To see the other types of publications on this topic, follow the link: Metallic uranium-plutonium fuel.
Journal articles on the topic 'Metallic uranium-plutonium fuel'
Create a spot-on reference in APA, MLA, Chicago, Harvard, and other styles
Consult the top 25 journal articles for your research on the topic 'Metallic uranium-plutonium fuel.'
Next to every source in the list of references, there is an 'Add to bibliography' button. Press on it, and we will generate automatically the bibliographic reference to the chosen work in the citation style you need: APA, MLA, Harvard, Chicago, Vancouver, etc.
You can also download the full text of the academic publication as pdf and read online its abstract whenever available in the metadata.
Browse journal articles on a wide variety of disciplines and organise your bibliography correctly.
1
Khorasanov, Georgiy, Dmitriy Samokhin, Aleksandr Zevyakin, Yevgeniy Zemskov, and Anatoliy Blokhin. "Lead reactor of small power with metallic fuel." Nuclear Energy and Technology 4, no. 2 (2018): 99–102. http://dx.doi.org/10.3897/nucet.4.30527.
Full textAbstract:
The possibility for obtaining a hard neutron spectrum in small reactor cores has been considered. A harder spectrum than spectra in known fast sodium cooled and molten salt reactors has been obtained thanks to the selection of relatively small core dimensions and the use of metallic fuel and natural lead (natPb) coolant. The calculations for these compositions achieve an increased average neutron energy and a large fraction of hard neutrons in the spectrum (with energies greater than 0.8 MeV) caused by a minor inelastic interaction of neutrons with the fuel with no light chemical elements and with the coolant containing 52.3% of 208Pb, a low neutron-moderating isotope. An interest in creating reactors with a hard neutron spectrum is explained by the fact that such reactors can be practically used as special burners of minor actinides (MA), and as isotope production and research reactors with new consumer properties. With uranium oxide fuel (UO2) substituted by metallic uranium-plutonium fuel (U-Pu-Zr), the reactors under consideration have the average energy of neutrons and the fraction of hard neutrons increasing from 0.554 to 0.724 MeV and from 18 to 28% respectively. At the same time, the one-group fission cross-section of 241Am increases from 0.359 to 0.536 barn, while the probability of the 241Am fission increases from 22 to 39%. It is proposed that power-grade plutonium resulting from regeneration of irradiated fuel from fast sodium cooled power reactors be used as part of the fuel for future burner reactors. It contains unburnt plutonium isotopes and some 1% of MAs which transmutate into fission products in the process of being reburnt in a harder spectrum. This will make it possible to reduce the MA content in the burner reactor spent fuel and to facilitate so the long-term storage conditions for high-level nuclear waste in dedicated devices.
2
Khorasanov, Georgiy, Dmitriy Samokhin, Aleksandr Zevyakin, Yevgeniy Zemskov, and Anatoliy Blokhin. "Lead reactor of small power with metallic fuel." Nuclear Energy and Technology 4, no. (2) (2018): 99–102. https://doi.org/10.3897/nucet.4.30527.
Full textAbstract:
The possibility for obtaining a hard neutron spectrum in small reactor cores has been considered. A harder spectrum than spectra in known fast sodium cooled and molten salt reactors has been obtained thanks to the selection of relatively small core dimensions and the use of metallic fuel and natural lead (<sup>nat</sup>Pb) coolant. The calculations for these compositions achieve an increased average neutron energy and a large fraction of hard neutrons in the spectrum (with energies greater than 0.8 MeV) caused by a minor inelastic interaction of neutrons with the fuel with no light chemical elements and with the coolant containing 52.3% of <sup>208</sup>Pb, a low neutron-moderating isotope. An interest in creating reactors with a hard neutron spectrum is explained by the fact that such reactors can be practically used as special burners of minor actinides (MA), and as isotope production and research reactors with new consumer properties. With uranium oxide fuel (UO<sub>2</sub>) substituted by metallic uranium-plutonium fuel (U-Pu-Zr), the reactors under consideration have the average energy of neutrons and the fraction of hard neutrons increasing from 0.554 to 0.724 MeV and from 18 to 28% respectively. At the same time, the one-group fission cross-section of <sup>241</sup>Am increases from 0.359 to 0.536 barn, while the probability of the <sup>241</sup>Am fission increases from 22 to 39%. It is proposed that power-grade plutonium resulting from regeneration of irradiated fuel from fast sodium cooled power reactors be used as part of the fuel for future burner reactors. It contains unburnt plutonium isotopes and some 1% of MAs which transmutate into fission products in the process of being reburnt in a harder spectrum. This will make it possible to reduce the MA content in the burner reactor spent fuel and to facilitate so the long-term storage conditions for high-level nuclear waste in dedicated devices.
3
Sato, I., H. Furuya, T. Arima, K. Idemitsu, and K. Yamamoto. "Behavior of metallic fission products in uranium–plutonium mixed oxide fuel." Journal of Nuclear Materials 273, no. 3 (1999): 239–47. http://dx.doi.org/10.1016/s0022-3115(99)00071-9.
