Showing papers in "Nuclear Technology in 1978"
TL;DR: In this paper, the theory of swelling is reviewed in terms of basic concepts and simulation and impurity effects, and the basic theory employs the formalism of chemical reaction rates, e.g., voids, dislocations, etc.
Abstract: The theory of swelling is reviewed in terms of basic concepts and simulation and impurity effects. The basic theory employs the formalism of chemical reaction rates. Efficiencies of voids, dislocat...
272 citations
TL;DR: In this article, the thermal-hydraulics of a Boiling Water Nuclear Reactor were investigated and shown to have thermal properties similar to those of a conventional nuclear Reactor.
Abstract: (1978). The Thermal-Hydraulics of a Boiling Water Nuclear Reactor. Nuclear Technology: Vol. 39, No. 3, pp. 330-330.
95 citations
TL;DR: In this paper, the electrochemical evolution of hydrogen from a molten solution of LiF--LiCl--LiBr containing small quantities of LiH and saturated with metallic lithium was demonstrated, and the evolved hydrogen was recovered from the melt by sweeping the hydrogen electrode with a circulating stream of argon and subsequently trapping the hydrogen from the argon with a hot titanium getter bed.
Abstract: The electrochemical evolution of hydrogen from a molten solution of LiF--LiCl--LiBr containing small quantities of LiH and saturated with metallic lithium was demonstrated. The evolved hydrogen is recovered from the melt by sweeping the hydrogen electrode with a circulating stream of argon and subsequently trapping the hydrogen from the argon with a hot titanium getter bed. It is found that by continually gettering the argon, 100 +- 2% of the hydride present in the molten salt, at a concentration of 1 wppM, is recoverable. Results of metallographic examinations of stainless-steel components in contact with the salt solution during the experiments indicate some evidence of surface attack (10 to 15 ..mu..m) and intergranular penetration (30 to 50 ..mu..m), but extrapolated corrosion rates are generally small (approx. 0.2 mm/yr). These results incorporated into a reevaluation of the molten-salt extraction process, as it applied to tritium recovery from a liquid-lithium fusion reactor blanket, indicate a more favorable processing capability than was previously expected.
46 citations
TL;DR: The presently practised fractionated dose schedule for cancer treatment usually extends over a long period causing hardships and economic difficulties to patients in developing countries who often have to travel long distances for their treatment.
Abstract: (1978). Modification of Radiosensitivity of Biological Systems. Nuclear Technology: Vol. 37, No. 1, pp. 85-86.
41 citations
TL;DR: In this article, equilibrium thermodynamic calculations were performed on the Cs-U-Zr-H-I-O system that is assumed to exist in the fuel-cladding gap of light water reactor fuel under in-reactor, steam, and 50% steam--50% air conditions.
Abstract: Equilibrium thermodynamic calculations were performed on the Cs-U-Zr-H-I-O system that is assumed to exist in the fuel-cladding gap of light water reactor fuel under in-reactor, steam, and 50% steam--50% air conditions. The in-reactor oxygen potential is assumed to be controlled by either UO/sub 2+x/ + Cs/sub 2/UO/sub 4/ or Zr + ZrO/sub 2/. The important condensed phases in-reactor are UO/sub 2+x/, Cs/sub 2/UO/sub 4/, and CsI, and the major gaseous species are Cs, Cs/sub 2/, CsI, and Cs/sub 2/I/sub 2/. The presence of steam does not alter these species, although CsOH also becomes a major gaseous species. In a 50% steam--50% air mixture, the equilibrium condensed phases are U/sub 3/O/sub 8/ or UO/sub 3/ and Cs/sub 2/U/sub 15/O/sub 46/. Under a nonequilibrium situation where zirconium metal can react with iodine, ZrO/sub 3/ or liquid ZrI/sub 2/ is present, and the gaseous species ZrI/sub 3/ and ZrI/sub 4/ have large partial pressures.
37 citations
TL;DR: A control bank redesign and a modified control strategy have been developed to improve load follow capabilities in commercial pressurized water reactors (PWR) through the use of controlled moderator temperature reductions.
Abstract: A control bank redesign and a modified control strategy have been developed to improve load follow capabilities in commercial pressurized water reactors (PWRs). The use of controlled moderator temperature reductions adds to both load follow and return-to-power capability through an inherently negative moderator temperature coefficient. Using the new control strategy, a PWR can typically return from part power to at least a 90% power level at a 5%/min rate during daily load cycles, and also follow daily load cycles out to 98% of core life.
34 citations
TL;DR: In this paper, three general rules are developed by which long-term induced radioactivity can be minimized, and applied to two specific alloy systems, namely, stainless steels and TZM.
