Showing papers in "Nuclear Technology in 1979"
858 citations
TL;DR: In this article, the Chemical Thermodynamics of Actinide Elements and Compounds is discussed. But the authors focus on the chemical properties of actinide elements and compounds.
Abstract: (1979). The Chemical Thermodynamics of Actinide Elements and Compounds. Nuclear Technology: Vol. 42, No. 1, pp. 114-115.
397 citations
TL;DR: Sodium saturated bentonite clay compacted to a high density is a very efficient isolation for preventing radiotoxic nuclides from deeply buried canisters with radioactive wastes from reaching the biosphere.
Abstract: Sodium saturated bentonite clay compacted to a high density is a very efficient isolation for preventing radiotoxic nuclides from deeply buried canisters with radioactive wastes from reaching the biosphere. The main function of the bentonite, which is applied in the form of blocks between the rock and the canisters in large boreholes, is to provide a practically impervious barrier. The bentonite blocks take up water and swell so that they completely fill the space between rock and canisters. The swelling potential, which is permanent, makes the bentonite self-sealing. This means that rock joints, which may be opened, are sealed by extruding bentonite.
127 citations
40 citations
TL;DR: In this article, a process was developed for the production of dense fissile fuel kernels for the high-temperature reactor, where a concentrated uranyl nitrate solution with urea and ammonium was used.
Abstract: A process has been developed for the production of dense fissile fuel kernels for the high-temperature reactor. A method is used whereby a concentrated uranyl nitrate solution with urea and ammoniu...
38 citations
TL;DR: Experimental methods for detecting types of motion within and surrounding reactor cores were employed, including the dominating influence of noise analysis techniques and the associated theoretical analysis as mentioned in this paper, in order to detect different types of motions within and around reactor cores.
Abstract: Experimental methods for detecting types of motion within and surrounding reactor cores were employed, including the dominating influence of noise analysis techniques and the associated theoretical...
37 citations
TL;DR: The ELESIM code as mentioned in this paper models a single fuel element in a one-dimensional axisymmetric manner, and the constituent subroutines are physically based (rather than empirical) models, and include such phenomena as...
Abstract: The ELESIM code models a single fuel element in a one-dimensional axisymmetric manner. The constituent subroutines are physically based (rather than empirical) models, and include such phenomena as...
34 citations
TL;DR: In this paper, a method utilizing solvent extraction coupled with liquid scintillation spectrometry has been developed for the assay of uranium and thorium in fertilizers and phosphate-containing minerals and chemicals.
Abstract: A method utilizing solvent extraction coupled with liquid scintillation spectrometry has been developed for the assay of uranium and thorium in fertilizers and phosphate-containing minerals and chemicals. Trioctylphosphine oxide in toluene is used to extract uranium and thorium from a perchloric and nitric acid solution, with phosphate interference being suppressed by the addition of aluminum ion. The uranium and thorium are stripped from this solution, and uranium is separated from the thorium by selective reextraction of uranium into a scintillator with Adogen 364 (tertiary amine) sulfate. The thorium remaining in the aqueous is reextracted into another scintillator with (primary) 1-nonyldecylamine sulfate. Both nuclides are counted separately in a high-resolution liquid scintillation spectrometer. The sensitivity of the counting method is enhanced by the use of pulse-shape rejection of the beta--gamma background. Results indicate a detection threshold of 0.0038 pCi of uranium (1.1 part/10/sup 8/) with a 1000-min counting time. Reproducibility of +- 2.5% was found at the 50-ppM level. For thorium detection, thresholds are 4 part/10/sup 13/ for the same counting time with +- 3.0% average recovery of /sup 230/Th and 7 part/10/sup 8/ of /sup 232/Th.
33 citations
TL;DR: In this paper, an external gelation process was proposed to produce dense ThO2 and (Th,U)O2 microspheres for refabrication of high-temperature reactor fuel elements.
