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The 1999 evaluation for A = 125 mass chain (1999Ka26) has been revised using experimental results from decays and reactions. Adopted values for the level and decay properties are tabulated. Inconsistencies and discrepancies are noted.
Generally, passive gamma spectroscopy (PGS) is known as a useful technique for nondestructive measurement of burnup and nuclide production in normal spent nuclear fuel. However, PGS can be used not only for normal spent fuel, but also for molten fuel generated in severe accidents involving core meltdowns. In such cases, it is estimated by the radioactivity ratios of specific nuclides. At the time of the decommissioning of the Three Mile Island's unit 2 reactor, fission products (FPs) such as Cs{sup 134}, Cs{sup 137}, Ce{sup 144}, Eu{sup 154}, Ru{sup 106}, and Sb{sup 125} were concerned. The radioactivity ratios are estimated by numerical simulations. However, there are various uncertainties in numerical simulations. In this study, we evaluated the effect of the difference in fission yield data. We compared the productions of concerned nuclides of Three Mile Island's unit 2 reactor by using Tokyo-tech and JENDL FP Fission Yields Data File 2011(FPY- 2011). Tokyo-tech is newly created by Tsubakihara et al. at Tokyo Institute of Technology. FPY-2011 is currently the latest fission yield data in Japan. In the concerned nuclides, the relative difference (Tokyo-tech/FPY-2011 -1) in the production of Sb{sup 125} is 4.5% and that of Eu{sup 154} is -5.6%. The relative difference in the production of Sb{sup 125} is found to be caused by the difference in the independent yields of Sn{sup 125}, In{sup 125}, Cd{sup 125}, Ag{sup 125}, and Pd{sup 125}. The relative difference in Eu{sup 154} is found to be caused by the relative difference in the production of Sm{sup 152} and the relative difference in the cumulative yield of Eu{sup 153}.
Experimentally measured isotopic concentrations of well characterized spent nuclear fuel (SNF) samples have been collected and analyzed by previous researchers. These sets of experimental data have been used extensively to validate the accuracy of depletion code predictions for given sets of burnups, initial enrichments, and varying power histories for different reactor types. The purpose of this report is to present the diversity of data in a concise manner and summarize the current accuracy of depletion modeling. All calculations performed for this report were done using the Oak Ridge Isotope GENeration (ORIGEN) code, an internationally used irradiation and decay code solver within the SCALE comprehensive modeling and simulation code. The diversity of data given in this report includes key actinides, stable fission products, and radioactive fission products. In general, when using the current ENDF/B-VII.0 nuclear data libraries in SCALE, the major actinides are predicted to within 5% of the measured values. Large improvements were seen for several of the curium isotopes when using improved cross section data found in evaluated nuclear data file ENDF/B-VII.0 as compared to ENDF/B-V-based results. The impact of the flux spectrum on the plutonium isotope concentrations as a function of burnup was also shown. The general accuracy noted for the actinide samples for reactor types with burnups greater than 5,000 MWd/MTU was not observed for the low-burnup Hanford B samples. More work is needed in understanding these large discrepancies. The stable neodymium and samarium isotopes were predicted to within a few percent of the measured values. Large improvements were seen in prediction for a few of the samarium isotopes when using the ENDF/B-VII.0 libraries compared to results obtained with ENDF/B-V libraries. Very accurate predictions were obtained for 133 Cs and 153 Eu. However, the predicted values for the stable ruthenium and rhodium isotopes varied as much as 10% of the measured values, and 109 Ag was consistently over-predicted by as much as 170%. In general, there is a larger uncertainty for modeling radioactive fission products when compared to either the actinides or the stable fission products in SNF. The relative C/E ratios ranged from a few percent for 137 Cs up to 60% and 100% for 106 Ru and 125 Sb, respectively. Limited or no radioactive fission products data exist in the current data sets for reactor types other than PWRs and BWRs. More work is needed in obtaining a greater diversity of radioactive fission product data. While performing this survey, issues leading to inconsistencies in nuclear fission yield data were discovered that specifically impacted the fission product noble gases. Emphasis was given to this legacy data, and corrective actions were taken as described in this report. After the fission yield data were corrected, the stable xenon and krypton fission products were predicted to within 5% of their measurements. However, preliminary results not explicitly given in this report indicate that the relative C/E ratio for the radioactive isotope 85 Kr varied as much as 10%. Due to the complex migration and the difficulty in measuring noble gases in the fuel, a more thorough investigation is needed to understand how accurately depletion codes can calculate these gas concentrations.
