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P Division monthly report, September 1949

Description: This progress report discusses the activities of the P Division for the month of September, 1949. The B and F piles operated at 275 megawatts (MW) and the D pile at 305 MW throughout the month except for cutages listed under Area Activities. A total of 60.22 tons of metal, at an average concentration of 396 megawatt days/ton (MWD/ton) was discharged from the piles during the month. The 105-H Building was accepted from the Construction Division on September 28 with certain exceptions noted under the Operating Experience section of this report. At month end P Division operating personnel are making preparations for activation of the H pile. On September 28 the operation of the 300 Area oxide burning process was reduced from a two shift to a one shift five day operation. This change in schedule was possible as a result of working off the backlog of uranium oxide. The shipment of 200 tons of canned slugs to Building 105-DR for storage was completed on September 9 and the shipment of 250 tons of canned slugs to Building 105-H for the initial loading charge was completed on September 26.
Date: October 6, 1949
Creator: Lee, E.P.
Partner: UNT Libraries Government Documents Department

Compilation of data on 51 ruptured slugs

Description: The following tabulation includes information on all uranium slug failures which have occurred through September 26, 1951. The four suspect slugs which are listed were discharged from tubes which gave strong indication of containing ruptures. Although no obvious rupture could be found among the slugs from these tubes, the listed pieces exhibited defects which may be incipient ruptures.
Date: October 4, 1951
Creator: O'Keefe, D.P.
Partner: UNT Libraries Government Documents Department

Vernatherm functional test, Test Project No. 14

Description: The Vernatherm unit is a temperature sensitive capsule which translates temperature change to mechanical movement of a brass plunger. A change in phase of a hydrocarbon contained in a cylinder caused by temperature change, causes movement of the brass plunger. The units are available for various temperature ranges. The adaptation considered was to monitor the outlet water temperature from individual process tubes in the ``G`` pile. The purpose of the test was to determine the magnitude of error, if any, that is induced in a standard Vernatherm unit of known calibration when subjected to gamma irradiation. The accuracy of the units prior to irradiation was within 1.25 F. An examination of the calibration curves shows that a hysteresis effect in the hydrocarbon of the unit causes the curve of descending temperature to be displaced from the ascending temperature curve. The effects of irradiation were to decrease the accuracy to within 3.75 F. After a total exposure of 17,416 {times} 10{sup 4}R at 1.1 MEV. Since the exposure of 11,441 {times} 10{sup 4}R is equivalent to approximately 2,000 months exposure in the pile at 250 MW they can conclude that the effect of radiation is not detrimental since the increase in accuracy is only within 0.75 F.
Date: October 31, 1951
Creator: Smith, J. L.
Partner: UNT Libraries Government Documents Department

Special hazards report - I E fuel loads

Description: This report has been prepared in answer to the request from the AEC contained in the letter of October 1, 1957, from A. T. Gifford, HOO to A. B. Greninger. As requested, the report is of a summary nature and a more complete discussion of many of the points considered will be found in the references listed. The report is directed primarily at C reactor but some discussion of the other reactors is also included. A description of the proposed utilization of I E slugs in C reactor together with the associated power increase schedule is presented below. The reasons for changing to the I E element are presented together with a comparison of solid and I E slugs in the C reactor. The changes being made in C reactor under CG 600 are described. The operational characteristics of the C reactor using solid and I E elements are compared and finally the nuclear safety status of all of the Hanford reactors assuming I E loadings is reviewed.
Date: October 15, 1957
Creator: Brown, J.H.; Fullmer, G.C.; Trumble, R.E. & VanWormer, F.W.
Partner: UNT Libraries Government Documents Department

Pressurized Water Reactor Program Technical Progress Report for the Period July 15, 1954 to August 26, 1954

Description: This progress report has 2 parts. Part 1 PWR Engineering covers the following topics: (1) power plant analysis and systems; (2) power plant components and component materials and tests; and (3) reactor and auxiliaries. Part 2 PWR Development covers: (1) fuel element development; (2) metallurgy of core materials; (3) materials application development; (4) chemistry development; (5) irradiation effects; and (6) reactor physics.
Date: October 31, 1957
Partner: UNT Libraries Government Documents Department

Radiation damage study on the lithium hydride SNAP shield

Description: Radiation damage may occur to the lithium hydride shields as a result of the reaction Li{sup 6}(n, {alpha})H{sup 3}. There is evidence in the literature indicating both the existence and absence of radiation damage to the SNAP shields. It is believed that there is a high probability that there will be damage and that it will adversely affect the properties of the shield. This damage may take the form of: (1) volume expansion of the hybrids, (2) void formation within the hybrids, and (3) gas pressure build-up in the shield container. Based upon the results of experiments with lithium fluoride, which may serve as a model for the hydride, there appears to be a threshold neutron dose which volume expansion effects can not be removed by annealing. Similarly, above the threshold dose, intercrystalline voids, formed as a result of radiation damage, appear to increase in size with increasing temperature. It has been established that at the SNAP shield operating conditions, essentially all of the hydrogen formed will recombine with free lithium. The helium atoms, however, remain trapped interstitially, in intercrystalline voids, or along subgrain boundaries. Appreciable amounts of helium gas are not released until the melting point of the hydride is approached. An insignificant portion of the hydrogen in the shield is lost by permeation of the stainless steel shield container at the SNAP 10 operating conditions. 23 refs.
Date: October 4, 1961
Creator: Doctor, R.D.
Partner: UNT Libraries Government Documents Department