Abstract: For refabrication of high-temperature reactor fuel elements, a process for producing dense ThO2 and (Th,U)O2 microspheres has been developed. The process is an external gelation process and takes full advantage of the gelation features of thorium hydroxide for formation of particles. Unlike other external gelation methods, neither viscosity increase by adding other substances (e.g., organic polymers) to the broth nor drop formation in organic liquids is employed. The method uses few process steps and only simple process equipment. A pilot plant has been constructed that operation with one nozzle can produce 1 kg/h of (Th,U)O2 microspheres 500 µm in diameter.
32 citations
TL;DR: Borosilicate glasses containing actual high-level Savannah River Plant waste were leached in static, distilled water as discussed by the authors, with leach rates based on 90Sr, 137Cs, and plutonium analyses.
Abstract: Borosilicate glasses containing actual high-level Savannah River Plant waste were leached in static, distilled water. Leach rates based on 90Sr, 137Cs, and plutonium analyses were similar: 10−5 to ...
TL;DR: A computer code CAT (Computer Automated Tree) has been developed to automatically produce fault trees from decision tables to demonstrate the potential utility of such an automated approach to fault tree construction once a basic set of general decision tables have been developed.
Abstract: A systematic methodology for the construction of fault trees based on the use of decision tables has been developed. These tables are used to describe each possible output state of a component as a set of combinations of states of inputs and internal operational or T states. Two methods for modeling component behavior via decision tables have been developed, one inductive and one deductive. These methods are useful for creating decision tables that realistically model the operational and failure modes of electrical, mechanical, and hydraulic components as well as human interactions inhibit conditions and common-cause events. A computer code CAT (Computer Automated Tree) has been developed to automatically produce fault trees from decision tables. A simple electrical system was chosen to illustrate the basic features of the decision table approach and to provide an example of an actual fault tree produced by this code. This example demonstrates the potential utility of such an automated approach to fault tree construction once a basic set of general decision tables has been developed.
TL;DR: In this article, it is shown that a denatured MSR with full-scale on-site fuel reprocessing appears to be capable of break-even breeding, but it is doubtful that sufficient enhancement could be achieved to make the systems suitable for deployment other than at secure sites.
Abstract: Molten-salt reactors (MSRs), because of the fluid nature of the fuel, appear to provide an attractive approach to efficient fuel utilization in the thorium-/sup 233/U cycle as well as a means for limiting the availability of plutonium and the general proliferation risks associated with nuclear power generation. High-enrichment /sup 233/U systems could, in principle, be operated with positive breeding gains to effectively eliminate plutonium as a nuclear fuel. However, such systems would be proliferation sensitive. Concept modifications (short of denaturing the uranium fuel) can be conceived to enhance the proliferation resistance of high-enrichment MSRs, but it is doubtful that sufficient enhancement could be achieved to make the systems suitable for deployment other than at secure sites. Denaturing the uranium in an MSR introduces some plutonium into the fuel cycle and generally degrades its breeding performance. Nevertheless, a denatured MSR with full-scale on-site fuel reprocessing appears to be capable of break-even breeding. In addition, the plutonium (most of which is consumed in situ) would be of poor quality and could never be isolated from all other undesirable nuclides. Thus, such systems would provide for efficient utilization of uranium resources in a proliferation-resistant environment while limiting the amount of plutonium (and transplutoniummore » actinides) that would have to be handled as waste. The development of commercial MSRs by early in the 21st century appears to be technologically feasible.« less
TL;DR: In this paper, a novel fuel processing concept based on the high-temperature precipitation of ''actinide--nitrides'' from a liquid tin solution is proposed, which will allow for fission product removal to be performed entirely within the device at high burnup.
Abstract: The laser fusion driven actinide waste burner (LDAB) system investigated uses partitioned fission power reactor generated actinide wastes dissolved in a molten tin alloy as feed material (or fuel). A novel fuel processing concept based on the high-temperature precipitation of ''actinide--nitrides'' from a liquid tin solution is proposed. This concept will allow for fission product removal to be performed entirely within the device at high burnup. No attempt has been made to optimize this system, but potential performance is impressive. The equilibrium LDAB design consumes 7.6 MT/y of actinide waste. This corresponds to the waste output from 136 light water reactors (1000 MW (electric)). The mean life of an actinide atom in the LDAB is only 4.5 y; and actinides, once charged to the LDAB, might be reprocessed fewer times during irradiation than in previously proposed systems.
