Both of these reactor types operate with a thermal neutron spectrum and can also be fueled with mixed oxide MOX containing mixtures of uranium oxide and plutonium oxide. Alternatively, the spent fuel can be reprocessed, separating the reprocessed plutonium for use in MOX or possibly in breeder reactors once these have been successfully developed and built on a commercial scale.
MOX can be fabricated using such reprocessed plutonium or using weapons-grade plutonium WPu ; the separated fission products high-level radioactive waste, HLW are then usually vitrified and stored for eventual emplacement in a mined geological repository. These differences are important in that they determine how much plutonium can be loaded into MOX fuel elements and what fraction of the reactor core can use MOX fuel rods in place of LEU. These quantities in turn determine how many reactors are required for what period of time to dispose of a given quantity of excess WPu.
To give guidance for such estimates, one should be reminded that roughly one ton of either plutonium or U is consumed to generate one gigawatt year of electricity. The significant differences between plutonium and uranium-based reactor fuels are the following:. Delayed neutrons are important since their presence makes it possible to control the reactivity in a reactor through the mechanical insertion of control rods.porcelaintile.org/includes/3945.php
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Thus control rod performance must be enhanced if plutonium fuels are used. Since plutonium has a higher absorption cross section for thermal neutrons compared to U , the average energy of the neutron spectrum in reactors using plutonium fuels is higher. Thus the capture of neutrons by control rods is reduced if plutonium fuels are used, thus further reducing the effectiveness of the control rods.
Pu has a peak in its absorption called a resonance at neutron energies near 0. This has the result that under certain conditions the temperature coefficient of reactivity of the reactor can become positive, as the temperature increases more neutrons are being pushed into the resonance region, increasing reactivity still further. The above differences have to be taken into account in reactor design initially intended for the use of MOX or for modifying existing reactors for the use of MOX.
The effects noted above. It may appear that the potentially larger plutonium content of spent fuel is a negative factor arguing against the use of MOX for disposing of WPu. This is clearly not the case for two reasons: Second, the total amount of plutonium contained in spent fuel, if MOX fabricated from WPu is used, is less, relative to the total amount of the original WPu put into the fuel, plus the plutonium produced in the spent fuel had the same amount of electric power been generated from LEU.
In short, the plutonium content of the spent fuel is not a useful discriminant among alternate disposition approaches. Some currently operating reactors are already designed for using MOX for all their fuel elements and others not now burning MOX have been demonstrated to be usable for full MOX cores. It incorporates the additional control rods and increased neutron absorber in the coolant required to permit full MOX operation.
Actual commercial experience with MOX operations exists only in Europe since reprocessing of spent fuel is not practiced licensed in the United States or Canada and since use of MOX based on surplus WPu remains a matter for the future. In total about Extensive experience with a large variety of parameters, such as fuel composition and varying degrees of burn-up, has accumulated.
These German reactors have been licensed for MOX fractions up to 50 percent. Table B-1 gives an overview of these reactors. The actual percentage of MOX use has been considerably less than that shown in the table. The reason is that not enough fuel was available. In principle, it could also be possible to license the BWRs for.
This has not happened in the past because of the limited MOX production capacities. Table B-2 gives an overview of the BWRs. For some plants, adjustments in case of changing quality of the plutonium or the uranium are licensed. Jahrbuch der Atomwirtschaft Yearbookof the Atomic Economy. Belgium and Japan are planning to begin loading MOX fuel in commercial reactors later in this decade.
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Details of this favorable experience were presented in considerable detail at the GAAC workshop. In the United States and Canada, no reactors are licensed for MOX operation but the licensing process had been pursued in the past. If it were decided to use MOX in American or Canadian reactors, the licensing process could be relatively expeditious. While actual operating experience with MOX had been restricted to partial MOX loads and while most existing LWRs were originally perceived to permit only partial loadings, recent analyses by the major U.
These findings are only preliminary and would have to be put on a firm technical basis for an actual licensing procedure. In view of the above, it is difficult to forecast precisely how many reactor years of MOX use would be required to dispose of the nominal tons of surplus Russian WPu or the corresponding amount of U. If full MOX loading were possible, then four to six reactors could handle tons in about 20 years. Partial loadings would increase the number of required reactors or the time needed for a campaign accordingly. Extensive calculations on specific combinations of actual reactors, assuming various percentages of plutonium in MOX, and assuming various fractions of MOX fuel loading in the reactors, are available in the literature.
Estimates are contained in the Russian-German Study described in 1. Reactor-Related Options Washington, D. National Academy Press, ]. There is general agreement that the cost of fueling a reactor with MOX fabricated from free WPu today would be somewhat higher than fueling the same reactor with fully paid-for LEU; fully-paid for means including the costs of mining, processing, enriching, and fabricating the LEU fuel. Thus the use of WPu as a substitute for LEU in existing reactors is likely to require a subsidy despite the inherent fuel value of WPu.
The cost impact of such a differential on the total cost of electricity is minimal, however. The reason is that MOX fabrication, due to the radioactivity, toxicity, and safeguarding requirements, has to be much more automated than that for LEU and requires additional shielding and other facilities and highly secure procedures.
A partially completed MOX fabrication facility exists at the U. Measured spectrum of the unspiked SRM sample using the direct method. Note that for the isotope ratio evaluation the peak centroid of the isotopes of interest flat-top peak was selected. Other aliquots of the diluted working solutions were taken to determine the Pu and Pu amount content by isotope dilution ICP-MS using Pu spike.