Full text
4
Moiseenko, V., and S. Chernitskiy. "Nuclear Fuel Cycle with Minimized Waste." Nuclear and Radiation Safety, no. 1(81) (March 12, 2019): 30–35. http://dx.doi.org/10.32918/nrs.2019.1(81).05.
Full textAbstract:
A uranium-based nuclear fuel and fuel cycle are proposed for energy production. The fuel composition is chosen so that during reactor operation the amount of each transuranic component remains unchanged since the production rate and nuclear reaction rate are balanced. In such a ‘balanced’ fuel only uranium-238 content has a tendency to decrease and, to be kept constant, must be sustained by continuous supply. The major fissionable component of the fuel is plutonium is chosen. This makes it possible to abandon the use of uranium-235, whose reserves are quickly exhausted. The spent nuclear fuel of such a reactor should be reprocessed and used again after separation of fission products and adding depleted uranium. This feature simplifies maintaining the closed nuclear fuel cycle and provides its periodicity. In the fuel balance calculations, nine isotopes of uranium, neptunium, plutonium and americium are used. This number of elements is not complete, but is quite sufficient for calculations which are used for conceptual analysis. For more detailed consideration, this set may be substantially expanded. The variation of the fuel composition depending on the reactor size is not too big. The model accounts for fission, neutron capture and decays. Using MCNPX numerical Monte-Carlo code, the neutron calculations are performed for the reactor of industrial nuclear power plant size with MOX fuel and for a small reactor with metallic fuel. The calculation results indicate that it is possible to achieve criticality of the reactor in both cases and that production and consuming rates are balanced for the transuranic fuel components. In this way, it can be assumed that transuranic elements will constantly return to such a reactor, and only fission products will be sent to storage. This will significantly reduce the radioactivity of spent nuclear fuel. It is important that the storage time for the fission products is much less than for the spent nuclear fuel, just about 300 years.
5
Prathibha, T., K. Rama Swami, S. Sriram, and K. A. Venkatesan. "Interference of Zr(IV) during the extraction of trivalent Nd(III) from the aqueous waste generated from metallic fuel reprocessing." Radiochimica Acta 108, no. 7 (2020): 543–54. http://dx.doi.org/10.1515/ract-2019-3220.
Full textAbstract:
AbstractA metallic alloy of uranium–zirconium and uranium–plutonium–zirconium has been proposed as a fuel for fast reactors, owing to the possibility of achieving high breeding ratio in a short span of time. About 6–10 wt.% of zirconium has been added to these actinide fuels to increase the melting temperature and thermal-mechanical stability. Aqueous reprocessing of the spent metallic fuel generates the high-level liquid waste (HLLW) that contains about 60 % of the total zirconium from the fuel. In view of this, the extraction behavior of a trivalent representative ion, Nd(III) in the presence of Zr(IV) was studied from nitric acid medium using the candidate ligands proposed for trivalent actinide separation from HLLW, such as N,N,N′N′-tetraoctyldiglycolamide (TODGA), and N,N-di-octyl-2-hydroxyacetamide (DOHyA). The extraction was studied as a function of nitric acid concentration, zirconium and neodymium concentration and Nd(III) to Zr(IV) ratio. The findings of dynamic light scattering (DLS) and ATR-FTIR spectral techniques were used for understanding the complex chemistry of Zr(IV) extraction under different conditions. Poor extraction of nitric acid, smaller aggregate size, no third phase formation during the extraction of Zr(IV) and Nd(III) and other unique solvent properties favor the DOHyA molecule in n-dodecane as a solvent for partitioning of trivalent actinides from HLLW generated from metallic fuel reprocessing.
6
Okunev, Viacheslav. "The concept of a fast reactor with liquid metal fuel in tungsten capsules." E3S Web of Conferences 411 (2023): 01013. http://dx.doi.org/10.1051/e3sconf/202341101013.
Full textAbstract:
The concept of a dual-purpose high-power nuclear reactor is proposed. One of the goals is the production of electricity, the other is the production of high-potential thermal energy. It is proposed to use liquid fuel based on waste uranium and plutonium extracted from the spent fuel of VVER reactors (purified from the 238Pu isotope). The fuel is in sealed tungsten capsules. Lead extracted from thorium ores is used to cool the reactor. The electrical power of the reactor is 3.3 GW. The layout of the reactor is identical to the BREST-OD-300 reactor under construction. The analysis of emergency modes from among ATWS (anticipated transients without scram) is carried out. The reactor is reliable and safe. The maximum temperature of a high-temperature reactor coolant is close to the boiling point of lead. By the nature of the change in the maximum temperatures of the core components, the reactor occupies an intermediate position between a reactor with solid metallic fuel and a reactor with cermets based on UN-PuN and metal uranium nanopowder.
