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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
open access

Supporting Analysis for Thermal Suitability of Fuel Elements for SM-1A Core I Loading

Description: A recommended SM-1A Core I loading chart was derived from available, metallurgically acceptable elements at the SM-1A and SM-1 sites. The derivation was based on local thermal and hydraulic considerations of minimum elementto- element coolant channel clearances. These clearances were determined from field inspection measurements of outer fuel plate spacing, as modified by analytical calculations of plate ripple growth during exposure to reactor operating thermal stresses. (auth)
Date: January 10, 1962
Creator: Brondel, J. O.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

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
open access

Evaluation of Heat Rejection Systems for PL-3

Description: An investigation was made of heat rejection systems for use in the PL-3 nuclear power plant, designed for placement at Byrd Station, Antarctica. It was concluded that the glycol-coupled surface condenser and air blast cooler combination appears to be suited for PL-3 plant requirements and operating conditions. (auth)
Date: March 1, 1962
Creator: Thurnau, C. J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Spent Fuel Transfer, Storage and Shipment for PL-3

Description: In refueling development studies performed on PL-3 Phase I design, several methods of fuel transfer, storage, and shipment were investigated. An evaluation of the relative merits of the systems and designs under study, as applied to either the BWR or PWR concepts, is made and optimum designs are selected. An analysis of spent fuel shipping cask shielding requirements is presented, along with recommendations for future study in this area. (auth)
Date: March 1, 1962
Creator: Hauenstein, G. C. & Pomeroy, D. L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

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
open access

Detailed Stress Analysis of SM-1 Steam Generator Tube Sheet

Description: The detailed stress analysis of the SM-1 steam generator tube sheet showed it to be safe from strain cycling damage. However, the pressure stresses were greater than the yield strength during the hydrostatic test. The differential between pressure stresses and yield strength indicates that some initial deformation may have taken place in the tube sheet. (auth)
Date: July 11, 1962
Creator: Busuttil, J. J. & Chittum, R. A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Steady State and Transient Thermal and Hydraulic Analysis of SM-2 Termination Report

Description: Thermal characteristics of the SM-2 core were analyzed at steady state and loss of flow conditions. For steady state operation, the steady state code STDY-3 was used. For transients during-a loss of flow acident, ART-02, a onedimensional code, was used. This analysis indicated the SM-2 core is safe from burnout under steady state operation at design power level (28 Mw(t)) because no nucleate boiling exists, and the minimum burnout ratio is above 2.0. The core is safe from burnout under loss of … more
Date: September 1, 1961
Creator: Segalman, I. & Bradley, P. L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Economic Analysis of Replacement Cores for SM and PM Type Reactors

Description: An economic analysis is presented for the fabrication of replacement cores for SM and PM type reactors, including analysis of various core types and core fabrication technologies. The analysis indicates that major savings are possible by utilizing Type 3 cores (40-mil plates, 25 wt% UO/sub 2/, welded assembly) in all SM and PM type reactors, and that significant savings are possible by multiple core procurement and reprocessing, and relaxation of cobalt and tantalum requirements in Type 347 sta… more
Date: October 1, 1961
Creator: Wilder, A. S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Research and Development Reports for Sodium to Sodium Intermediate Heat Exchanger and Sodium to Water Steam Generator

Description: Results are presented of research and development work performed in conjunction with the 70-Mw design of a sodium-to- sodium intermediate heat exchanger and a sodium-to-water steam generator. Kanigen plating was substituted for Inconel overlays. A program to evaluate this plating was undertaken. Elimination of tube end ferrules, mechanical behavior of sine wave tubes, tube-to- tube sheet welded connections, metallurgical examination of bimetallic tubes, transition weld test, bayonet tube test, … more
Date: October 30, 1960
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Comparison of Gaseous Waste Handling Systems for PL-3

Description: Gaseous waste handling systems are compared for direct and indirect boiling water and pressurized water reactors for PL-3 application. Areas that are common to the various concepts are not discussed since they do not enter into a comparison study. The major differences present are in the handling of active gases released to or held in the primary system coolant. These gases which could be present, their possibIe release from the system, and the necessary processing requirements are discussed in… more
Date: February 28, 1962
Creator: Noble, J. H. & Duke, E. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Mathematical Analysis of Rippling of Type 1 Fuel Plates. Part 1

Description: Rippling phenomena due to heating in fuel plates of SM and PM type reactors are investigated analytically using small deflection theory of plates. Temperature variations across the width of the plate are accounted for. Detailed calculations are conducted for simply supported plates. It is found that within the limitations imposed by small deflection theory that the amplitude of the plate ripples induced by the heating is directly proportional to the initial amplitude. (auth)
Date: July 1, 1962
Creator: Beck, S. D. & Miller, J. V.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Primary Shield Optimization Survey for the PL-3 Reactor

