Showing papers in "Nuclear Technology in 1976"
TL;DR: In this paper, nuclear tracks in solids (Principles and Applications) nuclear technology: Vol. 30, No. 1, pp. 91-92, were discussed and discussed in detail.
Abstract: (1976). Nuclear Tracks in Solids (Principles and Applications) Nuclear Technology: Vol. 30, No. 1, pp. 91-92.
973 citations
TL;DR: In this paper, a range of property values that are useful for evaluating range of applicability of low-atom ceramics materials is presented, including carbon, SiC, Be2C, B4C, TiC, Al2O3, and BeO.
Abstract: The (low-atomic-number ceramic) materials carbon, SiC, Be2C, B4C, TiC, BN, Si3N4, Al2O3, and BeO provide a range of property values that are useful for evaluating range of applicability of low-atom...
160 citations
TL;DR: In this paper, a mathematical model for predicting the dynamic response of the H. B. Robinson pressurized water reactor plant was formulated and compared with results from measurements made during full-power operation of the plant.
Abstract: A mathematical model for predicting the dynamic response of the H. B. Robinson pressurized water reactor plant was formulated and compared with results from measurements made during full-power operation of the plant. The model was based on the basic conservation laws for neutrons, mass, and energy; design data from the safety analysis report were used to evaluate the necessary coefficients. The model included representations for point kinetics, core heat transfer, piping, pressurizer, and the steam generator. The experiment involved perturbations in control rod position and main steam valve opening. Periodic binary input signals and step inputs were used. Theoretical and experimental frequency responses were obtained from the model and the test data. The comparison showed that the model was capable of good predictions for reactivity perturbations and fair predictions for steam valve perturbations. A method was also demonstrated for using the test data for at-power determination of the differential...
82 citations
TL;DR: In this article, a load-follow demonstration was conducted on Consolidated Edison's Indian Point, Unit 2 plant located in Buchanan, New York, in August and September of 1974.
Abstract: Comprehensive load-follow demonstrations were conducted on Consolidated Edison’s Indian Point, Unit 2 plant located in Buchanan, New York, in August and September of 1974 The purpose was to examin
60 citations
TL;DR: In this paper, nuclear cross sections for displacements and post-short-term cascade annealing defects are derived from nuclear kinematics calculations of primary atomic recoil energy distributions and the...
Abstract: Neutron cross sections for displacements and post-short-term cascade annealing defects are derived from nuclear kinematics calculations of primary atomic recoil energy distributions and the...
41 citations
TL;DR: The TRIGA fuel was developed around the concept of inherent safety as discussed by the authors, and a core composition was sought that had a large prompt negative temperature coefficient of reactivity such that if all the available excess reactivity were suddenly inserted into the core, the resulting fuel temperature would automatically cause the power excursion to terminate before any core damage resulted.
Abstract: TRIGA fuel was developed around the concept of inherent safety. A core composition was sought that had a large prompt negative temperature coefficient of reactivity such that if all the available excess reactivity were suddenly inserted into the core, the resulting fuel temperature would automatically cause the power excursion to terminate before any core damage resulted. Experiments have demonstrated that zirconium hydride possesses a basic neutron-spectrum- hardening mechanism to produce the desired characteristic. Additional advantages include the facts that ZrH has a good heat capacity, that it results in relatively small core sizes and high flux values due to the high hydrogen content, that it has excellent fission-product retentivity and high chemical inertness in water at temperatures up to 100$sup 0$C, and that it can be used effectively in a rugged fuel element size. Tens of thousands of routine pulses to the range of 500 to 800$sup 0$C peak fuel temperatures have been performed with TRIGA fuel, and a core was pulse-heated to peak fuel temperatures in excess of 1100$sup 0$C for hundreds of pulses before a few elements exceeded the conservative tolerances on dimensional change. (auth)
38 citations
TL;DR: The safety incentives for separating and eliminating various elements from high-level radioactive waste prior to final geologic isolation have been examined in this article, and the study required evaluation of numero...
Abstract: The safety incentives for separating and eliminating various elements from high-level radioactive waste prior to final geologic isolation have been examined. The study required evaluation of numero...
