The Irradiated Rupture Prototype (IRP) has been used for rupture testing irradiated NPR prototype fuel elements. Most of the tests have been made to determine the rupture effect of different reactor exposures, fuel element geometries and water cooldown rates following the start of the rupture. This report summarizes the results obtained to date, mentions where information is lacking and gives further tests scheduled for the IRP.

The instructions described in this procedure are typical of the operation of Test Cell A relative to the KIWI-B4A. Operation of Test Cell A relative to the NRX reactor will require modifications dictated by specific test requirements. Under NRX conditions, it will be the responsibility of the test cell manager to evaluate the capabilities of Test Cell A in terms of given test requirements and then set forth detailed checklists which will be compatible with the test requirements.

This report details activities of the N-Reactor Department during the month of August 1963.

Americium (atomic weight 241) grows into plutonium as a result of the 13-year half-life beta decay of Pu{sup 241}. An appreciable quantity of Am{sup 241} has.grown into the scrap that accumulated following shut-down of the Recuplex facility, and which will be processed during the initial period of operation of the plutonium reclamation facility. Current interests in trans-plutonium research and in isotopic heat sources make it desirable to consider the recovery of this Am{sup 241}. The americium contained in scrap that is processed through the reclamation facility should appear almost quantitatively in the aqueous waste stream (CAW). Subsequent processing of the CAW through the DBBP extraction column (CW) of the waste treatment facility should result in further separation of plutonium and americium by extracting some of the plutonium while most of the americium remains in the aqueous waste stream (CWW). The fact that a low free nitric acid concentration will favor the extraction of americium into TBP-type solvents can be used to recover americium from the large volume, high salt concentration CWW stream. The flowsheets in this document present the chemical conditions for effecting this recovery.

These criteria delineate objective, bases, and functional requirements governing preparation of design of the bauxite-sulfuric acid feed facilities installed in the 183-KE and KW Buildings. These facilities produces the chemical coagulant used in the treatment of Columbia River water in the K Area water plants and thus replaces the existing liquid alum feed systems used for this purpose.

The report is divided into: Present reactor fuel production, N-RD production fuels, N-fuel development, and current reactor fuel development.

The objectives of this production test are: (1) to authorize the irradiation of enriched uranium hot-die-size diffusion-bonded fuel elements; (2) to provide information for further evaluation of the hot-die-size process; (3) to evaluate the relative corrosion behavior of the enriched hot-die-sized fuel as compared to the corresponding Al-Si fuel model; and (4) to determine the gross dimensional stability of the fuel under irradiation. This test authorizes irradiation of 19 monitor columns of enriched self-support fuel elements in the C Reactor. The columns will contain 28 fuel elements each; each of the columns will contain 14 fuel elements fabricated by the hot-die-size diffusion-bonded process and 14 elements of the standard Al-Si type. Ten of the columns will contain, in alternate positions in the downstream half of the charge, ``paired`` Al-Si and hot-die-size elements -- elements containing cores from adjacent ingot rod positions. The remaining nine columns will contain hot-die-sized and Al-Si elements, paired by ingot only, in alternate positions in the downstream half of each column. The upstream portion of all charges will contain hot-die-sized and Al-Si elements in alternate column positions. All of the columns of this test will be irradiated to an average column exposure of 1000 Mwd/t. The …

This document defines codes which are for use with the data processing program developed for calculating isotopic analyses from raw data out put of a thermal ionization mass spectrometer.

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This document defines the objectives, bases, and functional requirements that shall govern the preparation of detail design of the gravity fed water supply to reactor storage basins for all eight reactors. In the event that appears advisable and feasible to discharge all metal from the reactors into the metal storage basins, it would be necessary to add water to the storage basins to prevent overheating of the fuel elements. Without the addition of cool water the storage basin water would soon start to boil and evaporate, eventually exposing the metal to the air. Existing facilities do not permit assurance that sufficient water can be added to the storage basins for the required period of time to protect a complete discharge of fuel elements in the storage basin if the area is left unattended and pumps are shut down. A positive gravity fed system to the metal storage basins from existing supplies of stored water shall be provided by this project. This new gravity fed system, once it is started, shall operate unattended and shall supply adequate water to the storage basins for the required period of time. It shall not be dependent on electric or steam-driven pumps for its continuous …

In response to an inquiry by Euratom, estimates of the costs for the production of 85 kg of 15% Pu-240 metal and for the production of 15 kg of 25% Pu-240 oxide at HAPO were recently requested by the Atomic Energy Commission. In connection with this inquiry, comments were requested regarding the possibility of establishing and measuring product plutonium oxide ``sinterability`` parameters. This report presents estimates of the costs for separations processing by the Chemical Processing Department for the cases of interest. Process ground rules and any necessary assumptions are explained. In addition, the problems of measuring oxide ``sinterability`` are briefly discussed. Highlights of this study were previously transmitted to the Irradiation Processing Department for incorporation in a formal reply to the original Commission request. Subsequently, some adjustments in ground rules have caused cost changes which are included in this report.

This report briefly describes the circumstances involving the accidental release of Iodine 131 into the atmosphere from the Purex Plant at Hanford in September, 1963.

