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Preliminary Design Study for a Sodium-Graphite-Reactor Irradiation Facility

Description: The results of an investigation to integrate a Na/sup 24/ irradiation processing facility with an operating sodium graphite reactor are presented. An irradiation facility incorporated into a reference SGR (Hallam Nuclear Power Facility, Hallam, Nebraska) is described. Development of the facility application, preliminary design criteria and capital and operating costs are discussed. Recommendations for further development of the technology and economics of this type of irradiation facility are included. (auth)
Date: January 31, 1959
Creator: Thompson, D.S. & Benaroya, V.
Partner: UNT Libraries Government Documents Department

Na{sup 24} ACTIVITY IN THE HFIR PRIMARY COOLANT WATER

Description: Sodium-24, produced by the Al (n, alpha )Na/sup 24/ reaction, has been found in the coolant of reactors using aluminum-clad fuel elements. Methods of calculating the activity due to scdium-24 are outlined. This activity was calculated to be normally 1.4 x 10/sup 5/ to 6 x 10/sup 5/ disintegrations per min per ml. The Na/sup 24/ isotope disintegrates with 2.76- snd 1.38-Mev gammas. (B. O. G.)
Date: June 1, 1960
Creator: McLain, H A
Partner: UNT Libraries Government Documents Department

An Improved Gamma Detector Using Gamma Moderation

Description: Theoretical and experimental results of the effect of gamma moderation are described. A 6.7-fold enhancement factor was obtained with a 1/8-in-thick sodiumiodide crystal when using gamma moderation and a Co/sup 60/ gamma source. A counter was designed and built to make the most use of gamma moderation The results indicate that this counter can be made energy independent over a wide region Though the cost of the counter is about 1/3 that of a 1-in-diam,, 1-in- thick sodiumiodide crystal plus photomultiplier, the counting efficiency for gammas such as Na/sup 24/ is over two times greater. (auth)
Date: September 1, 1959
Creator: Fox, R.
Partner: UNT Libraries Government Documents Department

A Tritiated-Water Detector with U-232/Th-228 Source

Description: The detection capabilities of the new U-232/Th-228 source are comparable to those of the Na-24 source. The main benefit in using the new source is the ease of operation. Elimination of the neutron activation step required for Na-24 sources saves about 24 hours in planning, scheduling, and executing. With the new U-232/Th-228 source, the monitor can be put in operation in less than 15 minutes. The long half-life of the U-232/Th-228 source also eliminates the need to record calibration and measurement times, as required for decay corrections when using a Na-24 source.
Date: May 29, 2001
Creator: Baumann, N.P.
Partner: UNT Libraries Government Documents Department

{sup 208}Tl and {sup 24}Na Gamma Sources for Tracing soil Water Movement with Deuterium

Description: The heavy water soil water tracer technique is based on the reaction of high-energy gamma photons with deuterium to indicate the tracer's location. The high-energy gamma sources {sup 208}Tl and {sup 24}Na were compared in their effectiveness to produce neutrons from heavy water. In tests with vials of heavy water at various locations in 208-liter (55-gallon) drums of dry and moist soil, {sup 24}Na gave 23 percent higher neutron count rates than {sup 208}Tl, but this was more than offset by its shorter effective half-life (15 hours vs. 1.9 years).
Date: August 29, 2001
Creator: Hawkins, R.H.
Partner: UNT Libraries Government Documents Department

RADIOACTIVITY OF NUCLEAR REACTOR COOLING FLUIDS

Description: Methods were developed for analysis of cooling water for impurities, radioisotopes, etc., and experimental results are presented for the ORNL Research Reactor. The theory of nuclear reactions in a water-cooled reactor is discussed at length, and equations were developed which allows predictions of equilibrium conditions from nonequilibrium measurements. The equations were verified experimentally by work on the ORNL Research Reactor and can be extended to other reactors. The origins of Na/sup 24/, Cd, and fission product activities are discussed, and the possibility of fuel element rupture detection by delayed neutron measurements is considered. (D.L.C.)
Date: October 1, 1961
Creator: Ward, J.C.
Partner: UNT Libraries Government Documents Department