Description: A detailed study of four reactor and shield configurations was made. Two basic reactor types, the boiling water and pressurized water reactors were considered. Shield materials of lead-water and iron-water were used with varying thicknesses for determining the optimum shield configuration for the PL-3 reactor. Also presented is a survey of available shielding codes. (auth)
Date: June 28, 1962
Creator: Scoles, J. F. & Crouch, A. N.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Single Element Flow Tests for Type 3 (SM-2) Fuel Elements in SM-1, SM-1A, and PM-2A Cores

Description: Channel-to-channel flow distribution within Type 3 (SM-2, stationary and control rod fuel elements modified for use in the SM-1, SM1-1A, and PM-2A core support structures and control rod tubes was measured in single element flow testing. Plots of channel-to-channel flow distribution and element pressure drop at various element flow rates are given. Flow distribution for the top-orificed SM-1A and PM-2A stationary elements was within the plus or minus 12% deviation from element average utilized … more
Date: November 27, 1961
Creator: Krause, P. S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Dynamic Simulation of Multi-Pass Pressurized Water Nuclear Power Plants by Analog Computer Techniques

Description: A kinetic model of the primary loop of a multi-pass pressurized water reactor power plant is developed to evaluate, by analog computer techniques, the transient response characteristics under conditions of steam generator load and reactor control rod perturbations. Using the 2-pass 28 Mw(t) SM-2 reactor as a typical plant, transient behavior patterns are illustrated and examined for a variety of load inputs, variations in plant constants, and analog model simplifications. (auth)
Date: June 1, 1961
Creator: Brondel, J. O.
Item Type: Refine your search to only Thesis or Dissertation
Partner: UNT Libraries Government Documents Department
open access

Thermal Stress Testing of Type 1 Fuel Plates

Description: Thermal stress tests on Type 1, SM-1A Core II fuel ele-ment sections were performed to study plate distortion and determine its dependency on temperature distribution, temperature differential, initial flatness, and ripple length. Test results will be correlated with the analytical model and used to predict ripple growth in other plate-type fuel elements. The tests showed that ripple growth is dependent on initial flatness of the plate and that the characteristic shape of ripples is maintained … more
Date: June 27, 1962
Creator: Gebhardt, F. G.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Selection of Core Design No. 1 for Type 5 Replacement Cores in SM-1 and SM-1A

Description: Nuclear and thermal analyses were performed to determine the characteristics of the Type 5 core in the SM-1 and SM-1A reactor plants as a function of geometry and composition. The following nuclear properties were investigated: core energy release, maximum midlife reactivity, average fuel burnup fraction, B-10 reactivity coefficient, and power distribution. Thermal parameter surveys determined the effects of channel thickness and power distribution upon the DNBR, nominal and hot channel thermal… more
Date: July 1, 1962
Creator: Davidson, S. L. & Paluszkiewicz, S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Stress Analysis of the SM-1A Reactor Vessel

Description: The stress analysis performed on the SM-lA reactor vessel and cover is presented. The msximum combined stress (51,360 psi in compression) occurs in the vessel cover during a 50 deg F/hr transient. A fatigue analysis of these stresses indicated that they could be applied safely at least 2500 times, and since the vessel is expected to receive less than 900 cycles, it should not suffer any fatigue damage. (auth)
Date: March 30, 1962
Creator: McLaughlin, D. W.; Rowekamp, B. J.; Chittum, R. A. & Aitken, C. C.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Nuclear Analysis of Various Spert-III Critical Experiments

Description: Editor please delete 26456.<><DSN>16:026457<ABS>Work done in the P122 reactor control actuator area is summarized. Actuators were required to radially position the absorber blades in the core of the reactor. The P122C1 was a subsonic power plant and temperatures were low enough to permit the use of hydraulics in the actuator area. The program was reoriented and the power plant designated P122C3 which was a supersonic version of the folded flow power plant. The ambient temperature at maximum pow… more
Date: April 27, 1961
Creator: Paluszkiewicz, S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Supporting Analysis and Derivation of Dimensional Tolerance Specifications for Core II of SM-1A & PM-2A

Description: A method is presented for translating inspection measurements of fuel plate spacing to obtain minimum coolant channel clearances under reactor operating conditions. Considerations of fuel plate ripple growth and the inspection procedure used are included. The method is applied to establish dimensional tolerance specifications used for the procurement of SM-1A and PM-2A Core II. (auth)
Date: November 1, 1961
Creator: Brondel, J. O.
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
open access

Stress Analysis of the PM-2A Reactor Vessel

Description: The stress analysis performed on the PM-2A reactor vessel and cover is discussed. The maximum combined stress (51,000 psi) occurred in the studs after reaching steady-state conditions. A fatigue analysis indicated that this stress could be safely applied 2500 times, and since the studs do not approach 2500 cycles from initial stud tightening to steady-state conditions, they should not suffer any fatigue damage. (auth)
Date: May 14, 1962
Creator: Rowekamp, B. J.; McLaughlin, D. W.; Chittum, R. A. & Aitken, C. C.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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