Irradiation-induced moderator distortion has plagued the Hanford reactors from their earliest days until the present. The first-manifestation of graphite distortion was in the form of expansion of the B, D and F stacks. It was found that expansion was very temperature-dependent and could be reversed by raising the graphite operating temperature. The DR, H, C and K Reactors were designed with provisions to minimize, or accommodate expansion. It was not until after the K Reactors had been designed that graphite contraction was recognized as a threat to stack life. Under the present operating conditions, contraction is largely dependent upon neutron flux levels. To date, this study has been largely concerned with the K Reactors, since the vast bulk of the distortion data available is from the Ks. This study not only attempts to explain the mechanisms by which the K Reactors arrived at their present state of distortion but also provides a basis for forecasting future distortion. An understanding of the nature and magnitude of the forces at work within the stacks will enhance the possibility of devising means of conteracting the present trends.

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Production Test-080 authorized the conduct of certain physics tests during the deactivation of D Reactor. The primary objective of the test was to measure certain physics parameters which could be used to check current computational models. In this report are presented a brief chronology of the tests and a tabulation of the experimental data. In general, data processing has not progressed beyond simple reduction of the raw experimental data to a form more easily used in future analytical comparisons and evaluations. Tests included a test for excess reactivity of the reactor by calibration of a control rod, a test for evaluation of experimental technique for measuring the effect of safety systems, tests for data on space-dependent reactor transients, coupling coefficients, etc.

An apparatus to study rare-gas fission-product release from nuclear fuel materials during postirradiation heating was developed. Xenon and krypton fission gases escaping from a small specimen during heating at constant temperature are measured using a continuous radioactivity monitor and charcoal adsorption traps. The rhodium-wound furnace is capable of operation at 1600 deg C. Helium carrier gas is purified by activated alumina, copper, and zirconium traps, and the oxygen and moisture contents of the gas are monitored continuously. The operating procedure and data are presented for a typical heating experiment in which fused uranium dioxide was studied. (auth)

This study proposes improvements in the process water piping adjacent to the front and rear faces of these reactors. This report covers the external piping of the reactors from the incoming valve pit to the inlets of the outgoing retention basins.

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Fuel element corrosion studies designed to and in selecting and evaluating SM-2 fuel element welding techniques are discussed. Tests on type 347 ss plate type fuel elements welded by the selected tungsten inert gas technique showed good corrosion integrity of specimens under a variety of conditions including a 500-hr test under simulated SM-2 conditions of flow and coolant chemistry. (auth)

This report talks about the Aerospace Nuclear Safety

The reference letter outlined an initial proposal for irradiating limited categories of substandard fuel elements in the upstream positions of the reactor process tubes. As Phase 2 of this program, a comprehensive study has been made of all current fuel reject categories with respect to use in the upstream low rupture potential positions. This report describes additional ``reject`` fuel categories, the estimated cost savings, rupture potential, and procedures for identifying and handling pieces to fulfill the combined requirements of Phase 1 and 2 of this program. Future methods for grading fuel elements to reduce rupture potential in the downstream reactor positions are also outlined in this report.

The WPE progress report for September, 1962 documents reasons for suspecting nitrogen as a possible cause of weapons fabrication difficulties and loss of production caused by radio-graphic rejects. This is a summary and documentation of the ``Nitride Problem`` and related subjects concerning fabrication metal. It contains the background information leading to details published in the Weapons Process Engineering Progress Reports. This summary may provide a better understanding of the problem, since little of the background was suitable for publication in the WPE progress reports.

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As the first part of a study to evaluate the economics of the various steps leading to and including the permanent disposal of high-activity liquid and solid radioactive waste, costs of interim liquid storage of acid and alkaline Purex and Thorex wastes were estimated for storage times of 0.5 to 30 years. A 6- ton/day plant was assumed, processing 1500 tons/year of uranium converter fuel at a burnup of 10,000 Mwd/ton and 270 tons/year of thorium converter fuel at a burnup of 20,000 Mwd/ton. Tanks of Savannah River design were assumed, with stainless steel construction for acid wastes and mild steel construction for neutralized wastes. The operating cycle of each tank was assumed to consist of equal filling and emptying periods plus a full (or dead) period. With interim storage time defined as filling time plus full time, tank costs were minimum when full time was 40 to 70% of the interim storage time, using present worth considerations. For waste storage times of 0.5 to 30 years, costs ranged from 2.2 x 10/sup -3/ to 9.5 x 10/sup -3/ mill/kwh/sub e/ for acid wastes and from 1.7 x 10/sup -3/ to 5.1 x 10/sup -3/ mill/kwh/sub e/ for neutralized wastes. …

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This document outlines the Westinghouse Astronuclear Laboratory (WANL) specification and procedures for acceptance testing of accelerometers. All accelerometers that are to be used for reactor instrumentation shall meet the requirements of this specifications before being accepted.

The radial temperature around a sodium reactor heater assembly submerged in water is calculated using a model of the heater cross section found by conformal mapping. Thermocouple readings were also analyzed. When the heat flux is 5 x 10/sup 5/ Btu/hr-ft/sup 2/, a radial temperature drop of about 680 deg C across the center of the thermocouple well is calculated and found to be within 6% of the experimental value. Since most of this drop is across the 0.001-in. helium gap between the heater and its sleeve, it is concluded that the thermocouple will have to be bonded to the sleeve for dependable reading of true sleeve temperature. Drawings of the heater assembly and thermocouple are given. (D. L. C.)

This report is an interpretation of data (flow, pressure, temperatures within a process tube during events affecting single tubes) as applied to the most severe (rapid) K-reactor transients which are credible. Analyses indicate that no fuel channel burnout will result from a BPA power loss to the process pumps.

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Irradiation of the initial test in this program involving ten tubes of alternately charged nickel-plated C-64 alloy clad test elements and X-80001 alloy control elements has been successfully completed. The test indicated that the nickel-plate spalling problem has been resolved, as no significant spalling or flaking was observed during the post-irradiation examination. The second test in this program will be to verify that the nickel-plate integrity problem has been solved by irradiating a pilot loading (up to 100 charges) of fuel elements which have been nickel-plated on a production basis.

Available pressure data on the B and D downcomers and approach piping is tabulated herein. All readings were taken with an air bubbler system; readings given in psi were taken on a Bourdon-type pressure-vacuum gauge, readings given in inches of water were read on an oil-filled manometer. In each case, fluctuations occurred and the mean value was estimated by eye. Data are presented as a function of flow rate and temperature.

An investigation was made to determine the structural behavior of selected components of the Oak Ridge Research Reactor for increased power level conditions. It was found that a reactor cooling water outlet temperature of 150 deg F will cause severe plastic strain cycling in the aluminum housings for the large test facilities. Increasing the reactor cooling water flow rate of 21,000 gpm will cause plastic deformations in certain reaons of the core box. These latter deformations can be tolerated, but the full implications asscciated with any change in pressure differential must be understood before adopting the above flow rate. (auth)

Maintenance of poiarization of polarized protons in a linear accelerator is known to be feasible. Circular accelerators present a different problem, and the investigation of the interaction of orbit dynamics and particle polarization in general is undertaken. The equation of motion of the spin vector of a charged particle in a magnetic field as formulated by Bargman, Michel, and Telegdi is utilized in the study of depolarization for several accelerators. High values of depoiarization are obtained, and means for avoiding such depolarization are suggested. (D.C.W.)

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