36 citations
TL;DR: In this paper, the authors surveyed the experience with the high-strength ferritic steels and prepared ingots of 26 selected compositions, including high-nickel Alloy 800, for the Clinch River Breeder Reactor Plant steam generators.
Abstract: Extensive knowledge and acceptability of pertinent properties, wide fabrication experience, and code acceptance have led to selection of 2$sup 1$/ $sub 4$ Cr--1 Mo steel for the Clinch River Breeder Reactor Plant steam generators. Limitations of this alloy indicate that further development of high- strength ferritic steels containing 9 to 12 percent chromium and the high-nickel Alloy 800 could lead to superior materials, and programs to develop these materials have started. Combustion Engineering has surveyed the experience with the high-strength ferritic steels and prepared ingots of 26 selected compositions. Charpy V-notch tests and metallography have been used to characterize these alloys, and optimum welding rod compositions for these alloys are under development. Westinghouse-Tampa is undertaking a program to gain code acceptance of Alloy 800. A program has been set up to provide the information required for design and fabrication of reliable components. Progress has been made on characterization, the role of tertiary creep in failure, and the development of welding processes. The Heppenstall Company is demonstrating its process for manufacturing large high-quality ingots. (auth)
33 citations
TL;DR: In the course of neutronically analyzing and comparing several fission blanket concepts, this work has demonstrated that fusion-fission hybrids can be designed to meet a broad spectrum of fissile- breeding and fusion-energy-multiplying requirements.
Abstract: Fusion-fission hybrid concepts are viewed as subcritical fission reactors driven and controlled by high-energy neutrons from a laser-induced fusion reactor. Blanket designs encompassing a substanti...
26 citations
TL;DR: The effects of titanium additions up to 0.6 wt percent on the irradiation-induced swelling and changes in creep-rupture properties were investigated in this paper, where samples were irradiated in the Experimental Breeder Reactor II at temperatures in the range from 450 to 700/sup 0/C to a maximum neutron fluence of 7.8 x 10/sup 26/ n/m/sup 2/ (>0.1 MeV).
Abstract: The effects of titanium additions up to 0.6 wt percent on the irradiation-induced swelling and changes in creep-rupture properties were investigated. Samples were irradiated in the Experimental Breeder Reactor II at temperatures in the range from 450 to 700/sup 0/C to a maximum neutron fluence of 7.8 x 10/sup 26/ n/m/sup 2/ (>0.1 MeV). In annealed material, the irradiation-induced swelling exhibited a minimum in the range 0.2 to 0.4 wt percent titanium. The minimum in swelling was directly attributable to a minimum in the concentration of voids. Samples irradiated in the 20 percent cold-worked condition exhibited slight densification at 3.0 x 10/sup 26/ n/m/sup 2/ (>0.1 MeV) at both 500 and 600/sup 0/C. A small density decrease (0.23 percent) occurred during irradiation to 6.6 x 10/sup 26/ n/m/sup 2/ (>0.1 MeV). Postirradiation creep-rupture ductility was a maximum for alloys containing 0.23 and 0.33 wt percent titanium. The observed swelling behavior in the annealed material is thought to be associated with changing amounts of titanium and carbon in solution in the austenite as the total titanium concentration is increased. The improved ductility is attributable to a decreased tendency for grain boundary crack formation and appears to be associated with removal ofmore » sulfur and possibly other impurities from solution in the austenite.« less
Journal Article•
TL;DR: The OPROD has facilitated prompt response to varying operating conditions and the investigation of a conflicting relationship between the thermal limitation and the cycle length.
Abstract: The OPROD computer code has been developed to generate a long-term control rod program, a series of control rod patterns that optimize a cycle length within various operational constraints. In the algorithm, the optimization problem is decomposed into two hierarchies. In the inner loop, a time-invariant target power distribution is assumed, and a control rod pattern is determined so as to best fit the power distribution to the target within the constraints at each burnup step. The target is then improved in the outer loop to achieve a longer cycle length. The code consists of two major parts: a three- dimensional boiling water reactor (BWR) core simulator and MAP, the method of approximate programming. It readily generates a long-term control rod program of BWRs without trial search by core-management engineers. The OPROD has therefore facilitated prompt response to varying operating conditions and the investigation of a conflicting relationship between the thermal limitation and the cycle length. (auth)
TL;DR: Aluminum was added to the niobium core, and in various quantities to the copper-tin bronze, of composite wires that were reacted to form Nb/sub 3/Sn layers as discussed by the authors.
