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Safety rod latch inspection

Description: During an attempt to raise control rods from the 100 K reactor in December, one rod could not be withdrawn. Subsequent investigation revealed that a small button'' in the latch mechanism had broken off of the lock plunger'' and was wedged in a position that prevented rod withdrawal. Concern that this failure may have resulted from corrosion or some other metallurgical problem resulted in a request that SRL examine six typical latch mechanisms from the 100 L reactor by use of radiography and met… more
Date: February 1, 1992
Creator: Leader, D.R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Design configuration of GCFR core assemblies

Description: The current design configurations of the core assemblies for the gas-cooled fast reactor (GCFR) demonstration plant reactor core conceptual design are described. Primary emphasis is placed upon the design innovations that have been incorporated in the design of the core assemblies since the establishment of the initial design of an upflow GCFR core. A major feature of the design configurations is that they are prototypical of core assemblies for use in commercial plants; a larger number of the … more
Date: May 1, 1980
Creator: LaBar, M.P.; Lee, G.E. & Meyer, R.J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Rapid quenching of molten lithium-aluminum jets in water under loss-of-control-rod-cooling conditions

Description: A series of fifteen tests were performed to investigate the thermal interactions between molten LiAl control rod material and water under conditions prototypic of the loss-of-control-rod-cooling (LCRC) accident scenario. The experimental parameters such as melt mass, stream diameter, melt temperature and flowrate, water depth and water temperature were controlled or varied to agree with analytically determined conditions, thus insuring prototypicality of the experiments and applicability of the… more
Date: January 1, 1992
Creator: Greene, G. A.; Finfrock, C. C.; Schwarz, C. E.; Allison, D. K. & Hyder, M. L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Summary of HTGR (high-temperature gas-cooled reactor) benchmark data from the high temperature lattice test reactor

Description: The High Temperature Lattice Test Reactor (HTLTR) was a unique critical facility specifically built and operated to measure variations in neutronic characteristics of high temperature gas cooled reactor (HTGR) lattices at temperatures up to 1000{degree}C. The Los Alamos National Laboratory commissioned Pacific Northwest Laboratory (PNL) to prepare this summary reference report on the HTLTR benchmark data and its associated documentation. In the initial stages of the program, the principle of th… more
Date: October 1, 1989
Creator: Newman, D.F.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Aging assessment of BWR control rod drive systems

Description: This study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assess the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to… more
Date: January 1, 1991
Creator: Greene, R.H.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Force analysis of the advanced neutron source control rod drive latch mechanism

Description: The Advanced Neutron Source reactor (ANS), a proposed Department of Energy research reactor currently undergoing conceptual design at the Oak Ridge National Laboratory (ORNL), will generate a thermal neutron flux approximating 10{sup 30} M{sup {minus}2}{emdash}S{sup {minus}1}. The compact core necessary to produce this flux provides little space for the shim safety control rods, which are located in the central annulus of the core. Without proper control rod drive design, the control rod drive … more
Date: January 1, 1989
Creator: Damiano, B.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Additional information for impact response of the restart safety rods

Description: WSRC-RP-91-677 studied the structural response of the safety rods under the conditions of brake failure and accidental release. It was concluded that the maximum impact loading to the safety rod is 6020 pounds based on conservative considerations that energy dissipation attributable to fluid resistance and reactor superstructure flexibility. The staffers of the Defense Nuclear Facility Safety Board reviewed the results and inquired about the extent of conservatism. By request of the RESTART tea… more
Date: October 14, 1991
Creator: Yau, W. W. F.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Substitute safety rods: Physics design and NTG calibration

Description: Under certain assumed accident conditions, an SRS reactor may loose most of its bulk moderator while maintaining flow to fuel assemblies. If this occurs immediately after operation at power, components normally dependent on convective heat transfer to the moderator will heat up with the possibility of melting that component. One component at risk is the safety rod. Tests have shown that the current cadmium safety rod, which contains aluminum as well as cadmium, can fail at temperatures only sli… more
Date: July 1, 1991
Creator: Baumann, N.P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Tensile and burst tests in support of the cadmium safety rod failure evaluation