SURVEY OF THE RADIATION LEVELS IN THE CONTAINMENT VESSEL OF THE ENRICO FERMI ATOMIC POWER PLANT. PART V. GAMMA RADIATION LEVELS ON THE OPERATING FLOOR OF THE CONTAINMENT BUILDING. a. LEVELS ABOVE THE EQUIPMENT COMPARTMENT. Technical Memorandum No. 16

Description: The results are presented of a survey of calculated gamma-ray levels at many points on the surface of the operating floor of the containment building for the Enrico Fermi reactor. That portion of the floor surveyed lies directly above the equipment compartment. The calculations were made with the aid of an IBM-650 electronic computer. The main source of radioactivity which gives rise to gamma radiation above the floor is the radioactive sodium-24 in the primary coolant system. This system was considered to be completely filled with sodium, and activated to an equilibrium activity of 0.05 curies/cc, which corresponds to infinite reactor operation at 500 megawatts power. No fission product contamination was considered for these calculations. The operating floor is 5 feet thick and of concrete and steel. The results of the survey indicate that above the equipment compartment the surface dose on the operating floor will in no case exceed 0.9 mr/hr at the expected full operating power of 430 megawatts. Included as appendices are derivations and methods of corrections from one set of concrete and steel thicknesses to another. (auth)
Date: December 22, 1959
Creator: Chaltron, W.F. & Hungerford, H.E.
Partner: UNT Libraries Government Documents Department

Calibration of direct nuclear activation diagnostics for measuring intense proton, lithium, and fluorine beams

Description: In the light-ion-beam fusion program at Sandia, an intense Li beam is being developed to drive inertial confinement fusion targets. Since beam purity is an important issue, direct nuclear activation diagnostics based on thick target yields of the reactions {sup 7}Li(p, n){sup 7}Be, {sup 10}B(p,{alpha}){sup 7}Be, {sup 19}F({sup 7}Li, d){sup 24}Na, and {sup 11}B({sup 19}F,2p){sup 28}Mg have been developed to measure the proton, Li, and F content of the beam. Target materials used were LiF and BN. To calibrate these diagnostics, a Van de Graaff accelerator was used to measure the thick target yields as a function of ion energy for each of the ion and target material combinations. Each target material was also irradiated by carbon ions to assess the importance of any possible competing reactions. Results of these calibration studies are presented in this paper.
Date: July 1, 1996
Creator: Ruiz, C.L.; Cooper, G.W.; Chambers, G. & Schmidlapp, F.A.
Partner: UNT Libraries Government Documents Department

Neutron Yield Measurements via Aluminum Activation

Description: Neutron activation of aluminum may occur by several neutron capture reactions. Four such reactions are described here: {sup 27}Al + n = {sup 28}Al, {sup 27}Al(n,{alpha}){sup 24}Na, {sup 27}Al(n, 2n){sup 26}Al and {sup 27}Al(n,p){sup 27}Mg. The radioactive nuclei {sup 28}Al, {sup 24}Na, and {sup 27}Mg, which are produced via the {sup 27}Al + n = {sup 28}Al, {sup 27}Al(n,{alpha}){sup 24}Na and {sup 27}Al(n,p){sup 27}Mg neutron reactions, beta decay to excited states of {sup 28}Si, {sup 24}Mg and {sup 27}Al respectively. These excited states then emit gamma rays as the nuclei de-excite to their respective ground states.
Date: December 8, 1999
Partner: UNT Libraries Government Documents Department

[Calibration methods for neutron diagnostics at Omega]. DOE/NLUF final report

Description: The investigation of Calibration Methods for Neutron Diagnostics at Omega is still underway. The data shown here is a compilation of measurements taken at Omega during the time of the grant. The data set has been updated with additional information taken this year. The neutron yield from an ICF event was determined by measuring the activity of an aluminum sample activated by target-produced DT neutrons. The radioactive nuclei {sup 24}Na and {sup 27}Mg, which were produced via the {sup 27}Al(n,{alpha}){sup 24}Na and {sup 27}Al(n,p){sup 27}Mg direct reactions, beta decay to excited states of {sup 24}Mg and {sup 27}Al respectively. These excited states then emitted gamma rays as the nuclei de-excite to their respective ground states. The gamma rays are detected and counted. From their numbers the neutron yield is determined.
Date: January 15, 1998
Creator: Padalino, S.
Partner: UNT Libraries Government Documents Department