Abstract: Three general rules are developed by which long-term induced radioactivity can be minimized. These rules are then applied to two specific alloy systems--stainless steels and the molybdenum alloy, TZM. A particular steel, Tenelon, containing neither nickel nor molybdenum, is especially attractive. It is found that the principles of both elemental substitution and isotopic tailoring can reduce the long-term radioactivity levels by orders of magnitude compared to normal Type 316 stainless steel. A comparison of long-term activity levels in such systems as the liquid-metal fast breeder reactor, fusion with standard structural alloys, and fusion with steel alloys designed for low activity quantitatively shows the potential advantage of fusion in this area. The influence of isotopic tailoring on gas production rates is also discussed. The calculations on radioactivity indicate that with proper attention to the choice of materials and isotopes, long-term radioactivity in fusion devices can be made so low as to either eliminate concern over long-term storage or allow recycling within a few human generations.
30 citations
TL;DR: In this paper, a mathematical model for ex-core detector response in PWR configurations is presented, based on the use of a spatial weighting function that is independent of core power.
Abstract: A mathematical model for ex-core detector response in pressurized water reactor (PWR) configurations is presented, based on the use of a spatial weighting function that is independent of core power...
30 citations
TL;DR: The gas circulation in a gas centrifuge due to temperature differences, differential rotation and injection, and removal of fluid at the ends, as well as temperature gradients at the cylinde...
Abstract: The gas circulation in a gas centrifuge due to temperature differences, differential rotation and injection, and removal of fluid at the ends, as well as due to temperature gradients at the cylinde...
29 citations
27 citations
TL;DR: In this article, the activation integrals with the differential neutron spectrum provided an assessment of the nuclear data, and the fission rates to fission product activations in the foil materials was also used to measure cumulative fission yields for /sup 235/U and /sup 238/U.
Abstract: Integral tests have been made for 27 nuclear activation cross sections required for neutron dosimetry at accelerator-based irradiation facilities. Thirteen materials were irradiated in a well-defined geometry using neutrons produced by stopping 14- to 16-MeV deuterons in a thick beryllium target. The neutron spectrum was measured by time-of-flight spectrometry. Comparison of the activation integrals with the differential neutron spectrum provided an assessment of the nuclear data. A small fission chamber was used to monitor the irradiation, map flux gradients, and study attenuation and scattering effects. Comparison of the fission rates to fission product activations in the foil materials was also used to measure cumulative fission yields for /sup 235/U and /sup 238/U.
TL;DR: To prevent diversion of nuclear material from power production to weapon production either by a nation or by clandestine groups within a nation, the nuclear fuel cycle must be proliferation-resourced.
Abstract: To prevent the diversion of nuclear material from power production to weapon production either by a nation or by clandestine groups within a nation, the nuclear fuel cycle must be proliferation-res...
TL;DR: A feasibility study has demonstrated that very strong, durable, relatively impermeable cylinders may be prepared by hot pressing combinations of cements with simulated radioactive waste solids as mentioned in this paper, which is a technique that is similar to the one described in this paper.
Abstract: A feasibility study has demonstrated that very strong, durable, relatively impermeable cylinders may be prepared by hot pressing combinations of cements with simulated radioactive waste solids. Whi...
TL;DR: In this article, chemical getters can be used to scavenge tritium from inert gases, which has a high reactivity and is relatively inexpensive, and it is a good...
Abstract: Chemical getters can be used to scavenge tritium from inert gases. Cerium, which forms a hydride with a low dissociation pressure, has a high reactivity and is relatively inexpensive. It is a good ...
TL;DR: In this paper, the migration of Uranium and Thorium was discussed and discussed in terms of their exploration significance and their importance in the history of nuclear technology. But this work was limited.
Abstract: (1978). Migration of Uranium and Thorium—Exploration Significance. Nuclear Technology: Vol. 38, No. 3, pp. 449-449.
TL;DR: In this article, the authors used uranyl nitrate solution with the uranium enriched to 93.17% /sup 235/U. The concentration ranges from 54.89 to 369.96 g U/l. Both material and geometry properties are so well determined as to reduce greatly any contribution to a theoretical/experimental discrepancy attributable to the experiment.
Abstract: Seventy-six benchmark critical conditions are reported. Both material and geometry properties are so well determined as to reduce greatly any contribution to a theoretical/experimental discrepancy attributable to the experiment. The program uses uranyl nitrate solution with the uranium enriched to 93.17% /sup 235/U. The concentration ranges from 54.89 to 369.96 g U/l. Unreflected experiments are reported, as well as measurements within thick-walled cubical reflector shells composed of such common materials as concrete and plastic. For experiments using a single tank, the diameter of the tank ranged from 27.88 to 50.69 cm, and arrays of up to 16 cylinders have containers of two diameters: 16.12 and 21.12 cm. Containers composed of aluminum or stainless steel are studied. For all these parameters, the critical heights range from 17.13 to 110.20 cm.
