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Startup and Initial Testing of SM-1 Core II With Special Components

Description: The loading operation for SM-1 Core II is described. Results of startup physics measurements (Test A-300 (Series) and fission product iodine monitoring in the primary coolant are given. The SM-1 Core II initial loading progressed satisfactorily, fulfilling the predictions of the zero power experiment performed at the Alco Criticality Facility. The initial cold clean five rod bank position was 6.53 in.; the initial hot, no xenon, five rod bank position was 9.62 in.; the initial hot, equilibrium … more
Date: February 28, 1962
Creator: Moote, F. G. & Schrader, E. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Measurements and Changes on SM-1 Core II During Period October 1, 1961 to May 30, 1962

Description: Tests at the SM-1 reactor are reported for the period October 1, 1961, to May 31, 1962. Loading changes were made in SM-1 Core II during the scheduled semiannual shutdowns in October to November 1961 and April to May 1962. Core physics tests include control rod bank calibrations, bank position at several temperature and xenon poison conditions vs core changes and energy release, shutdown neutron source decay and startup channel testing, and critical rod positions for stuck rod configurations. S… more
Date: July 1, 1962
Creator: Motte, F. G.; Best, W. C. & Kortheuer, J. D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Evaluation of Wire Scanner for SM-1

Description: Preliminary design concepts are presented for a wire scanner for experimentally evaluating spatial variations of neutron flux in the SM-l reactor core. Results of a literature search and determination of optimum criteria for flux mapping the core in minimum time dictated requirements for design concepts and specifications. The utility of both manually instrumented and automatically instrumented wire scanners was analyzed with respect to rapidity of measurement, selectivity of detector location,… more
Date: November 22, 1961
Creator: Kemp, S. N.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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CONCEPTUAL DESIGN AND ECONOMIC EVALUATION OF A STEAM-COOLED FAST BREEDER REACTOR

Description: A conceptual design and economic evaluation of 300 and 40 MW/.sub e/ steam-cooled fast breeder reactor power plants were performed. A reactor core composed of U-Pu oxide rod-type fuel elements clad with Inconel-X and surrounded by a blanket of depleted UO/sub 2/ fuel was studied in some detail. Reactor breeding ratios of from 1.27 to 1.5 and overall system doubling times of from 20 to 30 years are achievable. For the near term (1967) 300 MW/sub e/ plant, an energy cost of 7.6 mills/kwh is estim… more
Date: November 15, 1961
Creator: Sofer, G.; Hankel, R.; Goldstein, L. & Birman, G.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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SRE CONTROL ROD SHIELDING REQUIREMENTS

Description: Data taken on radiation traverse of the Mark 1 control rod were analyzed. Future radiation levels for all SRE control and safety rods were predicted from this. The shielding necessary to ship a complete rod and that necessary to protect a person doing maintenance work on these rods were calculated. The unshielded gamma dose rate 1 cm from the surface of the most highly activated portion of the control rod was calculated to be 5.0 x 10/sup 4/ r/hr 14 days after shutdown following an extended pow… more
Date: October 22, 1957
Creator: Whittum, H.O.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Hallam Nuclear Power Facility Preoperational Test Completion Report Dry Criticality

Description: A dry criticality test was carried out to determine the minimum critical mass of the HNPF Core without sodium. A subcritical calibration of the central control rod was performed and the relative reactivity worths of the inner ring of six control rods were determined. The extrapolated critical loading for the various plots after each incremental fuel loading with all rods out is shown. A tabulation is presented of multiplication data taken throughout the dry critical test. In order to find the r… more
Date: February 24, 1962
Creator: Kempt, H. C. & Corcoran, W. P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Transformer failure and common-mode loss of instrument power at Nine Mile Point Unit 2 on August 13, 1991

Description: On August 13, 1991, at Nine Mile Point Unit 2 nuclear power plant, located near Scriba, New York, on Lake Ontario, the main transformer experienced an internal failure that resulted in degraded voltage which caused the simultaneous loss of five uninterruptible power supplies, which in turn caused the loss of several nonsafety systems, including reactor control rod position indication, some reactor power and water indication, control room annunciators, the plant communications system, the plant … more
Date: October 1, 1991
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

ORR Operations for Period April 1960-April 1961

Description: Conversion of the ORR from 20 to 30 Mw operating level was achieved during July 1960 after a scheduled shutdown for completion of a new cooling system. Operating time reached a high of 82% during the last quarter of 1960. The first quarter of 1961 showed an operating time of 80% despite 2 shutdowns and some additional down time to repair mechanisms associated with the shim rods. Changes were made in the south engineering test facility for GCR test loops. Detailed data taken from quarterly opera… more
Date: October 20, 1961
Creator: Cox, J.A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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COOLING OF THE HFIR BERYLLIUM REFLECTOR FOLLOWING A REACTOR SCRAM OR AN ELECTRICAL POWER OUTAGE

Description: Thermal stresses in the HFIR beryllium reflector were computed for the unlikely case where the reactor is scrammed with a simultaneous loss of coolant flow and for the case following an electrical power outage where the reactor power level and the coolant flow rate are reduced simultaneously. For the case where the reactor is scrammed with a sudden loss of the coolant flow, the resulting maximum tensile thermal stress following the scram is 22,500 psi. In case of an electrical power outage, the… more
Date: December 12, 1961
Creator: McLain, H. A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

The Mixed Waste Management Facility closure and expansion at the Savannah River Site

