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Quarterly progress report on high-temperature gas-cooled reactor safety studies sponsored by the NRC Division of Reactor Safety Research for October--December 1975

Description: Research progress is briefly reported for the following tasks: reheater and steam generator model development, development of the nuclear steam supply system simulation code, core simulation for enemergency cooling analysis, core auxiliary cooling system model development, and computer code implementation.
Date: February 1, 1976
Creator: Sanders, J.P.
Partner: UNT Libraries Government Documents Department

Monthly progress report for January 1976 for the HTGR safety studies for the Division of Systems Safety, U.S. Nuclear Regulatory Commission

Description: Progress is reported in the areas of thermal analysis of HTGR systems, axial mode spacing affects in flow distribution calculations, and performance calculations for the core auxiliary heat exchanger.
Date: February 1, 1976
Creator: Sanders, J.P.
Partner: UNT Libraries Government Documents Department

Monthly progress report for December 1975 for the HTGR safety studies for the Division of Technical Review U.S. Nuclear Regulatory Commission

Description: Brief highlights are presented for the following activities: HEXERI code development; FLOOIS code development; analysis of thermal barrier cover plates; and determination of physical property values for helium-air mixtures. (JWR)
Date: January 1, 1976
Creator: Sanders, J.P.
Partner: UNT Libraries Government Documents Department

Summary progress report for fiscal year 1976 and the transition quarter describing technical assistance work for the Division of Systems Safety, U. S. Nuclear Regulatory Commission. [HTGR]

Description: The report reviews briefly the HTGR core analytical methods that were developed during the course of the program. The features of these analytical methods are compared with methods used to perform similar analyses, and examples of the use of these methods are cited. Included are discussions of HEATUP (a computer code for the thermal analysis of an LOFC accident in an HTGR), HEATING 5 (an IBM 360 heat-conduction code), CCCM (a coupled conduction-convection model for core thermal analysis), FLODIS (a computer model to determine the flow distribution and thermal response of the Vrain reactor), and HEXEREI 2 code development. (DG)
Date: January 3, 1977
Creator: Sanders, J. P.
Partner: UNT Libraries Government Documents Department

Location of test bundle instrumentation and anticipated experimental values for the CFTL AG-1 bundle

Description: The placement of instrumentation within the CFTL (Core Flow Test Loop) AG-1 test section to meet the following objective is described. The objectives are threefold: (1) to provide values for the evaluation of the performance of the test section, (2) to compare the experimental data with values determined by pretest calculations to indicate the approach to conditions that can lead to a bundle failure, and (3) to acquire data during testing that will form a data base for subsequent use in the verification of computational procedures used in the licensing of the Gas Cooled Fast Reactor. Anticipated values for the various instruments have been determined using the computational procedure, SAGAPO, modified for the AG-1 geometry. These results are used as the basis for the specification of differential pressure cells and the range of readings anticipated from the thermocouples. Part of the results for the full flow, full power case are presented.
Date: January 1, 1980
Creator: Sanders, J.P.
Partner: UNT Libraries Government Documents Department

PRELIMINARY STUDY OF INTERMEDIATE-ENERGY RESEARCH REACTORS

Description: A limited investigation was made of possible reactor configurations that would contuin a region having a high intermediate-energy (1 to 10/sup 3/ ev) flux relative to the fast and thermal flux. All calculations were performed with a multigroup, multiregion reactor program (GNU-II). Consideration of a single- region, unreflected sphere indicated that the desired flux distribution could not be generated in a region containing fuel because of its resonance absorption. A flux-trap configuration was investigated with a cyclindrical iron core surrounded by a fuel annulus. Calculations indicated that the fast flux would not be depressed significantly in the central region unless the dimensions were large. Consideration was then given to an iron reflector for an infinite slab reactor. Good agreement was obtained between reported experimental and calculated relaxation lengths for the flux at different energies. It was found that the desired flux-energy distribution was approached at a point some 20 to 30 cm ixto the iron slab from the fuel; however, the magnitude of the flux relative to that in the fuel region was low. (auth)
Date: September 1, 1959
Creator: Sanders, J.P.
Partner: UNT Libraries Government Documents Department

Thermal response of core and central-cavity components of a high-temperature gas-cooled reactor in the absence of forced convection coolant flow. [NATCON code]

Description: A means of determining the thermal responses of the core and the components of a high-temperature gas-cooled reactor after loss of forced coolant flow is discussed. A computer program, using a finite-difference technique, is presented together with a solution of the confined natural convection. The results obtained are reasonable and demonstrate that the computer program adequately represents the confined natural convection.
Date: September 1, 1976
Creator: Whaley, R. L. & Sanders, J. P.
Partner: UNT Libraries Government Documents Department

