This system will be undergoing maintenance July 5th between 8:00AM and 12:00PM CDT.

Search Results

Advanced search parameters have been applied.
open access

Experiment Operations Plan for a Loss-of-Coolant Accident Simulation in the National Research Universal Reactor Materials Test 2

Description: A loss-of-coolant accident (LOCA) simulation program is evaluating the thermal-hydraulic and mechanical effects on pressurized water reactor (PWR) test fuel bundles. This Experiment Operation Plan (EOP) Addendum 2, together with the referenced EOP, describes the desired operating conditions and additional hazards review associated with the four-part MT-2 experiment. The primary portions of the experiment, MT-2.2 and MT-2.3, will evaluate the following: 1) the mechanical deformation of pressurized fuel rods subjected to a slow LOCA, using reflood water for temperature control, that is designed to produce cladding temperatures in the range from 1033 to 1089K (1400 to 1500°F) for an extended time, and 2) the effects of the deformed and possibly failed cladding on the thermal-hydraulic performance of the test assembly during simulated LOCA heating and reflooding. The secondary portions of the experiment, MT-2.1 and MT-2.4, are intended to provide thermal-hydraulic calibration information during two-stage reflood conditions for 1) relatively low cladding temperatures, <839K (1050°F), on nondeformed rods, and 2) moderately high cladding temperatures, <1089K (1500°F), on deformed rods.
Date: September 1, 1981
Creator: Russcher, G. E.; Barner, J. O.; Hesson, G. M.; Wilson, C. L.; Parchen, L. J.; Cunningham, M. E. et al.
Partner: UNT Libraries Government Documents Department
open access

Beginning-of-Life Data Report for the Instrumented Fuel Assembly (IFA)-527

Description: This report presents beginning-of-life (BOL) data from the first four months of operation of the six-rod instrumented fuel assembly (IFA)-527 in the Halden Boiling Water Reactor (HBWR), Halden, Norway. This assembly is the last in a series of U.S. Nuclear Regulatory Commission (NRC)-sponsored tests to verify steady-state fuel performance computer codes. IFA-527 contains five identical rods with high-density stable fuel pellets and 0.23-mm diametral gaps and one rod with similar fuel pellets but with a 0.06-mm diametral gap. All six rods were xenon-filled to provide simulation of the effects of fission gas and to enhance the observable effects of fuel cracking and relocation on fuel temperatures. The assembly operated successfully from July 1, 1980, to August 15, 1980; and then the reactor was shut down until September 10, 1980. Sometime during the shutdown, four of the six rods suffered pressure boundary failure. The decision was made to restart the reactor to collect operating data with failed rods. This report presents both pre- and postfailure data for IFA-527.
Date: September 1, 1981
Creator: Lanning, D. D.
Partner: UNT Libraries Government Documents Department
open access

FUEL PERFORMANCE IMPROVEMENT PROGRAM Power-Ramp Testing and Postirradiation Examination of PCI- Resistant LWR Fuel Rod Designs

Description: This report describes the power-ramp testing results from 10 fuel rods irradiated in the Halden Boiling Water Reactor (HBWR), Halden, Norway. Tne work is part of the Fuel Performance Improvement Program (FPIP), which is sponsored by the U.S. Department of Energy (DUE) and is conducted through the joint efforts of Consumers Power Company, Exxon Nuclear Company, lnc., and Pacific Northwest Laboratory. The objective of the FPlP is to identify and demonstrate fuel concepts with improved pellet-cladding interaction (PCl) behavior that will be capable of extended burnup. The postirradiation examination results obtained from one nonramped rod are also presented. The power-ramping behavior of three basic fuel rod types--rods with annular-pellet fuel, sphere-pac fuel, and dished-pellet (reference) fuel--are compared in terms of mechanisms known to promote PCl failures. The effects of graphite coating on the inside cladding surface and helium pressurization in rods witn annular fuel are also evaluated .
Date: September 1, 1982
Creator: Barner, J. O. & Guenther, R. J.
Partner: UNT Libraries Government Documents Department
open access

