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Safety Analysis of the Storage and Handling of the SRE Fuel

Description: "The most hasardous conditions relating to the storage and handling of the SRE fuel are those in which water and fuel form a possible critical assembly. Therefore, this memorandum is primarily devoted to an analysis of the conditions necessary for a critical mixture of water and SRE fuel. Certain special geometries are then analyzed for their safeness."
Date: November 17, 1954
Creator: Williams, R. O.
Partner: UNT Libraries Government Documents Department

POWER REACTOR FUEL REPROCESSING: MECHANICAL PHASE

Description: An occurrence at the SRE which caused approximately 7 fuel elements to part when being pulled from the reactor complicated reprocessing, however the bulk of the elements may still be reprocessed as planned. New or reworked pants for the mechanical decladder were supplied by the fabricators for rebuilding in ORNL shops. Administrative actions dictated by budgetary considerations include reduction of UO/sub 2/ pellet order to the quantity needed this fiscal year, reduction of Mark III prototype fuel element to 12, stoppage of all work by REE Division personnel, and stoppage of all work not related to SRE fuel element reprocessing. (auth)
Date: September 1, 1959
Creator: Klima, B.B.
Partner: UNT Libraries Government Documents Department

NaK FREE CONVECTION COOLED SHAFT FREEZE SEAL FOR SRE PUMPS

Description: An investigation was conducted to determine the feasibility of using an NaK free-convection-cooled shaft freeze seal on the SRE main sodium pumps. The use of sodium in this application instead of tetralin eliminates the carburization hazard present should the seal coolant-(tetralin) leak into the sodium system. Results and recommendations are included. (J.R.D.)
Date: June 18, 1959
Creator: Perez, F.
Partner: UNT Libraries Government Documents Department

CONTROL OF OXYGEN CONCENTRATION IN A LARGE SODIUM SYSTEM

Description: Data on the performances of two types of cold traps in the 50,000-lb radioactive sodium system at the SRE are tabulated. The rates were determined when trap inlet oxygen concentrations were at 8 to 10 parts per million. Oxygen concentration was readily controlled to 8 ppm using a cold trap. Extraction of oxygen from sodium by zirconium at 1200 deg F (hot trapping) reduces the concentration below the limit of detection, i.e., oxide solubility saturation temperature below 225 deg F. The theoretical limit for the equilibrium oxygen concentration was calculated to be less than 7 x 10/sup -6/ ppm. The observed extraction rate of 0.009 lb oxygen/hr was one-half of the rate predicted from material behavior studies. (auth)
Date: December 1, 1959
Creator: Hinze, R B
Partner: UNT Libraries Government Documents Department

MEASUREMENT OF THE SRE AND KEWB PROMPT NEUTRON LIFETIME USING RANDOM NOISE AND REACTOR OSCILLATION TECHNIQUES

Description: The prompt neutron lifetime of the SRE was measured by both the oscillation and random noise techniques. Measurement by use of the oscillation technique gave a prompt neutron lifetime of (5 25 plus or minus 0 35) x 10/sup - 4/ sec for a calculated beta of 7 x 10/sup -3/. The measured noise response indicated a lifetime of (5.25 plus or minus 0.7) x 10/sup -4/ sec. Both measured values are in agreement with the calculated value of 5 x 10/sup -4/ sec. Four experiments utilizing the noise analysis technique were performed to determine the prompt neutron lifetime of the KEWB. All four experiments gave results which agreed within 3%, For an estimated beta of 8 x 10/sup -3/, the measured value obtained was (7.8 plus or minus 0.3) x 10/sup -5/ sec. This is in reasonable agreement with both the energy independert calculated value of 6.6 x 10/sup -5/ see and the value of 6.2 x 10/sup -5/ sec obtained from the experimental inhour equation The oscillation technique has been found to be better suited for lifetime determinations in reactors where the prompt neutron break frequency is less than 5 cps. Reactor noise analysis is more suitable for reactors which have prompt neutron lifetime break frequencies above 20 cps. (auth)
Date: October 15, 1959
Creator: Griffin, C.W. & Lundholm, J.G. Jr.
Partner: UNT Libraries Government Documents Department

