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Control and Dynamics Performance of a Sodium Cooled Reactor Power System

Description: Introduction: Objectives and Method of Approach. High plant efficiencies can be realized without excessively high core temperatures and high coolant pressures by the use of liquid metal coolant. In an attempt to prove the feasibility of liquid sodium as a reactor coolant ALCO Products, Inc., under sponsorship of the Atomic Energy Commission, is undertaking a design study of three vital system components: the intermediate exchanger, the boiler, and the superheater. Since, in the past programs,… more
Date: 1960
Creator: Alco Products (Firm).
Item Type: Refine your search to only Report
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Evaluation of Kanigen, Electroless Nickel Plating for Steam Side of a Sodium Component Steam Generator

Description: Introduction: This is a final report on the evaluation of Kanigen electroless nickel plating for surfaces in contact with water and steam i a sodium heated AISI Type 316 stainless steel steam generator. The purpose of the coasting was to afford protection from stress corrosion cracking originating on the water-steam side of the unit. It has been concluded that the kanigen coating does not afford adequate protection for the services condition intended. This work was performed as part of the r… more
Date: February 15, 1961
Creator: Alco Products (Firm).
Item Type: Refine your search to only Report
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Intermediate Heat Exchanger and Steam Generator Final Design. Volume 1. Thermal & Mechanical Design

Description: This technical report represents the final design for a sodium to sodium intermediate heat exchanger and a sodium to water steam generator. The intermediate heat exchanger is designed for operation in a nuclear power plant using liquid sodium as the primary and secondary coolant. Since the primary fluid coming fro the reactor is radioactive, the purpose of the IHX is to transfer heat to a nonradioactive fluid which then goes to the steam generator. Because of radioactivity the unit will be encl… more
Date: September 30, 1960
Creator: Alco Products (Firm).
Item Type: Refine your search to only Report
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Intermediate Heat Exchanger and Steam Generator Final Design. Volume 2. Chemical & Stress Analysis

Description: Introduction: This volume deals principally with the chemical analysis and the stress analysis for a sodium to sodium intermediate heat exchanger and a sodium to water steam generator. The work presented is an extension and modification of the analysis presented in the preliminary design report. The chemical analysis covers the sodium cover gas system and the effects of sodium-water reactions in the event of a leak in the steam generator. Considerable design work was done in an effort to mainta… more
Date: September 30, 1960
Creator: Alco Products (Firm).
Item Type: Refine your search to only Report
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Intermediate Heat Exchanger and Steam Generator Final Design. Volume 3. Specifications

Description: Introduction: Sodium Components Material Specifications. Twenty-three material, inspection and welding specification are presented for the various parts of both the intermediate heat exchanger and steam generator. Tables indicate the applicable parts and assemblies to which these specifications shall apply. For other parts, where the material requirements are not severe, the ASTM or other indicated specifications shall apply.
Date: September 30, 1960
Creator: Alco Products (Firm).
Item Type: Refine your search to only Report
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Intermediate Heat Exchanger and Steam Generator Final Design. Volume 4. Operation & Maintenance

Description: This technical report contains the operation and maintenance specifications for the intermediate heat exchanger and the steam generator. The report contains eight sections: (1) General Information, (2) Shipping and Installation, (3) Operation Procedures, (4) Scram and Casualty Shutdowns, (5) Leaks, (6) Instrumentation and Control, (7) Maintenance, and (8) four Appendixes (a) Boiler Water Chemistry Recommendations, (b) Final Concept Drawings, (c) Industrial Nucleonics Literature on Liquid Level… more
Date: September 30, 1960
Creator: Alco Products (Firm).
Item Type: Refine your search to only Report
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Scale "Up or Down" Analysis for Prototype Test

Description: Introduction: In conjunction with the final design and development of a 70 MW sodium intermediate heat exchanger and a sodium steam generator, an analysis is required which can be used as a basis for a determination to scale up or scale down the designs. Included in this analysis are those considerations leading to the recommendation of the best prototype test unit and to some of the limits imposed on scaling up or down when considering future applications of designs other than those actually … more
Date: May 1, 1960
Creator: Alco Products (Firm).
Item Type: Refine your search to only Report
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Test Program for 30 Megawatt Prototype Sodium Intermediate Heat Exchanger and Steam Generator

Description: The designer's concept of a test program for the 30-Mw prototype intermediate heat exchanger and steam generator designed and fabricated as part of the Sodium Components Development Program is presented. The performance data will serve to verify the thermal design, or allow application of improved techniques to future designs, give an improved basis for stress analysis in design of future units, and demonstrate the capability and limitations of the units in relation to the performance specifica… more
Date: September 14, 1962
Creator: Alco Products (Firm).
Item Type: Refine your search to only Report
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Tube-To-Tubesheet Welding Development Programs to 30 Megawatt Prototype Sodium Intermediate heat Exchanger and Steam Generator

