During the course of the ZPE-2 experimental program additional measurements were performed such as the evaluation of various absorber section compositions and reactivity studies designed to facilitate analytical techniques

The Alco Products Criticality Facility, site location, and operating procedures are described in detail, including the handling of fissionable material and the operating procedures for the safe performance of critical experiments.

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 …

The purpose of this study is to determine the time required for blowdown of the primary system when the rupture disc fails, and the effect of blowdown on the control rods during scram.

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.

Calculations used in previous studies are being modified with the results reported in this study.

A general non-uniform burnup program has been developed to determine the lifetime of the APPR-1. Calculations are used to determine performance using two one dimensional multiregion burnout calculations

The methods used to treat partially and fully inserted APPR-1 control rods are described

On the basis of recent experiment with APPR-1 fuel elements manufactured by ORNL,hot channel factors have been derived from the observed dimensional deviations

Developing a test program to review reaching criticality with the minimum number of fuel elements

A basic kinetic model of the primary system coolant loop is developed for the Army Package Power Reactor-1

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.

This program using the IBM 650 electronic data processing machine is used primarily as a tool in order to prepare input data for the IBM 704 code, entitle ATBAC. The code calculates steady state thermal characteristics of a plate type fuel element for both a nominal and adverse channel

A method has been developed for using the IBM 650 Electronic Data Processing Machine to obtain detailed information concerning thermal and hydraulic conditions within a plate type reactor channel when the coolant in the channel is present in both vapor and liquid phases

An analytical prediction of the burnout distributions in particular SM-1 fuel elements and absorber sections after 10.5 MWYR of core energy release is given

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.

7 rods were used in the APPR-1 rather than 5. The critical experiments and calculations from the use of the rods, including the capability of any four to shut down the cold reactor are examined in this study.

There is a possibility of dimensional instability in the B-10-iron APPR-1 control rods due to helium production. This effect will be evaluated in this study.

A detailed analysis of hydraulic and mechanical forces affecting control rod drop time has been made to determine a mathematical expression for rod position in terms of elapsed time after initiation of a scram.

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.

This study was conducted to determine the operating and performance characteristics of the ORNL concept of control rod drive mechanism as applied to APPR-1

A brief discussion of the core characteristics of four uranium fuel, pressurized-water type Army Package Power Reactors is presented

This report covers fuel element corrosion studies conducted from April,1959 through July,1960, designed to aid in selecting and evaluating SM-2 fuel element welding techniques

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.