FLOWSHEET EVALUATION FOR THE DISSOLVING AND NEUTRALIZATION OF SODIUM REACTOR EXPERIMENT USED NUCLEAR FUEL

Description: This report includes the literature review, hydrogen off-gas calculations, and hydrogen generation tests to determine that H-Canyon can safely dissolve the Sodium Reactor Experiment (SRE; thorium fuel), Ford Nuclear Reactor (FNR; aluminum alloy fuel), and Denmark Reactor (DR-3; silicide fuel, aluminum alloy fuel, and aluminum oxide fuel) assemblies in the L-Bundles with respect to the hydrogen levels in the projected peak off-gas rates. This is provided that the number of L-Bundles charged to the dissolver is controlled. Examination of SRE dissolution for potential issues has aided in predicting the optimal batching scenario. The calculations detailed in this report demonstrate that the FNR, SRE, and DR-3 used nuclear fuel (UNF) are bounded by MURR UNF and may be charged using the controls outlined for MURR dissolution in a prior report.
Date: October 30, 2012
Creator: Daniel, W. E.; Hansen, E. K. & Shehee, T. C.
Partner: UNT Libraries Government Documents Department

PRELIMINARY RADIATION SURVEY OF THE SRE

Description: An investigation was conducted to determine radiation dosags rates associated with operation of the Sodium Reactor Experiment at 8.3 Mw. These values were extrapolated to full-power operation. Hadiation survey results are tabulated. (J.R.D.)
Date: September 18, 1957
Creator: Hale, J.P.
Partner: UNT Libraries Government Documents Department

Hanford site sodium management plan

Description: The Hanford Site Sodium Management Plan, Revision 1, provides changes to the major elements and management strategy to ensure an integrated and coordinated approach for disposition of the more than 350,000 gallons of sodium and related sodium facilities located at the DOE`s Hanford Site
Date: September 25, 1995
Creator: Guttenberg, S.
Partner: UNT Libraries Government Documents Department

Sodium Graphite Reactor Quarterly Progress Report for July-September, 1954

Description: Reactivity calculations have been performed for the steady-state Pu feedback technique outlined in the previous progress report. A full-scale power plant study was initiated, based on sodium-graprite technology. A twin-core power plant is now considered to be the most promising configuration. Several design drawings are given of such a reactor, using slightly enriched U to produce Pu amd electrical power. The thermal neutron flux distribution in a cluster of 6 U rods was measured, and the results are compared with previous measurements for 7 rod cluster. The average thermal cycling of hollow U slug elements was begun. Results are given for 500 cycles between 100 and 500 deg C. A series of powder- compacted U alloys were thermal cycled between 200 and 700 deg C. Data on the transfur of radioactivity from Zr by Na has been obtained from a capsule of the first series of three miniharps. Fe, Al, and Cu were immersed in toluene end irradiated at 150 deg F in the MTR-Gamma canal. Toluene is being considered as a shield coolant for the SRE. The effect of 1-Mev electron irradlation on terphenyls was also studied. A venting tube arrangement has been designed for the Zr-canned graphite moderator. A number of thermal insulating brick amd fiber materials were sublected to liquid Na to study deterioration effects. The materials tested were JohnsManville Brick C-16 (Sil-O-Cel mortar), Superex Paste, and Eagle-Pitcher Mineral Wool. Encouraging results were obtained in an efiort to evaluate the effectiveness of Na decontamination by liquid ammonia. Pressure drop and flow characteristics of the latest design SRE fuel element have been completed. Design details of the 2-speed control rod drive assembly are given. Other aspects of the reactor control system, including design and component fabrication, are discussed. Gamma dose rates at the surface of the top shield ...
Date: December 1, 1954
Creator: Siegel, S. & Inman, G. M.
Partner: UNT Libraries Government Documents Department

Design, Fabrication and Installation of SRE Fuel Element Guide Assembly 7519-44741