The United States produces a large amount of cladding material from used nuclear fuel, expected to approach 1,000 MT/year for the next 50 years. The recovery of zirconium from the cladding material would represent a ∼50% savings on the geologic waste disposal requirement, and savings of >$40 M/year due to the cost of hafnium-free zirconium. Previous simulation studies have shown that in the irradiated cladding, the major activation products found are {sup 125}Sb, {sup 60}Co, and {sup 93m}Nb, and the major fission product found is {sup 137}Cs. These radioisotopes make it impossible to treat the cladding as low level waste. In a chlorination method, zirconium cladding reacts with Cl{sub 2} at 350-380 deg. C to produce mainly ZrCl{sub 4}. The zirconium tetrachloride has a relatively low boiling point (sublimes at 330 deg. C) which allows its separation from other products, and so it can be treated as low level waste for disposal or recycling. This chlorination approach has been proposed for the reprocessing of zirconium. However, previous work has observed element impurities (Fe, Sn, Cr, Cs, Sb, and Nb) in the recovered ZrCl{sub 4} meaning that additional steps are necessary in a chlorination process previous to the reduction of ZrCl{sub 4} to the metallic form. The chlorination process for zirconium reprocessing has been highly encouraged as it has proven more efficient than previous mechanical processes, redox processes, and other chemical methods such as iodination. The overall goal of this project is to develop the road map for a chlorination process that will yield high purity zirconium. The current study focuses on the recovery of ZrCl{sub 4} from non-irradiated alloys used as nuclear fuel cladding. (authors)
Destructive post-irradiation examination was performed on AGR-1 fuel Compact 4-1-1, which was irradiated to a final compact-average burnup of 19.4% FIMA (fissions per initial metal atom) and a time-average, volume-average temperature of 1072°C. The analysis of this compact focused on characterizing the extent of fission product release from the particles and examining particles to determine the condition of the kernels and coating layers. The work included deconsolidation of the compact and leach-burn-leach analysis, visual inspection and gamma counting of individual particles, metallurgical preparation of selected particles, and examination of particle cross-sections with optical microscopy, electron microscopy, and elemental analysis. Deconsolidation-leach-burn-leach (DLBL) analysis revealed no particles with failed TRISO or failed SiC layers (as indicated by very low uranium inventory in all of the leach solutions). The total fractions of the predicted compact inventories of fission products Ce-144, Cs-134, Cs-137, and Sr-90 that were present in the compact outside of the SiC layers were <2×10 -6 , based on DLBL data. The Ag-110m fraction in the compact outside the SiC layers was 3.3×10 -2 , indicating appreciable release of silver through the intact coatings and subsequent retention in the OPyC layers or matrix. The Eu-154 fraction was 2.4×10 -4 , which is equivalent to the inventory in one average particle, and indicates a small but measurable level of release from the intact coatings. Gamma counting of 61 individual particles indicated no particles with anomalously low fission product retention. The average ratio of measured inventory to calculated inventory was close to a value of 1.0 for several fission product isotopes (Ce-144, Cs-134, and Cs-137), indicating good retention and reasonably good agreement with the predicted inventories. Measured-to-calculated (M/C) activity ratios for fission products Eu-154, Eu-155, Ru-106, Sb-125, and Zr-95 were significantly less than 1.0. However, as no significant release of these fission products from compacts was noted during previous analysis of the AGR-1 capsule components, the low M/C ratios are most likely an indication of a bias in the inventories predicted by physics simulations of the AGR-1 experiment. The distribution of Ag-110m M/C ratios was centered on a value of 1.02 and was fairly broad (standard deviation of 0.18, with values as high as 1.42 and as low as 0.68). Based on all data gathered to date, it is believed that silver retention in the particles was on average relatively high, but that the broad distribution in values among the particles represents significant variation in the inventory of Ag-110m generated in the particles. Ceramographic analysis of particle cross-sections revealed many of the characteristic microstructures often observed in irradiated AGR-1 particles from other fuel compacts. Palladium-rich fission product clusters were observed in the IPyC and SiC layers near the IPyC-SiC interface of three Compact 4-1-1 particle cross-sections. In spite of the presence of fission product clusters in the SiC layer, no significant corrosion or degradation of the layer was observed in any of the particles examined.