Description: Special welding techniques for joining Inconel weld overlays on type 316 stainless steel tubesheets and channels are described. The program for the development of the required welding procedures are divided into three specific programs: overlaying Inconel filler Metal 82 on type 316 stainless steel, welding type 316 stainless steel tubes to a type 316 stainless steel tubesheet, and welding Inconel-type 316 stainless steel composite tubes to Inconel weld overlaid type 316 tubesheets. Shock tests… more
Date: 1963
Creator: Alco Products (Firm).
Item Type: Refine your search to only Report
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SM-1 Research and Development Program: Long-lived Induced Activity Buildup During SM-1 Core I Lifetime. Task XVIII, Phase I

Description: Abstract: The results of activity buildup studies in the SM-1 performed during Core I lifetime (June 3, 1957 to April 28, 1960) are reported. Data are presented on the extent, nature, and mechanism of the buildup of long-lived gamma emitting nuclides in the reactor primary system. Radiation levels after reactor shutdown are presented, as well as mathematical equations used to account for the observed activity levels. The data have shown that Co60 is the major contributor to radiation levels in… more
Date: November 30, 1960
Creator: Bergmann, C. A.; Bergen, C.; Cox, J. F.; Chupak, J. & Grant, L. G.
Item Type: Refine your search to only Report
open access

Interim Report of Nuclear Analysis Performed on SM-2 Core and Vessel : September 1, 1958 to December 31, 1959.

Description: Abstract: This technical report contains a description of the nuclear analysis performed upon the SM-2 core and vessel for the period September 1, 1958 to December 31, 1959. Calculations are given for core reactivity, power distributions, lifetime, reactor control, kinetics, radiation problems, fuel and poison burn-ups, and the nuclear effects of poisons, temperature, and geometry. Wherever possible, experimental data is included in order to test the validity of the analytical models. The SM-… more
Date: May 27, 1960
Creator: Bobe, P. E.; Birken, S. H.; Byrne, B. J.; Clancy, E. F. & Fried, B. E.
Item Type: Refine your search to only Report
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Investigation of Local Boiling of SM-1

Description: Abstract; SM-1 Reactor Core I Rearranged and Spiked, and Core II with Special Components were analyzed under various off-design conditions to induce nucleate boiling. The steady state code, STDY-3, written for the thermal analysis of pressurized water cores, was employed for the analysis. The code performs a complete steady state parallel channel thermal analysis for both nominal and hot channels. Thermal characteristics of individual elements were investigated while changing the parameters of… more
Date: June 20, 1961
Creator: Bradley, P. L.
Item Type: Refine your search to only Report
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Hazards Report for SM-1 Core Temperature and Flow Instrumentation (Task XIV) Covering Special Test Procedures.

Description: Abstract: Test procedures for special tests involving in-core SM-1 temperature and flow instrumentation are described (Task XIV Package Tests). These tests involve in-core steady state flow and temperature measurements, loss of flow transients, load transients, reduced primary system pressure operations and reduced element flow. The thermal and hydraulic conditions prevailing in these tests, including steady state and transient burnout rations, are developed. The effects of reduced system pres… more
Date: February 28, 1962
Creator: Bradley, P. L. & Coombe, J. R.
Item Type: Refine your search to only Report
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SM-2 Full Scale Flow Studies Termination Report

Description: Abstract: Hydrodynamic flow studies were conducted on a full scale model of the SM-2 reactor vessel and core. Test fluid was water at 200 psi and 200 degree F. Test facilities, model, and instrumentation design are discussed. Flow distribution in the stationary fuel elements, lattices, and control rods of the second pass was investigated. Pressure losses through the various core components were measured and are compared with calculated values. Observed over-all pressure drop was 71 feet of wate… more
Date: July 30, 1961
Creator: Christenson, J. A.; Richards, W. M. S. & Davidson, S. L.
Item Type: Refine your search to only Report
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Hazards Summary Report for the SM-1 Core Temperature and Flow Instrumentation: Task XIV

Description: Abstract; This technical report describes the changes in the SM-1 incurred by the experiment, Core Temperature and Flow Instrumentation (Task XIV), and evaluates the possible hazard involved in these changes. Temperature and flow measurements will be taken on a Task XIV instrumented stationary fuel element, instrumented control rod fuel element and other selected points in the SM-1 core to provide data on the core steady state and transient performance. The hazards evaluation consists of a nu… more
Date: March 30, 1961
Creator: Coombe, J. R.; Brondel, J. O.; Lee, D. H. & Matthews, F. T.
Item Type: Refine your search to only Report
open access