7
Nikitin, D. I., I. B. Polovov, and O. I. Rebrin. "ELECTROREFINING OF URANIUM ALLOYS CONTAINING PALLADIUM AND NEODYMIUM IN 3LiCl–2KCl–UCl<sub>3</sub> MELTS." Расплавы, no. 3 (May 1, 2023): 316–28. http://dx.doi.org/10.31857/s0235010623030052.
Full textAbstract:
The technology of pyrochemical processing of mixed nitride uranium-plutonium spent fuel, realizable at the experimental and demonstration energy complex of the site of the Siberian Chemical Plant, includes several operations with the ultimate goal of isolating the target fission products. It’s planned to use the electrofining of the products of the previous stage, metallized spent nuclear fuel, аs the penultimate stage of processing. It’s necessary to determine the processes and technological modes of electrolytic refining of alloys modeling the product of this stage of the processing module to implement electrolytic refining. This paper presents the results of electrofining of model alloys (simulating the raw materials of the stage of electrofining processing) on an enlarged laboratory electrolyzer. The initial parameters of uranium refining processes in melts based on 3LiCl–2KCl–UCl3 were determined earlier. The basic parameters of refining were the use of electrolyte 3LiCl–2KCl–UCl3 (10.1 wt % UCl3) and conducting experiments at 550°C. Uranium alloys containing palladium and neodymium were prepared by direct fusion of uranium metal, PdAP-1 grade palladium metal powders and neodymium metal (99.99%) in a medium of high-purity argon (99.998%). The data obtained showed that at a temperature of 550°C, cathode precipitates are typical dendritic forms of alpha-uranium in rhombic syngony with a tendency to needle formation with an increase in cathode current density. An increase in the company time and cathode current density leads to a decrease in the current output due to short-circuiting of the electrodes with cathode sediment needles or metal shedding from the cathode. The modes of the cathode process have been experimentally refined as a result of electrofining. When electrofining alloys U–Pd(1.59 wt %), U–Pd(1.62 wt %), U–Pd(1.54 wt %), U–Pd(1.58 wt %)–Nd(5.64 wt %), U–Pd(1.84 wt %)–Nd(6.49 wt %), U–Pd(1.79 wt %)–Nd(6.54 wt %), uranium cathode precipitates were obtained, which were subjected to chemical analysis, which showed the high purity of the resulting metallic uranium, as well as the absence of metallic palladium and molybdenum in it. The palladium purification coefficient exceeds 5000, the neodymium purification coefficient exceeds 1000, which meets the requirements for purification from fission products at this stage of pyrochemical processing of spent fuel. Palladium accumulates in anode slime, while the bulk of neodymium passes into the molten electrolyte.
8
Kuzina, Yu, D. Klinov, G. Mikhailov, A. Sorokin, and V. Alekseev. "COMPLEX OF EXPERIMENTAL FACILITIES FOR DESIGN AND SAFETY JUSTIFICATION OF FAST REACTORS WITH LIQUID METAL COOLANTS." PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS 2021, no. 4 (2021): 172–94. http://dx.doi.org/10.55176/2414-1038-2021-4-172-194.
Full textAbstract:
To substantiate the safety and characteristics of fast reactors with liquid metal coolants, a complex of more than 20 stands of various profiles and purposes, well equipped with modern measuring instruments, including hydrodynamic, thermohydraulic and technological stands, has been created at SSC RF - IPPE. In addition, JSC “SSC RF - IPPE” has a complex of fast physical stands, including two critical stands - BFS-1 and the world's largest physical stand BFS-2. The article presents the characteristics and the possibility of stands designed for research in the field of hydrodynamics, heat transfer and coolant technology in support of design solutions, safety improvement and testing of equipment elements and assemblies of operating and planned installations with fast reactors with sodium, lead and lead-bismuth coolants, as well as for accelerator-controlled systems and thermonuclear fusion, low-power nuclear power plants for space: - Hydrodynamic stands - “SGDI” (air), “V-2” (air), “SGI” (water), “V-200” (water), “GDK” (air). - Thermal-hydraulic liquid metal stands - “6B” (Na, Na-K), “AR-1” (Na, Na-K), “Pluton” (Na), “SPRUT” (Na, Na-K, Pb, Pb-Bi, water). - Technological liquid metal stands - “Protva-1” (Na), “Protva-2” (Na), “PUSHM” (Na), “Armatura” (Na), “IK-MT” (Na), “SID” (Na), “BTS” (Na), “TT-1M” (Pb), “TT-2M” (Pb-Bi), “LIS-M” (Li). A large-scale sodium test stands “SAZ” is under construction, which allows testing full-scale prototypes of equipment and its elements to substantiate existing and future projects of fast sodium reactors. The BFS complex of physical stands is the world's only experimental tool for full-scale modeling of the cores of nuclear reactors of various types (of any composition, geometry and configuration). The materials and construction of the stands allow simulating the core, breeding zones, reflectors and in-core shielding, as well as elements of fuel cycles and storage facilities for spent nuclear fuel and radioactive waste. Reactor materials of the stands (metallic plutonium, oxide and metallic highly enriched uranium with enrichment of 36% and 90% in uranium-235, hundreds of tons of fertile materials, construction materials, various coolants) make it possible to assemble both complex full-scale models of fast reactors, and benchmarks, experiments for which are carried out to correct neutron-physical constants and improve computational methods.