Abstract: Calculations are made of 2.5-MeV neutron yields via D--D reactions from various beam line components, including beam targets and a neutralizer gas cell, and of x-ray generation from accelerator columns in deuterium beam injectors of various energies from 40 to 200 keV, with accelerator powers ranging from 2 to 20 MW per beam line. The calculated neutron intensities from the neutral beam injector systems for present and future fusion research are in the range between 10/sup 11/ and 10/sup 13/ n/s and warrant biological shielding in most cases, even for low duty cycle operation. X radiation from the accelerator columns becomes a health physics concern only for injectors with energy higher than approx. 100 keV.
TL;DR: In this article, the microstructural response of fast breeder reactor fuel to accident transients has been analyzed, and fuel response can be classified as either basically brittle or basically ductile in nature.
Abstract: The microstructural response of fast breeder reactor fuel to accident transients has been analyzed. Based on experimental results, fuel response can be classified as either basically brittle or basically ductile in nature. In the analysis, the type of response is assumed to be determined by the behavior of grain boundary fission gas. The transient variables taken into consideration are the temperature, heating rate, the mean gas content per bubble, mean bubble spacing in the grain boundary, and the stresses resolved normal to grain boundaries containing gas bubbles. By calculating the rate at which a grain boundary bubble grows as a sharp crack and comparing it to the rate of bubble growth by mass transport, a criterion is established to predict the characteristic response of a fuel sample to a specified thermal transient. A swelling threshold time is also determined for the case of ductile fuel behavior. Tensile stresses applied to the grain boundary are shown to enhance brittle behavior, and compressive stresses are shown to enhance ductile behavior. When average values of the relevant variables are extracted from a number of fission gas release and direct electric heating experiments and are used in the above calculation, fuel behavior predictions formore » these tests are found to correspond well with the experimental results.« less
TL;DR: In this article, an apparatus was developed that utilizes light ions to simulate the effect of a fusion reactor first wall environment on the creep properties of metals and alloys, and the creep apparatus included a wi...
Abstract: An apparatus was developed that utilizes light ions to simulate the effect of a fusion reactor first wall environment on the creep properties of metals and alloys. The creep apparatus includes a wi...
TL;DR: In this paper, the authors present a cost-risk-benefit analysis of partitioning long-lived nuclides from waste and transmuting them to shorter-lived or stable Nuclides, and show that the use of tributyl phosphate (TBP) followed by extraction with a bidentate organophosphorous extractant (DHDECMP) appears to be the most efficient for removing actinides from saltmore waste.
Abstract: The US Department of Energy awarded Oak Ridge National Laboratory a program to develop a cost-risk-benefit analysis of partitioning long-lived nuclides from waste and transmuting them to shorter lived or stable nuclides. Two subtasks of this program were investigated at Rocky Flats. In the first subtask, methods for solubilizing actinides in incinerator ash were tested. Two methods appear to be preferable: reaction with ceric ion in nitric acid or carbonate-nitrate fusion. The ceric-nitric acid system solubilizes 95% of the actinides in ash; this can be increased by 2 to 4% by pretreating ash with sodium hydroxide to solubilize silica. The carbonate-nitrate fusion method solubilizes greater than or equal to 98% of the actinides, but requires sodium hydroxide pretreatment. Two additional disadvantages are that it is a high-temperature process, and that it generates a lot of salt waste. The second subtask comprises removing actinides from salt wastes likely to be produced during reactor fuel fabrication and reprocessing. A preliminary feasibility study of solvent extraction methods has been completed. The use of a two-step solvent extraction system - tributyl phosphate (TBP) followed by extraction with a bidentate organophosphorous extractant (DHDECMP) - appears to be the most efficient for removing actinides from saltmore » waste. The TBP step would remove most of the plutonium and > 99.99% of the uranium. The second step using DHDECMP would remove > 99.91% of the americium and the remaining plutonium (> 99.98%) and other actinides from the acidified salt waste. 8 figures, 11 tables.« less
TL;DR: In this article, the authors examined the interactions of trace quantities of cesium, plutonium, neptunium, and americium in aqueous solutions with rocks from formations that may be suitable for waste repositories.
