Abstract: The most adverse power distribution and power peaking factors have been developed for Type 3 replacement cores for SM-1, SM-1A, and PM-2A. Also included is the 37 element Type 3 core for SM-1, which is known as SM-1 Core III and is a prototype of a Type 3 core for PM-2A. The power distributions were compared with measured distributions obtained on type 3 fuel elements in SM-1, SM-1A and PM-2A core configurations. Using the measured data as a standard, correction factors were obtained and applied to the analytical predictions of the most adverse power distribution.

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.

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 specifications for which they were designed. Welding techniques for type 316 stainless steel are described. The specifications and operating conditions of the units are given along with instrumentation drawings showing test equipment design and arrangement.

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 are described which attempt to assimilate the most drastic thermal transient that could occur in both units and to evaluate the effect of the resulting stresses on the Inconel overlay and the tube-to-tubesheet welds. (N.W.R.)

Abstract: A cyclic stress analysis of the SM-1 primary system was carried out. Problems encountered in the fabrication of PM-2A Core II and SM-lA Core II are described, and the results of an examination of damaged SM-lA Core I stationary fuel elements reported. A preliminary study of the radiation damage to SM-1 reactor vessel was made and the possibility of annealing the vessel discussed. Performance analyses are presented for five cores: SM-1 Core, SM-1 Core 1 rearranged and spiked, SM-1 Core II with special components, PM-2A Core 1, and SM- 1A Core 1. Preliminary critical experiments were made with SM-2 elements in a SM- 1 core configuration and nuclear and thermal analyses of the use of SM-2 elements in SM-1, SM-1A, and PM-2A completed. A throttling steam calorimeter was selected for measuring moisture carry-over on the PM-2A steam generator. Test procedures for evaluating the shielding of the SM-1, SM-lA, and PM-2A plants are summarized. Radiochemical and chemical analyses of SM-1 coolant and crud are summarized, and methods of activity control discussed. Preliminary results of studies of the properties of reactor pressure vessels under irradiation and no irradiation conditions are summarized briefly.

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 primary pressure or inlet temperature to introduce the phenomenon of nucleate boiling in the the core. Reduction of system pressures to 1000, 800, and 600 psia and increasing core inlet temperatures to 465 and 500 degree F were studied as the means to induce boiling in the core. This analysis indicates that SM-1 Core I Rearranged and Spiked can be safely operated at the reduced pressure of 910 psia without introducing extensive boiling in the core. SM-1 Core II with Special Components can be operated at 800 psia or at an inlet temperature of 500 degree F at 1200 psia.

Abstract: 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 accident, ART-02, a one-dimensional code, was used. This analysis indicates the SM-2 core is safe from burnout under steady state operation at design power level (28 tMW) because (1) no nucleate boiling exists, and (2) the minimum burnout ratio is above 2.0. The core is safe from burnout under loss of flow transient because the minimum burnout ratio in the hottest element channel of 1. 82 is above the minimum design criteria of 1. 5.

Abstract: A detailed description of the PM-2A reactor, primary system and instrumentation is presented. A general description of the secondary system and plant arrangement at Camp Century, Greenland is included. The results of the startup physics, shielding, thermal and hydraulics and radiochemistry test are presented. the analysis performed on the shielding and thermal and hydraulic tests is included. The shielding problem at startup and the shielding modification made to the plant are described along with the radiation measurements made in these areas.

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.

Abstract: This technical report covers the investigation of 17-4 PH steel components used in the SM-1, SM-1A and PM-2A nuclear power plants, which consist of control rod drive racks, pinions, seal shafts and associated parts. Evidence supporting the continued use of 17-4 PH steel in these applications is presented. Alco's current procurement and processing specifications for 17-4 PH steel are included. The evidence presented is in the form of stress calculations, test and environmental data and the results of non-destructive and destructive metallographic analyses.

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.