INGRES: a computer code for the rate of air ingress into an HTGR following a design-basis depressurization accident

Description: The computer program INGRES was written to calculate the rate of air ingress into the prestressed concrete reactor vessel after a design-basis depressurization accident in a high-temperature gas-cooled reactor. The model includes the free convection loop that can occur in a cold-leg break, the expansion and contraction air exchange mechanisms, and the conversion of oxygen to carbon monoxide. Results are presented for the 2000-MW(t) Summit Power Station and the 3000-MW(t) Fulton Generating Station and are compared to computational results provided by the General Atomic Company. The results agree reasonably well even though some differences exist in the two models. (auth)
Date: December 1, 1975
Creator: Reid, R.L. & Sanders, J.P.
Partner: UNT Libraries Government Documents Department

High temperature, high pressure gas loop - the Component Flow Test Loop (CFTL)

Description: The high-pressure, high-temperature, gas-circulating Component Flow Test Loop located at Oak Ridge National Laboratory was designed and constructed utilizing Section III of the ASME Boiler and Pressure Vessel Code. The quality assurance program for operating and testing is also based on applicable ASME standards. Power to a total of 5 MW is available to the test section, and an air-cooled heat exchanger rated at 4.4 MW serves as heat sink. The three gas-bearing, completely enclosed gas circulators provide a maximum flow of 0.47 m/sup 3//s at pressures to 10.7 MPa. The control system allows for fast transients in pressure, power, temperature, and flow; it also supports prolonged unattended steady-state operation. The data acquisition system can access and process 10,000 data points per second. High-temperature gas-cooled reactor components are being tested.
Date: January 1, 1984
Creator: Gat, U.; Sanders, J.P. & Young, H.C.
Partner: UNT Libraries Government Documents Department

Determination of friction factors and heat transfer coefficients for flow past artifically roughened surfaces

Description: A critical review of the assumptions, theoretical foundations, and supporting experimental evidence for the analytical procedures in current use for evaluation of the effects of artificial surface roughening on friction factor and Stanton number is provided. Recommendations are given concerning the application of these procedures to rough rod bundles. A new method is demonstrated for determination of the slope and intercept of the universal logarithmic dimensionless velocity distribution law for fully rough flow past roughened surfaces without the need for experimental measurement of the velocity profile. The slope is shown to vary with the nature of the roughened surface and to deviate significantly from the slope for turbulent flow past smooth walls in some cases. It is further shown that the intercept, which is a boundary condition equivalent to the roughness parameter for friction, is independent of the width of the velocity profile. A similar method is developed for determination of the slope and intercept of the temperature distribution law, but additional experimental investigation is required before the efficacy of this application can be conclusively established.
Date: November 1, 1979
Creator: Hodge, S.A.; Sanders, J.P. & Klein, D.E.
Partner: UNT Libraries Government Documents Department

Operating experience with gas-bearing circulators in a high-pressure helium loop

Description: A high-pressure engineering test loop has been designed and constructed at the Oak Ridge National Laboratory for circulating helium through a test chamber at temperatures to 1000/sup 0/C. The purpose of this loop is to determine the thermal and structural performance of proposed components for the primary loops of gas-cooled nuclear reactors. Five MW of power is available to provide the required gas temperature at the test chamber, and an air-cooled heat exchanger, rated at 4.4 MW, serves as a heat sink. This report contains results of tests performed on gas-bearing circulators.
Date: January 1, 1987
Creator: Sanders, J.P.; Gat, Uri & Young, H.C.
Partner: UNT Libraries Government Documents Department

Gas erosion of impeller housing in the operation of a high-temperature, high-pressure helium circulator

Description: Three gas-bearing circulators are installed in series in a high-pressure, high-temperature loop to provide helium flow up to 0.47 m/sup 3//s at a total head of 78 kJ/kg. The design pressure is 10.7 MPa, and temperatures of 1000/sup 0/C can be obtained in the test section. The inlet temperature to the circulators is limited to 450/sup 0/C. During a routine examination of the circulator, deep V-shaped grooves were found in the stationary surface of this cavity. At the same time, a very fine, dark particulate was observed in crevices of the housing. At first it was assumed that the grooves were formed by particulate erosion; however, examination of the grooves and discussions with persons experienced with large circulator operation changed this opinion. Erosion caused by particulate is characteristically rounded on the bottom and has a greater width to depth aspect than the V-shaped grooves, which were observed. Analysis of the particulate indicated that it was essentially the material of the housing that had undergone reactions with impurities in the circulating gas. It was subsequently concluded that the impeller housing had not been heat treated in a sufficiently oxidizing atmosphere after machining to form an adherent oxide coating. This suboxide coating was eroded by the shear forces in the gas. The exposed layer of metal was then further oxidized by the impurities in the gas, and these layers of oxide were successively eroded to produce the grooves. This erosion problem was eliminated by machining a ring of the same material, heat treating it to form an adherent stable oxide, and bolting it in place in the cavity.
Date: January 1, 1987
Creator: Sanders, J.P.; Heestand, R.L. & Young, H.C.
Partner: UNT Libraries Government Documents Department