PARAMETRIC REACTIVITY TRANSIENT ANALYSES FOR THE FFTF NUCLEAR PROOF TEST REACTOR

Description: Fault tree techniques have been used to identify possible failure paths within the NPTR which could lead to core disassembly. The analysis o f the various faults has led to formulation of design requirements, protective system requirements, and administrative restraints required to prevent accidents from these faults. Transient analyses were performed using the heat transfer-nuclear kinetics codes, Nutiger-II, FORE-II, and MELT-II . To verify results, intercomparison studies were made between the codes. The codes were i n good general agreement. Each code was found to exhibit different advantages and disadvantage. Inherent reactivity feedback effects were assessed in the analysis. With the assumed core parameters, there appears to be sufficient Doppler to prolong a nuclear transient to allow protective action to prevent fuel from melting. The use of average values of the feedback coefficients smeared over the entire core does not appear to be an acceptable method with spacially dependent temperatures. In the thermal analysis, the fuel pin gap coefficient and sodium film coefficient do not appear to be highly sensitive parameters for transient analysis. Power transients resulting from reactivity insertions of from 2$/sec to 20$/sec have been examined in detail. Sodium will be molten before fuel melting occurs for accidents within this range. For the smaller ramp rates (< 4$/sec), sodium nay even reach vaporization temperatures before any fuel melts. Power transients terminated by effective protective action were investigated. It i s believed p o s s i b l e t o design a scram system, with the . present state of the art, to prevent sodium from melting for a reactivity ramp up to at least 6$/sec. This same system would prevent fuel melting for a reactivity ramp up to 15$/sec. Sodium thermal expansion will play a very important role in a core disassembly. When the average sodium …
Date: January 1, 1970
Creator: Schade, A. R.
Partner: UNT Libraries Government Documents Department
open access

High Burnup Effects Program A State-of-the-Technology Assessment

Description: Various analytical models and empirical correlations describing the fission gas release phenomenon were examined. An evaluation was made of the current pertinent experimental data on the subject of high burnup fission gas release. Data reported by individual investigators were compared and evaluated in relation to their applicability to the content and scope of the High Burnup Effects Program. These evaluations then form the bases for defining the data needs, and the selection of variables to be studied in this program.
Date: June 1, 1982
Creator: Rising, K. H.; Bradley, E. R.; Williford, R. E. & Freshley, M D.
Partner: UNT Libraries Government Documents Department
open access

FAST FLUX TEST FACILITY CONCEPTUAL COMPONENT DESIGN DISCRIPTION FOR THE REACTOR NUCLEAR CONTROL COMPONENTS No. 33

Description: This document describes the nuclear control component of the Fast Test Reactor (FTR) . The Fast Test Reactor is t he central test bed of the Fast Flux Test Facility (FFTF). The Reactor Nuclear Control Components described in this document consist of neutron absorbing elements which are moved axially into and out of the general area of the active portion of the core. The electromechanical drive housings are mounted on the reactor cover and are part of the pressure boundary of the vessel and cover during reactor operation. Drive lines (extension shafts) extend through the cover and are connected to the poison element. The element is guided in the core throughout its travel by a flow duct which resides in the c or e grid plate receptacle. Sodium coolant from the primary loop is used to cool the neutron absorbing rods. This Conceptual Component Design Description (CCDD) suppor t s and expands the requirements of the Overall Conceptual Systems Design Description . Specifically , it establishes the reactor nuclear control design requirements; describes the current reference design of this control; defines safety, operational · and maintenance principles; indicates unresolved problems ; and identifies the principal interfaces affecting the reactor control component functional requirements and design configuration. Included are functions and design requirements, a physical description of the system, safety considerations, principles of operation, and maintenance principles.
Date: October 1, 1969
Partner: UNT Libraries Government Documents Department
open access

