"The purpose of this report is to examine the safety aspects of PNPF restart on continued operation, after completion of the core cleanup and system modifications."

This report summarizes the results and conclusions of a study made to evaluate the merits of using zirconium hydride as a solid moderator in an integral boiling water-nuclear superheat reactor of the pressure vessel type.

Abstract: Uranium monocarbide shows considerable potential for use as a fuel in high temperature, high power density, nuclear power reactors. As Atomics International is proposing its use in sodium graphite type reactors, it was necessary to develop a process for fabricaing sodium bonded uranium carbide fuel elements.

Introduction: This report has been prepared to document the extensive analytical work required to design the control system for the Sodium Components Test Installation (SCTI), constructed for the Atomic Energy Commission at the Atomics International field laboratory.

Abstract: A fuel development program has been conducted, to investigate the behavior of U-10 wt% Mo alloy at the fission rates and temperatures anticipated in Core I of the Hallam Nuclear Power Facility.

Abstract: One hundred gram quantities of uranium dioxide, irradiated to burnup as great as 21,000 Mwd/MTU and previously reprocessed by AIROX (Atomics International Reduction Oxidation), were refabricated into high density pellets and encapsulated for re-irradiation.

Abstract: Atomics International conducted a fuel irradiation program to investigate and define the behavior of U-10 Mo fuel at the temperatures and fission rates anticipated in Hallam Nuclear Power Facility (HNPF) Core I.

Introduction: The development and qualification of the system acceptance test heaters and the reactor simulator heater are described in this progress report.

Introduction: The purpose of this experiment is to determine the loss of electrical power output due to operating many unequally-heated cesium diodes, connected in series and parallel circuits.

From abstract: The response of the reactor to forced oscillations has been investigated with the pile oscillator technique and the data are presently being analyzed.

Abstract: The following report covers the final design of the modular steam generators.

From summary: The full 140 element loading of the core was completed on October 10, 1962. At this point, critical operation was begun for operator training and post-critical testing purposes.

Introduction: A series of critical and subcritical experiments have been carried out as part of the Advanced Sodium Graphite Reactor (ASGR) Critical Experiments program to provide physics information which would be directly applicable to the AGSR concept.

Abstract: An analytical and experimental effort toward a better understanding of the mechanisms involved in nuclear power reactor loss-of-pressure accidents is presented.

Abstract: This report presents an analysis of coolant boiling in sodium graphite reactors.

Abstract: Perforated BeO plates were thermal shock tested to evaluate the effect of: (1) localized temperature variations adjacent to the perforations, and (2) radial gradients across the entire plate.

Abstract: Two in-pile capsules containing mock-up uranium carbide fuel rods were subjected to a total of five neutron bursts in the TREAT facility with initial reactor periods ranging from 0.26 to 0.10 sec.

Abstract: This report reviews various means which have been used to control experimental conditions in irradiation experiments, and presents the concept of local flux control.

Abstract: The structural design criteria and the system basic loads for the SNAP 2 compact power unit are presented.

Abstract: This report describes in detail two designs of a nominal 6-kwe Nuclear Power Plant (NPP), one using thermoelectrics for power conversion and the other using the Mercury-Rankine cycle NPP.

Abstract: A new concept for continuously monitoring neutron flux has been successfully tested.

Abstract: Irradiated fuel elements from the Organic Moderated Reactor Experiment (OMRE) first core loading have been examined and evaluated to determine: (1) the stability of the floating plate fuel element design, (2) the stability of the stainless steel clad UO2 - stainless steel cermet core fuel plates under irradiation and exposure to the organic coolant, (3) the extent and nature of deposits on the fuel element services, and (4) the distribution of burnup in the fuel elements.

Abstract: This report outlines the general techniques developed for replacement of PRE in-cell equipment or equipment components.

Abstract: This report presents a description of the Organic Moderated Reactor Experiment (OMRE), of the hazards associated with this experiment, and all of the safeguards taken to ensure the safety of the operating personnel and the population of the surrounding area.