Abstract: To aid in assessing the suitability of geologic disposal, the authors have examined the interactions of trace quantities of cesium, plutonium,, neptunium, and americium in aqueous solutions with rocks from formations that may be suitable for waste repositories Results indicate that many geologic formations are barriers to the movement of these elements in flowing water However, reactions that retard element migration are varied and do not lend themselves to simplified descriptions In experiments with plutonium and americium, kinetics of reactions were seen to differ for each trace element and rock studied In rock infiltration experiments with radioactive cesium, plutonium, neptunium, and americium, often most of the activity moved slowly compared to the water stream, but small quantities of the trace elements moved downstream from the main peaks of activity because of the slow reaction rates seen in static experiments, or possibly because of multiple speciation, colloid formation, movement of particles with adsorbed nuclides, or other causes 2 figures, 7 tables
TL;DR: In this paper, a number of SS-316 right circular cylindrical shells of varying lengths have been analyzed using two-dimensional discrete ordinates transport methods together with first and last flight particle estimators to aid in the design of neutron collimators for the Tokamak Fusion Test Reactor.
Abstract: A number of SS-316 right circular cylindrical shells of varying lengths have been analyzed using two-dimensional discrete ordinates transport methods together with first and last flight particle estimators to aid in the design of neutron collimators for the Tokamak Fusion Test Reactor. In the TFTR the 14-MeV neutron source has a very large spatial extent and the collimators must be designed to allow spectral measurements that refer to only a small spatial region of this extended source. The analysis identifies the 14-MeV neutrons from scattering in the SS-316 immediately adjacent to the collimator opening as the dominant contributor to detector background. Collimator lengths greater than 0.60 m were found sufficient to attenuate uncollided background neutrons for reasonable source-detector distances. The lower energy (less than 13.8 MeV) neutron background and gamma background were not found to be significant.
TL;DR: In this paper, the placement of liquid metals (lithium, lead, and a Pb--Li eutectic Pb/sub 4/Li) between the first wall and the source of neutrons has been considered as a mechanism for extending first wall lifetimes in inertial confinement fusion reactors.
Abstract: The placement of liquid metals (lithium, lead, and a Pb--Li eutectic Pb/sub 4/Li) between the first wall and the source of neutrons has been considered as a mechanism for extending first wall lifetimes in inertial confinement fusion reactors. This scheme is called the Internal Spectral Shifter and Energy Converter (ISSEC). All three liquid metals have been shown to reduce the radiation damage in the Type 316 stainless-steel structural first wall and thus increase the first wall lifetime. An overall conclusion of the study is that the Pb--Li eutectic ISSEC has better characteristics than both pure lead and lithium ISSECs, and for best results it should be used at thicknesses ranging from 45 to 65 cm.
TL;DR: It has been shown that design and maintenance errors are the predominant causes of CMFs and the large differences between nuclear safety and aircraft system CMF rates are shown to be generally explicable and illuminating in connection with the means of preventing or reducing the probability ofCMFs.