Abstract: Aluminium was added to the niobium core, and in various quantities to the copper-tin bronze, of composite wires that were reacted to form Nb/sub 3/Sn small amounts of aluminium in the bronze enhance the growth rate of Nb/sub 3/Sn layers; aluminium in the core, and greater amounts in the bronze displacing some of the tin, cause a reduction in growth rate. Layer thickness is a function of (reaction time)/sup 0/./sup 67/. Microprobe analysis revealed the presence of aluminium in the reacted layers only for specimens with aluminium additions to the core and in substantial quantities to the matrix. Critical current densities are primarily a function of reacted layer thickness. Specimens in which some aluminium was successfully incorporated in thin (1- to 1.5-..mu..m) layers of Nb/sub 3/Sn showed maximum current densities around 10/sup 9/A/m/sup 2/ in transverse fields of 16T.
TL;DR: In this article, the changes in the critical current I/sub c/ of multifilament Nb/sub 3/Sn following several fission-reactor-neutron irradiations at approximately 60/sup 0/C have been investigated as a function of applied transverse magnetic field up to 160 kG.
Abstract: The changes in the critical current I/sub c/ of multifilament Nb/sub 3/Sn following several fission-reactor-neutron irradiations at approximately 60/sup 0/C have been investigated as a function of applied transverse magnetic field up to 160 kG. Increases in I/sub c/ below 10/sup 18/ n/cm/sup 2/ (E greater than 1 MeV) show a strong field dependence, relative changes being larger as the field increases. These increases are attributed to increases in H/sub c2/ brought about by irradiation-induced increases in the normal-state resistivity of the superconductor. For doses greater than 10/sup 18/ n/cm/sup 2/, sharp decreases in I/sub c/ are observed, but the behavior of I/sub c/ is qualitatively identical for all fields from 40 to 160 kG. Therefore, data obtained at the more easily attainable lower magnetic fields are directly applicable to the high-field regions in this high-fluence regime. However, for fluences below approximately 10/sup 18/ n/cm/sup 2/, magnetic-field-dependent measurements are required to determine the response of the superconductor to the neutron irradiation.
TL;DR: Vanadium-base alloys, including V--10 percent Cr, V--20 percent Ti and VANSTAR-7, have been irradiated in the Experimental Breeder Reactor II in the temperature range of 400 to 800/sup 0/C, mainly to a fluence of 1.5 x 10/sup 22/n/cm/sup 2/ (greater than 0.1 MeV) as mentioned in this paper.
Abstract: Vanadium-base alloys, V--10 percent Cr, V--20 percent Ti, and VANSTAR-7, alloys with potential for fusion reactor application, have been irradiated in the Experimental Breeder Reactor II in the temperature range of 400 to 800/sup 0/C, mainly to a fluence of 1.5 x 10/sup 22/ n/cm/sup 2/ (greater than 0.1 MeV). Swelling determined both from immersion density measurements and void distribution data obtained by transmission electron microscopy showed that the V--20 percent Ti was completely resistant to void formation for these irradiation temperatures and for the highest fluence achieved, 6 x 10/sup 22/ n/cm/sup 2/. Voids formed in both the V--10 percent Cr and VANSTAR-7 alloys, but only the V--10 percent Cr, irradiated at 690 and 805/sup 0/C, showed technologically significant swelling, near 1 percent. Swelling in this alloy at lower temperatures and in VANSTAR-7 at all temperatures was below 0.1 percent. Dislocation structures were complex in all three alloys. In the V--20 percent Ti, the scale of the dislocation network coarsened with increasing irradiation temperature. In the other two alloys, the scale of the damage, both dislocation and void components, was similar for irradiation at 496 and 580/sup 0/C, but coarsened considerably to produce similar structures for irradiations at 690more » and 805/sup 0/C. In many cases, detail of the microstructure was obscured by strongly diffracting zones that are believed to be impurity related. Of the three alloys examined, V--20 percent Ti possesses the greatest swelling resistance for the irradiation temperatures and fluences achieved and thus is judged to have the greatest potential for use in fusion reactors.« less
TL;DR: Metallic molybdenum, Mo-Ru-Rh-Pd alloys, barium, zirconium, and tungsten have been added to uranium and uranium-plutonium oxides by co-precipitation and mechanical mixture techniques as mentioned in this paper.