Dosimetry investigation of the recuplex accident

Description: At 10:59 AM (PST), Saturday, April 7, 1962 a criticality accident occurred in a plutonium waste chemical recovery facility at the Hanford Atomic Products Operation, operated for the Atomic Energy Commission by the General Electric Company. Four men were hospitalized but were released after medical observation and after estimates of the radiation doses received were available. This report describes the dosimetry investigation that was made following the accident. This investigation was facilitated by the fact that all employees affected had personnel dosimeters in their possession when the incident occurred. The interpretation of the data supplied by these dosimeters was supplemented by information gathered by techniques that were developed in connection with other accidents. Below, the available information is first presented and then applied in a discussion of the dosimetry of the people involved in the accident.
Date: August 22, 1962
Creator: Roesch, W.C.; Gamertsfelder, C.C.; Larson, H.V.; Watson, E.C. & Nielsen, J.M.
Partner: UNT Libraries Government Documents Department

RADIOLOGICAL HAZARDS FROM RUPTURE OF THE SECONDARY COOLANT SYSTEM OF THE 10 Mw ESCR

Description: The hazards study was made to determine the radiation level from the secondary sodium lines of the l0-Mw ESCR, and to evaluate the corresponding radioactive concentration in the secondary loop and the maximum permissible effective activation flux in the intermediate heat exchanger. The results are presented graphicallyn the radiation level from the coolant lines during normal operation as a function of line diameter and Na/sup 4/ concentration; the effective thermal neutron activation flux level in the intermediate heat exchanger as a function of the Na/sup 4/ concentration in the secondary loop and the ratio of the time spent in the flux field to the time spent in making one cycle; average Na concentration in the reactor room atmosphere for operative and inoperative ventilation systems, assuming that all the sodium in the secondary system is released to the room and burns; and the downwind concentration of Na/ sup 24/ resulting from the release of all the secondary sodium on an open pad area exterior to the reactor building. An analysis of the results shows that in the event of an accident the toxicological hazards are more severe than the radiological hazards. Recommendations are given for minimizing the toxological hazards. (B.O.G.)
Date: January 19, 1960
Creator: Piccot, A.R.
Partner: UNT Libraries Government Documents Department

SPALLATION OF ALUMINUM BY 28-Gev PROTONS

Description: In the bombardment of Al foils with 28-Bev protons the following cross sections (in mb) for production of the indicated nuclides were obtained: Be/sup 7/ , 7.9 plus or minus 0.5; F/sup 18/, 6.0 plus or minus 0.3; C/sup 11/, 4.7 plus or minus 0.2; Na/sup 22/, 9.8 plus or minus 0.6. N/sup 13/, 1.18 plus or minus 0.07; Na/su p 24/, 8.3 plus or minus 0.5; O/sup 15/, 3.6 plus or minus 1.0, Mg/sup 27/, 0.067 plus or minus 0.006. An upper limit for the Ne/ sup 24/ eross section of 0.6 mb was obtained. The Mg/sup 27/ cross section has been corrected for secondary reactions and represents only production by primary protons, presumably by the (p,p pi /sup +/) reaction. The measurements were made relative to the production of C/sup 11/ activity in polyethylene and polystyrene foils. The production of Be/sup 7/ in the plastic foils was also observed and corresponds to a cross section of 7.7 plus or minus 0.4 mb per carbon nucleus. By comparison of these cross sections with those at lower bombarding energies, no significant trend of the excitation functions is observable. (auth)
Date: August 1, 1962
Creator: Cumming, J.B.; Friedlander, G.; Hudis, J. & Poskanzer, A.M.
Partner: UNT Libraries Government Documents Department

Analysis of Radiation From Hnpf Cold Traps and Primary Sodium Pumps During Removal and Shipping