Presentation to the Atomic Energy Commission and the Air Force, June 14, 1962

Description: This volume contains the charts and backup material presented to the Atomic Energy Commission and Air Force on June 14, 1962 concerning General Electric's Nuclear Materials and Propulsion Operation (formerly the Aircraft Nuclear Propulsion Department), during its work on the development of a nuclear power plant for manned aircraft.
Date: October 1, 1962
Partner: UNT Libraries Government Documents Department

Tensile and stress rupture tests of S8DR Hastelloy-N heats: ORNL verification tests

Description: In connection with the ORR Hastelloy-N irradiation experiments, a limited number of tensile tests and uniaxial and biaxial stress-rupture tests on S8DR Hastelloy-N heats were conducted at AI to determine the effect of some of the ORNL test conditions. The tests performed at AI were in parallel to the ORNL control tests. The results showed that the effect of the test condition variations between the two tests were generally insignificant. The effect of the test conditions is discussed.
Date: October 5, 1967
Creator: Lee, S.K.
Partner: UNT Libraries Government Documents Department

FAST FLUX TEST FACILITY. Design and development quality assurance requirements for the FFTF

Description: The document is presented to provide general management requirements for Pacific Northwest Laboratory (PNL) and contractor design and development quality assurance programs to assure the required quality level of the various items required for the FFTF. The document is applicable as imposed by the contract to FFTF contractors and subcontractors. The document is also applicable to PNL design and development activities related to the FFTF.
Date: October 23, 1968
Creator: Albert, W.G.
Partner: UNT Libraries Government Documents Department

Status of irradiations performed by fuel and target irradiation technology for BNW as of September 30, 1969

Description: This report itemizes the irradiations performed by Testing and Irradiation Services for Battelle-Northwest. It lists the material being awaiting disposition and material shipped during the report period. Information consists of: TISR No.; request number; target material; piece number; operating time; CMK absorbed; charge date; location; exposure to date-NVT; discharge, date and time; and shipping date.
Date: October 9, 1969
Creator: Barker, L. V.
Partner: UNT Libraries Government Documents Department

Dissipation of the reactor heat at the Savannah River Plant

Description: The effluent cooling water from the heat exchangers of the Savannah River nuclear reactors is cooled by natural processes as it flows through the stream beds, canals, ponds, and swamps on the plant site. The Langhaar equation, which gives the rate of heat removal from the water surface as a function of the surface temperature, air temperature, relative humidity, and wind speed, is applied satisfactorily to calculate the cooling that occurs at all temperature levels and for all modes of water flow. The application of this equation requires an accounting of effects such as solar heating, shading, mixing, staging, stratification, underflow, rainfall, the imposed heat load, and the rate of change in heat content of the body of water.
Date: October 1, 1971
Creator: Neill, J.S. & Babcock, D.F.
Partner: UNT Libraries Government Documents Department

Characterization of an energy source for modeling hypothetical core disruptive accidents in nuclear reactors. First interim report. [LMFBR]

Description: The expansion characteristics of the detonation products of a high-explosive energy source used to simulate the pressure-volume change relationships for sodium-vapor expansions during hypothetical core disruptive accidents in a Fast Test Reactor were determined experimentally. Rigid cylinder-piston experiments performed at two scales (ratio 1:3) were undertaken to determine a pressure-volume relationship as a function of source mass and expansion environment. Some of these measurements were compared with code calculations for the source.
Date: October 1, 1972
Creator: Cagliostro, D J & Florence, A L
Partner: UNT Libraries Government Documents Department

Superheater hydraulic model test plan

Description: The plan for conducting a hydraulic test on a full scale model of the AI Steam Generator Module design is presented. The model will incorporate all items necessary to simulate the hydraulic performance characteristics of the superheater but will utilize materials other than the 2-1/4 Cr - 1 Mo in its construction in order to minimize costs and expedite schedule. Testing will be performed in the Rockwell International Rocketdyne High Flow Test Facility which is capable of flowing up to 32,00 gpm of water at ambient temperatures. All necessary support instrumentation is also available at this facility.
Date: October 1, 1973
Creator: Gabler, M. & Oliva, R.M.
Partner: UNT Libraries Government Documents Department

AI reference LMFBR steam-generator development

Description: The Design Data Sheets summarize the key parameters being used in the design and analysis of the AI Prototype LMFBR Steam Generator. These Data Sheets supplement SDD-097-330-002, Steam Generator System, 1450 psi Steam Conditions. This document will serve as the baseline design data control until a GE/RRD approved steam generator specification with ordering data is received.
Date: October 12, 1973
Creator: Anderson, T.L.
Partner: UNT Libraries Government Documents Department