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Identification of Curium-242 in Irradiated Neptunium

Description: Technical report. From Abstract : "Curium-242 was experimentally identified as a minor product of the irradiation of neptunium-237 in a nuclear reactor." From Introduction : "Determinations of the half-life, alpha energy, and daughter product of the purified curium nuclide were made in this study to confirm the identity as curium-242."
Date: August 1961
Creator: Carothers, Glenn A.
Partner: UNT Libraries Government Documents Department
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Surge suppressors for the PRTR process tube flow meters

Description: Each tube of the PRTR is provided with flow monitoring equipment consisting of a venturi flow meter in the inlet piping, sensing lines containing valves, and a Panellit flow transmitter. The flow transmitter does three things: converts the pressure drop signal of the venturi to a visual readout; provides an electrical signal for recording; and provides a signal to the safety circuit which causes a reactor scram should the flow increase or decrease beyond pre-set valves. After startup of the PRTR, it was found that the readings of flow meters on those process tubes which connect near the inlet of the bottom ring headers were fluctuating excessively. As an interim measure during the power tests at low reactor powers, the meter fluctuations were reduced by throttling the valves in the sensing lines from the flow venturi to the flow meter. This was recognized as being questionable for a permanent solution since this practice introduces an unknown and variable lengthening of the response characteristics of the meter. An experimental program was therefore undertaken to determine the degree of valve throttling which might be appropriate for fluctuation suppression and to device other and better methods of suppression. The experiments show that throttling of valves in the flow transmitter sensing lines is not a satisfactory way of decreasing the fluctuations. The valves must be closed to about 1/4 open before they decrease fluctuations significantly. Further closure to about 1/8 open causes an excessive lengthening of the response characteristics of the transmitter. This provides only a very narrow range of valve positions which are permissible and effective. The increased hydraulic flow resistance of 3--5 foot lengths of 1/16-inch tubing in the sensing line causes the flow transmitter system to become nearly critically damped. This will eliminate the oscillatory behavior of the readings without causing …
Date: August 11, 1961
Creator: Hesson, G. M.; Thorne, W. L. & Batch, J. M.
Partner: UNT Libraries Government Documents Department
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Serious Radiation Event - Hanford Capabilities

Description: The need for an industrial plant to know its capabilities to rescue personnel and minimize damage to vital operating equipment in the event of a serious accident has long been recognized. Although Hanford`s experience has been without a serious radiation event as characterized in this report, it is prudent to have an objective assessment of the capabilities which do exist and an analysis of the general capabilities which the plant should possess to maximize the effectiveness of rescue action in such an event and to limit the consequences to personnel, vital equipment, and the surrounding environs. It in the purpose of this report to: (1) make an assessment of the capabilities which are currently in place at Hanford to handle the multitude of serious problems arising from a serious radiation event; and (2) define the level of capability which the Hanford Atomic Products Operation should possess to minimize the consequences of such an accident. It is not the intent of this presentation to discuss the possibilities, probabilities, prevention or courses of serious radiation events.
Date: August 15, 1961
Creator: Keene, A. R.; Unruh, C. M.; Backman, G. E. & Carter, L. A.
Partner: UNT Libraries Government Documents Department
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Supplement A to design of PT-IP-263-A-FP, evaluation of chemically nickel plated fuel elements

Description: 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.
Date: August 22, 1961
Creator: Clinton, M. A. & Hodgson, W. H.
Partner: UNT Libraries Government Documents Department
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Hazards review Phase 1 fission product processing in B-plant

Description: Future operation of B-Plant for Phase I fission product processing involves may of the hazards associated with current chemical Processing Department activities in both the Purex and Redox Plants. The specific B-Plant hazards under Phase I operations are discussed in this report from the standpoint of comparable Purex and Redox hazards.
Date: August 24, 1961
Creator: Michels, L. R. & Zahn, L. L.
Partner: UNT Libraries Government Documents Department
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Report to the working committee

Description: Topics include: uranium quality and technology, fuel assembly technology, irradiation experience, E-N program, projection fuel program, overbore fuel program, Ni-plated Al-clad fuel elements, fuel element development, NPR program.
Date: August 29, 1961
Creator: Stringer, J. T. & Minor, J. E.
Partner: UNT Libraries Government Documents Department
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Production test IP-431-A expansion of overbore test facilities C Reactor

Description: The objective of this production test is to authorize the installation of at least 40 additional large size Zircaloy process tubes in graphite channels that have been enlarged to 2.275 inches at C Reactor and to charge these tubes with large diameter (CVIN) fuel elements to obtain preliminary conversion ratio data and further qualitative data regarding the irradiation behavior of large diameter fuel element designs.
Date: August 11, 1961
Creator: Van Wormer, F. W.
Partner: UNT Libraries Government Documents Department
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Measurement procedures for purified birch