Abstract: Metallic molybdenum, Mo-Ru-Rh-Pd alloys, barium, zirconium, and tungsten have been added to uranium and uranium-plutonium oxides by co-precipitation and mechanical mixture techniques. This material...
TL;DR: In this paper, a series of neutron irradiation experiments were conducted on annealed and 20% cold-worked Type 316 stainless steel in a high-flux mixed-spectrum fission reactor to simulate a controlled thermostat.
Abstract: Results of a series of neutron irradiation experiments conducted on annealed and 20% cold-worked Type 316 stainless steel in a high-flux mixed-spectrum fission reactor to simulate a controlled ther...
TL;DR: In this article, the authors analyzed fracture growth behavior with respect to fracture mechanics criteria and showed that singleedge-notch cantilever and compact tension specimens gave comparable crack growth rates for a given stress intensity factor range.
Abstract: Fatigue crack propagation rates at 24, 427, and 593°C were not affected by specimen thickness within the range 76 to 254 mm Also, singleedge-notch cantilever and compact tension specimens gave comparable crack growth rates for a given stress intensity factor range Crack growth behavior was analyzed with respect to fracture mechanics criteria
TL;DR: In this paper, various options and trade-offs in the nuclear design of the blanket/shield for a Tokamak Experimental Power Reactor (TEPR) are investigated, and the nuclear performance of various material compositions is studied.
Abstract: The various options and trade-offs in the nuclear design of the blanket/ shield for a Tokamak Experimental Power Reactor (TEPR) are investigated. The TEPR size and cost are particularly sensitive to the blanket/shield thickness, $delta$/sub BS/, on the inner side of the torus. Radition damage to the components of the superconducting magnet and refrigeration power requirements set lower limits on $delta$/sub BS/. These limits are developed in terms of TEPR design parameters such as the wall loading, duty cycle, and frequency of magnet anneals. The study of the nuclear performance of various material compositions shows that mixtures of tungsten, or tantalum, or stainless-steel alloys and boron carbide require the smallest $delta$/sub BS/ for a given attenuation. This $delta$/sub BS/ has to be doubled if the low induced activation materials graphite and aluminum are used. The space problems are greatly eased in the Argonne National Laboratory ANL-TEPR reference design by using two separate segments of the blanket/shield. The inner segment occupies the region of the high magnetic field, uses very efficient attenuators (tungsten- or tantalum- or stainless-steel-boron carbide mixtures), and is only 1 m thick. The outer more » blanket/shield is 131 cm and consists of an optimized composition of stainless steel and boron carbide. For the design parameters of 0.2 MW/m$sup 2$ neutron wall loading and 50 percent duty cycle, the reactor components can operate satisfactorily up to (a) 10 yr for the stainless-steel first wall, (b) 10 yr for the superconductor composite after which magnet warmup becomes necessary, and (c) 30 yr for the Mylar insulation. Nuclear heat generation rates in the blanket/ shield and magnet are well within the practical limits for heat removal. « less
TL;DR: In this article, various sample sizes, reflector sizes and shapes, and reflector materials were examined to determine their effect on pulse-height and pulse-shape resolution in alpha liquid-scintillation spectroscopy.