Abstract: There can be no doubt that difficulties have been generally experienced in assessing the impact of common-mode failures (CMFs) on the reliability of safety systems involving redundancy This certainly became clear in a review of the available literature carried out as part of the study of CMFs Consequent to studying CMFs in the nuclear, aviation, and chemical industries, it was possible to generally define CMFs and to produce a comprehensive scheme of classification The latter has been used in the analysis of data from these industries, concentrating on particular redundant nuclear safety and aircraft systems It has been shown that design and maintenance errors are the predominant causes of CMFs This is important since these reflect on the tasks and organizations that produce the redundancy systems The large differences between nuclear safety and aircraft system CMF rates are also shown to be generally explicable and illuminating in connection with the means of preventing or reducing the probability of CMFs These undoubtedly require serious consideration if the reliability of nuclear safety systems is not to be dominated by CMFs The study reported has led to further work relating CMF control and modeling that is described elsewhere and is also still inmore » progress« less
TL;DR: In this article, the tritium permeability through Fe-2¼Cr-1 Mo steel (Croloy) steam generator material has been measured for the surface-clean metal and for the metal oxidized in steam at 755 K (482°C).
Abstract: The permeability of tritium through Fe-2¼Cr-1 Mo steel (Croloy) steam generator material has been measured for the surface-clean metal and for the metal oxidized in steam at 755 K (482°C). The temp...
TL;DR: The Electric Power Research Institute is presently supporting research on radiation control techniques for light water nuclear power plants Nine different contractors working on 12 different plants were employed by the EPRI in this article.
Abstract: The Electric Power Research Institute is presently supporting research on radiation control techniques for light water nuclear power plants Nine different contractors working on 12 different contr
TL;DR: In this paper, models for cesium and iodine release from light water reactor (LWR) fuel rods defected in steam were formulated based on experimental fission product release data from several types of defected LWR fuel rods.
Abstract: Models for cesium and iodine release from light water reactor (LWR) fuel rods defected in steam were formulated based on experimental fission product release data from several types of defected LWR...
TL;DR: Bayes' theorem is demonstrated as a means for incorporating in the prediction of the availability performance of new generations of turbine blades the limited operational data on the new blades along with the experience of the earlier generation and the knowledge of the design changes.
Abstract: Bayes' theorem is used to quantify the impact of new evidence in three energy-related decision problems. The first problem concerns the risk of radioactivity release during the railroad transport of spent nuclear fuel. This history of shipments thus far is shown to make it highly unlikely that the frequency of release is on the order of 0.001 or greater per shipment. The second and third applications involve predicting the availability performance of new generations of turbine blades. Bayes' theorem is demonstrated as a means for incorporating in the prediction the limited operational data on the new blades along with the experience of the earlier generation and the knowledge of the design changes.
TL;DR: At the Savannah River Plant, stress corrosion of carbon steel storage tanks containing alkaline nitrate radioactive waste is controlled by specification of limits on waste composition and temperatu... as discussed by the authors.
Abstract: At the Savannah River Plant, stress corrosion of carbon steel storage tanks containing alkaline nitrate radioactive waste is controlled by specification of limits on waste composition and temperatu...
TL;DR: A cesium trap containing reticulated vitreous carbon and operating at 190°C has been used at the Experimental Breeder Re... as discussed by the authors, where carbon has been shown to be an effective getter for cedium in liquid sodium.
Abstract: Carbon has been shown to be an effective getter for cesium in liquid sodium. A cesium trap containing reticulated vitreous carbon and operating at 190°C has been used at the Experimental Breeder Re...
TL;DR: In this paper, an analysis of the uncertainty in the prediction of dose to an individual's thyroid due to the transport of radioactive molecular iodine, /sup 131/I/sub 2/, from air through the pasture-cow-milk pathway was made.
Abstract: An analysis was made of the uncertainty in the prediction of dose to an individual's thyroid due to the transport of radioactive molecular iodine, /sup 131/I/sub 2/, from air through the pasture-cow-milk pathway. This analysis was facilitated by the adoption of a model consisting of a multiplication of several factors represented by lognormal distributions of values. Results indicate there is a 64, 50, or 23% chance of the annual dose to an individual's thyroid not exceeding the mean, median, or most probable doses, respectively. However, these results are tentative as a result of the limited amount of data available for annual average dose assessments. The suggestion is made that consideration be given to adopting a probabilistic approach to determining an acceptable probability of an individual receiving a dose that exceeds a limiting value.