REACTOR PHYSICS QUARTERLY REPORT JANUARY, FEBRUARY, MARCH 1970

Description: The objective of the Reactor Physics Quarterly Report is to inform the scientific community in a timely manner of the technical progress made on the many phases of reactor physics work within the laboratory. The report contains brief technical discussions of accomplishments in all areas where significant progress has been made during the quarter.
Date: May 1, 1970
Creator: Schmid, L. C.; Clayton, E. D. & Heineman, R. E.
Partner: UNT Libraries Government Documents Department
open access

FRAPCON-2: A Computer Code for the Calculation of Steady State Thermal-Mechanical Behavior of Oxide Fuel Rods

Description: FRAPCON-2 is a FORTRAN IV computer code that calculates the steady state response of light Mater reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, deformation, and tai lure histories of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (a) heat conduction through the fuel and cladding, (b) cladding elastic and plastic deformation, (c) fuel-cladding mechanical interaction, (d) fission gas release, (e} fuel rod internal gas pressure, (f) heat transfer between fuel and cladding, (g) cladding oxidation, and (h) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat transfer correlations. FRAPCON-2 is programmed for use on the CDC Cyber 175 and 176 computers. The FRAPCON-2 code Is designed to generate initial conditions for transient fuel rod analysis by either the FRAP-T6 computer code or the thermal-hydraulic code, RELAP4/MOD7 Version 2.
Date: January 1, 1981
Creator: Berna, G. A; Bohn, M. P.; Rausch, W. N.; Williford, R. E. & Lanning, D. D.
Partner: UNT Libraries Government Documents Department
open access

Prototypic Thermal-Hydraulic Experiment in NRU to Simulate Loss-of-Coolant Accidents

Description: Quick-look test results are reported for the initial test series of the Loss-of-Coolant Accident (LOCA) Simulation in the National Research Universal {NRU) test program, conducted by Pacific Northwest Laboratory (PNL) for the U.S. Nuclear Regulatory Commission (NRC). This test was devoted to evaluating the thermal-hydraulic characteristics of a full-length light water reactor (LWR) fuel bundle during the heatup, reflood, and quench phases of a LOCA. Experimental results from 28 tests cover reflood rates of 0.74 in./sec to 11 in./sec and delay times to initiate reflood of 3 sec to 66 sec. The results indicate that current analysis methods can predict peak temperatures within 10% and measured quench times for the bundle were significantly less than predicted. For reflood rates of 1 in./sec where long quench times were predicted (>2000 sec}, measured quench times of 200 sec were found.
Date: April 1, 1981
Creator: Mohr, C. L.; Hesson, G. M.; Russcher, G. E.; Marsh, R. K.; King, L. L.; Wildung, N. J. et al.
Partner: UNT Libraries Government Documents Department
open access

PAR Loop Schedule Review

Description: The status of the overall design, fabrication, and installation of the component items of the PAR loop experiment in the ORR is reviewed.
Date: January 15, 1958
Creator: Schaffer, Jr. & W.F.
Partner: UNT Libraries Government Documents Department
open access

PAR Loop Schedule Review

Description: The schedule for the installation of the PAR slurry loop experiment in the South Facility of the ORR has been reviewed and revised. The design, fabrication, and installation is approximately two weeks behind schedule at this time due to many factors; however, indications are that this time can be made up. Design is estimated to be 56 per cent complete, fabrication 27 per cent complete-and installation 11 per cent complete:
Date: March 31, 0958
Creator: Schaffer, Jr. & W.F.
Partner: UNT Libraries Government Documents Department
open access

Reactor Safety Research Programs

Description: This document summarizes the work performed by Pacific Northwest laboratory from October 1 through December 31, 1979, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, lspra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.
Date: August 1, 1980
Creator: Dotson, CW
Partner: UNT Libraries Government Documents Department
open access