Abstract: Various sample sizes, reflector sizes and shapes, and reflector materials were examined to determine their effect on pulse-height and pulse-shape resolution in alpha liquid-scintillation spectometry. A section of a metal sphere coated with a diffuse-white reflective material was found to have the best characteristics for both pulse-height and pulse-shape resolution. Although sample volumes as large as 10 ml could be tolerated when used with reflectors to accommodate them, the best results were obtained with 1-ml samples and smaller reflectors. Comparison of two types of pulse-shape discrimination circuitry for separating alpha and beta-gamma pulses indicated that a zero-crossover method was superior to a constant fraction timing method. The combination of these improved detectors with solvent extraction methods of incorporating the sample in the scintillator and pulse-shape discrimination allows alpha spectrometry with a background as low as 0.01 count/min and an energy resolution as high as 5.5%. (auth)
TL;DR: In this paper, the amount and chemical and radiochemical composition of fuel element crud deposits from a number of modern commercial pressurized water reactors were measured and the results from the measurements were reported.
Abstract: Measurements have been made of the amount and chemical and radiochemical composition of fuel element crud deposits from a number of modern commercial pressurized water reactors. Results from the Po...
TL;DR: In this paper, LiNiobate plates intended for application as sensors for acoustical monitoring of reactors were subjected to thermal-neutron irradiation near room temperature and showed that the crystal had become highly disordered, as shown by the back-reflection Laue picture.
Abstract: Lithium niobate plates intended for application as sensors for acoustical monitoring of reactors were subjected to thermal-neutron irradiation near room temperature. By 8 x 10$sup 19$n/cm$sup 2$, a crystal had become highly disordered, as shown by the back-reflection Laue picture, by loss of piezoelectric response, and by loss of optical birefringence. This transformation appears to be a metamictization. Changes in optical absorption, tritium loss, postirradiation lithium migration, surface crazing, and effects of postirradiation storage at room temperature and of heating to 140$sup 0$C are also observed. (auth)
TL;DR: In this article, the authors proposed the construction of an intense Li(d,n) neutron source at Brookhaven National Laboratory, which is based on the stripping reaction of energetic deuterons on a flowing liquid-lithium target.
Abstract: Brookhaven National Laboratory has proposed the construction of an intense Li(d,n) neutron source. The neutron production process is based on the stripping reaction of energetic deuterons on a flowing liquid-lithium target. The resulting neutron fluxes of greater than 10/sup 14/ n/(cm/sup 2/ sec) are well collimated in the forward direction providing approximately 1 liter of experimental volume for a 100-mA deuteron beam at approximately 30 MeV. The neutron energy spectrum is centered at approximately 14 MeV and extends from 8 to 20 MeV at FWHM. Models to calculate the radiation damage effectiveness of this neutron spectrum were developed. These show good agreement with the radiation damage expected in a fusion reactor model (BENCH) both in terms of dpa and helium production and recoil energy probabilities. The facility consists of a drift-tube-type linear accelerator producing the 30-MeV deuteron beam. This beam comprising two components (D/sup +/ and D/sup -/ ions) will be directed to the experimental area where it will be stopped on flowing liquid-lithium targets. The two different ion species will provide for the availability of two separate and independent experimental caves.
Journal Article•
TL;DR: The Dragon Project has conducted creep/corrosion tests in air and in impure helium on a wide range of structural and experimental steels and high-temperature alloys as discussed by the authors.
Abstract: To evaluate performance of materials in a high-temperature reactor, Dragon Project has conducted creep/corrosion tests in air and in impure helium on a wide range of structural and experimental steels and high-temperature alloys. These included 11 casts of austenitic steels tested in helium with impurity levels controlled at 50 to 100 μ at H2, 25 to 50 µ at CO, 3 to 8 µ at CH4, and 0.5 to 3 µ at H2O in a total pressure of 1.8 atm. Tests were conducted at 650 to 800°C for times up to 15 000 h. For materials based on 9 to 17% nickel and 15 to 18% chromium, surface corrosion rates were lower in steels containing 0.16 to 0.7% niobium than in those with similar levels of titanium or those of AISI Type 316 stainless steel. Subsurface intergranular oxidation and carburization were also noted in the niobium-free steels. Depths of intergranular oxidation ranged up to 200 µm, depending on strain, time, and temperature. In AISI Type 316 stainless steel, carburization was noted for depths up to 1.3 mm after 1...