Hazards Evaluation of the SM-1 Penetrated Gasket

Description: Abstract: This technical report describes the as-constructed SM-1 penetrated gasket designed for SM-1 Core and Flow Instrumentation (Task XIV). This report supplements APAE No. 79, The Summary Hazards Report for Task XIV, and evaluates the effects of a postulated failure of this gasket. The effects of failure on the Maximum Credible Accident are determined and conclusions and recommendations for the use of this gasket are made.
Date: September 8, 1961
Creator: Coombe, J. R.; Gebhardt, F. G. & James, B.
Item Type: Refine your search to only Report
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Hazards Report for SM-1 Core II Without the SM-1 Core I High Burnup Elements and With the PM-1-M-2 Element

Description: Abstract: The removal of both SM-1 Core I high burnup elements from SM-1 Core II and the insertion of the PM-1-M-2 element and the SM-1 Core I spare element in SM-1 Core II is discussed. Nuclear and thermal characteristics of Core II with these changes are presented and conclusions related to the changes in the hazard potential are made. If the core change indicated by this report is made, local peaking factors will be decreased and burnout ratios will be increased. This, of course, in itself… more
Date: October 7, 1961
Creator: Coombe, J. R.; Lee, D. H. & Mathews, F. T.
Item Type: Refine your search to only Report
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Hazards Report for SM-1 Core II With the SM-1 Core II High Burnup Elements Replaced with SM-1 Core I Spare Elements

Description: Abstract: The removal of both SM-1 Core I high burnup elements from the SM-1 Core II and the insertion of two SM-1 Core I spare elements i their places are discussed. Nuclear and thermal characteristics of Core II with the change are presented and conclusion related to the change in hazard potential are made. If the core change indicated by this report is made, local peaking factors will be decreased and burnout ratios will be increased. This, of course, in itself leads to a more conservative… more
Date: October 9, 1961
Creator: Coombe, J. R.; Lee, D. H. & Matthews, F. T.
Item Type: Refine your search to only Report
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Hazards Report for Insertion of the PM-1-M-2 Element in the SM-1 Core II

Description: Abstract: This technical report describes the Martin Co. PM-1-M-2 test element and analyzes the potential hazard incurred by its inclusion in the SM-1 Core II. A nuclear analysis develops power distributions and reactivity effects. Hydraulic and thermal analyses develop anticipated burnout heat flux ratios. An evaluation of the risk involved with the inclusion of this element is presented. In view of the narrow margin by which the PM-1-M-2 test element meets the minimum burnout ratios as def… more
Date: September 1, 1961
Creator: Coombe, J. R.; Scoles, J. F.; Brondel, J. O. & Lee, D. H.
Item Type: Refine your search to only Report
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Hazards Report for the SM-1 Core II With Special Components

Description: Abstract: This technical report describes the changes incurred in the SM-1 by the insertion of the SM-1 Core II and special components. The special components consist of impact specimens, a boron gradient rod, SM-2 elements, a PM-1-M element, and high burnup SM-1 Core I elements. The change in hazards, due to operation of SM-1 with Core II and the special components is evaluated. The analysis indicates there is no change in hazards.
Date: March 30, 1961
Creator: Coombe, J.; Lee, D.; Segalman, I. & Robertson, R.
Item Type: Refine your search to only Report
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Hazards Report for PM-2A Core II

Description: Abstract: This technical report describes the changes incurred in the PM-2A by the planned insertion of PM-2A Core II and the replacement of the startup and check sources. PM-2A Core II components were fabricated to specifications very nearly identical to those of PM-2A Core I. The essential difference in the cores is the boron loading which permits PM-2A Core II to meet a "one-stuck rod criteria" at beginning of life. This core has been subjected to a zero power experiment and loading procedur… more
Date: June 21, 1962
Creator: Coombe, John R. & Stephenson, L. D.
Item Type: Refine your search to only Report
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Thermal Analysis of SM-1 Core III

Description: Abstract This technical report covers the thermal analysis performed on the SM-1 Core III for both steady state and transient conditions is reported. SM-1 Core III will be used as a test for Type 3 elements in a PM-2A Core. The steady state analysis indicated minimum departure from nucleate boiling ratios (DNBR) for both design and scram conditions above the minimum criteria of 1.5. Local nucleate boiling was noted in the hot internal channels and lattice passage at scram power conditions. Loss… more
Date: June 29, 1962
Creator: Davidson, S. L.
Item Type: Refine your search to only Report
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Hazards Report for the SM-1 Core II Without Special Components

Description: Abstract: This technical report describes the changes incurred in the SM-1 by the insertion of the SM-1 Core II without special components. The SM-1 Core II components were made to specifications very nearly identical to those of SM-1 Core I. The differences consist of europium absorber sections, internal europium flux suppressors in the control rod fuel elements, and low impurity cladding. Each of the SM-1 Core II components with the exception of the five absorber sections new in SM-1 Core … more
Date: April 19, 1961
Creator: Gallagher, J. G.
Item Type: Refine your search to only Report
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