TL;DR: The results obtained from approximately 4000 h of operation of an electrochemical oxygen meter developed at General Electric Company for measuring oxygen activities in liquid sodium are described in this article, where several meters were tested in a flowing sodium system, where the oxygen concentrations were controlled by a cold trap.
Abstract: The results obtained from approximately 4000 h of operation of an electrochemical oxygen meter developed at General Electric Company for measuring oxygen activities in liquid sodium are described. In the present design, a small ceramic cup (ThO/sub 2/ with 7 to 8 wt % Y/sub 2/O/sub 3/) has been brazed to a metal tube that is inserted into the liquid sodium via a standard sodium flange. The reference electrode used in the present design consists of In/In/sub 2/O/sub 3/. Several meters were tested in a flowing sodium system, where the oxygen concentrations were controlled by a cold trap. At 500/sup 0/C, the oxygen electromotive force (emf) output can be expressed as emf/sub (V)/ = 0.269 + (160/T) (+-0.003 V), where T is the cold-trap temperature in K. The temperature coefficient of the meters can be expressed as (dE/dT) = 0.51 +- (0.02) mV//sup 0/C.
TL;DR: The role foreseen for the SECURE (Safe Environmentally Clean Urban REactor) nuclear district heating plant is to provide the baseload heating needs of primarily the larger and medium size urban centers that are outside the range of waste heat supply from conventional nuclear power stations.
Abstract: The role foreseen for the SECURE (Safe Environmentally Clean Urban REactor) nuclear district heating plant is to provide the baseload heating needs of primarily the larger and medium size urban centers that are outside the range of waste heat supply from conventional nuclear power stations. The rationale of the SECURE concept is that the simplicity in design and the inherent safety advantages due to the use of low temperatures and pressures should make such reactors economically feasible in much smaller unit sizes than nuclear power reactors and should make their urban location possible. It is felt that the present design should be safe enough to make urban underground location possible without restriction according to any criteria based on actual risk evaluation. From the environmental point of view, this is a municipal heat supply plant with negligible pollution. Waste heat is negligible, gaseous radioactivity release is negligible, and there is no liquid radwaste release. Economic comparisons show that the SECURE plant is competitive with current fossil-fueled alternatives. Expected future increase in energy raw material prices will lead to additional energy cost advantages to the SECURE plant.
TL;DR: The trapping or retention efficiency of impregnated charcoal beds for the vapor of methylradioiodide tagged with 131I was determined under various flow conditions by radioactive counting of each of the columns as discussed by the authors.
Abstract: The trapping or retention efficiency of impregnated charcoal beds for the vapor of methylradioiodide tagged with 131I was determined under various flow conditions by radioactive counting of each of
TL;DR: In this paper, the technical feasibility of a coprocessing concept involving recovery of all the actinides in the spent fuel as a product group has been analyzed and it has been shown that this can be accomplished b
Abstract: The technical feasibility of a coprocessing concept involving recovery of all the actinides in the spent fuel as a product group has been analyzed It has been shown that this can be accomplished b
TL;DR: This article found a significant correlation between the increased concentrations of 210Pb in glacier ice and periods of nuclear tests in both hemispheres and found that nuclear tests were correlated with increased concentrations.
Abstract: In widely separated locations in both hemispheres, we have found a significant correlation between the increased concentrations of 210Pb in glacier ice and periods of nuclear tests. The concentrati...
TL;DR: The physical and nuclear properties of EU2O3 were reviewed, evaluated, and compared with those of the other potential control materials for breeder reactors as discussed by the authors, with Europia with a 90% theoretical density.
Abstract: The physical and nuclear properties of EU2O3 were reviewed, evaluated, and compared with those of the other potential control materials for breeder reactors. Europia with a 90% theoretical density ...
TL;DR: In this article, a computer model based on available materials property data has been developed to predict the lifetimes of first wall structural materials under a variety of reactor conditions, including temperature and integrated neutron wall loading.
