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Analysis of Extended Zero Power Experiments on the Army Package Power Reactor : ZPE-2

Description: Introduction: This report is principally concerned with analysis of measurements taken on the APPR-1 core during the course of the extended Zero Power Experiments (ZPE-2). The bulk of these measurements are reported in APAE No. 21. There are some additional measurements reported in APAE Memo 115. In addition to the analysis of the ZPE-2 data some re-evaluation has been made of a few of the results obtained from the first set of Zero Power Experiments (ZPE-1). The ZPE-1 measurements are reported in APAE No. 8. During the course of analysis work it became apparent that a considerable amount of basic experimental data had been taken on the APPR-1 core. It seemed worthwhile to organize this report in such a fashion that other investigators could make maximum use of this data. It provides excellent opportunity for individuals and groups interested in basic reactor reactor analysis problems to check calculational techniques. An attempt has been made to include all of the fundamental information concerning the material content and geometry of the APPR-1. This material is in included in the Appendices. In addition, cross-section files and group constants have been listed rather extensively in order that other investigators could compare results presented in this report with those obtained by other techniques. Listing are generally given to six places although not all of these places are significant. This was done so as to avoid round off error.
Date: May 7, 1958
Creator: Byrne, B. J. & Oby, P. V.

Analysis of Zero Power Experiments on SM-1 Core II and SM-1A Core I

Description: Abstract: An analysis of SM-1 Core II and SM-1A Core I zero power experiments was made by comparing these cores to each other and to AM-1 Core I on the basis of critical bank positions, bank calibrations and available chemical analyses of the fuel plate compositions. The effects of replacing boron absorbers by europium absorbers upon rod worth and stuck rod conditions were studied. Comparisons of measured and calculated power distributions were made. It was concluded that both SM-1 Core II and SM-1A Core I contain nearly identical B-10 loading of 17.79 grams, compared to the best estimate of 15.75 grams for SM-1 Core I. The available data indicates that all three cores possess similar nuclear characteristics.
Date: October 5, 1960
Creator: Paluszkiewicz, S.

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.

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).

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.

Decontamination Program Task II. Volume II, Evaluation of Chemical Agents for Nuclear Reactor Decontamination

Description: Abstract: The caustic permanganate-rinse decontamination treatment was investigated. Loop and metallurgical studies were performed to determine optimum operating conditions as well as the metallurgical effects of the treatment. A treatment with 10 percent sodium hydroxide and 5 percent potassium permanganate solution followed by a rinse with a 5 percent ammonium citrate, 2 percent citric acid and 1/2 percent Versene solution was chosen for the decontamination of a stainless steel steam generator. Decontamination factors of greater than 50 were obtained in loop tests using the above treatment. Corrosion and metallurgical results indicated a total penetration of less than 0.01 mil on annealed Type 304 stainless steel with no evidence of any deleterious effects.
Date: February 15, 1959
Creator: Zegger, John L. & Pancer, Guyon P.

Design Criteria for Irradiated Vessels Task 6.0 Summary Report

Description: Abstract: This technical report presents design criteria to prevent the brittle fracture of ferritic reactor vessels that cold occur as a result of the rise in NDT caused by fast neutron irradiation. The criteria require that maximum principal stress in the vessel does not exceed 18 percent of yield stress at temperatures below NDT + 60 degree F. Under certain conditions the allowable stress may be based on the irradiated yield stress. A discussion of brittle fracture and an explanation of the criteria are included.
Date: September 29, 1961
Creator: McLaughlin, D. W.

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.

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 research and development program for the United States Atomic Energy Commission sodium Components Design Project.
Date: February 15, 1961
Creator: Alco Products (Firm).

Experiments and Analysis for SM-1 Core II With Special Components

Description: Abstract: This technical report contains a summary of analytical and experimental work performed on SM-1 Core II, with special components is presented. The effects of these special assemblies upon power distribution and core reactivity were calculated and compared to experimental measurements. A thermal analysis was conducted to determine steady state and transient performance of the special test components of the core as well as some of the hotter standard Core II components. Experimental work discussed includes individual reactivity effects of all the special elements and the total effect of all of the elements. Power mappings were also made and are reported.
Date: January 1, 1961
Creator: Lee, D. H.; Robinson, R. A. & Segalman, I.

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.

