date: October 1, 1977
Creator: Domanus, H.M.; Schmitt, R.C. & Sha, W.T.
Partner: UNT Libraries Government Documents Department

Posttest REALP4 analysis of LOFT experiment L1-3A

Description: This report presents selected results of posttest RELAP4 modeling of LOFT loss-of-coolant experiment L1-3A, a double-ended isothermal cold leg break with lower plenum emergency core coolant injection. Comparisons are presented between the pretest prediction, the posttest analysis, and the experimental data. It is concluded that pressurizer modeling is important for accurately predicting system behavior during the initial portion of saturated blowdown. Using measured initial conditions rather than nominal specified initial conditions did not influence the system model results significantly. Using finer nodalization in the reactor vessel improved the prediction of the system pressure history by minimizing steam condensation effects. Unequal steam condensation between the downcomer and core volumes appear to cause the manometer oscillations observed in both the pretest and posttest RELAP4 analysis.
Date: October 1, 1977
Creator: White, J.R. & Holmstrom, H.L.O.
Partner: UNT Libraries Government Documents Department

Preliminary results of CSTF aerosol behavior test, AB-1. [LMFBR]

Description: A large-scale aerosol behavior test (AB-1) was performed in the Containment Systems Test Facility (CSTF) containment vessel using sodium oxide aerosol generated by a pool of sodium burning in air. The purpose was to characterize the aerosol properties and compare the experimental results with computer code predictions. The 20-meter high CSTF vessel is by far the largest ever used in aerosol agglomeration studies and is approximately half-scale of large commercial reactor containment buildings for the parameters affecting particle agglomeration and settling.
Date: October 1, 1977
Creator: Hilliard, R.K.; McCormack, J.D.; Hassberger, J.A. & Muhlestein, L.D.
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

Reactor safety research programs. Quarterly progress report, 1 July--30 September 1977

Description: HTGR safety evaluation included studies on fission product release; materials, chemistry, and instrumentation; structural evaluation; and analytical safety evaluation. LMFBR work included accident sequence studies, technical coordination of structural integrity, and SSC code development and validation. LWR safety studies included thermal/hydraulic analysis; THOR code development; and appraisal of analysis development in foreign countries.
Date: October 1, 1977
Creator: Kouts, H.J.C. & Kato, W.Y.
Partner: UNT Libraries Government Documents Department

Safety status of HTGR structural ceramics

Description: The objective of this study is to assess the safety status of structural ceramics proposed for use in an HTGR. In this report the mechanical and physical properties of the proposed structural ceramics are reviewed, and the current design and material specifications are discussed. Potential safety problem areas are identified. Analytical models for creep deformation, devitrification, compatibility, fatigue, and thermal insulation stability are based on published work in the literature. The predictions made are generally based upon extrapolations to HTGR conditions. The data indicate that creep deformation in silicon nitride and alumina ceramics is acceptable if the purity and composition are properly controlled. Devitrification and creep in fused silica used in the core support structure and the thermal barrier system could be excessive. Thermodynamic data indicate that silicon nitride will be incompatible with the graphite seat used in the design. Under normal reactor operating conditions, or during start-up and shutdown, static fatigue and thermal shock effects in all three members of the core support structural ceramics could cause reductions in safety factors.
Date: October 1, 1977
Creator: Wei, G.C. & DiStefano, J.R.
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

Study of nuclei far from stability with TRISTAN II at the High Flux Beam Reactor at Brookhaven

Description: The ISOL facility TRISTAN II is described, and its expected capabilities on-line to the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory are discussed. In particular, the range of isotopes expected to be available and possible experimental studies of the short-lived fission-product isotopes are described. Background information consisting of an overview of the study of nuclei far from stability and a survey of existing ISOL facilities is presented in order to aid in evaluating the present status of the study of nuclei far from stability in general and the TRISTAN II facility in particular.
Date: October 1, 1977
Creator: Wohn, F.K.
Partner: UNT Libraries Government Documents Department

Assessment of a small pressurized water reactor for industrial energy

Description: An evaluation of several recent ERDA/ORNL sponsored studies on the application of a small, 365 MW(t) pressurized water reactor for industrial energy is presented. Preliminary studies have investigated technical and reliability requirements; costs for nuclear and fossil based steam were compared, including consideration of economic inflation and financing methods. For base-load industrial steam production, small reactors appear economically attractive relative to coal fired boilers that use coal priced at $30/ton.
Date: October 4, 1977
Creator: Klepper, O. H.; Fuller, L. C. & Myers, M. L.
Partner: UNT Libraries Government Documents Department

Review of FFTF and CRBRP control rod systems designs

Description: The evolution of the primary control rod system design for FFTF and CRBR, beginning with the initial choice of the basic concepts, is described. The significant component and systems tests are reviewed together with the test results which referenced the development of the CRBR primary control rod system design. Modifications to the concepts and detail designs of the FFTF control rod system were required principally to satisfy the requirements of CRBR, and at the same time incorporating design refinements shown desirable by the tests.
Date: October 4, 1977
Creator: Pitterle, T. A. & Lagally, H. O.
Partner: UNT Libraries Government Documents Department

Stress analysis of cylindrical pressure vessels with closely spaced nozzles by the finite-element method. Volume 1. Stress analysis of vessels with two closely spaced nozzles under internal pressure. [BWR; PWR; MULT-NOZZLE code]

Description: A finite-element computer program, MULT-NOZZLE, was developed for the stress analysis of cylindrical pressure vessels with two or three closely spaced reinforced nozzles. MULT-NOZZLE consists of two modules which may be operated independently. The first module, FEMG, automatically prepares a finite-element mesh including the nodal point coordinates, finite-element connectivities, mesh options, and boundary value specifications for input to the finite-element solution module SAP3M. SAP3M, which is a modified and improved version of the SAP3 computer program, computes the nodal point displacements and stress tensor components, and prints and/or stores the results for later postprocessing. The accuracy of the SAP3M module is demonstrated by comparison studies of two classical theory-of-elasticity problems: a simply supported beam and a thick-walled ring under internal pressure loading. A complete discussion of MULT-NOZZLE is presented in four volumes. Volume develops the finite-element idealization for pressure vessels with two idential radially attached closely spaced nozzles for internal pressure loading. The nozzles may be unreinforced or fully reinforced according to the rules of the ASME Boiler and Pressure Vessel Code and may be located in either a longitudinal or a transverse plane of the vessel. Validation of the program for analyzing this type of structure is demonstrated by the analysis of three two-nozzle pressure vessel models and comparison of results with experimental data. In general, quite satisfactory results were obtained.
Date: October 4, 1977
Creator: Tso, F.K.W.; Bryson, J.W.; Weed, R.A. & Moore, S.E.
Partner: UNT Libraries Government Documents Department