Tests on Models of Nuclear Reactor Elements - Studies of Diffusion

Description: To estimate the distribution of temperature in the proposed nuclear reactor, one must determine a coefficient of eddy diffusivity and devise a suitable method of computing the heat transfer. Measurements of diffusion in a model of a blanket element for the proposed reactor indicated a gross eddy diffusion coefficient of about 0.002 (? .0005) ft{sup 2}/sec. Thus, the apparent eddy diffusion for the test conditions is about 200 times the molecular diffusivity of water and about. twice that of the liquid sodium. Even approximate methods of applying this result require an elaborate calculation if the primary characteristics of the flow system are to be taken into account. The dispersion of dye in flowing water provided an indication of the diffusion in the model tests. The presence and arrangement of the rods, the effect on the flow?of the spiral wire spacers, and the existence of a comparatively large area on which a laminar sub-layer develops made it impossible to get simple turbulence criteria like those obtained downstream from a screen. Although the results are consequently somewhat unsystematic, they do establish reliably the approximate magnitude of the coefficient of eddy diffusivity. The data were obtained from both line and sectional traverses, the two results being approximately equal. Preliminary data were also obtained for a core element for which {epsilon} ~ 0.003, only slightly less than for the blanket element. Determination of the diffusion coefficient makes it possible to compute the temperature for an array of spatially variable heat sources, as occur in any element. Because of the extreme complexity of the problem, two alternative simplifying assumptions are proposed., In one, the heat sources are assumed to be concentrated along their axes. In the other, the heat is assumed to pass to the fluid immediately at the surface of each circular rod and …
Date: March 1, 1957
Creator: McNown, J. S.; Yih, C. S.; Yagle, R. A.; O&#x27 & Dell, W. W.
Partner: UNT Libraries Government Documents Department
open access

FAST FLUX TEST FACILITY CONCEPTUAL FACILTY DESIGN DESCRIPTION FOR THE INERT GAS CELL EXAMINATION FACILITY NO. 71

Description: The purpose of this Conceptual Facility Design Description (CFDD) is to provide a technical description of the Inert Gas Cell Examination Facility such that agreement with RDT on a Conceptual Design can be reached . The CFDD also serves to establish a common understanding of the facility concept among all responsible FFTF Project parties including the Architect Engineer and Reactor Designer. Included are functions and design requirements, a physical description of the facility, safety considerations, principles of operation, and maintenance principles.
Date: December 12, 1968
Partner: UNT Libraries Government Documents Department
open access

A New 2D-Transport, 1D-Diffusion Approximation of the Boltzmann Transport equation

Description: The work performed in this project consisted of the derivation, implementation, and testing of a new, computationally advantageous approximation to the 3D Boltz- mann transport equation. The solution of the Boltzmann equation is the neutron flux in nuclear reactor cores and shields, but solving this equation is difficult and costly. The new “2D/1D” approximation takes advantage of a special geometric feature of typical 3D reactors to approximate the neutron transport physics in a specific (ax- ial) direction, but not in the other two (radial) directions. The resulting equation is much less expensive to solve computationally, and its solutions are expected to be sufficiently accurate for many practical problems. In this project we formulated the new equation, discretized it using standard methods, developed a stable itera- tion scheme for solving the equation, implemented the new numerical scheme in the MPACT code, and tested the method on several realistic problems. All the hoped- for features of this new approximation were seen. For large, difficult problems, the resulting 2D/1D solution is highly accurate, and is calculated about 100 times faster than a 3D discrete ordinates simulation.
Date: June 17, 2013
Creator: Larsen, Edward
Partner: UNT Libraries Government Documents Department
open access

FAST FLUX TEST FACILITY MONTHLY INFORMAL TECHNICAL PROGRESS REPORT MAY 1969

Description: This report was prepared by Battelle-Northwest under Contract No. AT(45-1)-1830 for the Atomic Energy Commission, Division of Reactor Development and Technology, to summarize technical progress made in the Fast Flux Test Facility Program during May 1969 .
Date: June 6, 1969
Creator: Astley, E. R. & Cabell, C. P.
Partner: UNT Libraries Government Documents Department
open access