In order to minimize the sample manipulation and the generated waste, extraction chromatography was selected to separate Pu from U. Heating of the sample can accelerate the reactions and help to reduce bubble formation during the column separation. The load and wash solution were collected together in a mL PE vial, resulting in about 4. The time of the chemical separation was registered as the reference date for age dating. Note that Np and Am can also be recovered with the proposed sample preparation.
In parallel to the spiked sample, a blank and an aliquot of the unspiked sample were also subjected to the chemical separation and a forthcoming analysis by ICP-MS. The separations were done in duplicates. Such a separation factor is sufficient for higher burn-up Pu samples. Measured spectrum of the unspiked SRM sample after the chemical separation. Other aliquots of the starting diluted working solutions were also taken to determine the Pu amount content by isotope dilution ICP-MS using Pu spike.
The spiked samples were also subjected to the chemical separation.
The n Pu , n Pu , n Pu and n Pu amount contents needed for the age dating calculations were calculated using the measured Pu isotopic composition of the separated unspiked Pu sample and the Pu amount content obtained by isotope dilution ICP-MS. Concentrations of isotopes of interest necessary for the production date calculation were experimentally determined according to the isotope dilution method. The measured amount contents of the required U and Pu isotopes were used to calculate the model age of the material according to Eq.
The age calculations were performed with commercially available software, GUM Workbench [ 25 ]. The dominant uncertainty components were the measured isotope ratios in the spiked sample and the spike concentrations, except for the Pu and U measurements, where the dominant uncertainty contributors were the blank level and the Pu mathematical correction, respectively.
The results are given relative to a common reference date of 11 March date of the plutonium separation. The results obtained with and without the chemical separation agree well within measurement uncertainty. Also the ages obtained by the different chronometers give identical results concordant ages. The achievable uncertainties by the direct method for such old materials are comparable to the results measured after chemical separation.
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- Plutonium age dating (production date measurement) by inductively coupled plasma mass spectrometry!
In our separation and measurement scheme the residual Pu signal contribution was in the same order of magnitude as the progeny U signal. Measured age dating results of the plutonium certified reference materials using the different chronometers. The measured production dates of the SRM and samples agree well with the known, archive dates of production January and in December , respectively.
This agreement implies that the U separation from the Pu material was complete during the purification step. Calculated production date results of the plutonium certified reference materials for the different chronometers. The obtained absolute production dates allow for the comparison with previously reported results in the literature Fig.
For ease of comparison only the values for SRM are shown here, which has been studied more extensively. The reported production dates were calculated as the difference of the reference date and the measured ages. The uncertainties shown in Fig. As the exact date of the purification by day was not available from the NBL archive, and it is given only as January , an uncertainty of half a month was assigned to this date. Comparison of the measured and reported production dates for SRM standard [ 10 , 17 ].
The reported production dates are in relatively good agreement with one another and also with the known, archive purification date. Note that the production dates based on different chronometers reported by the same groups are in better agreement than the absolute production dates reported by the different groups. Two rapid measurement methods have been developed for the determination of the age and the respective production date of plutonium materials by ICP-MS for nuclear forensics and safeguards.
The other method, which involves the chemical separation of U, is particularly suited for age dating of freshly separated Pu materials with low amount of U decay products, for Pu samples with low Pu abundance e. The minimum amount of sample needed for the analysis is in the sub-nanogram range. The methods have been applied for the age dating measurement of two Pu isotopic reference materials SRM and , and the experimentally determined production dates are in good agreement with the known, archive date of material purification.
Small discrepancies between the literature values have been found for the SRM , which may be related to analytical measurement issues or the differences in the age calculations e.
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National Center for Biotechnology Information , U. Journal of Radioanalytical and Nuclear Chemistry. J Radioanal Nucl Chem. Published online Sep 4. Author information Article notes Copyright and License information Disclaimer. Received Jun 5. Abstract This paper describes rapid methods for the determination of the production date age dating of plutonium Pu materials by inductively coupled plasma mass spectrometry ICP-MS for nuclear forensic and safeguards purposes.
Age dating, Radiochronometry, Plutonium, Nuclear safeguards, Nuclear forensics. The ratio of the daughter nuclide amount relative to the amount of its parent nuclide can be calculated as follows: Experimental Reagents and materials All labware was thoroughly cleaned before use. Analysed plutonium reference materials The measured Pu samples in the present study are the NBS and certified reference materials, certified for Pu isotopic composition.
Instrumentation and analytical measurements The Pu and U isotopic measurements were carried out using a double-focusing magnetic sector ICP-MS equipped with a single electron multiplier Element 2, Thermo Electron Corp. Open in a separate window. Age measurement by ICP-MS after chemical separation In order to minimize the sample manipulation and the generated waste, extraction chromatography was selected to separate Pu from U.
Data evaluation Concentrations of isotopes of interest necessary for the production date calculation were experimentally determined according to the isotope dilution method. Conclusions Two rapid measurement methods have been developed for the determination of the age and the respective production date of plutonium materials by ICP-MS for nuclear forensics and safeguards. The state of nuclear forensics. Application of lead and strontium isotope ratio measurements for the origin assessment of uranium ore concentrates.
Origin assessment of uranium ore concentrates based on their rare-earth elemental impurity pattern. Attribution of uranium ore concentrates using elemental and anionic data. Production date determination of uranium-oxide materials by inductively coupled plasma mass spectrometry.
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