High-temperature gas-cooled reactor safety studies. Progress report for January 1, 1974--June 30, 1975

Description: Progress is reported in the following areas: systems and safety analysis; fission product technology; primary coolant technology; seismic and vibration technology; confinement components; primary system materials technology; safety instrumentation; loss of flow accident analysis using HEATUP code; use of coupled-conduction-convection model for core thermal analysis; development of multichannel conduction-convection program HEXEREI; cooling system performance after shutdown; core auxiliary cooling system performance; development of FLODIS code; air ingress into primary systems following DBDA; performance of PCRV thermal barrier cover plates; temperature limits for fuel particle coating failure; tritium distribution and release in HTGR; energy release to PCRV during DBDA; and mathematical models for HTGR reactor safety studies.
Date: July 1, 1977
Creator: Cole, T. E.; Sanders, J. P. & Kasten, P. R.
Partner: UNT Libraries Government Documents Department

Evaluation of the General Atomic codes TAP and RECA for HTGR accident analyses

Description: The General Atomic codes TAP (Transient Analysis Program) and RECA (Reactor Emergency Cooling Analysis) are evaluated with respect to their capability for predicting the dynamic behavior of high-temperature gas-cooled reactors (HTGRs) for postulated accident conditions. Several apparent modeling problems are noted, and the susceptibility of the codes to misuse and input errors is discussed. A critique of code verification plans is also included. The several cases where direct comparisons could be made between TAP/RECA calculations and those based on other independently developed codes indicated generally good agreement, thus contributing to the credibility of the codes.
Date: April 4, 1978
Creator: Ball, S.J.; Cleveland, J.C. & Sanders, J.P.
Partner: UNT Libraries Government Documents Department

Inelastic analysis of two plates under deformation dependent loads

Description: Cover plates are used in current designs for high temperature gas-cooled reactors to compress the mineral fiber insulation against the inside of the liner of the prestressed concrete pressure vessel. In the upper plenum, these plates are hexagonal and specified as carbon steel; in the lower cross ducts, the plates are square and made of Hastelloy X. The General Atomic Company has specified both damage and safety limit criteria for these plates. These plates have been analyzed at these limits using the inelastic finite element computer program EPACA. The results indicate that the total strains for the square plate were within the specified values; however, the maximum deformations at the free corners indicate separation from the insulation and failure to achieve one of the design requirements. Since no material data were available for carbon steel at the limiting temperatures, it was assumed that the hexagonal plates were constructed of 2$sup 1$/$sub 4$ percent Cr--1 percent Mo material. Although this material was found to produce satisfactory performance, extrapolation of available information would lead to the conclusion that the performance of carbon steel plates would not be satisfactory at the specified conditions. (auth)
Date: February 1, 1976
Creator: Iskander, S.K.; Collins, C.W. & Sanders, J.P.
Partner: UNT Libraries Government Documents Department

Core-support performance test in the component-flow test loop

Description: The CFTL is a closed-circuit, out-of-pile loop circulating helium at temperatures and pressures anticipated in gas-cooled reactors at flows sufficiently large to perform engineering-scale tests. It has the present capability for fast data acquisition and the control and measurement of gaseous impurities, and it has the potential to perform controlled rapid transients in pressure, flow, and power. The initial HTGR component to be tested in the CFTL is the core support structure for the prismatic bed HTGR. This structure has vertical posts mating with post seats, each with spherical curvatures of different radii. At the point of contact, Hertzian stress concentrations are produced. Under the load of the weight of the core plus the pressure gradient, the graphite will deform until the stress is below its yield limit. The Core Support Performance Test (CSPT) will subject this interface, using actual materials and geometry, to impure helium at HTGR operating temperatures, pressures, and flows under a simulated structural load. The concentration of water, hydrogen, and carbon dioxide will be controlled so that six months of test operation will simulate 40 years of reactor operation. The specification of this concentration is based on existing studies involving small graphite samples exposed to a variety of conditions at a few atmospheres. The extrapolation to concentrations that will duplicate both the amount and the nature of the corrosion is based on the oxidation kinetics of the Gadsby equation as parameterized by Velasquez. Ports are provided in the test vessel for in situ inspection of the graphite during the test period. Post-test examination of the structure will be used to correlate its performance with available computational methods.
Date: January 1, 1982
Creator: Sanders, J.P.; Grindell, A.G. & Eatherly, W.P.
Partner: UNT Libraries Government Documents Department