Description: The reactor safety rods may be subjected to high temperatures due to gamma heating after the core coolant level has dropped during the ECS phase of hypothetical LOCA event. Accordingly, an experimental safety rod testing subtask was established as part of a task to address the response of reactor core components to this accident. This report discusses confirmatory separate effects tests conducted to support the evaluation of failures observed in the safety rod thermal tests. As part of the fail… more
Date: February 1, 1992
Creator: Thomas, J.K.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

A simple model for radial expansion reactivity in LMRs (liquid metal reactors)

Description: Presented in this report is a simple analytical model developed for evaluating the radial expansion reactivity in small modular liquid metal reactors (LMRs). The present model is based on a non-leakage representation of the effective neutron multiplication factor. The resultant analytical expression for the radial expansion reactivity is simple and can be used directly in a system code for safety analyses. Applications of the present model to PRISM and SAFR resulted in a good agreement with the… more
Date: January 1, 1988
Creator: Cheng, H.S. & Van Tuyle, G.J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

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 r… more
Date: October 4, 1977
Creator: Pitterle, T. A. & Lagally, H. O.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Tests of a Vapor-Space Feed Nozzle for Calcining Aluminum Nitrate Solutions in a Fluidized Bed

Description: An investigation was conducted to determine the performance of a vapor- space feed nozzle which sprayed aluminum nitrate solution on the surface of a fluidized bed calciner. Results indicate that this type of feed system is satisfactory for calcining aqueous wastes from the processing of spent aluminum- type nuclear fueIs. Process and product control were achieved by adjusting the volumetric ratio of the air to the liquid fed to the nozzIe. The results obtained at various operating conditions a… more
Date: October 31, 1962
Creator: Lohse, G. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Control rod trip failures. Salem 1: the cause, response and potential fixes

Description: Although the failure to trip or scram represents a single class of initiators, the actual events of each transient are operationally unique and require individual human response. The operational team's reaction to the challenge can be successful, in very short response times, and without complete diagnosis of the event's root cause. This underscores the need for a better basic understanding of the team response patterns in such cases, to allow designs, procedures, and training to take advantage… more
Date: January 1, 1983
Creator: Hall, R.E.; Boccio, J.L. & Luckas, W.J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Summary report on four foot septifoil cooling experiment

Description: Cooling parameters for some of the SRS reactor internal components are computed using the Transient Reactor Analysis Code, TRAC.'' In order to benchmark the code, the Safety Analysis Group of SRL requested an experiment to provide measurements of cooling parameters in a well defined physical system utilizing SRS reactor component(s). The experiment selected included a short length of septifoil with both top and bottom fittings containing five simulated control rods in an unseated'' configuratio… more
Date: October 1, 1991
Creator: Randolph, H. W.; Collins, S. L.; Verebelyi, D. T. & Foti, D. J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Self-actuating reactor-shutdown system. [LMFBR]

Description: A control system for the automatic or self-actuated shutdown or scram of a nuclear reactor is described. The system is capable of initiating scram insertion by a signal from the plant protection system or by independent action directly sensing reactor conditions of low-flow or over-power. Self-actuation due to a loss of reactor coolant flow results from a decrease of pressure differential between the upper and lower ends of an absorber element. When the force due to this differential falls belo… more
Date: June 4, 1981
Creator: Barrus, D.M.; Brummond, W.A. & Peterson, L.F.
Item Type: Refine your search to only Patent
Partner: UNT Libraries Government Documents Department
open access

Substitute safety rods: Physics of operation and irradiation

Description: Under certain assumed accidents, an SRS reactor may lose most of its bulk moderator while maintaining flow to fuel assemblies. If this occurs immediately after operation at power, components normally dependent on convective heat transfer to the moderator will heat up with the possibility of melting that component. One component at risk is the currently used cadmium safety rod. A substitute safety rod consisting solely of sintered B{sub 4}C and stainless steel has been designed which is capable … more
Date: November 18, 1991
Creator: Baumann, N.P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Nuclear reactor control apparatus. [FBR]