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This Cost Improvement Proposal recommends an increase in the length of enriched fuel elements (for H Reactor only) from 6.64 in. to 8.1 in. A cost improvement would result as follows: Annual Out-of-Pocket Cost Savings, $121,000; Annual Outage Time Savings -- 65 Hours, worth 4,550 MD and Annual Production Level Gain -- 7,000 MWD. Since the Production Fuels Section and off-site suppliers are already equipped to handle elements of this length, the cost to make the change should be insignificant compared to the annual savings.

This report is divided into: metallic fuels technology, metallic fuel development, plutonium physical metallurgy, plutonium mechanical properties, and plutonium mechanical metallurgy.

Several evaluations of Harvey ``O`` size natural I&E aluminum components have been made. The performance of components received prior to May 1961, was unsatisfactory. To improve component quality, Harvey made major process modifications. Components processed at HAPO after July 1, 1961, were received subsequent to Harvey`s modifications, and met specifications. Procurement and evaluation of an additional 150,000 Harvey OIIIN components was authorized. Evaluation of the 5 first 150,000 Harvey OIIIN improved components suggested partial qualification. This document reports and makes recommendations based on the test data for Harvey improved components as of August 31, 1963.

A computer program was written to calculate the initial ratio of isotopes which should be used in an in-core neutron flux detector to obtain the maximum use lifetime where the useful lifetime is defined as that period of time during which the sensitivity of the detector remains within specified limits. Input data for the program include cross sections of the isotopes to be used. In order that various reactor environments might be considered, Westcott cross sections have been employed. Westcott cross sections assume that the neutron spectrum in a thermal reactor can be characterized by the temperature, T, and by a spectral parameter, r. An experimental technique was developed for measuring these parameters. The evaluation of the parameter, r, is based on a standard technique comparing the relative radioactivity induced in bare and cadmium-covered cobalt samples. The determination of the neutron temperature, T, was made using a mass spectrometer to determine the isotopic changes in uranium and plutonium samples as a function of.exposure in the neutron flux. This is a new technique which appears to be both simple and accurate. The irradiations were made in a water-cooled irradiation facility located in the KE Hanford production reactor. Based on the results …

The 04T lithium-aluminum target elements that were irradiated at Hanford during the early part of 1963 indicated that a longer element would be more desirable from the standpoint of tritium production. The Production Fuels Sections was requested in late August, 1963, to manufacture a test load of ``O`` size elements that would be five inches in length. This document presents the manufacturing information that is necessary to make a finished target element five inches in length that will approximate the ``O`` size geomery with the existing supply of AlSi type aluminum components.

When the ``stop-gap`` rate-of-rise meter was installed on all 8 reactors by Amendment 1 to CGI-806 it was intended to replace it with a more satisfactory design when possible. Successful completion of the development program now permits installation of equipment of sufficient reliability to be utilized in the safety circuit. This document defines the program necessary to provide power rate-of-rise protection in the safety circuits of B, C, D, DR, F, H and KE Reactors, provides justification for the project., and presents estimates of the cost and schedule required to accomplish the program.

A half-plant low pH test began at D Reactor on March 19, 1963. The purpose of the test was to provide quantitative data on the reduction in aluminium corrosion obtained by lowering reactor coolant pH from 7.0 to 6.6. The benefits of lower pH will be monitored by ex-reactor tube examinations) in-reactor wall thickness measurements, coupons, and visual examination pleas weight loss measurements of fuel elements. This report presents the results of the visual examination and weight loss measurements on 18 columns of fuel elements irradiated during the test

An analytical chemistry program was conducted to determine total U-233 and U-232/U-233 ratio in the irradiated thorium oxide. This report summaries the results for samples exposed for one month and briefly describes the analytical procedures.

Fabrication studies to produce high density solid solutions of 80% UC-- 20% RaC, with reproducible structure, composition, and density, were completed. Two types of material were produced: (U/sub 0.8/Pu/0.2/)C/sub 0.95/, single-phase monocarbide pellets with average densities of 12.8 g/cm/sup 3/ (94% of theoretical), sintered at 1950 deg C; and (U/sub 0.8/Pu/0.2/) C/sub 0.95/ + 0.1 wt% Ni sintering aid, major monocarbide and minor amount of sesquicarbide pellets, with average densities of 13.1 g/cm/sup 3/ (96.5% of theoretical) sintered at 1550 deg C. Bar-shaped thermal expansion specimens were fabricated of UC, prior to fabrication of similar (U,Pu)C specimens. UC pellets were fabricated for electropolishing and liquid-metal bonding studies. Chemical analysis procedures were established, and checked, for plutonium, nitrogen, and oxygen. Chemical analysis procedures for carbon are being estsblished. The experimental setups for the measurement of coefficient of thermal expansion, thernial stability, melting point, and fuel cladding compatibility were completed. Testing of thermal expandsion, vapor pressure, and melting point standards was initiated. UCliquid metal-tantalum compatibility tests were completed, and (U,Pu)C-liquid metal-tantalum compatibility specimens were prepared. These tests are to help in choosing a liquid-metal bond for the thermal conductivity test. The pre-installation tests for the safe performance of the high temperature measurements …

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