Tests on Models of Nuclear Reactor Elements - Head Losses in Core Sub-Assemblies

Description: Losses have been determined for flow through models of various proposed core sub-assemblies as part of a study of the elements of a nuclear reactor. Six core sections and two? axial blanket sub-assemblies have been compared on the basis of drop in piezometric. head or pressure drop. The core sub-assemblies are composed of an entrance nozzle, a lower axial blanket section, the core section, an upper axial blanket section, and a short section for the handling lug. The four parts of the sub-assembly other than the core section are designated as the axial blanket sub-assembly. In each core section there are 144 rods within a container which has a square cross-section. The primary differences between one core section and another are the means qf supporting and spacing the rods. Bars or wires wrapped in spirals around the rods were used as well as a series of grids made up of wires and supported at the four corners. Also, in one core an, inner wall was used to provide an annular flow passage which helps to reduce the difference in temperature at the inner and outer walls of the core. The two axial blanket sub-assemblies tested are similar except ?that the second model is characterized by more gradual transitions in changes of cross section. Other parts of this study of the elements of a nuclear reactor have been described in two previous reports dealing with head losses in complete blanket subassemblies and with diffusion studies.
Date: July 1, 1957
Creator: McNown, J. S.; Yagle, R. A. & Spengos, A.
Partner: UNT Libraries Government Documents Department
open access

Analysis of Credible Accidents for Argonaut Reactors

Description: Five areas of potential accidents have been evaluated for the Argonaut-UTR reactors. They are: • insertion of excess reactivity • catastrophic rearrangement of the core • explosive chemical reaction • graphite fire • fuel-handling accident. A nuclear excursion resulting from the rapid insertion of the maximum available excess reactivity would produce only 12 MWs which is insufficient to cause fuel melting even with conservative assumptions. Although precise structural rearrangement of the core would create a potential hazard, it is simply not credible to assume that such an arrangement would result from the forces of an earthquake or other catastrophic event. Even damage to the fuel from falling debris or other objects is unlikely given the normal reactor structure. An explosion from a metal-water reaction could not occur because there is no credible source of sufficient energy to initiate the reaction. A graphite fire could conceivably create some damage to the reactor but not enough to melt any fuel or initiate a metal-water reaction. The only credible accident involving offsite doses was determined to be a fuel-handling accident which, given highly conservative assumptions, would produce a whole-body dose equivalent of 2 rem from noble gas immersion and a lifetime dose equivalent commitment to the thyroid of 43 rem from radioiodines.
Date: April 1, 1981
Creator: Hawley, S. C.; Kathern, R. L. & Robkin, M. A.
Partner: UNT Libraries Government Documents Department
open access

Experiment Operations Plan for a Loss-of-Coolant Accident Simulation in the National Research Universal Reactor Materials Tests 1 and 2

Description: A loss of Coolant Accident (LOCA) simulation program is evaluating the thermal-hydraulic and mechanical effects of LOCA conditions on pressurized water reactor test fuel bundles. This experiment operation plan for the second and third experiments of the program will provide peak fuel cladding temperatures of up to 1172K (1650{degree}F) and 1061K (1450{degree}) respectively. for a long enough time to cause test fuel cladding deformation and rupture in both. Reflood coolant delay times and the reflooding rates for the experiments were selected from thermal-hydraulic data measured in the National Research Universal (NRU) reactor facilities and test train assembly during the first experiment.
Date: September 1, 1981
Creator: Russcher, G. E.; Wilson, C. L.; Marshall, R, K.; King, L. L.; Parchen, L. J.; Pilger, J. P. et al.
Partner: UNT Libraries Government Documents Department
Back to Top of Screen