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Additional Experiments on Stored Energy in BNL Reactor Graphite (open access)

Additional Experiments on Stored Energy in BNL Reactor Graphite

In the memorandum entitled "Stored Energy in BNL Reactor Graphite", dated February 25, 1953, there is described an experiment conducted by Gurinsky's group to determine the energy per gram of irradiated graphite released in a 200°C anneal. Similar experiments were subsequently conducted by W. Kosiba, differing from the original in two particulars: a) Instead of two graphite samples, one normal, and one irradiated, Kosiba used only an irradiated sample which he heated first to release the stored energy, and then again after the energy was released. In this way, he obtained time against temperature curves for both normal and irradiated graphite from the same sample. (These curves are graphed for each run in Figs. 1 thru 5.) b) The vycor tubing used in the original experiment was not used by Kosiba. Five runs of this experiment were selected, Runs 4P, 13, 36, and 40 at furnace temperatures of 200°C, and Run 45 at a furnace temperature of 400°C.
Date: August 3, 1953
Creator: Mulhern, T.
System: The UNT Digital Library
The Distribution of Thermal Neutrons in a Slug with Thick End Caps (open access)

The Distribution of Thermal Neutrons in a Slug with Thick End Caps

The distribution of thermal neutrons in a W slug having a one centimeter aluminum end cap has been calculated on the basis of simple diffusion theory. It is found that the average neutron density, and therefore the power output, at the end of the slug is about 34% higher than the density far from the end cap. This result agrees well with the recent Argonne pile experiments (CP-1729).
Date: August 3, 1944
Creator: Wilkins, J. Ernest, Jr., 1923-2011
System: The UNT Digital Library
HRT Reactor Hazards (open access)

HRT Reactor Hazards

Several potential hazards that have been recognized and anticipated in the design and fabrication of the pressure vessel in the Homogeneous Reactor Test are discussed. These hazards results from the high operating pressure and temperature of the reactor, the exposure of the reactor vessel material to potential embrittlement and other affects of fast-neutron irradiation, and the need for containment of corrosive flowing liquids. The steps taken in recognition of these hazards are also discussed. The applicability of present codes to the reactor vessel fabrication is considered. Additional fields are suggested where recommended practices developed by code writing bodies could assist in development-type reactor design and fabrication.
Date: August 3, 1956
Creator: Miller, E. C.
System: The UNT Digital Library
Isotopic sources of secondary radiation : second interim technical report covering the period from March 1 to July 1, 1959 (open access)

Isotopic sources of secondary radiation : second interim technical report covering the period from March 1 to July 1, 1959

Summary. During the past year the work carried out in this program has included experiments on the x-ray output produced by the separated fission product beta sources, the interpretation of such experiments for the effective design of high level secondary x-ray sources, and the design and testing of prototype sources of secondary radiation for specific industrial applications. In this report results of work carried out during the past fours months are reported: (1) analysis of beta-excited x-rays; (2) performance of Kr85 prototype x-ray source; and (3) design of high level Pm147 x-ray sources.
Date: August 3, 1959
Creator: Voyvodic, L.
System: The UNT Digital Library
Laboratory Survey of Deoxidants for Uranium Chips (open access)

Laboratory Survey of Deoxidants for Uranium Chips

Summary: "Six different types of solutions of various concentrations were used to deoxidize uranium chips on a laboratory scale. None of the solutions tested appears to be more desirable than 50% nitric acid for deoxidizing uranium chip on a production scale."
Date: August 3, 1950
Creator: Kattner, W. T.
System: The UNT Digital Library
Melting Point of Th-U-C Fuel Elements (open access)

Melting Point of Th-U-C Fuel Elements

From the point of view of predicting melting behavior of fuel elements containing fission products after 50 percent burn-up, the fuel can be considered to consist of 2000 moles Th, 150 moles U, 55 moles of rate earth metal, 31 moles of Zr, 25 moles of Mo, 20 moles of Rh-Ru-Tc, and 15 moles of alkaline earth metal. All other fission products are present in too small amounts to have any important effect upon the melting point or will have vaporized. However, the presence of alkali metal vapor should be considered.
Date: August 3, 1959
Creator: Brewer, Leo, 1919-2005
System: The UNT Digital Library
The Roll Cladding of Uranium With Aluminum (open access)

The Roll Cladding of Uranium With Aluminum

Report discussing a study regarding rolling as a technique for making uranium fuel elements that are flat-plate and aluminum-clad.
Date: August 3, 1954
Creator: Saller, Henry A.; Paprocki, Stan J. & Delaney, J. F.
System: The UNT Digital Library
Temperature and Heat Flow in a Graphite Electrode (open access)

Temperature and Heat Flow in a Graphite Electrode

In making various experimental heat studies a power producing metal slug is simulated by a slug with a graphite rod electrode of 3/8" diameter inserted lengthwise through it. There is a helium filled annular space between the graphite and the inner surface of the slug cylinder. Radiant heat passes from the electrode to the metal; with proper adjustment of the electrode current the slug in the steady state will therefor "produce" the same amount of energy from its exterior surface as it would under operating conditions. The question arises, however, as to how uniform the electrode temperature is along its length. And also, in some cases one end of the electrode is embedded in the slug metal; it is then desirable to know how much heat flows by conduction from the electrode into the slug.
Date: August 3, 1944
Creator: Schlegel, R.
System: The UNT Digital Library