Description: The design, fabrication, and installation of a fuel-element guide assembly for the Sodium Reactcr Experiment are described. Improvements in this assembly over other models are outlined, and recommendations for the testing of SRE core elements are presented. (T.F.H.)
Date: November 25, 1960
Creator: Meise, E. R. & Gower, G. C.
Partner: UNT Libraries Government Documents Department

RADIAL THERMAL AND FAST NEUTRON FLUX DISTRIBUTIONS IN THE SODIUM REACTOR EXPERIMENT (SRE) AND IN THE TITLE I CONFIGURATION OF THE HALLAM NUCLEAR FACILITY (HNPF)

Description: The thermal neutron flux distributions for the Sodium Reactor Experiment and the Hallam Power Reacter radial shields are calculated by three different methods. The method giving the highest fluxes is used to calculate conservative values of the heat generation rates in these shields. (T.F.H.)
Date: March 27, 1959
Creator: Legendre, P.J.
Partner: UNT Libraries Government Documents Department

FIRST SODIUM REACTOR EXPERIMENT (SRE) TEST OF HALLAM NUCLEAR POWER FACILITY (HNPF) CONTROL MATERIALS

Description: An experiment was conducted in the SRE to measure temperatures and neutron flux levels in and near a boron-containing simulated control rod. The data are being used to check analytical methods developed for prediction of control rod heat generation rates and maximum temperatures in this type of control rod in the Hallam Nuclear Power Facility. The maximum observed temperatures with a reactor power level of 20 Mw were 1363 deg F for a boron-- nickel alloy ring having a 0.105-in. radial clearance with the thimble and 1100 deg F for a boron -nickel alloy ring having a 0.020-in. radial clearance. The maximum temperature difference between the coolant and the control rod was 473 deg F. It is concluded that the expected greater heat generation rates in the Hallam reactor would prohibit the use of boron-containing absorber materials in a combined a him-safety rod. (auth)
Date: June 1, 1959
Creator: Arneson, S.O.
Partner: UNT Libraries Government Documents Department

HEAT TRANSFER ANALYSIS AND DESIGN OF A PLUGGING INDICATOR SYSTEM FOR SRE

Description: The analysis was performed on a system comprising a counterflow, concentric-pipe economizer, heat exchanger, flowmeter, plug, and connecting pipe. The system was assumed to be at some initial temperature equal to the inlet sodium temperature and suddenly loses heat to a medium in the heat exchanger. Design and operating data are presented. A cooling rate curve is given where the nitrogen flow rate is decreased when the plug temperature reaches 400 deg F. The time variation of minimum temperatures is given for various values of thermal capacitance with constant equilibrium temperature, and the economizer parameter with constant equilibrium temperatures and thermal capacitance. The variation in heat exchanger parameter with economizer parameter for a constant equilibrium minimum temperature of 250 deg F, and a constant inlet temperature of 750 deg F is indicated. (B.O.G.)
Date: April 1, 1955
Creator: Sletten, H. L.
Partner: UNT Libraries Government Documents Department

SODIUM REACTOR EXPERIMENT (SRE) SHIELDING EVALUATION FOR THERMAL NEUTRON STREAMING AT REACTOR VESSEL COOLANT PIPE PENETRATIONS

Description: The experimental program performed in the SRE auxiliary and main primary galleries was part of a program to determine the adequacy of the shielding configuration for the SRE. The work discussed in this report is concerned with analysis of neutron streaming at coolant pipe penetrations of the reactor vessel, analysis of the shielding required, testing and evaluation of recommended shielding, and measurement and correlation of neutron streaming in labyrinths with theory. The activation analysis method using zinc sheets which was developed for the program of determining thermal neutron streaming in the SRE primary galleries was proven to be versatile, accurate, and reliable. A modified form of the theoretical method of Price, Horton, and Spinney, used to determine neutron scattring through labyrinths, was found to agree favorably with the experimental results obtained from the SRE primary galleries. The theoretical attenuation method used no determine the neutron shield configuration installed in the auxiliary primary gallery was found to give an overestimate of the actual attenuation properties of this shield. The neutron shield configuration installed in the auxiliary primary gallery proved to be adequate in reducing the thermal neutron streaming flux to an acceptable level. It is concluded that both SRE primary galleries are now adequately shielded to prevent excessive neutron- induced activation of the components and equipment located therein. (auth)
Date: October 31, 1959
Creator: Anderson, F. D.
Partner: UNT Libraries Government Documents Department