Fast Neutron Flux Measurements and Analysis in SM-1 and PM-2A Core and Vessel Mockups

Description: Abstract: This technical report contains a summary of experimental and analytical work performed on cold clean SM-1 Core II with special components and PM-2A Core I each with thermal shield and vessel mockups. The purpose of this work was to define the neutron intensity and spectra to which the pressure vessels are subjected in order to assess the effect of irradiation on the vessel lifetimes. The radial distributions of the energy dependent neutron flux per unit power in the core and vessel mockups are presented. The uncertainties associated with the measurements are also given. In addition to graphical and tabular presentations of the test results, discussions are provided on the experimental techniques and theory. A review of the analytical models is provided in addition to the calculational results. A special analytical study of the the PM-2A Core I power distribution and absolute power level determination is given. A comparison of the experimental and analytical results is made and conclusions and recommendations presented.
Date: March 31, 1962
Creator: Sontheimer, K. C.; Kemp, S. N.; Lois, L.; Clancy, E. F. & McCool, W. J.

Fission Product Activity in SM-1 Core I Primary System and Surface Contamination on SM-1 Type Fuel Elements. Task XVIII, Phases 2 and 3

Description: Abstract; The fission product data obtained during SM-1 Core I operation (June 1957 - May 1960) is reviewed briefly and interpreted. Evidence is presented to indicate that a fuel element defect was responsible for the high fission product activity level observed in the primary coolant. Relative escape coefficients are calculated and the defect size estimated. Anticipated fission product levels during SM-1 Core II and SM-1A Core I operation are estimated from alpha surface contamination data on completed fuel elements. The importance of in-line sampling for monitoring fission product activity is stressed as well as the need for failed fuel element detection methods.
Date: February 29, 1961
Creator: Hasse, Robert A. & Zegger, John L.

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.

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 defined by Alco Products, Inc., it is recommended that if time permits that critical facility design verification be accomplished. The PM-1-M-2 test element meets the minimum requirements for insertion in SM-1 Core II and in view of the importance of this element to the PM-1 and PM-3A program, should be considered for insertion.
Date: September 1, 1961
Creator: Coombe, J. R.; Scoles, J. F.; Brondel, J. O. & Lee, D. H.

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 procedures have been developed at the Alco Critical Facility. The nuclear and thermal and hydraulic characteristics are essentially identical to those of Core I and the replacement of the startup and check sources represent no increase in the potential for or magnitude of a hazardous situation.
Date: June 21, 1962
Creator: Coombe, John R. & Stephenson, L. D.

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 estimate of core safety. There is no conceivable reason why this perturbation may not be safely made in the SM-1 Core II.
Date: October 9, 1961
Creator: Coombe, J. R.; Lee, D. H. & Matthews, F. T.

Hazards Report for SM-1 Core II With the SM-1 Core II With the Silver-Cadmium-Indium Control Rod Absorber Section

Description: Abstract: In the March-April 1962 shutdown of SM-1 Core II, the SM-28 element will be re-inserted in SM-1 Core II and an SM-1 Core I element will be removed. An SM-1 Core II europium absorber will be replaced by a Ag-Cd-In absorber, and surveillance specimens will be inserted above the core support structure. Analysis of these changes concludes that re-insertion of the SM-2B stationary element and insertion of surveillance specimens do not affect hazards potential previously defined for SM-1. Replacement of the europium absorber by the Ag-Cd-In absorber will have negligible effect on reactivity control worth of the rod. The absorber meat section is encapsulated to prevent exposure of silver alloy to the primary coolant; postulated release of silver due to a cladding defect, after 2 years irradiation in SM-1, would not cause a hazard such as to restrict access to the vapor container. Possibility of steam formation in the air gap between the absorber core and cladding, causing a cladding failure, is remote. Deformation of the absorber section sufficient to cause the rod to stick, would not impair the ability of the other rods to shut down the reactor safely.
Date: March 15, 1962
Creator: Stephenson, L. D.

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 leads to a more conservative estimate of core safety. There is no conceivable reason why the perturbation may not be safely made in the SM-1 Core II.
Date: October 7, 1961
Creator: Coombe, J. R.; Lee, D. H. & Mathews, F. T.

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 pressure and flow on the burnout ratios are determined as are the expected stuck rod conditions when Task XIV test elements are installed. The effect on the maximum credible accident is included and a recommendation to conduct these Task XIV package tests is made.
Date: February 28, 1962
Creator: Bradley, P. L. & Coombe, J. R.

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.

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.

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 I were subjected to a Zero Power Experiment at the Alco Critical Facility. The results of this experiment indicate that the SM-1 Core II will have nuclear characteristics very similar to that of the SM-1 Core I. Since SM-1 Core II will be operated with the same mode of rod control, in the same core support structure, and with the same primary coolant flow conditions, the thermal characteristics should be essentially identical to that of SM-1 Core I. Also, all kinetic characteristics of SM-1 Core II should be identical to those of SM-1 Core I. This report demonstrates that there is no increase in potential for a hazardous situation at SM-1 due to the replacement of SM-1 Core I by SM-1 Core II.
Date: April 19, 1961
Creator: Gallagher, J. G.