Description: Since start-up of the neptunium recovery and purification program at HAPO, the neptunium-has been subjected to the following process steps: Production of crude neptunium nitrate by the Redox and Purex plants; Production of purified neptunium nitrate, using an anion exchange process, in research and development facilities operated by CPD`s Process Chemistry Operation; Production of neptunium oxide (NpO{sub 2}) by Process Chemistry. Mixing the NpO{sub 2} powder with aluminum powder and incorporation in an aluminum-clad-target element, done by the Plutonium Metallurgy Operation of Hanford Laboratories; and Irradiation of the target elements in the Savannah River reactors; and Chemical reprocessing of the irradiated target elements at Savannah River to recover the desired plutonium-238 and the unconverted neptunium. As frequently happens to between-contractor transfers, neptunium material balances between Hanford`s NpO{sub 2} and Savannah River`s recovered neptunium have sometimes been poorer than expected. Because of wide spread interest in this problem, it was deemed desirable to put on record a description of the measurement techniques which have been used at Hanford for NpO{sub 2} and the improvements which have been made in the analytical methods. The measurement methods used for purified nitrate are also described.
Date: August 14, 1961
Creator: Buckingham, J. S.
Partner: UNT Libraries Government Documents Department
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Post irradiation examination of transverse cracked rupture from 2986 DR (Rm 427)

Description: An O III N, I and E dingot element exhibited several transverse cracks when it failed in tube 2986 DR. Detailed examination in the Radiometallurgy Laboratory was requested to determine the cause of failure and obtain the ingot number. The element was so badly damaged that it was impossible to determine the exact cause of failure. Reaction between the coolant and fuel occurred at the internal surface around the spire over the total length of the element. Reaction was the greatest near the midpoint where the transverse cracks occurred and was least in the male end. Metallography of both longitudinal and transverse sections revealed that the metal quality of the fuel vas good. Cracks which originated from the oxide extended in both the longitudinal and transverse directions. The fuel near the core, which was insulated by oxide, was annealed but had not been heated into the uranium beta phase. One transverse crack which appeared to have water entry was observed approximately two inches from the male end. The ingot number was UZ 5751 B. After removal of the cladding at the ends, the uranium was bright dip etched and no defects were observed in the surface metal.
Date: August 28, 1961
Creator: Gruber, W. J.
Partner: UNT Libraries Government Documents Department
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Radiochemistry for the rupture of Zircaloy-2 clad heavy-walled tubular fuel element in KER Loop-3

Description: On the 0000 -- 0800 shift, December 9, 1960, the delayed neutron monitor on KER Loop 3 gave a high coolant activity signal indicating a possible fuel element failure in this loop. KE Reactor was shut down immediately thereafter. This report is being written to summarize the events pertinent to this KE Reactor scram and to discuss the results and significance of data from analyses on coolant and coupon samples taken from the KER Loop 3 system.
Date: August 21, 1961
Creator: Perrigo, L. D.
Partner: UNT Libraries Government Documents Department
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Production test IP-412-AI: B and C reactors export system test

Description: Purpose of this test was to determine the adequacy of the export system for supplying flow to a dual reactor area under simulated emergency conditions.
Date: August 2, 1961
Creator: Benson, J. L. & Jones, S. S.
Partner: UNT Libraries Government Documents Department
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Radiochemistry for the rupture of a Zircaloy-2 clad thermocoupled fuel element in KER Loop-1 on May 12, 1961

Description: On the 1600--2400 shift, May 12, 1961, the delayed neutron monitor on KER Loop 1 gave a high coolant activity signal indicating a possible fuel element failure in this loop. KE Reactor was shut down immediately thereafter. This report is being written to summarize the events pertinent to the KE Reactor scram and to discuss the results and significance of data from the analyses of coolant samples taken from the KER Loop-1 System.
Date: August 18, 1961
Creator: Perrigo, L. D.
Partner: UNT Libraries Government Documents Department
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Daily data sheets, UT-2 test, PT-IP-310-A-FP, D reactor

Description: This report contains daily data of the D reactor`s: outlet temperature; panelist dial; panelist base; front header pressure; and downstream probe thermocouple readings.
Date: August 17, 1961
Creator: Clinton, M. A.
Partner: UNT Libraries Government Documents Department
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Analysis of radiation induced stresses in C Reactor graphite