The earthquake and tsunami on March 11, 2011 severely damaged the Fukushima Dai-ichi nuclear plant leading to the most severe nuclear incident since Chernobyl. Ongoing operations to cool the damaged reactors at the site have led to the generation of highly radioactive coolant water. This is currently mainly treated to remove Cs-137 and Cs-134 and passed through a reverse osmosis (RO) unit to reduce the salinity before being cycled back to the reactors. Because only the Cs isotopes are removed, the RO reject water still contains many radioactive isotopes and this has led to the accumulation of over 200,000 cubic meters (52 million gallons) of extremely contaminated water which is currently stored on site in tanks. EnergySolutions, in partnership with Toshiba, were contracted to develop a system to reduce 62 isotopes in this waste down to allowable levels. This was a significant technical challenge given the high background salt content of the wastewater, the variation in aqueous chemistry of the radioactive isotopes and the presence of non-active competing ions (e.g. Ca and Mg) which inhibit the removal of isotopes such as Sr-89 and Sr-90. Extensive testing was performed to design a suitable system that could meet the required decontamination goals. These tests were performed over a 6 month period at facilities available in the nearby Fukushima Dai-ni laboratory using actual waste samples. This data was then utilized to design a Multi Radioactive Nuclides Removal System (MRRS) for Fukushima which is a modified version of EnergySolutions' proprietary Advanced Liquid Processing System (ALPS)'. The stored tank waste is fed into a preliminary precipitation system where iron flocculation is performed to remove a number of isotopes, including Sb-125, Ru-106, Mn-54 and Co-60. The supernatant is then fed into a second precipitation tank where the pH is adjusted and the bulk of the Mg, Ca and Sr precipitated out as carbonates and hydroxides. After passing through a cross-flow ultrafiltration membrane, the permeate then goes through a total of 14 fixed ion exchange and adsorbent columns followed by a disposable polishing column to polish the residual isotopes down to allowable levels. At the end of the system, the effluent is filtered for a final time to removal any particulates that may have been picked up from the media columns and then stored prior to analysis. (authors)
Directional correlation of first forbidden beta group and gamma ray in decay of antimony at four beta energies
Thermoelectric materials utilize the Seebeck effect to convert heat to electrical energy. The Seebeck coefficient (thermopower), S, depends on the free (mobile) carrier concentration, n, and effective mass, m*, as S ~ m*/n 2/3 . The carrier concentration in tellurides can be derived from 125 Te nuclear magnetic resonance (NMR) spin-lattice relaxation measurements. The NMR spin-lattice relaxation rate, 1/T 1 , depends on both n and m* as 1/T 1 ~(m*) 3/2 n (within classical Maxwell-Boltzmann statistics) or as 1/T1~(m*) 2 n 2/3 (within quantum Fermi-Dirac statistics), which challenges the correct determination of the carrier concentration in some materials by NMR. Here it is shown that the combination of the Seebeck coefficient and 125 Te NMR spin-lattice relaxation measurements in complex tellurides provides a unique opportunity to derive the carrier effective mass and then to calculate the carrier concentration. This approach was used to study Ag x Sb x Ge 50–2x Te 50 , well-known GeTe-based high-efficiency tellurium-antimony-germanium-silver thermoelectric materials, where the replacement of Ge by [Ag+Sb] results in significant enhancement of the Seebeck coefficient. Thus, values of both m* and n derived using this combination show that the enhancement of thermopower can be attributed primarily to an increase of the carrier effective mass and partially to a decrease of the carrier concentration when the [Ag+Sb] content increases.
This report documents the work done under NEUP grant to examine the capability of novel chalcogels and some binary metal chalcogenides as a host matrix for the capture of gaseous iodine and the feasibility of their iodine-laden materials to be converted into a permanent waste form. The presented work was conducted over last two years. A number of novel chalcogels Zn 2 Sn 2 S 6 , Sb 4 Sn 4 S 12 , NiMoS 4 , CoMoS 4 , antimony sulfide (SbS x ) chalcogels, silver functionalized chalcogels and binary metal sulfides (Sb 2 S 3 ) were developed and studies for their iodine absorption efficacies. A new and simple route was devised for the large scale preparation of antimony sulfide chalcogel. The chalcogel was obtained by treating Sb 2 S 3 with Na 2 S in the presence of water followed by addition of formamide. The obtained gels have a low-density sponge like network of meso porous nature having BET surface area of 125 m 2 /g. The chalcogels, silver functionalized chalcogel and the binary metal sulfides were exposed to iodine vapors in a closed container. Silver-functionalized chalcogels and Sb 2 S 3 powders showed iodine uptake up to 100 wt%, the highest iodine uptake of 200 wt% was observed for the SbS-III chalcogel. The PXRD patterns of iodine-laden specimens revealed that iodine shows spontaneous chemisorption to the matrix used. The iodine loaded chalcogels and the binary chalcogenides were sealed under vacuum in fused silica ampoules and heated in a temperature controlled furnace. The consolidated products were analyzed by PXRD, energy dispersive spectroscopy (EDS), UV-Vis and Raman spectroscopy. The final products were found to be amorphous in most of the cases with high amount (~4-35 wt%) of iodine and aapproximately ~60- 90 % of the absorbed iodine could be consolidated into the final waste form. Alginate reinforced composite scaffolds with SbS/SnS chalcogels and Sb 2 S 3 bulk powder were also fabricated aiming to study their efficacy as host matrices in capturing the gaseous molecular iodine in dynamic mode from nuclear spent fuel. The obtained composites looks robust in comparison to their respective pristine chalcogels and Sb 2 S 3 bulk powder.
