This report addresses the control design review of P&W quarterly for July to September 1960.

In 1955 the Fuels Development Operation began irradiation testing of fuel elements in high temperature water. It was assumed that if a new reactor were built at Hanford, it would be cooled by high-temperature, pressurized water. Corrosion tests showed that aluminum-clad production fuel elements could not be used in high-temperature water. Therefore, while work to improve the resistance of aluminum to high-temperature water proceeded, the Fuel Design Operation began irradiation of stainless steel- and Zircaloy-2-clad fuel elements. During 1956 and 1957, stainless steel-clad elements were tested in the Materials Testing Reactor (MTR), Hanford H Reactor Loop, and the KE Reactor Recirculating (KER) Loops. During 1957, a coextrusion method for cladding uranium rods with Zircaloy-2 was developed. The first irradiation of Zircaloy-2-clad fuel from an off-site supplier began in late 1958. The objective of the irradiation was to study the dimensional stability of the fuel rods and a seven-rod fuel assembly. Two coextruded, seven-rod elements were irradiated in KER Loop l.

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In Hanford reactors the reactivity gain upon loss of coolant water is an important factor in the speed of control requirements. The reactivity gain in the cold, clean reactor is determined from experiments, but additional effects must be taken into account if the gain in the operating reactors is to be obtained. One of these effects is the change in Xenon and Samarium poisoning with neutron temperature, which is discussed here. Earlier work on the relationship of operating limits to the reactivity gain upon loss of coolant is given in Reference 1. Work on this problem is continuing by Reactor Physics, IPD, but the newer work is not yet documented. In earlier calculations, the neutron temperature could only be guessed. Recent measurements of neutron temperatures have indicated the magnitude of the neutron temperature change upon loss of water. This document interprets the data of Reference 2 in terms of the change in Xenon and Samarium poison to be expected on water loss under typical operating conditions.

Under simplifying assumptions, an expression is obtained relating the integrated kinetic energy transferred to moderator atoms to the thermal neutron exposure in adjacent fuel. Reactor-dependent factors are explicit in the expression which make it possible to compare graphite contraction data from different reactors or to estimate graphite damage in a new reactor based oil observed damage in existent reactors. Graphite temperature, which may be an important factor in graphite contraction, is not considered as one of the variables in this report. Also, the damage dependency on neutron energy is not dealt with in great detail because of the relatively unknown facts in this regard. A particular application of concern at the present is to determine the pertinent factors relating to fast neutron exposure in the NPR to assist in determining the integrated damage over the life of the pile.

The described information consists of fuel element specifications, irradiation system parameters, operating conditions, and failure observations to include rupture identification.

A significant reduction in non-equilibrium losses is possible through the use of poison splines for reactivity and heat distribution control during reactor startups. The curves presented show the results of an analysis of recent KE Reactor poison spline startup usage. These curves demonstrate the magnitude of gains possible at other reactors through the use of poison splines for turnaround control.

This document consists of length and annealing data sheets.

In order to predict graphite damage rates for the NPE, it is necessary to have some knowledge of the neutron fluxes at various energy levels. In particular it is important to know flux levels relative to those existing where graphite samples have been irradiated experimentally. Three-group flux levels have been calculated for regions far from boundaries in the core of the NPR and of a K-reactor. The results reported here have been obtained from more detailed calculations using better cross section data than used in obtaining a previous set of values (for NPR only). The calculations of effective group cross sections for the reactor lattice still are not entirely satisfactory however, and it is hoped that some improvement can be made. Since any revision of the cross sections will affect the calculated flux levels, the values reported here should be regarded as preliminary. It is felt that the relationship between NPR and K fluxes should be fairly satisfactory, since identical calculational methods were used for both.

The purpose of this report is to present laboratory data concerning thermal and hydraulic changes which occur in a low power process channel during a flow interruption. The experiments were conducted in the Heat Transfer Laboratory of Thermal Hydraulics Operation.