Description: Figures are included which gives the necessary information on C Reactor for analysis of stress buildup in the graphite tube blocks.
Date: August 14, 1961
Creator: Yoshikawa, H. H.
Partner: UNT Libraries Government Documents Department
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Chemical Processing Department Monthly Report: July 1961

Description: This report, for July 1961 from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; employee relations; weapons manufacturing operation; and safety and security.
Date: August 21, 1961
Creator: Hanford Atomic Products Operation. Chemical Processing Department.
Partner: UNT Libraries Government Documents Department
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Heat transfer experiments simulating a failure of the inlet piping to a BDF reactor process tube

Description: Laboratory heat transfer experiments were conduced to investigate fuel element temperatures which could result from coolant flow loss following a failure of the inlet piping to a process tube at a B, D, F, DR, or H reactor. The results are reported herein. Failure of the inlet coolant piping between the front header and the process tube on a reactor would stop the normal flow of cooling water to the fuel elements. Such a failure should immediately initiate a reactor shutdown, but the only means of removing the heat released during the post-shutdown period would be by reverse flow of hot water from the rear cross header. The subject experiments were conducted to determine what rear header pressure would be required to achieve adequate cooling of a BDF type reactor fuel assembly following such a piping rupture. Experimental studies were previously reported concerning failure of inlet piping to a K reactor geometry. The analytical techniques and experimental procedures used previously were also used in the present experiments.
Date: August 16, 1961
Creator: Waters, E. D. & Kreiter, M. R.
Partner: UNT Libraries Government Documents Department
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Post irradiation examination of self support elements with severe in-reactor cladding corrosion (RM425)

Description: Three self support, natural uranium, I & E production fuel elements, which exhibited serious in-reactor cladding corrosion were selected from several tubes of discharged pieces from PT-IP-272-A-FP. The elements were transferred to the Radiometallurgy Laboratory for detailed examination to determine thickness of the remaining cladding and measure any irradiation induced dimensional changes in the fuel cores. Each element had four support tabs attached to each end and was from Parent Lot KT.009. The fuel geometry of all three was C IV NS.
Date: August 28, 1961
Creator: Gruber, W. J.
Partner: UNT Libraries Government Documents Department
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Coextrusion Process Specifications

Description: These specifications, prepared for the N-reactor fuel, provide the framework on which the overall control and guidance of the manufacturing process shall be built.
Date: August 1, 1961
Partner: UNT Libraries Government Documents Department
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Speed-of-control limits, 105-DR

Description: The current operating conditions at 105-DR, because of the recent Ball 3X drop, dictate temporary lower speed-of-control limits. While is si still not expected that this reactor will be limited by the speed-of- control criterion, it is felt that the limit curves have been temporarily sufficiently reduced as to warrant publication of former limits. Limit curves as a function of weighted graphite temperature for the continuous and block discharge schemes are included in this document. The two sources of the reduction in the speed-of-control limit comes from two sources: (1) While the limit is commonly expressed in MW of pile power level, it is actually a function of maximum tube power, (2) The poisoning effect of the balls lodged in the stack has been estimated at about 15 mk.
Date: August 1, 1961
Creator: Nechodom, W. S.
Partner: UNT Libraries Government Documents Department
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Process tube expansion: 105-C conversion to self-supported fuel elements

Description: A method of reducing gas leakage and rear fate maintenance is to fix the rear gunbarrel and process tube to the shield and eliminate the need for a rear expansion bellows. This arrangement is being considered as part of the C-reactor conversion to self-supported fuel element program. The fixed rear gunbarrel will require that all hydraulic load and thermal expansion elongation of the tube be taken through the front bellows. In the event that tubes freeze in their channels, forces due to thermal expansion and hydraulic pressure will be transmitted to the rear shield if the tube and rear gunbarrel are fixed to the shield. Calculated maximum forces which can be transmitted to the shield are given in this report. The probability and amount of restraint which can occur in the tube are discussed. Finally, methods of insuring that the restraint does not cause excessive forces on the rear shield are discussed.
Date: August 9, 1961
Creator: Mollerus, F. J. Jr.
Partner: UNT Libraries Government Documents Department
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Ultrasonic and resistance welding of ZR supports to ZR fuel element cladding

Description: Tube-tube type fuel elements require support devices to maintain concentricity both between the element components and between element and process tube. A variety of support devices have been conceived. all require welding of a tab or other attachment device to the Zircaloy clad. Ultrasonic and resistance welding were evaluated for this application. Investigation showed ultrasonic welding to be not acceptable in its present state. Resistance welding proved satisfactory. This paper will briefly cover the work done in evaluating the two methods of attachment.
Date: August 8, 1961
Creator: Steinkamp, W. I.
Partner: UNT Libraries Government Documents Department
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