Search Results

open access

Decontamination Program Task II. Volume 1, Contamination and Decontamination in Nuclear Power Reactors

Description: Abstract: A survey of the problem of reactor system contamination by radioactive material and methods that have been employed to remove the material was carried out. Following this survey, an investigation of chemical solutions was undertaken to find one which might be successfully employed in the decontamination of a stainless steel steam generator. From a preliminary screening, the most promising chemical method from the view point of minimum corrosion and maximum decontamination is a caustic permanganate treatment followed by and acid rinse.
Date: February 13, 1959
Creator: Zegger, John L. & Pancer, Guyon P.
Partner: UNT Libraries Government Documents Department
open access

Decontamination Program Task II. Volume III, Recommended Procedure for Decontamination of a Stainless Steel Steam Generator

Description: Abstract: A decontamination procedure for a stainless steel steam generator similar to the APPR-1 using a fill-flush application of a caustic permanganate-citrate combination solution is recommended. The isolation of the steam generator is to be accomplished by means of specially designed plugs at the reactor vessel outlet and at the primary coolant pumps. Anticipated results, including corrosion rates and decontamination factors, are presented.
Date: February 13, 1959
Creator: Pancer, Guyon P. & Zegger, John L.
Partner: UNT Libraries Government Documents Department
open access

SM-1 (APPR-1) Research and Development Program : Final Report on Short-Lived and Fission Product Activity in the SM-1 Primary Coolant, Task III

Description: Abstract: The primary coolant of the SM-1 (APPR-1) was analyzed for short-lived and fission product activities. Manganese-56 was found to be the predominant non-fission product nuclide contributing to the short lived activity. Fission products were found in the coolant. It is concluded these fission products originate from a defects in the cladding and from surface contamination of the fuel elements.
Date: March 10, 1959
Creator: Brown, William S. & Hasse, Robert A.
Partner: UNT Libraries Government Documents Department
open access

Theory of Asymmetric Arrays of Control Rods in Nuclear Reactors

Description: Introduction: Seldom does the actual arrangement of control elements in a nuclear reactor confers to the ideal and convenient mathematical array. In order to achieve shim control. safety and regulation, it is desirable to design with rods of different sizes and materials. With given fuel element arrangement, typically in square or hexagonal lattice spacing, there will be rods located at different distances form the center of the core and from each other. As the reactor operates, absorbers will be withdrawn, leaving further asymmetries in the location of those remaining. It is the purpose of this report to develop in detail a two-group diffusion theory with as complete generality as possible. The method is as yet restricted to the unreflected core, or to the reflected core by use of reflector savings and bare equivalent geometries.
Date: April 25, 1959
Creator: Murray, Raymond L.
Partner: UNT Libraries Government Documents Department
open access

SM-1 Research and Development Program : Final Report on Fission Product Activity in the SM-1 Primary Coolant, Task XIII

Description: Abstract: Fission product measurements were made on the SM-1 primary coolant. The airborne activity observed during the sampling of the primary system was identified. An analysis was made on the primary coolant for alpha activity and on the secondary water for fission production iodine.
Date: June 30, 1959
Creator: Hasse, Robert A.
Partner: UNT Libraries Government Documents Department
open access

SM-1 Research and Development Program, Activity Buildup Program Task 1 : final report February 1958 to June 1959

Description: Abstract: The results of activity buildup studies in the SM-1 (APPR-1) performed from February 1958 to January 1959 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. Mathematical equations to describe the activity buildup are derived. Radiation levels after reactor shutdown are presented, as well as the predicted radiation levels at the end of core life.
Date: August 10, 1959
Creator: Brown, William S.; Bergen, C. Richard.; Bergmann, Carl A.; Chupak, Julius.; Fitzsimmons, Susanne R. & Grant, Louis G.
Partner: UNT Libraries Government Documents Department
open access

SM-2 Critical Experiments : CE-1

Description: Abstract: Critical experiment studies were performed, varying the parameters U235, B10 and metal to water ratio, in the SM-2 7 x 7 core configuration with 38 stationary elements and seven control rods of the SM-1 (APPR-1) type. An experimental mock-up of the SM-1 was assembled using the basic SM-2 fuel plates. Excellent agreement between the SM-1 boron loading, determined by chemical analysis, and the SM-1 mock-up boron loading, for equivalent bank positions, was noted. Several SM-2 mock-ups, cold clean and midlife, were assembled and studied with regard to reflector effects, flow divider effects, relative control rod array worths, critical rod configurations, and relative power distributions. The results of these experiments indicate as satisfactory a U235 loading of 36.4 Kg and a B10 loading of 63.4 grams for the SM-2. Attention is drawn to numerous power peaks present in the active core. The open seven control rod array has a slight reactivity advantage over the closed seven array and consequent minor disadvantage with respect to "stuck rod" criteria.
Date: November 30, 1959
Creator: Noaks, J. W.; McCool, W. J.; Robinson, R. A.; Schrader, E. W. & Weiss, S. H.
Partner: UNT Libraries Government Documents Department
open access