9
Heidet, F., and J. Roglans-Ribas. "CORE DESIGN ACTIVITIES OF THE VERSATILE TEST REACTOR – CONCEPTUAL PHASE." EPJ Web of Conferences 247 (2021): 01010. http://dx.doi.org/10.1051/epjconf/202124701010.
Full textAbstract:
The Versatile Test Reactor (VTR) is a new fast spectrum test reactor being developed in the United States under the direction of the US Department of Energy, Office of Nuclear Energy. The VTR mission is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. This includes neutron irradiation capabilities which would support alternate coolants including molten salt, lead/lead-bismuth eutectic mixture, gas, and sodium. The VTR aims at addressing most of the needs of the various stakeholders, which is primarily composed of advanced reactor technologists, developers and vendors, as well as a number of others interested parties. Design activities are underway targeting a first criticality date by 2026, with General Electric recently joining the project to contribute to the VTR plant design. Current efforts are focused on all aspects of the VTR design, with the core design being at the center of the initial steps. The VTR is currently proposed as a 300 MWth sodium-cooled fast reactor able to reach peak fast flux levels in excess of 4.0x1015 n/cm2-s (and total flux level of about 6.0x1015 n/cm2-s). In this configuration, it is using ternary metallic fuel with reactor-grade plutonium and 5% low-enriched uranium.
10
Bondarenko, G. G., G. S. Bulatov, K. N. Gedgovd, D. Yu Lyubimov, and M. M. Yakushkin. "Effect of the electron decay of metallic fission products on the chemical and phase compositions of an irradiated uranium-plutonium fuel." Russian Metallurgy (Metally) 2009, no. 5 (2009): 426–30. http://dx.doi.org/10.1134/s0036029509050115.
Full text
11
Khvostov, S. S., O. A. Golosov, E. V. Nikitina, E. A. Karfidov, N. V. Glushkova, and Yu P. Zaikov. "POSSIBILITIES OF NEUTRON ACTIVATION ANALYSIS FOR STUDYING THE CORROSION BEHAVIOR OF METALLIC MATERIALS IN MOLTEN SALTS." Расплавы, no. 6 (November 1, 2023): 644–51. http://dx.doi.org/10.31857/s0235010623060038.
Full textAbstract:
For the BREST-OD-300 reactor facility [1, 2], the technology for evaluating mixed nitride uranium-plutonium spent economical fuel is being determined [3–9]. To separate MNUP SNF from fuel claddings made of materials with high radiation resistance – ferritic-martensitic steel EP-823 [10–16], it is planned to use pyrometallurgical grades of “soft chlorination” [17]. When alloying and impurity elements of steel EP-823 are dissolved in molten salts of eutectic composition based on lithium and potassium chlorides, the melt will be contaminated. For the same reason, the formation of volatile compounds will occur, with their further mass transfer from hot to cold sections of process equipment. When studying the corrosion behavior of metals and alloys in liquid media, the problem often arises of determining small amounts of dissolution products in solution. This problem arises, for example, the rate of dissolution of microimpurities. The sensitivity of the usual, traditional methods used in corrosion testing such as mass loss or colorimetric determination of corrosion products in solution is often insufficient to make appropriate measurements. In these cases, the most effective is the use of the radiochemical method of neutron activation analysis based on. qualitative and quantitative determination of chemical elements, based on the measurement of the radiation characteristics of radionuclides formed during the irradiation of materials with neutrons. This paper presents the results of a study of the corrosion behavior and mass transfer of corrosion products of EP-823 steel pre-irradiated in the IVV-2M research nuclear reactor in molten salts 2KCl–3LiCl and 2KCl–3LiCl–PbCl2 at temperatures of 500 and 650°C for 24 h. It is shown that the method of neutron activation analysis can be used to study the corrosion behavior of EP-823 steel in molten salts of various compositions.
12
Bondarenko, G. G., G. S. Bulatov, K. N. Gedgovd, D. Yu Lyubimov, and M. M. Yakushkin. "Effect of the electron decay of metallic fission products on the chemical and phase compositions of an uranium-plutonium fuel irradiated by fast neutrons." Russian Metallurgy (Metally) 2011, no. 11 (2011): 1074–78. http://dx.doi.org/10.1134/s0036029511110036.
Full text
13
Shankar, A. Ravi, K. Thyagarajan, and U. Kamachi Mudali. "Corrosion Behavior of Candidate Materials in Molten LiCl-KCl Salt Under Argon Atmosphere." Corrosion 69, no. 7 (2013): 655–65. http://dx.doi.org/10.5006/0746.