Abstract: A computer model based on available materials property data has been developed to predict the lifetimes of first wall structural materials under a variety of reactor conditions. The model combines the materials property data with the appropriate ranges of limiting criteria to establish design lifetimes as functions of such relevant parameters as temperature and integrated neutron wall loading. Empirical equations developed from existing literature data were used to interpolate and extrapolate the required materials properties over the desired ranges. The present effort has concentrated on the evaluation of two candidate structural materials, namely, Type 316 stainless steel and a vanadium-base alloy (V--15% Cr--5% Ti). Curves have been derived that show the estimated lifetime and life-limiting property as a function of temperature for a specified set of design criteria, e.g., maximum swelling of 4%, minimum uniform elongation of 1%, and total creep strain of <1%, for an applied stress of 103 MPa (15 ksi). The results obtained indicate a much longer design lifetime for the vanadium-base alloy than for stainless steel under the conditions of interest. The computational model has been incorporated into the Tokamak Power Plant Systems Program at Argonne National Laboratory.
TL;DR: One-dimensional neutronic and photonic calculations have been carried out using the discrete-ordinates code ANISN to compare the nuclear performance of blanket and shield designs proposed for use in a nuclear power station as mentioned in this paper.
Abstract: One-dimensional neutronic and photonic calculations have been carried out using the discrete-ordinates code ANISN to compare the nuclear performance of blanket and shield designs proposed for use i...
TL;DR: In this article, a new seawater desalination method, Nord-Aqua Vacuum Evaporation, which utilizes waste heat at a very low temperature, has been developed, where the requisite vacuum is obtained by a barometric column and siphon, and the dissolved air is removed from the vacuum by water flows.
Abstract: A new seawater desalination method, Nord-Aqua Vacuum Evaporation, which utilizes waste heat at a very low temperature, has been developed. The requisite vacuum is obtained by a barometric column and siphon, and the dissolved air is removed from the vacuum by water flows. According to test results from a pilot plant, the process is operable if the waste heat exists at a temperature 7/sup 0/K higher than ambient. The pumping energy that is then required is 38 kJ/kg, or 1.5 percent of the heat of vaporization of water. Calculations reveal that the method is considerably superior economically to conventional distilling methods.
TL;DR: In this paper, the maximum concentrations and discharge rates of radioactive nuclide chains migrating through an adsorbing medium were derived in terms of non-dimensional parameters, and conditions were established that quantify the reconcentration effort of radionuclide chain migration.
Abstract: Expressions are developed for the maximum concentrations and discharge rates of radioactive nuclide chains migrating through an adsorbing medium. These expressions are presented in terms of nondimensional parameters. Conditions are established that quantify the reconcentration effort of radionuclide chain migration in terms of the nondimensional parameters. This approach provides a relatively simple mechanism for determining an upper bound to the concentrations or release rates of parent and daughter radionuclides in groundwater.
TL;DR: In this article, joint Monte Carlo calculations have been carried out using the three-dimensional radiation transport code, MORSE, to estimate the nuclear heating and radiation damage in the toroidal field (TF) coils adjacent to a 28 x 68 cm/sup 2/ rectangular neutral beam injector duct that passes through the blanket and shield of a D-T burning Tokamak reactor.
Abstract: Adjoint Monte Carlo calculations have been carried out using the three-dimensional radiation transport code, MORSE, to estimate the nuclear heating and radiation damage in the toroidal field (TF) coils adjacent to a 28 x 68 cm/sup 2/ rectangular neutral beam injector duct that passes through the blanket and shield of a D-T burning Tokamak reactor. The plasma region, blanket, shield, and TF coils were represented in cylindrical geometry using the same dimensions and compositions as those of the Experimental Power Reactor. The radiation transport was accomplished using coupled 35-group neutron, 21-group gamma-ray cross sections obtained by collapsing the DLC-37 cross-section library. Nuclear heating and radiation damage rates were estimated using the latest available nuclear response functions. The presence of the neutral beam injector duct leads to increases in the nuclear heating rates in the TF coils ranging from a factor of 3 to a factor of 196 depending on the location. Increases in the radiation damage also result in the TF coils. The atomic displacement rates increase from factors of 2 to 138 and the hydrogen and helium gas production rates increase from factors of 11 to 7600 and from 15 to 9700, respectively.
TL;DR: In this paper, an analysis of primary pipe rupture accidents in the Clinch River Breeder Reactor was carried out with Brookhaven National Laboratory modified versions of the DEMO code and the thermal transient responses of the core and radial blanket were calculated for a large number of initial conditions and plant configurations.
Abstract: Analyses of primary pipe rupture accidents in the Clinch River Breeder Reactor were carried out with Brookhaven National Laboratory-modified versions of the DEMO code. The thermal transient responses of the core and radial blanket were calculated for a large number of initial conditions and plant configurations. These include studies of variations of pipe break size and operating power. Pipe ruptures commencing from two-loop initial operating conditions are presented. The sensitivity of the thermal response to variations in particular parameters within the estimated ranges of their uncertainties has been studied. Conditions under which sodium boiling in the core occurs are delineated.