Reflected Reactor Kinetics

Description: The transfer function of a reflected reactor was derived from a model which is applicable to all reactors. The result is identical to a bare reactor transfer function at low frequencies, provided that the neutron lifetime includes the effect of the reflector. At high frequencies the reflector introduces in effect an additional group of delayed neutrons. A calculation of the effective neutron lifetime in the SRE from the theoretical results was in good agreement with the experimental value. (auth)
Date: March 1, 1963
Creator: Keaten, R. W. & Griffin, C. W.
Partner: UNT Libraries Government Documents Department

SRE CORE III FUEL ELEMENTS THERMAL ANALYSIS

Description: The initial 30-Mw SRE Core III loading will contain a total of 33 fuel elements. Of these, the central 7 fuel elements are test elements and the remaining 26 are driver elements; 4 of the 7 test elements are designated as standard test elements. These elements are identical to the driver elements with the exception of the active fuel length. The remaining 3 test elements are designated as special test elements and incorporate fuel rods of smaller diameter and increased enrichment to obtain higher burnup rates, greater specific power, and higher fuel temperatures. Geometry design data for the various elements considered in the analysis was obtained from Dwg. No. 650 deg F, the mixed-mean coolant outlet temperature for the core at 1200 deg F ( plus or minus 15 deg F), the maximum average temperature on the cladding at 1275 deg F, and the maximum fuel temperature in the range of 2000 to 2100 deg F. The thermal performance of the fuel elements was analyzed with the SORTD code and the Heating code. The results of the analysis are presented in tabular form. Axial temperature profiles are also presented for representative fuel elements. The core outlet coolant mixed-mean temperature was found to be 1185 deg F. For the special test, standard test, and highest powered driver elements the coolant outlet temperature is limited to values below 1200 deg F because of the specified maximum value of cladding temperature. For the remaining elements the coolant outlet temperature is 1200 deg F. Maximum fuel temperatures are at values less than 2000 deg F with the exception of the special test elements. These have maximum fuel temperatures in the range of 2000 to 2100 deg F. Recently, the highest powered driver element'' was redesignated as standard test element'' because of the high power ...
Date: February 27, 1964
Creator: Bergonzoli, F.
Partner: UNT Libraries Government Documents Department

PRELIMINARY DESIGN OF SRE-PEP CORE III VARIABLE ORIFICE

Description: A single, wide-range variable orifice design is developed, which provides a maximum orifice flow area of 2.0 in./sup 2/ and a minimum oriflce flow area of 0.2 in./sup 2/ The design is based on an overall operating core DELTA P of 4.0 psi and provides for a sodium flow range of 2.45 to 10.0 lb/sec. Bypass leakage around the orifice is analyzed and accounted for in the design. It is shown that the use of piston rings is not necesary to reduce this leakage. The orifice plug profile is calculated for constant temperature sensitivity; the maximum average sensitivity being 451 deg F/ in. for a 1.5 in. plug length. Fuel channel pressure drop is recalculated for both special and standard elements using the latest proposed designs. Maximum DELTA P including wide open variable orifice is 3.6 psi. (auth)
Date: November 21, 1963
Creator: Noyes, R.C.
Partner: UNT Libraries Government Documents Department

A Survey of the Hazards Involved in Processing Liquid Metal Bonded Fuels

Description: A survey of the character and magnitude of hazards involved in processing liquid metal bonded fuels was made and the scope of a preliminary experimental program outlined. Processing of SRE and CPPD fuels by mechanical decladding followed by controlled reaction of the collected methods. Simdlarly, shearing of PRDC fuel and controlled exposure of the Na in the severed portions to water appears more desfrable than chethical dissolution of the metallic cladding. (auth)
Date: August 14, 1961
Creator: Adams, J. B.
Partner: UNT Libraries Government Documents Department