Full textAbstract:
Pyrochemical reprocessing involves the use of molten LiCl-KCl (lithium chloride-potassium chloride) eutectic salt at 773 K for the recovery of uranium and plutonium from spent metallic fuel of fast breeder reactors. The materials selected for such corrosive environments should withstand high temperatures and at the same time offer good corrosion resistance. The present work discusses the corrosion behavior of candidate materials like 2.25Cr-1Mo steel (UNS K21590), 9Cr-1 Mo steel (UNS K90941), Ni-based alloy 600 (UNS N06600), Ni-based alloy 625 (UNS N06625), and Ni-based alloy 690 (UNS N06690) in molten LiCl-KCl eutectic salt at 873 K for various durations under ultrahigh-purity argon atmosphere. Corrosion behavior of partially stabilized zirconia (PSZ) coating on candidate materials also was evaluated. Weight-loss results indicated that the corrosion resistance of the materials increased in the following order: 2.25Cr-1Mo &gt; 9Cr-1 Mo &gt; Ni-based alloys &gt; PSZ coating. PSZ-coated specimens showed better corrosion resistance in molten LiCl-KCl salt when compared with uncoated specimens; however, accidental ingression of oxygen and moisture could result in premature spallation of the coating. Scanning electron microscopy (SEM) examination and grazing incidence x-ray diffraction (GIXRD) analysis of exposed Cr-Mo steels and Ni-based alloys exhibited dealloyed surfaces and corrosion product regions rich in Cr, indicating preferential leaching of Cr. The paper highlights the results of the present investigation.
14
Grabezhnoy, Vladimir A., Viktor A. Dulin, Vitaliy V. Dulin, and Gennady M. Mikhailov. "On the determination of neutron multiplication by the Rossi-alpha method." Nuclear Energy and Technology 7, no. 3 (2021): 253–57. http://dx.doi.org/10.3897/nucet.7.74156.
Full textAbstract:
Introduction. This work contains the results of determining the prompt neutron multiplication factor in the subcritical state of a one-core BFS facility, obtained by the neutron coincidence method, for which the influence of the error in the βeff in determining the multiplication factor turned out to be insignificant. The core of the facility consisted of rods filled with pellets of metallic depleted uranium, 37% enriched uranium dioxide and 95% enriched plutonium, sodium, stainless steel and Al2O3. Stainless steel served as a reflector. Methods. In contrast to the inverse kinetics equation solving (IKES) method, which is convenient for determining reactor subcritical states, the neutron coincidence method practically does not depend on the error in the value of the effective fraction of delayed neutrons βeff. If in the IKES method the reactivity value is obtained in fractions of βeff, i.e., from the measurement of delayed neutrons, the neutron coincidence method is based on the direct measurement of the value (1 – kσp)2, where is the effective multiplication factor by prompt neutrons. The total multiplication factor is defined as keff = kσp + βeff. If, for example, keff ≈ 0.9 (which is typical for determining the fuel burnup campaign), then it is the error in determining kσp that is the main one in comparison with the error in βeff. Thus, a 10% error in βeff of 0.003–0.004 (typical for plutonium breeders) will make a contribution to the error 1 – keff equal to 1 – kσp + βeff ≈ 0.00035, i.e., approximately 0.35%, but not 10%, as in the IKES method. Rossi-alpha measurements were carried out using two 3He counters and a time analyzer. The measurement channel width Δt was 1.0 μs. From these measurements, the value of the prompt neutron multiplication factor was obtained. In this case, the space-isotope correlation factor for the medium with a source was calculated using the following values: Φ(x) – solutions of the inhomogeneous equation for the neutron flux and Φ+(x) – solutions of the ajoint inhomogeneous equation. Results. The authors also present a comparison of the results of the Rossi-alpha experiment and measurements of the BFS-73 subcritical facility by the standard IKES method in determining the multiplication factor value. The data of the IKES method differ insignificantly from the results of the Rossi-alpha method over the entire range of changes in the subcriticality with an increase in the subcriticality of the BFS-73 one-core facility. Conclusion. It was impossible to apply the neutron coincidence method to fast reactors; however, the method turned out to be quite workable on their models created at the BFS facility, which was successfully demonstrated in this study.
15
Bondarenko, G. G., A. V. Androsov, G. S. Bulatov, K. N. Gedgovd, D. Yu Lyubimov та M. M. Yakunkin. "Effect of the β decay of metallic fission products on the chemical and phase compositions of the uranium–plutonium nitride nuclear fuel irradiated by fast neutrons". Russian Metallurgy (Metally) 2016, № 9 (2016): 879–83. http://dx.doi.org/10.1134/s0036029516090068.
Full text
16
Grabezhnoy, Vladimir A., Viktor A. Dulin, Vitaliy V. Dulin, and Gennady M. Mikhailov. "On the determination of neutron multiplication by the Rossi-alpha method." Nuclear Energy and Technology 7, no. (3) (2021): 253–57. https://doi.org/10.3897/nucet.7.74156.