The United States produces over 2000 metric tons of used nuclear fuel material per year, of which 25% is irradiated fuel cladding, and it is expected to increase for the next 50 years. Previous simulation studies have shown that in the irradiated cladding, the major activation products found are {sup 125}Sb, {sup 60}Co, and {sup 99m}Nb, and the major fission product found is {sup 137}Cs. These radioisotopes make it impossible to treat the cladding as low level waste. A chlorination process has been proposed for zirconium reprocessing as it has been proven more efficient than previous mechanical processes, redox processes, and other chemical methods (e.g. iodination). In the chlorination process, zirconium alloy cladding reacts with Cl{sub 2} at 350- 380 deg. C to produce mainly ZrCl{sub 4}. The zirconium tetrachloride has a relatively low sublimation point (331 deg. C) that allows its separation from other products, and so it can be treated as low level waste for disposal or recycling. However, previous work has observed element impurities (iron, tin, chromium, cesium, antimony, and niobium) in the recovered ZrCl{sub 4}, meaning that additional steps are necessary in a chlorination process previous to the reduction of ZrCl, to metallic Zr. The emphasis of this research summary is on the progress that has been made to isolate and characterize the impurities produced after chlorination of three previously characterized non-irradiated zirconium alloys (Zircaloy-2, Zircaloy-4, and a commercial Zr-Nb alloy) widely used as cladding materials.
We have investigated the carrier spin relaxation in GaInNAsSb/GaNAsSb/GaAs quantum well (QW) by time-resolved photoluminescence (PL) measurement. The sample consists of an 8-nm-thick GaIn{sub 0.36}N{sub 0.006}AsSb{sub 0.015} well, 5-nm-thick GaN{sub 0.01}AsSb{sub 0.11} intermediate barriers and 100-nm-thick GaAs barriers grown by molecular beam epitaxy on a GaAs(100) substrate. The spin relaxation time and recombination lifetime at 10 K are measured to be 228 ps and 151 ps, respectively. As a reference, we have also obtained a spin relaxation time of 125 ps and a recombination lifetime of 63 ps for GaInNAs/GaNAs/GaAs QW. This result shows that crystal quality is slightly improved by adding Sb, although these short carrier lifetimes mainly originate from a nonradiative recombination. These spin relaxation times are longer than the 36 ps spin relaxation time of InGaAs/InP QWs and shorter than the 2 ns spin relaxation time of GaInNAs/GaAs QW.
Excellent structural stability, high operating voltage, and high capacity have made Na 3 V 2 (PO 4 ) 2 F 3 a promising cathode material for sodium-ion batteries. However, high-temperature battery performances and heat generation measurements have not been systematically reported yet. Carbon-coated Na 3 V 2 (PO 4 ) 2 F 3 @MWCNT (multi-walled carbon nanotube) samples are fabricated by a hydrothermal-assisted sol-gel method and the electrochemical performances are evaluated at three different temperatures (25, 45, and 55 °C). Furthermore, the well-crystallized Na 3 V 2 (PO 4 ) 2 F 3 @MWCNT samples exhibit good cycling stability at both low and high temperatures; they deliver an initial discharge capacity of 120–125 mAhg -1 at a 1 C rate with a retention of 53 % capacity after 1,400 cycles with 99 % columbic efficiency. The half-cell delivers a capacity of 100 mAhg -1 even at a high rate of 10 C at room temperature. Furthermore, the Na 3 V 2 (PO 4 ) 2 F 3 @MWCNT samples show good long-term durability; the capacity loss is an average of 0.05 % per cycle at a 1 C rate at 55 °C. Additionally, ionic diffusivity and charge transfer resistance are evaluated as functions of state of charge, and they explain the high electrochemical performance of the Na 3 V 2 (PO 4 ) 2 F 3 @MWCNT samples. In-situ heat generation measurements reveal reversible results upon cycling owing to the high structural stability of the material. Excellent electrochemical performances are also demonstrated in the full-cell configuration with hard carbon as well as antimony Sb/C anodes.