TL;DR: In this paper, the flow velocity and direction in wire-spaced rod bundles were performed in air flow with a 37-rod subassembly that consisted of rods 31.6 mm in diameter, 3464 mm long, 6.0 mm in spacer wire diameter, and 1100 mm in wire wrapping pitch.
Abstract: Measurements of the flow velocity and direction in wire-spaced rod bundles were performed in air flow with a 37-rod subassembly that consisted of rods 31.6 mm in diameter, 3464 mm long, 6.0 mm in spacer wire diameter, and 1100 mm in wire wrapping pitch. The fine flow velocity and its direction were measured in both interior and wall subchannels in two sections of axial direction. Circular flow and decrease of axial flow velocity were observed in the subchannel containing the spacer wire. The variation of the transverse flow rate within one spacing wire pitch, which occurred between two adjacent subchannels, was obtained from the measured value to develop the computer program for the subchannel analysis.
TL;DR: Inconel-600 and Incoloy-800 cracks intergranularly at mildly anodic potentials and transgranularally at reduced potentials at open circuit potentials, respectively.
Abstract: High-temperature electrochemical tests have resulted in the stress corrosion cracking of Inconel-600 and Incoloy-800 (registered trademarks, International Nickel Company), and Type 304 stainless steel in caustic solutions. Results show that stress corrosion cracking of these alloys can be prevented or accelerated by varying their electrochemical potential. To a certain extent, the same effect can be achieved by altering the gas atmosphere above the test solution from a pure nitrogen cover gas to a mixture of 5 percent H$sub 2$ and 95 percent N$sub 2$. The effect of the cover gas can then be negated by adjusting the specimen's electrochemical potential either to cause or to inhibit stress corrosion cracking. Some specifics of the test results reveal that in deoxygenated caustic solutions, Inconel-600 cracks intergranularly at mildly anodic potentials; Incoloy-800 cracks transgranularly at reduced potentials (at or near the open circuit potential) and intergranularly at highly oxidizing potentials; and cracking is mixed (transgranular/intergranular) for Type 304 stainless steel at or near the open circuit potential. The severity of cracking for both Inconel-600 and Incoloy-800 in deoxygenated caustic solutions is reduced by giving the materials a simulated post-weld heat treatment (1150$sup 0$F for 18 h). Test results on Inconel-600 show that high-carbonmore » (0.06 percent) material cracks less severely than low-carbon (0.02 percent) material, in both the simulated post-weld heat-treated condition and the mill-annealed condition. (auth)« less
TL;DR: Simultaneous auger electron spectroscopy and ion sputtering have been used to measure the sputter yield, S (atom/ion), for Ar/sup +/ on carbon, tungsten, niobium, and silver in the energy range from 0.5 to 1.5 keV as discussed by the authors.
Abstract: Simultaneous auger electron spectroscopy and ion sputtering have been used to measure the sputter yield, S (atom/ion), for Ar/sup +/ on carbon, tungsten, niobium, and silver in the energy range from 0.5 to 1.5 keV and for H/sup +/ on tungsten, carbon, and silver at 11 keV. All measurements were performed on thin films, ranging in thickness from 150 to 6000 A, which were maintained at room temperature during bombardment. These films were produced by vacuum vapor deposition, and the thicknesses were measured by surface profilometry. The auger electron signals were used to determine the time required to etch through a film; from these measurements and a knowledge of the ion current density, the sputter yield was determined. For Ar/sup +/, 0.7 less than or equal to S less than or equal to 5.1 and for H/sup +/, 0.004 less than or equal to S less than or equal to 0.04 for the various materials studied in this energy range. Agreement with earlier experimental results is generally within +-25 percent.
TL;DR: In this article, hydrogen re-emission and scanning electron microscopy measurements were conducted on a series of cold-worked Types 316 and 302 stainless-steel specimens implanted with 20-keV protons at tempera...
Abstract: Hydrogen gas re-emission and scanning electron microscopy measurements were conducted on a series of cold-worked Types 316 and 302 stainless-steel specimens implanted with 20-keV protons at tempera...