Full textAbstract:
Introduction. This work contains the results of determining the prompt neutron multiplication factor in the subcritical state of a one-core BFS facility, obtained by the neutron coincidence method, for which the influence of the error in the β<sub>eff</sub> in determining the multiplication factor turned out to be insignificant. The core of the facility consisted of rods filled with pellets of metallic depleted uranium, 37% enriched uranium dioxide and 95% enriched plutonium, sodium, stainless steel and Al<sub>2</sub>O<sub>3</sub>. Stainless steel served as a reflector. Methods. In contrast to the inverse kinetics equation solving (IKES) method, which is convenient for determining reactor subcritical states, the neutron coincidence method practically does not depend on the error in the value of the effective fraction of delayed neutrons β<sub>eff</sub>. If in the IKES method the reactivity value is obtained in fractions of β<sub>eff</sub>, i.e., from the measurement of delayed neutrons, the neutron coincidence method is based on the direct measurement of the value (1 – k<sub>σ</sub><sub>p</sub>)<sup>2</sup>, where is the effective multiplication factor by prompt neutrons. The total multiplication factor is defined as k<sub>eff</sub> = k<sub>σ</sub><sub>p</sub> + β<sub>eff</sub>. If, for example, k<sub>eff</sub> ≈ 0.9 (which is typical for determining the fuel burnup campaign), then it is the error in determining k<sub>σ</sub><sub>p</sub> that is the main one in comparison with the error in β<sub>eff</sub>. Thus, a 10% error in β<sub>eff</sub> of 0.003–0.004 (typical for plutonium breeders) will make a contribution to the error 1 – k<sub>eff</sub> equal to 1 – k<sub>σ</sub><sub>p</sub> + β<sub>eff</sub> ≈ 0.00035, i.e., approximately 0.35%, but not 10%, as in the IKES method. Rossi-alpha measurements were carried out using two <sup>3</sup>He counters and a time analyzer. The measurement channel width Δt was 1.0 μs. From these measurements, the value of the prompt neutron multiplication factor was obtained. In this case, the space-isotope correlation factor for the medium with a source was calculated using the following values: Φ(x) – solutions of the inhomogeneous equation for the neutron flux and Φ<sup>+</sup>(x) – solutions of the ajoint inhomogeneous equation. Results. The authors also present a comparison of the results of the Rossi-alpha experiment and measurements of the BFS-73 subcritical facility by the standard IKES method in determining the multiplication factor value. The data of the IKES method differ insignificantly from the results of the Rossi-alpha method over the entire range of changes in the subcriticality with an increase in the subcriticality of the BFS-73 one-core facility. Conclusion. It was impossible to apply the neutron coincidence method to fast reactors; however, the method turned out to be quite workable on their models created at the BFS facility, which was successfully demonstrated in this study.
17
Hull, Gregory, Hugues Lambert, Kiran Haroon, et al. "Quantitative prediction of rare earth concentrations in salt matrices using laser-induced breakdown spectroscopy for application to molten salt reactors and pyroprocessing." Journal of Analytical Atomic Spectrometry, 2021. http://dx.doi.org/10.1039/d0ja00352b.
Full textAbstract:
Pyroprocessing of spent nuclear fuels is an electrochemical separation method where spent metallic fuel is dissolved in a molten salt bath to allow uranium (U) and plutonium (Pu) to be isolated from fission products (FPs) and other impurities.
18
Morss, Lester R., Carol J. Mertz, A. Jeremy Kropf, and Jennifer L. Holly. "Properties of Plutonium-Containing Colloids Released from Glass-Bonded Sodalite Nuclear Waste form." MRS Proceedings 713 (2002). http://dx.doi.org/10.1557/proc-713-jj6.6.
Full textAbstract:
ABSTRACTIn glass-bonded sodalite, which is the ceramic waste form (CWF) to immobilize radioactive electrorefiner salt from spent metallic reactor fuel, uranium and plutonium are found as 20-50 nm (U,Pu)O2 particles encapsulated in glass near glass-sodalite phase boundaries. In order to determine whether the (U,Pu)O2 affects the durability of the CWF, and to determine release behavior of uranium and plutonium during CWF corrosion, tests were conducted to measure the release of matrix and radioactive elements from crushed CWF samples into water and the properties of released plutonium. Released colloids have been characterized by sequential filtration of test solutions followed by elemental analysis, dynamic light scattering, transmission electron microscopy (TEM), and X-ray absorption spectroscopy. This paper reports the composition, size, and agglomeration of these colloids. Significant amounts of colloidal, amorphous aluminosilicates and smaller amounts of colloidal crystalline (U,Pu)O2 were identified in test solutions. The normalized releases of uranium and plutonium were significantly less than the normalized releases of matrix elements, i.e., the CWF retains these radionuclides well.
19
"The fast-neutron breeder fission reactor: safety issues in reactor design and operation." Philosophical Transactions of the Royal Society of London. Series A, Mathematical and Physical Sciences 331, no. 1619 (1990): 409–18. http://dx.doi.org/10.1098/rsta.1990.0078.