Description: The expected maximum contamination of the HNPF cold traps and primary sodium pumps was determined along with the maximum dose rates from these components during removal and shipping. Suitable shielding for casks to be used in the removal operation and for shipping these components away from the reactor site is specified. Access to an unshielded cold trap is limited by high dose rates, i.e., 100 mr/hr at 120 ft, after 180 days decay time. A handling cask providing a radial shield of 3 in. of lead will provide adequate personnel protection for the removal operation, if 180 days decay time is allowed before the trap is removed. An additional 2.4 in. of lead is required for offsite shipment of the cask. This additional shielding can be added after the trap is removed from the reactor building. Dose rates from the cold trap after the shield plug is removed from the access hole are shown. If direct line-ofsight exposure is avoided, dose rates to personnel will be below 100 mr/hr at any position, and below 10 mr/hr at distances greater than 20 ft from the access hole. Dose rates from the cask during its travel away from the hole, will be below 100 mr/hr at distances from the cask greater than 10 ft and below 10 mr/hr at 35 ft, if the cask is raised no more than 3 in. from the floor during its travel. Remote, unshielded handling of a primary sodium pump is feasible, since dose rates would be 100 mr/hr at 28 ft and 10 mr/hr at 90 ft, after ten years of operation, and providing that 14 days decay time is allowed to eliminate activity from the Na/sup 24/ film clinging to the pump. Dose rates after only one year of operation would be lower by a ...
Date: December 15, 1959
Creator: Rhoades, W. A.
Partner: UNT Libraries Government Documents Department

Residual Radiation of the LRL 184-Inch Cyclotron

Description: Residual radioactivity at the Lawrence Radiation Laboratory 184-Inch Cyclotron was measured during November 1960. The study was conducted along three principal lines: (1) general survey of radiation levels in the cyclotron vault, (2) activation of foils placed near the cyclotron, and (3) gamma-ray spectra of the cyclotron gap region, including dee structure. Initial radiation levels were less than 8 r/hr which dropped to abcut 10 mr/hr after 48 hr. The observed activities induced in copper foils were Cu/sup 64/ and Co/sup 58/; in iron foils, Mn/sup 52/, Mn/sup 54/, and Mn/sup 56/; in aluminum foils Na/sup / 2>s/sup 4/ The gamma-ray spectra from the gap region included two intense long-lived peaks, at 510 and 810 kev, caused principally by Co/sup 58/. (auth)
Date: July 12, 1961
Creator: Boom, R. W.; Toth, K. S. & Zucker, A.
Partner: UNT Libraries Government Documents Department

Determination of neutron dose from criticality accidents with bioassays for sodium-24 in blood and phosphorus-32 in hair

Description: A comprehensive review of accident neutron dosimetry using blood and hair analysis was performed and is summarized in this report. Experiments and calculations were conducted at Oak Ridge National Laboratory (ORNL) and the University of Tennessee (UT) to develop measurement techniques for the activity of {sup 24}Na in blood and {sup 32}P in hair for nuclear accident dosimetry. An operating procedure was established for the measurement of {sup 24}Na in blood using an HPGe detector system. The sensitivity of the measurement for a 20-mL sample is 0.01-0.02 Gy of total neutron dose for hard spectra and below 0.005 Gy for soft spectra based on a 30- to 60-min counting time. The operating procedures for direct counting of hair samples are established using a liquid scintillation detector. Approximately 0.06-0.1 Gy of total neutron dose can be measured from a 1-g hair sample using this procedure. Detailed procedures for chemical dissolution and ashing of hair samples are also developed. A method is proposed to use blood and hair analysis for assessing neutron dose based on a collection of 98 neutron spectra. Ninety-eight blood activity-to-dose conversion factors were calculated. The calculated results for an uncollided fission spectrum compare favorably with previously published data for fission neutrons. This nuclear accident dosimetry system makes it possible to estimate an individual`s neutron dose within a few hours after an accident if the accident spectrum can be approximated from one of 98 tabulated neutron spectrum descriptions. If the information on accident and spectrum description is not available, the activity ratio of {sup 32}P in hair and {sup 24}Na in blood can provide information related to the neutron spectrum for dose assessment.
Date: June 1, 1993
Creator: Feng, Y.; Miller, L. F.; Brown, K. S.; Casson, W. H.; Mei, G. T. & Thein, M.
Partner: UNT Libraries Government Documents Department