Abstract: This technical report contains the description and results of an experimental program to evaluate the effect of utilizing Type 3 (SM-2) replacement cores in existing Army field plants SM-1, SM-1A and PM-2A. This program, conducted at the Alco Products Critical Facility, employed SM-2 mockup fuel elements similar in composition to Type 3 fuel elements to determine start-up characteristics of Type 3 cores in SM-1, SM-1A and PM-2A core configurations. measurements include comprehensive power distribution, temperature coefficients, initial critical bank positions, control rod calibrations, critical rod configuration and material coefficients, all obtained under cold, clean, core conditions. The 45 element SM-1 and SM-1A configuration with SM-2 mockup fuel elements contain 36.4 Kg U-235 and an estimated 67.9 gm B-10, while the 37 element PM-2A configuration with SM-2 mockup elements contains 30.0 Kb U-235 and an estimated 56 gm B-10.

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.

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."

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.

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.

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.

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 …

Abstract: An analysis and a summary are.given of the radioactivity buildup program at the Army Package Power Reactor during the last nine months. · Due to the wide fluctuation of water and crud results, only general interpretations can be made with this data. Metal test coupon data indicate a substantially greater buildup on Croloy 16-1 metal than on Type 304 stainless steel. Coupled with the decreased ability to remove the radioactivity buildup on Croloy 16-1 by conventional descaling techniques, the implication is that this metal might pose a serious problem for use as steam generator material. It was emphasized, however, that results to date ae only preliminary and extensive additional experimentation would be required to reach more definite conclusions.

Preface; Subsequent to the analysis in the body of this report metallurgical developments indicated that it would not be feasible to build the APPR-1 Core II with the increased boron loading specified herein. At the time of the issuance of this report the Core II boron loading is to be the same as that for Core I. At the time of procurement of the core in the fall, if metallurgical developments warrant, the increased boron loading will be employed. The loading discussed in the body of the report and the first three appendices is that designed to meet the specifications outlined in the introduction. Appendix IV discusses changes in the loading to account for the various methods of employing boron.

Abstract: Neutron flux and gamma radiation measurements through the SM-1 primary shield were made at the startup of Core II in June 1961. They extend previous measurements (APAE-35) both vertically and horizontally in the primary shield and in the rod drive pit. Dose rate measurements on spent fuel elements under water are also reported.

Abstract: This technical report analyzes gamma dose rate and neutron measurements in their relation to the SM-1 shield design and is a continuation of previous shielding measurements and analyses reported in APAE-35 and APAE-35 Supplement 2. The data reported herein are spent fuel element and rod drive pit gamma dose rates. An analysis of gamma dose rates off the core midplane is presented and compared with test data.

Phase I - Plant Transient Analysis. Behavior of the basic and refined kinetic models differs only slightly. It is therefore suggested that the basic model be used in any studies where the improvement in fidelity attainable fro the refined model is not warranted by the complexities introduced by the addition of function generator to the analog circuitry and derivation of the function to be programmed. The parameter responses of both kinetic models appear to be essentially similar to those of the plant with the exception of the primary pressure. In the pressurizer analysis it was noted that the primary system pressure surges of the model should be higher than those of the plant because of the adiabatic steam compression assumed in the model derivation. the fact that the mode surge is very much greater indicates that the compression process is far from adiabatic. A more detailed and complex model of the pressurizer, one that evaluates heat transfer in the steam pocket boundaries, would therefore reduce the costly conservatism otherwise necessary in the specification of pressurizer vessel size. Phase II - Xenon Reactivity Transients. On the basis of this study an analog computer circuit has been presented which accurately represents bank …

Preface: 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 from the reactor is radioactive, the purpose of the IHX is to transfer heat to a nonradioactive fluid which then goes to a steam generator. Because of this activity the until will be enclosed in a concrete pit and will not be accessible during periods of operation. Immediately after shut down it will be necessary to allow time for radioactive decay before the unit will be accessible to personnel. Because of inaccessibility and possible long periods allowed for decay time, it is imperative that the unit give trouble free operation. During periods of shut down, the internals should have easy access for inspection and repair if necessary so that down time is held to a minimum. The general arrangement of the heat exchanger described in this report presents a conventional design utilizing known materials and existing methods of fabrication. In further consideration of all concepts, designs and analyses developed during this period of the program, it is felt that this preliminary design will provide an intermediate sodium heat exchanger of lower cost …