Description: Process wastes containing radioactive and hazardous constituents have been generated throughout the operational history of the Savannah River Site. Solid wastes containing low level radionuclides were buried in Low Level Radioactive Disposal Facility (LLRWDF). Until 1986, waste containing lead and cadmium was disposed of in the Mixed Waste Management Facility (MWMF) portion of LLRWDF. Between 1986 and 1990, waste containing F-listed hazardous rags were buried. Current Resource Conservation and … more
Date: January 1, 1992
Creator: Bittner, M.F. & Frye-O'Bryant, R.C.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Dynamic Properties of Heterogeneous Water Reactors

Description: The types of tests performed in SPERT-I, and the tests proposed for SPERT-II and -III, are described. These reactors are described, and factors influencing their dynamic behavior are discussed. The tests are classed as static, step, ramp, and oscillatory. The correlation between the test results and the reactor dynamic safety characteristics (stability, self-shutdown under excursion conditions, etc.) is investigated. (T.F.H.)
Date: July 20, 1961
Creator: Forbes, S. G. & Nyer, W. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Core cooling under accident conditions at the high flux beam reactor (HFBR)

Description: In certain accident scenarios, e.g. loss of coolant accidents (LOCA) all forced flow cooling is lost. Decay heating causes a temperature increase in the core coolant and the resulting thermal buoyancy causes a reversal of the flow direction to a natural circulation mode. Although there was experimental evidence during the reactor design period (1958--1963) that the heat removal capacity in the fully developed natural circulation cooling mode was relatively high, it was not possible to make a co… more
Date: January 1, 1991
Creator: Tichler, P.; Cheng, L. (Brookhaven National Lab., Upton, NY (USA)) & Fauske, H. (Fauske and Associates, Inc., Burr Ridge, IL (USA))
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Aging assessment of auxiliary feedwater systems

Description: A study of Pressurized Water Reactor Auxiliary Feedwater (AFW) Systems has been conducted by Oak Ridge National Laboratory (ORNL) under the auspices of the Nuclear Regulatory Commission's Nuclear Plant Aging Research Program. The study has reviewed historical failure experience and current monitoring practices for the AFW System. This paper provides an overview of the study approach and results. 7 figs.
Date: January 1, 1989
Creator: Casada, D.A. (Oak Ridge National Lab., TN (USA))
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Benchmarking of flowtran with Mark-22 mockup flow excursion test data from Babcock Wilcox

Description: Version 16.2 of the FLOWTRAN code with a Savannah River Site (SRS) working criterion (St=0.00455) for the onset of significant void (OSV) was benchmarked against power and flow excursion data derived from tests at the Babcock Wilcox Alliance Research Center test facility. This document presents analyses which show that FLOWTRAN accurately predicts the mockup test assembly thermal-hydraulic behavior during the steady state and LOCA transient conditions, and that FLOWTRAN with a Savannah River Si… more
Date: June 1, 1990
Creator: Chen, Juo-Fu.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

CORE REMOVAL COOLING SYSTEM-SECTION II. CORE I, SEED I. Test Results T- 641113. Section 2

Description: A test was performed on June 19, 1959 to determine the capacity of the Core Removal Cooling System for removing reactor decay heat under split-flow'' conditions. The system operated satisfactorily during this test; the pumps developed a flow of approximates 73 gpm at a total head of 254 ft water, as compared with their rated capacity of 75 gpm at a total head of 250 ft water. (D.L.C.)
Date: May 19, 1961
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Safety aspects of the US advanced LMR (liquid metal reactor) design

Description: The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the US program is to produce a standard, commercial ALMR, including the associated fuel cycle. This paper discusses the US regulatory framework for design of a… more
Date: January 1, 1989
Creator: Pedersen, D.R.; Gyorey, G. L.; Marchaterre, J. F.; Rosen, S. (Argonne National Lab., IL (USA); General Electric Co., San Jose, CA (USA); Argonne National Lab., IL (USA) et al.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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A review of the Los Alamos effort in the development of nuclear rocket propulsion

Description: This paper reviews the achievements of the Los Alamos nuclear rocket propulsion program and describes some specific reactor design and testing problems encountered during the development program along with the progress made in solving these problems. The relevance of these problems to a renewed nuclear thermal rocket development program for the Space Exploration Initiative (SEI) is discussed. 11 figs.
Date: January 1, 1991
Creator: Durham, F.P.; Kirk, W.L. & Bohl, R.J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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SUMMARY OF HRT RUNS 22, 23 AND 24

Description: In runs 22, 23, and 24 the HRT was operated with downward flow through the core at powers up to 5 Mw. Because of leakage past the lower core tank patch, uranium could not be prevented from entering tbe blanket, so the concentration was deliberately kept above 1 g U/kg D/sub 2/O. The core power was about 0.6 of the total. Nuclear power fluctuations were larger than with upward core flow, and the cause was investigated intensively in run 22 at powers up to 1.8 Mw. Run 22 was terminated after 778 … more
Date: March 1, 1962
Creator: Bauman, H.F.; Buchanan, J.R.; Engel, J.R.; Haubenreich, P.N. & Richardson, D.M.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

SM-1 Reactor Vessel Cover and Flange Stress Analysis

Description: The maximum stress calculated for the SMl-1 reactor vessel closure studs occurs during operation at full power. This value is 27,180 psi of which 19,800 psi is tension and 7380 psi bending. This stress does not include a stress concentration factor for effect of threads. It was eonservatively assumed the studs were initially tightened to a code allowable stress of 20,000 psi as specified in the ASME Code rather than the lesser stress obtained by the normal operating procedure. The maximum calcu… more
Date: February 19, 1962
Creator: Sayre, M. F.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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