The nuclear characteristics of the CETR are described. Core operating lifetime, control-rod worth, and power density distribution are discussed in relation to maximizing the core operating life. Other objectives of nuclear design are to minimize the power-density variation and to assure control of the reactor.

The adequacy of the Consolidated Edison Thorium Reactor (CETR) control rod drive line operation was evaluated and proven by the development program described in this report. This program also included evaluation of materials and fabrication processes.

The Consolidated Edison Thorium Reactor Control System has been designed to provide the operators with a reliable, safe and accurate means of controlling the plant. Manual and automatic positioning of control rods can be accomplished. The automatic plant controller has been designed to meet the design criteria under the representative CETR demanded load changes. Adaptability has been built into the plant controller by providing variable controller adjustments for variations in reactor characteristics. Extensive analog studies have been made of the various plant parameters and their effect on control was carefully evaluated. These analog studies have shown that the control system will perform in a satisfactory manner under the most adverse plant conditions.

Results of zero power tests to confirm the nuclear design of the Consolidated Edison Thorium Reactor Core are presented. The maximum core reactivity, control rod worths, and power distribution were measured.

This report describes the critical experiments with pin-type oxide fuel elements for the Consolidated Edison Thorium Reactor (CETR). This report also describes the measurements and gives a brief interpretation of the results in some cases

A series of exponential experiments were performed using Consolidated Edison Thorium Reactor oxide type fuel pins in a light water lattice. These experiments were performed in a new facility where measurements were made at temperatures up to 468 F. In addition, boric acid was added to the system in various concentrations. The main objective of these experiments was to obtain the change in material buckling for the system as operating conditions were changed from ambient temperature to high temperature.

The design functions and fabrication details for internal components of the CETR are presented and pertinent analytical stress studies are summarized. Functions of the internal components include proper orientation and support for the fuel elements, proper distribution of primary coolant within the reactor vessel, and the establishment of guide channels for the control rods.

The series of experiments described in this paper was performed to determine reactivity values of various patterns of fully inserted cruciform control blades. The experiments yielded the worth of fully inserted blade patterns in terms of their holddown characteristics.

The general purpose of this report is to present the major characteristics and design of the Consolidated Edison Thorium Reactor (CETR) core

A study has been performed to determine the corrosion product and water impurity activities in the CETR Chemical Process System. A simplified mathematical model is used to represent the physical situation; however, the magnitude of the activities, taken at the end of the core life to represent the most conservative case, is considered sufficiently accurate for engineering calculations.

Prototype testing and thorough design analyses have proven the control components for the Consolidated Edison Thorium Reactor are adequate to withstand loads imposed during operation and handling. Assumptions made in the design analyses and cycles run during prototype testing subjected the components to more rigid conditions that are anticipated during actual operation. Results indicate the control components meet their design requirements.

The development of swaging and vibratory compaction process for fabrication of clad UO2 fuel rods is described. The cost is less than 50% of that for an equivalent core fabricated from pellets. The irradiation testing of vibratory filled and swaged UO2 rods is reported for burn-ups from 2,000 to 12,000 Mwd/t UO2; the results indicate that the rods are capable of excellent performance power reactors.

Experiments performed by WAPD and others have shown that with fuel in an oxide form, it is possible to continue to operate a nuclear plant with fuel rods that have developed defects with only a small increase in the radiation contamination problem

The design of the fuel-element structure for the CETR is described. Component parts of the assembly comprise the fuel-bundle assembly, the fuel-can assembly, the upper transition, the fuel-element spring, and the upper nozzle and seal assembly.

The pressurizer system and its related design and operating data are described. The philosophy and analytical procedures used in the design of the pressurizer for the Consolidated Edison Thorium Reactor are summarized

Although there have been several previous investigations of neutron resonance absorption in thorium, knowledge in this area has not been sufficiently accurate nor extensive enough to satisfy the practical requirements of thorium breeder reactor design. The present experiment was performed in an effort to reduce this uncertainty by checking and extending the results of previous measurements of the geometric and temperature dependence of thorium resonance integrals. The experimental procedure was the determination of reactivity coefficients by measurements of reactor period in the Lynchburg Pool Reactor.

In the basic design for a two-fluid LMFRE, a seal has to be made between the metal core inlet and the graphite bottom reflector. To evaluate the effectiveness of such a method of sealing, a screening test apparatus was constructed to determine quickly the physical parameters and the best configuration for interface contact

The Irradiation Test Program for the Consolidated Edison Thorium Reactor was conducted to obtain confirmation of the stability and performance of the reference fuel element materials and of test assemblies under design operation conditions. The irradiated specimens were tested to determine mechanical properties and corrosion resistance to borated water.

This report provides an overview of the creation of the Liquid Metal Fuel Reactor Experiment program. It furthers the work by constructing a single loop to test all the components required for the 16 loop reactor. This utility loop was also constructed to provide a facility for testing various components such as valves and flow meters.

A monthly report summarizing work perform by Babcock & Wilcox personnel for the U.S. Atomic Energy Commission. In this monthly report information on the reactor design No. 10 had been selected as the LMFRE preliminary reference design, various shielding designs have bee studied, various calculations on multiregion code and reactivity have been continued, and other related to the LMFRE

A monthly report summarizing work performed by Babcock & Wilcox personnel for the U.S. Atomic Energy Commission. In this monthly report the R&D budget was reviewed, two-dimensional calculations were performed, analysis work was started on control rods and their housing, as well as continued work on the development of the LMFRE reactor.

A monthly report summarizing work perform by Babcock & Wilcox personnel for the U.S. Atomic Energy Commission. Preliminary design studies for nuclear reactor development and its components were reported.

A monthly report summarizing work perform by Babcock & Wilcox personnel for the U.S. Atomic Energy Commission. Included in this report was a study made regarding the possible use of beryllium as a core moderating material, calculations made regarding temperature distribution for various channel sizes and power levels, design parameters for initial operation of the LMFRE, and other work in the design and operation of the reactor.

A monthly report summarizing work perform by Babcock & Wilcox personnel for the U.S. Atomic Energy Commission. Initial contacts were made at National Reactor Testing Station and Hanford Works, design conditions for the proposed reactor vessels were reviewed, estimates of heating rates caused by radiation and the pluggage of fuel passages were computed, and design parameters for the LMFRE were established.