Determination of the slope and intercept of the universal velocity profile from pressure loss measurements

Description: A method is demonstrated for determination of the slope and intercept of the universal velocity distribution law for flow past roughened surfaces without the need for measurement of the velocity profile. The slope is shown to vary with the nature of the roughened surface and in some cases to deviate considerably from that for turbulent flow past smooth walls. It is further shown that the intercept, commonly known as the roughness parameter R(h/sup +/), is independent of the width of the velocity profile. The dependence noted by previous investigators was due to an assumption that the slope of the law of the wall for roughened surfaces is constant and equal to that for smooth surfaces.
Date: January 1, 1979
Creator: Hodge, S. A.; Sanders, J. P. & Conklin, J. C.
Partner: UNT Libraries Government Documents Department

Evaluation of error bands and confidence limits for thermal measurements in the CFTL bundle

Description: Surface cladding temperatures for the fuel rod simulators in the Core Flow Test Loop (CFTL) must be inferred from a measurement at a thermocouple junction within the rod. This step requires the evaluation of the thermal field within the rod based on known parameters such as heat generation rate, dimensional tolerances, thermal properties, and contact coefficients. Uncertainties in the surface temperature can be evaluated by assigning error bands to each of the parameters used in the calculation. A statistical method has been employed to establish the confidence limits for the surface temperature from a combination of the standard deviations of the important parameters. This method indicates that for a CFTL fuel rod simulator with a total power of 38 kW and a ratio of maximum to average axial power of 1.21, the 95% confidence limit for the calculated surface temperature is +- 45/sup 0/C at the midpoint of the rod.
Date: January 1, 1979
Creator: Childs, K. W.; Sanders, J. P. & Conklin, J. C.
Partner: UNT Libraries Government Documents Department

High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

Description: During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.
Date: June 1, 1983
Creator: Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E. & Sanders, J.P.
Partner: UNT Libraries Government Documents Department

Gas-cooled reactor programs. High-temperature gas-cooled reactor technology development program. Annual progress report, December 31, 1983

Description: ORNL continues to make significant contributions to the national program. In the HTR fuels area, we are providing detailed statistical information on the fission product retention performance of irradiated fuel. Our studies are also providing basic data on the mechanical, physical, and chemical behavior of HTR materials, including metals, ceramics, graphite, and concrete. The ORNL has an important role in the development of improved HTR graphites and in the specification of criteria that need to be met by commercial products. We are also developing improved reactor physics design methods. Our work in component development and testing centers in the Component Flow Test Loop (CFTL), which is being used to evaluate the performance of the HTR core support structure. Other work includes experimental evaluation of the shielding effectiveness of the lower portions of an HTR core. This evaluation is being performed at the ORNL Tower Shielding Facility. Researchers at ORNL are developing welding techniques for attaching steam generator tubing to the tubesheets and are testing ceramic pads on which the core posts rest. They are also performing extensive testing of aggregate materials obtained from potential HTR site areas for possible use in prestressed concrete reactor vessels. During the past year we continued to serve as a peer reviewer of small modular reactor designs being developed by GA and GE with balance-of-plant layouts being developed by Bechtel Group, Inc. We have also evaluated the national need for developing HTRs with emphasis on the longer term applications of the HTRs to fossil conversion processes.
Date: June 1, 1984
Creator: Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E. & Sanders, J.P.
Partner: UNT Libraries Government Documents Department

Investigations of postulated accident sequences for the Fort St. Vrain HTGR

Description: The systems analysis capability of the ORNL HTGR Safety analysis research program includes a family of computer codes: an overall plant NSSS simulation (ORTAP), and detailed component codes for investigating core neutronic accidents (CORTAP), shutdown emergency-cooling accidents via a 3-dimensional core model (ORECA), and once-through steam generator transients (BLAST). The component codes can either be run independently or in the overall NSSS code. Verification efforts have consisted primarily of using existing Fort St. Vrain reactor dynamics data to compare against code predictions. Comparisons of core thermal conditions made for reactor scrams from power levels between 30 and 50% showed good agreement. An optimization program was used to rationalize the difference between the predicted and measured refueling region outlet temperatures, and, in general, excellent agreement was attained by adjustment of models and parameters within their uncertainty ranges. However, more work is required to establish a unique and valid set of models.
Date: January 1, 1978
Creator: Ball, S.J.; Cleveland, J.C.; Conklin, J.C.; Hatta, M. & Sanders, J.P.
Partner: UNT Libraries Government Documents Department