Extended SM-2 Critical Experiments : CE-2

Description: Abstract: This technical report contains a description and results of a second series of critical experiments performed on the SM-2 core mock-up, as additional to the first series of experiments reported in APAE No. 54. The SM-2 core mock-up contains 36.4 kg U-235 and and estimated 67.9 gm B-10. The equivalent diameter and the active height are about 22 in.; the metal-to-water volume ration is 0.344. Data is presented on activation, reactivity, and stuck rod measurements. All measurements were conducted on the open seven control rod array employing 38 stationary fuel elements. Activation measurements consisted of neutron flux measurements using uranium fission foils for relative power distribution studies, the effect of flux suppressors on reducing power peaks, blocked coolant channel measurements, and gamma ray dose distribution. Reactivity measurements were performed to determine the effect f flow divider, flux suppressors and stimulated high temperature and pressure operation; b-10 loading in the SM-2 core; and core material coefficients. For the later, the worth in cents per gm or cents per cc was determined at simulated temperature of 510 degree F for B-10, U-235, stainless steel, and void. Stuck rod measurements were made to obtain an indication of the criticality margin in the event one or more control rods should stick in the operating position.
Date: June 30, 1961
Creator: McCool, W. J.; Robinson, R. A.; Weiss, S. H.; Raby, T. M.; Schrader, E. W. & Walthousen, L. D.
Partner: UNT Libraries Government Documents Department
open access

SM-1 Reactor Core Inspection at 2/3 Core Life : March 7, 1959 to May 17, 1959

Description: Abstract: This technical report is concerned with the program and results of the SM-1 reactor vessel head removal and core inspection at Fort Belvoir, Virginia from the period March 7, 1969 to May 17, 1959. It covers the operating procedures in detail and records the conditions found and problems encountered in order to make a record for reference in future work of this nature. The major objective of the program, to obtain irradiation data on the SM-1 type core, has been met. The boron control rod elements were found unsatisfactory for full core life, and europium oxide elements were placed in the core for future irradiation stability data. A major problem was experienced with the cracked pressure vessel head studs. The methods developed for the removal of the broken studs are presented. The complete metallurgical study of the stud failure from stress corrosion is included as Appendix B.
Date: January 13, 1960
Creator: Obrist, C. H.; Byrne, B. J.; Connolly, T. F.; Foley, D. D.; Lichtenberger, R. V.; Mackay, S. D. et al.
Partner: UNT Libraries Government Documents Department
open access

DuPont Prototype Safety and Control Rod Drive Testing

Description: Summary: Prototype testing of the safety and control rod drives indicated that both units functioned properly. No major problems were encountered during testing. Seal leakage data collected indicated that the seal units were performing satisfactorily. Scram times during both cold and hot testing were excellent and actually better than expected.
Date: April 25, 1960
Creator: VandeMark, G. M. & Krause, P. S.
Partner: UNT Libraries Government Documents Department
open access

Zero Power Experiments for the SM-1 Core II : Task XV

Description: Abstract: An element by element reactivity check for SM-1 Core II fuel elements and control rod absorber sections was performed and the burnable nuclear poison loading in the SM-1 Core II stationary fuel elements was established. An approach to criticality of the SM-1 Core II was performed by the inverse multiplication method and the critical rod bank position obtained as a function of fuel loading up to the full SM-1 Core II loading. Maximum and minimum core reactivity measurements were obtained by selective loading of stationary fuel elements and the total "excess K" for the core established. Power distribution measurements in the region of the core-reflector interface and the fuel-absorber interface in the control rod assemblies were performed. The effectiveness of europium flux suppressors in the top of control rod fuel elements and the power peaking in stationary elements adjacent to water gaps in control rod assemblies were measured. Survey measurements established the worth of spiking cold clean SM-1 cores with SM-2 elements, and of water holes in the SM-1 core which might be utilized as flux traps for materials irradiation.
Date: March 15, 1960
Creator: Robinson, R. A.; Weiss, S. H.; McCool, W. J. & Schrader, E. W.
Partner: UNT Libraries Government Documents Department
open access

SM-1 Research and Development. Task XV, Zero Power Experiments for SM-1 Core II and SM-1A Core I

