The April 7, 1962 criticality accident involving human exposures was the first to have occurred in any production facility at Hanford. The accidental nuclear excursion did not result in any mechanical damage or spread of contamination. Three employees received over-exposure to gamma and neutron radiation. None were fatally exposed and in each case the over-exposure was recognized promptly. Following an initial period of medical observation and testing, the men were released to work. They continued to be followed clinically. Clinical studies performed were hematological procedures including leukocyte chromosome aberrations, morphologically aberrant blood cells, bone marrow evaluations, blood chemistry determinations, amino acid excretion studies, seminal fluid, urinary gonadotropins and estrogen excretion studies, testicular biopsies and crystalline lens examinations. These studies, along with a brief description of the accident and of the dosimetry, are summarized in this report by those participating in the studies. In view of the dose ranges received in these cases, both the negative and positive findings are considered to be of unusual interest due to the lack of knowledge of effects following human exposures at these levels.

Assistance was requested in reviewing and comparing certain measures that could be effected immediately with little capital cost to render the plant less vulnerable to self-destruction in case of a war situation that made it impossible to man the water plants. Several courses of action have been considered with the assumptions detailed.

While it is obvious that a power surge in a Hanford reactor will result in temperature increases of the fuel and water, it is also true that there will be changes in the coolant flow rate and the pressures in each process tube. In addition, if the power increase is of sufficient magnitude, a self-induced flow reduction will occur as a result of pressure changes due to boiling within the process tube. The relations between power, temperatures, flow and pressure following a power surge are difficult to calculate with precise accuracy. However, a knowledge of these relations is important for incorporation with physics considerations to allow prediction of maximum fuel temperatures following a power surge and to specify detection instrumentation for proper reactor control. Fairly precise information of temperatures, flow, and pressure in a single process tube following a power surge can be obtained using an electrically heated model in the 189-D heat transfer laboratory. Obtaining such information for all possible conditions and for all the Hanford reactors would be a fairly large job. However, two cases were run with an existing full scale electrically heated model of a single column of fuel elements in a C reactor overbore process …

The purpose of this report is to transmit revised goal exposure plans for normal production material. These plans supersede previous recommendations. Inherent in these plans are the assumptions: (1) The plant must produce a product containing six percent Pu-240, and (2) product blending capabilities are such that there are no restrictions, other than economic, on Pu-240 buildup at an individual reactor. Due to recent assessed changes in metal performance and conversion of DR Reactor to bumper natural material, revision of recommended goal exposure plans is called for. During the past year a significant shift in the relative metal performance between reactor types occurred. You may have observed the low rupture frequency at the C and K reactors compared to the rupture frequency at the old reactors. Based on the observed experience, goal exposures should be raised for C and K material and lowered for old reactor material. The goal exposure plans recommended herein are based on present, July 1962, assessed rupture performance and the latest conversion tables, incremental metal costs and burnout cost data. The current reactor optimization program was used to determine optimum goal exposures.

The objective of the production test detailed in this report is to authorize testing in the 1706-KE single-pass facility to measure the corrosion of aluminum-clad fuel elements in process water containing 0.20 ppm of Quachrom Glucosate added in place of dichromate and to compare this corrosion to corrosion in normal process water. Tubes SP-5 and SP-6 will be run with 0.20 ppm of Quachrom Glucosate and no dichromate; tests will be made at both pH 6.6 and 7.0. Simultaneous comparison corrosion measurements will be made in two regular tubes, 4561-KE and 4660-KE, supplied with normal process water. The use of SP-1 and SP-2 as additional control tubes is authorized, but not planned. Effluent-activity samples will be taken. Premature discharge and special pick-up are authorized.

This document discusses activities of the Fuels Preparation Department during the month of December 1961. (FI)

This document details activities of the Fuels Preparation Department during the month of February 1962.

This report is divided into manufacturing (AlSi process, coextrusion) and engineering (present reactor fuels, new canning processes, N Reactor fuels).

This document details activities of the Fuels Preparation Department during the month of January 1962. (FI)

This document provides details of activities of the Fuels Preparation Department during the month of October 1962. (IF)

The geologic map of the Paiute Ridge (Papoose Lake SW) 7 1/2-minute quadrangle shows an area of about 4 square miles on the east side of Yucca Flat that is underlain by basaltic rocks. These rocks are estimated to be as much as 1,000 feet in places and occur as a capping on a small mesa and as lopoliths, igneous masses that in cross section are saucer-shaped. This area was further studied to determine the geologic features of the basaltic rocks in order to evaluate the potential use of this area for underground test sites in media other than tuff, alluvium, and granite, particularly in media of high density. The results of this work are set forth in this report. There are two possible test sites in basaltic rock in the area. Site 1 (within Nevada State coordinates E 709,500 and 710,500 ft and N 854,000 and 856,000 ft) is suitable for a shallow test. Site 2 (bounded by Nevada State coordinates E 707,000 and E 712,000 ft and N 848,000 and N 853,000 ft) could be used for a deep confined test in basaltic rock.

This document details activities of the Fuels Preparation Department during the month of May 1962.

This document details activities of the Fuels Preparation Department during the month of July 1962.

This document details activities of the Fuels Preparation Department during the month of November 1962.