Description: Nuclear reactor safety rod release apparatus comprises a ring which carries detents normally positioned in an annular recess in outer side of the rod, the ring being held against the lower end of a drive shaft by magnetic force exerted by a solenoid carried by the drive shaft. When the solenoid is de-energized, the detent-carrying ring drops until the detents contact a cam surface associated with the lower end of the drive shaft, at which point the detents are cammed out of the recess in the sa… more
Date: April 16, 1981
Creator: Sridhar, B.N.
Item Type: Refine your search to only Patent
Partner: UNT Libraries Government Documents Department
open access

Analysis of coolability of the control rods of a Savannah River Site production reactor with loss of normal forced convection cooling

Description: An analytical study of the coolability of the control rods in the Savannah River Site (SRS) K-Production Reactor under conditions of loss of normal forced convection cooling has been performed. The study was performed as part of the overall safety analysis of the reactor supporting its restart. The analysis addresses the buoyancy-driven flow over the control rods that occurs when forced cooling is lost, and the limit of critical heat flux that sets the acceptance criteria for the study. The obj… more
Date: January 1, 1992
Creator: Easterling, T.C.; Hightower, N.T. (Westinghouse Savannah River Co., Aiken, SC (United States)); Smith, D.C. & Amos, C.N. (Science Applications International Corp., Albuquerque, NM (United States))
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Advanced absorber assembly design for breeder reactors

Description: An advanced absorber assembly design has been developed for breeder reactor control rod applications that provides for improved in-reactor performance, longer lifetimes, and reduced fabrication costs. The design comprises 19 vented pins arranged in a circular array inside of round duct tubes. The absorber material is boron carbide; cladding and duct components are constructed from the modified Type 316 stainless steel alloy. Analyses indicate that this design will scram 30 to 40% faster than th… more
Date: January 1, 1980
Creator: Pitner, A.L. & Birney, K.R.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Failure of latch mechanism for motion control of safety rods

Description: During safety rod tests in K-reactor prior to startup, one safety rod could not be lifted because the button'' broke off and became lodged in the mechanism. Examination of the failed latch assembly along with other assemblies from both K-Area and L-Area revealed several missing buttons as well as severely deformed jaw hanger extensions.'' We participated in the investigation of the damage by request of the Reactor Restart Section. Based on our study of the latch mechanism, the modifications to … more
Date: January 16, 1992
Creator: Yau, W. W. F. & Leader, D. R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Transactions of the nineteenth water reactor safety information meeting

Description: This report contains summaries of papers on reactor safety research to be presented at the 19th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 28--30, 1991. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Res… more
Date: October 1, 1991
Creator: Weiss, A. J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Nuclear reactor control apparatus

Description: Nuclear reactor core safety rod release apparatus comprises a control rod having a detent notch in the form of an annular peripheral recess at its upper end, a control rod support tube for raising and lowering the control rod under normal conditions, latches pivotally mounted on the control support tube with free ends thereof normally disposed in the recess in the control rod, and cam means for pivoting the latches out of the recess in the control rod when a scram condition occurs. One embodime… more
Date: August 28, 1981
Creator: Sridhar, B.N.
Item Type: Refine your search to only Patent
Partner: UNT Libraries Government Documents Department
open access

Accuracy and Efficiency of a Coupled Neutronics and Thermal Hydraulics Model

Description: The accuracy requirements for modern nuclear reactor simulation are steadily increasing due to the cost and regulation of relevant experimental facilities. Because of the increase in the cost of experiments and the decrease in the cost of simulation, simulation will play a much larger role in the design and licensing of new nuclear reactors. Fortunately as the work load of simulation increases, there are better physics models, new numerical techniques, and more powerful computer hardware that w… more
Date: September 1, 2007
Creator: Mousseau, Vincent A. & Pope, Michael A.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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