This production test was designed to evaluate the suitability of low hydrogen dingot uranium as routine process material. Nine tubes of I and E fuel elements (6 dingot, 3 ingot) with 32 fuel elements in each tube, have recently been discharged at the C Reactor and this document contains the results of analyses made on the dimensional stability properties of this material.

This report discusses the events from a chemical standpoint following a total loss of coolant disaster will not be altered in the melting reactor by the introduction of N metal. The interdiffusion of uranium and aluminum will be the dominating reaction, causing the blockage and tying up of the lithium in UAl{sub 3} which does not melt until after the uranium does. Pressure from the swelling UAl{sub 3} will extrude uranium-aluminum and lithium into graphite weep holes and block interfaces. The migration of lithium by vaporization will not became appreciable until well over 2000{degrees}C, well beyond the time when uranium and UAl{sub 3} have melted. The eventual result will be a diffuse distribution of uranium, lithium, and aluminium in the lattice. The E-N pile has a larger excess over required control capacity than the uranium provided the large reactivity poison tied up in the lithium is not lost. Compared to the natural uranium pile, the gain of reactivity on loss of coolant is less and the net temperature coefficient in the dry pile remains negative to higher exposures. Furthermore, permanent pile poisoning during meltdown is accomplished via two mechanisms both lithium and uranium redistribution in the lattice produce large negative …

This study was undertaken to determine the most feasible and economical process or path for closing the uranium fuel cycle at HAPO, and to establish what benefits, other than improvement in FPD`s competitive position, would result from the selected closed fuel cycle. The study was separated into four phases; Phase I includes the selection and organization of plausible processes and the establishment of a realistic and effective evaluation procedure, Phase II will include evaluation and selection of an optimum path based on a medium range and a short range approach (5 to 10 years), Phase III will include evaluation and selection of an optimum path based on a long range approach (> 10 yrs.), and Phase IV will include refinement of previous work and issuance of a final report summarizing the study and the conclusions or recommendations which develop from the study. Phase I of the study has been completed. The purpose of this paper is to document the assembled data and the completed work.

Neptunium-237 is currently produced in the Hanford reactors at a rate of approximately .003 gms/MWD via the following reactions: (a) U{sup 238} (n,2n) U{sup 237}{sup {beta}}{yields} Np{sup 237}. (b) U{sup 235} (n,{gamma}) U{sup 236} (n,{gamma}) U{sup 237}{sup {beta}}{yields} Np{sup 237}. In order to calculate the buildup of Np{sup 237} via reaction (a), which accounts for the greater share of the formation of Np{sup 237}, the n,2n cross section for U{sup 238} must be known. An old value quoted by Arnold of 5.2 millifermis for an {open_quotes}effective{close_quotes} 2200 m/s value is not large enough to account for the observed Np{sup 237} yield by about a factor of two. Recent n,2n cross section measurements for U{sup 238} permit a newer calculation and the result is 11.2 mF, effective 2200 m/s value.

In this report, the equation (HW-55219-RD) expressing tube-wise rupture rate as a function of exposure, power, and temperature has been adjusted to include 1958 data.

Early in the development of an extended surface fuel element for use in the NPR, several 7-rod cluster fuel elements were irradiated to determine the dimensional stability of such geometries at high burnups. These elements were fabricated from small diameter uranium rods clad unbonded in stainless steel tubes and assembled in a rod cluster geometry by various support devices. Zircaloy clad fuel rods were not yet available, the stainless steel clad rods therefore served as a suitable material which would withstand high temperature water over a long period of time and maintain relatively high strength properties. The purpose of the irradiation detailed in this report was to determine the effect of high exposure on the swelling, dimensional stability, microstructure, and physical properties of uranium rods restrained unbonded in stainless steel. At the same time, this test was designed to evaluate the effect of fuel rods operating in a cluster geometry, to monitor the central core temperature of the uranium, to determine the stainless steel-uranium interface heat transfer bond coefficient, and to determine the average specific power of the assembled element. Goal exposure for this irradiation test was 3500 MWD/t.

A study of the stored energy buildup in the MTR reflector graphite and a program of controlled energy release is presented. Calculations, based on measurements of samples from the pebble zone show that an inadvertent spontaneous stored energy release would cause a temperature rise of 90 deg F in the pebble zone. The maximum transient structure temperatures resulting from a worst credible accidental release of energy would be less than allowable at present (except for possible damage to neutron detector chambers) but could exceed this value in five years. It is proposed that the stored energy be released by thermal annealing. The reflector graphite is heated by reducing the air flow and operating the reactor at low power until a temperature of 500 deg F is reached, at which point the reactor is scrammed. Normal cooling is provlded after 15 minutes at peak anneal temperature or if the temperature rises to 600 deg F. Health physics monitoring includes continuous measurement of particulate and of Ci/sup 4/ activity. Sustained oxidatlon, if it occurs, wlll be detected with a C0/sub 2/ monitor and controlled by smothering. An estimated 2 or 3 days of MTR operating time will be needed of which the …

The Fuel Development operation at Hanford uses a variety of in-reactor facilities to test experimental and prototypical fuel elements. High pressure-high temperature water loops in constant use are the 6 in. {times} 9 in. ETR core facility, the 3 in. {times} 3 in. ETR reflector facility and four front-to-rear loops in the 100-K East reactor at Hanford. Low pressure water cooled test facilities in use are located in the MTR the various Hanford reactors. Stainless steel sheathed the thermocouples 1/16-inch diameter insulated with MgO or Al{sub 2}O{sub 3} when properly fabricated and installed, will reliably measure in-reactor temperature up to 1000 C and at least 4000 MWD/T exposure. This report provides a brief description of some of the temperature monitored in-reactor experiments.