Description: Abstract: A zero power experiment on the SM-1 Core II included an element by element reactivity check of fuel elements and control rod absorber sections, and an estimate of burnable nuclear poison loading in stationary fuel elements. An approach to criticality was made by the inverse multiplication method, and critical rod bank position obtained as a function fuel loading up to full core loading. Minimum and maximum core reactivity measurements were obtained by selective loading of stationary fuel elements, and total excess K of the core was established. Power distribution measurements were taken in the regions of the core-reflector interface and the fuel-absorber interface in the control rod assemblies. The effectiveness of europium flux suppressors in the top of control rod fuel elements was determined, and power peaking was measured in stationary elements adjacent to control rod assembly water gaps. Survey measurements established the work of spiking clean, cold SM-1 cores with SM-2 elements and the work of water holes in the SM-1 Core. The reduced scope zero power experiment performed on SM-1A core I included an element by element uniformity check of stationary fuel elements, a core assembly test, comparison of Eu2O3 and B4C absorber sections, and development of initial core loading procedures for the SM-1A plant.
Date: October 12, 1960
Creator: Robinson, R. A.; Weiss, S. H.; McCool, W. J. & Schrader, E. W.
Partner: UNT Libraries Government Documents Department
open access

A Survey of the Effects of Neutron Irradiation on the Impact and Other Mechanical Properties of Pressure Vessel Steels for the SM-2 Reactor

Description: Abstract: This technical report summarizes the data obtained in a recent literature survey conducted to determine the effects of neutron irradiation on the impact and other mechanical properties of both ferritic steels and austenitic stainless steels. The survey was primarily aimed at obtaining sufficient data on the behavior of pressure vessel steels at high integrated neutron flux levels in order that a reference material of construction could be selected for the SM-2 (APPR-1B) reactor vessel. Materials studied in this literature survey included carbon and low alloy steels such as: ASTM A-212B, ASTM A-201, ASTM A-301B (CR-Mo), ASTM A-106 (coarse and fine grained), ASTM A-285, ASTM A-302B (Mn-Mo), ASTM A-353, ASTM A-203 Grade D, E-7016 carbon steel weld metal, USS Carilloy T-1, HY-65 and HY-80. In addition, Types 304 and 347 stainless steels were also investigated as representative austenitic materials which might be used in pressure vessel construction. A careful evaluation was made of the irradiation induced changes in the mechanical properties of the above materials. The ferritic steels were evaluated primarily on the basis of increases in transition temperature due to irradiation and decreases in the amount of maximum energy absorbed prior to ductile failure. Factors such as industrial experience, changes in other mechanical properties and the susceptibility of these materials to temper embrittlement were also considered. Austenitic stainless steels were evaluated on the basis of post-irradiation and low temperature impact strength and on irradiation induced changes in other mechanical properties. Based on available data, it is concluded that austenitic stainless steels are capable of resisting harmful property damage at integrated neutron fluxes > 1 Mev of at least t to 2 x 10(21) nvt. However, with the application of special reactor operating procedures, ASTM A212B will be satisfactory at integrated neutron fluxes up to 5 x 10(19) nvt.
Date: April 1, 1960
Creator: Kelleman, Richard William.
Partner: UNT Libraries Government Documents Department
open access

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, the nuclear reactor had been the major focus of attention, the development of the sodium cooled reactor and sodium pumps for this application are thought to need the less development than the heat exchanger equipment. Consequently, parallel design studies of the reactor, pumps, and other system components have not yet been initiated.
Date: 1960
Creator: Alco Products (Firm).
Partner: UNT Libraries Government Documents Department
open access

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 of flow transient results indicate DNBR's above 1.5, insuring that the core is safe from burnout. Bulk boiling was noted in the hot channels and lattice passage at scram power condition.
Date: June 29, 1962
Creator: Davidson, S. L.
Partner: UNT Libraries Government Documents Department
open access

Criteria for Evaluating Hazards Involved in Proposed Tests On and/or Modifications To the SM-1

Description: Abstract: This technical report elucidates principles of hazards evaluation. The concept of hazards potential is introduced and utilized to show how a reactor system perturbation will influence its nuclear safety. Literature relating to reactor safety is referenced to provide the sources of information required for hazards analysis and show how they influence a hazards evaluation. A checklist of items which should be considered in evaluating a change, test, or modification is presented.
Date: October 18, 1961
Creator: Scoles, J. F.
Partner: UNT Libraries Government Documents Department
open access

Hazards Report for Step Transient Loading Tests on the SM-1 : Task XII

Description: Abstract: This technical report evaluates hazards involved with SM- 1 plant response and system performance tests (Task XII), to determine the response of the SM-1 plant to step electrical and steam load changes. The report describes the changes in plant equipment and operating procedures for this task and evaluates these changes for possible additional hazards to those described in APAE No. 2, Revision 1, "Hazards Summary Report for the Army Package Power Reactor SM-1."
Date: February 1, 1962
Creator: Pomeroy, D. L.
Partner: UNT Libraries Government Documents Department
open access