Full textAbstract:
Today’s fast breeder reactors contain mixed uranium —plutonium oxide fuel and are cooled with liquid sodium. Their normal operational behaviour, including power transients, is similar to that of thermal reactors. The fact that the sodium density coefficient is positive is of no importance at normal operating temperatures because negative coefficients like Doppler or fuel expansion coefficients have compensating effects. Dangerous effects may arise when sodium boiling at much higher temperatures occur. It is shown that sodium boiling in most cases can be avoided by proper design of the reactor core. Energy releases associated with partial destruction of the core are discussed. The safety features of metallic fuel are briefly discussed, resulting in the statement that in principle, use of metallic fuel does not promise more positive safety features.
20
Iglesias, Luis, Jakub Kokinda, Daniel Serrano-Purroy, et al. "Dissolution of high burn-up spent nuclear fuel at high-pH." Radiochimica Acta, October 9, 2023. http://dx.doi.org/10.1515/ract-2023-0178.
Full textAbstract:
Abstract The release of radionuclides from high-burnup spent nuclear fuel (SNF) segments was studied at pH = 13.2 as well as the effect of the presence of calcium and silicon. The aim was to ascertain the dissolution of SNF in solutions corresponding to a high-level nuclear waste repository including concrete in different structural parts. The release of uranium at pH = 13.2 was higher than at pH = 8.4 in bicarbonate medium, while the presence of calcium resulted in a decrease of the uranium concentrations in solutions, assumed to be the consequence of the formation of a secondary solid phase such as Ca2U2O7. Caesium release was found higher at pH = 13.2 as well, but it was not influenced by the presence of Ca and Si at long term. On the other hand, actinide elements (plutonium, neptunium and americium) dissolution decreased at pH = 13.2, probably because of the formation of secondary solid phases. On the contrary, ruthenium and technetium release at pH = 13.2 was found to be much higher than the measured at lower pH, perhaps due to the higher dissolution kinetics of the metallic inclusions at such pH.
21
Gombert, Dirk, Joe Carter, Bill Ebert, Steve Piet, Tim Trickel, and John Vienna. "A Trade Study for Waste Concepts to Minimize HLW Volume." MRS Proceedings 1124 (2008). http://dx.doi.org/10.1557/proc-1124-q01-03.
Full textAbstract:
AbstractAdvanced nuclear fuel reprocessing can partition wastes into groups of common chemistry. This enables new waste management strategies not possible with the plutonium, uranium extraction (PUREX) process alone. Combining all of the metallic fission products in an alloy and the balance as oxides in glass minimizes high level waste (HLW) volume. Implementing a waste management strategy using state-of-the-art combined waste forms and storage to allow radioactive decay and heat dissipation prior to placement in a repository makes it possible to place almost 10x the HLW equivalent of spent nuclear fuel (SNF) in the same repository space. However, using generic costs based on preliminary studies for waste stabilization facilities and separations modules, this analysis shows that combining the non-actinide wastes and using only one glass waste form is the most cost-effective.
22
Sankar, Dwarapudi Bola, Rajeswari Seshadri, Kalaiyarasu Thirunavukkarasu, et al. "Radiochemical and chemical characterization of fuel, salt, and deposit from the electrorefining of irradiated U-6 wt% Zr in hot cells." Radiochimica Acta, February 14, 2024. http://dx.doi.org/10.1515/ract-2023-0203.
Full textAbstract:
Abstract Metal fuels are considered as the promising candidates for future fast breeder reactors. Pyro-chemical reprocessing is the ideal method for reprocessing spent metallic fuels due to the inherent process advantages. Electrorefining run was demonstrated in a hot cell facility with irradiated U-6 wt% Zr alloy at 500 °C using LiCl–KCl eutectic melt. In order to understand the behavior of the actinides and various fission products during high-temperature electrolysis, various process streams, viz., irradiated metal alloy fuel, the eutectic salt, and the cathode deposit were analyzed for the uranium, plutonium, and other fission product contents. Various methods employed for characterizing the process streams and the behaviors of some of the fission products during the electrolysis process are highlighted. The major gamma emitting radionuclides present in the irradiated fuel were 106Ru, 125Sb, 134Cs, 137Cs, 144Ce, and 154Eu. During electrorefining, cesium, cerium and europium were oxidized and dissolved in the molten media, whereas ruthenium and antimony remained in the anode basket. A minor contamination of zirconium was found in the cathode deposit.
23
Chenniappan, Thiagarajan, and Yuvarajan Devarajan. "Comprehensive review of surface contamination in nuclear waste waters: identification, quantification, and mitigation strategies." Kerntechnik, July 26, 2024. http://dx.doi.org/10.1515/kern-2024-0070.
Full textAbstract:
Abstract The safety and reliability of nuclear facilities hinge critically on addressing metallic surface contamination in nuclear waste waters. This contamination poses significant hazards to the environment, human health, and the structural integrity of equipment. Key contaminants include heavy metals such as lead, cadmium, and mercury from industrial processes, and radioactive isotopes like uranium, plutonium, and cesium, which present severe radiological risks due to their formation during nuclear reactions and fuel cycles. Corrosive chemicals further exacerbate the problem by promoting the accumulation of rust and other metallic compounds. Additionally, organic contaminants from equipment leaks and microbiological elements, including fungi and bacteria, can form biofilms that accelerate the corrosion process. The objective of this review is to evaluate the various techniques used to identify and quantify these contaminants on metal surfaces, such as surface sampling and microbiological analysis. By implementing appropriate mitigation measures based on these findings, it is possible to reduce risks and ensure the safety and operational integrity of nuclear plants. This comprehensive assessment aims to provide a framework for enhancing contamination management practices in nuclear facilities.
24
Johnson, S. G., M. Noy, T. DiSanto, and T. L. Barber. "Release of Neptunium, Plutonium, Uranium and Technetium from the Metallic Waste form from the Electrometallurgical Treatment Process." MRS Proceedings 713 (2002). http://dx.doi.org/10.1557/proc-713-jj11.73.
Full textAbstract:
ABSTRACTThis waste form is an alloy consisting of stainless steel with 15 wt% zirconium and acts as a host for the immobilization of radioelements that remain with the spent fuel cladding hulls following their treatment using an electrometallurgical treatment process. The results presented here are from 14, 34 and 90-day immersion tests conducted at 90 °C. These tests show that the release of uranium is considerably higher than that of all other major elements present (Fe, Cr, Ni, Zr), but that release of all constituents is comparable to or lower than that for borosilicate glass.
25
Yudintsev, Sergey V., Tatiana S. Ioudintseva, Andrey V. Mokhov, et al. "Study of Pyrochlore and Garnet-based Matrices for Actinide Wastes Produced by a Self-propagating High-temperature Synthesis." MRS Proceedings 807 (2003). http://dx.doi.org/10.1557/proc-807-273.
Full textAbstract:
ABSTRACTActinide-containing wastes are among the most dangerous for the environment. Such waste streams originate from reprocessing operations with irradiated nuclear fuel and conversion of weapons-grade plutonium metal into dioxide. The long-term toxicity of actinides derives from the presence of isotopes with half-life varying from hundreds of years (Am241) to tens of thousands (Pu239) or even millions of years (Np237). Therefore, these waste fractions need to be incorporated into durable crystalline host phases. The matrices have to incorporate substantial amounts of actinides, and possess chemical durability and resistance to radiation damage. Complex oxides with fluorite-derived and garnet lattices meet these requirements. Self-propagating high-temperature synthesis (SHS) based on exothermic oxidizing-reduction reactions may be used for production of these waste forms. This technology has the following advantages: absence of extrinsic heating sources, low energy requirements for equipment, high reaction velocity, simplicity of design of processing equipment, feasibility of remote-handling the processes, and lack of considerable amounts of facility decommission wastes. The basic features of the SHS technology are as follows: duration of initiation is 0.05–5.0 sec, temperature in a combustion wave is within 1500–3000 °K, and the velocity of advance of the combustion wave is 1–150 mm/s. Two sets of samples composed of pyrochlore and garnet-type phases were produced with SHS. The first of them corresponds to nominal pyrochlore formulation Y2Ti2O7 doped with various amounts of actinides: 10–30 wt.% UO2, 10 wt % PuO2, 10 wt % NpO2, or 9.5 wt % UO2 + 0.5 wt % Am2O3. The precursor was prepared from oxides of the base phases (TiO2, Y2O3, AnO2), an oxidizer (MoO3), and Ti. In the second set of runs, the target phase was garnet (Y2.8Gd0.2)(Al4.7Ga0.3)O12, where Gd3+ was used as a surrogate for Am3+. The initial batches were composed of MoO3, Y2O3, Gd2O3, Al2O3, Ga2O3, and metallic Al. Phase compositions of the samples have been determined by XRD and SEM/EDS. Samples of the first series are composed of major pyrochlore with minor metallic Mo. The samples with 10 wt % actinides do not contain any separate actinide oxide phase. In the samples with 20 and 30 % UO2 a separate uranium oxide phase was observed. SEM/EDS data allows determination of the limit of solid solution of the pyrochlore phase with respect to tetravalent actinide (U) as 12–14 wt. %. Principle phases in the second series were garnet:(Y2.82–2.88Gd0.13–0.14)(Al4.69–4.74Ga0.17–0.22Mo0.05–0.16)O12 and Mo-Al-Ga alloy. Small amounts of perovskite - (Y0.86Gd0.12)(Al0.93Ga0.05Mo0.03)O3, Mo, and Al oxides were also observed. Gd and Ga mainly entered in the garnet; small amounts of the elements were incorporated into perovskite (Gd), a metallic alloy, and perovskite (Ga).