SM-1 Reactor Vessel Closure Stud Investigation

Description: Abstract: This report presents metallurgical .analyses and corrosion tests conducted to determine :causes of failure of two SM- 1 reactor vessel studs due to fracture in the threaded areas, after 23 months' operation (March 1959). Conclusions of corrosion tests are: (1) failure of the studs was caused by stress corrosion cracking, (2) contributing factors to failure of the studs were improper control of heat treatment of the studs and absence of treatment for stress relief after thread grinding operation.
Date: March 23, 1962
Creator: Reali, D. L.
Partner: UNT Libraries Government Documents Department
open access

PM-2A Core II Zero Power Experiment

Description: Abstract: This technical report covers the zero power experiments performed on PM-2A Core II at the Alco Critical Facility. PM-2A Core II is the first replacement core for a portable pressurized reactor at Camp Century, Greenland. Core II is the same as Core I with the exception that Core II has an increased burnable poison (B-10) content. The zero power experiment consisted of fuel element uniformity test; core assembly test, development of an on-site loading procedure and an analysis of experimental data. Physical characteristics determined include distribution of fuel and B-10 in the fuel plates, minimum critical mass, control rod bank calibration, and integral rod worth. The report concludes with an analysis of the experimental data including estimated uniform and non-uniform burnup rates.
Date: March 30, 1962
Creator: Walthousen, L. D.; Mageean, J. V.; Raby, T. M. & Lee, D. H.
Partner: UNT Libraries Government Documents Department
open access

Army PWR Support and Development Program Six Months Summary Report : October 1, 1961 - March 31, 1962

Description: Abstract: Progress is reported on research and development tasks under the Program Plan for Engineering Support and Development of Army Pressurized Water Reactor Power Plants, Contract AT(30-1)-2639, during the six months' period October 1, 1061 to March 31, 1962.
Date: May 25, 1962
Creator: Dixon, M. H.
Partner: UNT Libraries Government Documents Department
open access

BWR Reference Design for PL-3

Description: Abstract: The natural circulation, direct cycle, boiling water reactor reference design presented in this technical report is the alternate to the preferred preliminary design developed under Phase I of the PL-3 contract. The report presents plant design criteria, summary of plant selection, plant description, reactor and primary system description, thermal and hydraulic analysis, nuclear analysis, control and instrumentation description, shielding description, auxiliary systems, power plant equipment, waste disposal, buildings and tunnels, services, operation and maintenance, logistics, erection, cost information and training program outline.
Date: February 28, 1962
Creator: Humphries, G. E.
Partner: UNT Libraries Government Documents Department
open access

PWR Preliminary Design for PL-3

Description: Abstract: The pressurized water reactor preliminary design presented in this volume is the preferred design developed under Phase I of the PL-3 contact. This technical report presents plant design criteria, summary of plant selection, plant description, reactor and primary system description, thermal and hydraulic analysis, nuclear analysis, control and instrumentation description, shielding description, auxiliary systems, power plant equipment, waste disposal, buildings and tunnels, services, operation and maintenance, logistics, erection, cost information and a training program outline.
Date: February 28, 1962
Creator: Humphries, G. E.
Partner: UNT Libraries Government Documents Department
open access

Hazards Summary Report for the Army Package Power Reactor : SM-1, Task XVII

Description: Preface. This technical report is an updated and revised version of the original SM-1 (APPR-1) Hazards Summary Report. The original report was issued July 1955, almost two years before construction of the SM-1 plant was completed. During that time interval there were numerous design changes. Consequently, the original report does not accurately describe the plant as built. This revision is written after the SM-1 has been operated 3 years. It describes the as-build plant and includes plant modifications and experience obtained during the first 3 years of SM-1 operation.
Date: May 1960
Creator: Rosen, S. S.
Partner: UNT Libraries Government Documents Department
open access

Shielding Requirements for the Army Package Power Reactor

Description: Abstract. The design, selection, and calculation of the Army Package Power Reactor shielding are described. The APPR-1, a prototype of a package reactor for remote locations, has a primary shield of iron and water. this shield has been adopted to permit fast erection and to provide low transported weight. economically, including transportation cost, the iron water shield is better than a lead water shield and is competitive with a concrete shield for a remote site. Because of the location at Fort Belvoir,Va., the shielding requirements for the APR-1 are considerably more stringent than those for a reactor at a remote base. Since the secondary shielding which surrounds the entire primary system must provide protection for personnel at any location outside the vapor container, concrete is provided for this need.
Date: May 1, 1956
Creator: Meem, J. L. (James Lawrence). & Fairbanks, F. B.
Partner: UNT Libraries Government Documents Department
Back to Top of Screen