The Containment of Fission Product Iodine in the Reprocessing of Lmfbr Fuels by Pyrochemical Reactions

Krumpelt, M.; Heiberger, J.J.; Maroni, V.A.; Steindler, M. J.

doi:10.13182/nt72-a16036

Cited by 3 publications

(3 citation statements)

References 3 publications

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“…On the other hand, mass spectrometry can rapidly detect elemental content at the ng/g level, with short measurement times and high e ciency for each sample. According to research ndings from Krumpelt et al [22] at the Argonne National Laboratory and research reports from the Idaho National Laboratory [23], the quantity of ssion product iodine in actual spent fuel is extremely low (at the µg/g level). Most of it exists as the I − anion in molten salts.…”

Section: Introductionmentioning

confidence: 99%

Rapid determination of iodide ion content in chloride molten salt by ascorbic acid reduction

Cheng,

Luo,

Geng

et al. 2024

Preprint

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Acidic conditions are typically used to dissolve fission product cations from the molten spent fuel. However, it is unfeasible to analysis iodide anion due to its memory effect. Proper pretreatment of the samples is therefore necessary. In the present work, an alternative method was developed to evaluate the efficacy of removing fission product iodine from molten salt spent nuclear fuel under acidic conditions. In the pretreatment stage, ascorbic acid with a mass fraction of 0.5wt% was used instead of dilute nitric acid with a volume fraction less than 2%. The iodide content in molten salt samples dissolved in different Cl/I concentration ratios was determined by ICP-MS and compared with ion chromatography(IC). The results showed that under 1 mg/L chloride ion concentration, the detection limit of this method was 0.255 µg/L, with a spike-and-recovery experience between 103.3% and 102.8% and a relative standard deviation of less than 3% (n = 6). When the concentration ratio of Cl/I in the molten salt was less than two orders of magnitude, the analytical value of this technique was comparable to the measurement result by IC. Furthermore, the accuracy of ICP-MS was superior to that of IC when the concentration ratio of Cl/I was higher than two orders of magnitude. Our results indicate that the improved pretreatment method with ascorbic acid can eliminate inaccuracies caused by the iodide memory effect, which is simple to operate and suitable for determining iodide content in chloride molten salt under acidic conditions.

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Section: Introductionmentioning

confidence: 99%

Rapid determination of iodide ion content in chloride molten salt by ascorbic acid reduction

Cheng,

Luo,

Geng

et al. 2024

Preprint

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“…Behavior of iodine in the head-end process was studied. Iodine is quantitatively taken up by the cover salt used in decladding and oxide reduction [30]. Two pyrochemical processes have been developed as production operations.…”

Section: Introductionmentioning

confidence: 99%

Pyrochemical Processing of LWR and L M F B R Fuels by the Salt Transport Method

Knighton

1978

Radiochimica Acta

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The Salt Transport Pyrochemical Process is being evaluated for coprocessing uranium and plutonium in spent LWR and LMFBR oxide fuel. The main objective is to develop a proliferation resistant process for recycle of uranium and plutonium. Diversion and proliferation resistance against weapons is enhanced by leaving part of the fission product elements with the uranium-plutonium product. The cladding (zircaloy or stainless steel) is removed from the oxide fuel by dissolution in molten zinc. The fuel oxides are then reduced by calcium metal in the presence of a Cu-Mg alloy and a CaCl 2 -CaF 2 salt. The subsequent fission product separations from uranium and plutonium are achieved by a series of molten salt-metal equilibrations. Vacuum distillation is used to recover the volatile solvent metals for recycle. The uranium and plutonium product with residual fission products is subsequently converted to oxide fuel. The chemistry of the Salt Transport step is described. The integration of the Salt Transport step into a conceptual flow sheet is described.

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“…In BREST-OD-300 reactor fuel elements, the gap between the fuel pellets and the steel jacket is filled with lead. There are several known methods for separating the stainless-steel jacket from ceramic pellets: mechanical cutting [1], melting steel at 1500 °C [2], and dissolving steel jackets in low-melting metal melts [3]. Mechanical cutting of fuel elements filled with congealed lead will not separate the fuel and the jacket.…”

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confidence: 99%

Extraction of Uranium Nitride Pellets During Extraction of Lead from Brest Reactor Fuel-Element Simulators by Melting

et al. 2005

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Spent fuel from the BREST-OD-300 fast reactor, which is now under development, is supposed to be reprocessed directly at the nuclear power plant by a water-free technology. The first operation is separation of the fuel (uranium and plutonium mononitride pellets) from the jacket. In BREST-OD-300 reactor fuel elements, the gap between the fuel pellets and the steel jacket is filled with lead. There are several known methods for separating the stainless-steel jacket from ceramic pellets: mechanical cutting [1], melting steel at 1500 °C [2], and dissolving steel jackets in low-melting metal melts [3]. Mechanical cutting of fuel elements filled with congealed lead will not separate the fuel and the jacket. Melting a steel jacket requires high temperatures and, aside from large amounts of energy and technical difficulties, it will lead to vaporization of the lead and premature destruction of the fuel. Dissolving in zinc will complicate the technology for opening fuel elements and will result in additional wastes.A method of separating the jacket from the fuel, including opening up of the fuel elements from the ends and extracting the pellets together with lead to be melted out by heating to 500°C, has been checked experimentally. Fuel-element simulators have been developed for check this method (Fig. 1). Fragments of simulators with different fill height were smelted. The extraction conditions, the completeness of the separation of the liquid lead and the pellets on filtering materials, and the degree to which the lead adheres to the surface of the steel tube and uranium mononitride pellets, were determined. Jackets, 9.4 × 0.4 in diameter, 7 × 0.3 mm thick, and 1101 mm long, made of ÉI-847 austenitic steel and ÉP-823 class ferrite martensite steel, were used for the experimental samples. The uranium (natural enrichment) mononitride pellet diameter and height were 7.9 × 12 and 5.8 × 10 mm. The gap between the fuel-element jacket and the fuel column was filled, on a special stand, with melted lead after pre-evacuation. This method gave a dense sublayer, and there was no discontinuity of the column of pellets and lead between them at the level of rsolution of the methods of analysis used. The filling of the fuel-element samples with lead was monitored by x-ray defectoscopy and eddy currents, whose sensitivity in detecting defects has been confirmed by investigations of the kernel by a destructive method. The samples were cut into fragments on screw-cutting lathe with a hard-allow cutter.A laboratory setup was designed, built, and assembled to perform investigations of the smelting. The setup contains a furnace-heated receptacle for the lead melted out and the pellets with a filtering barrier to separate them (nonrusting wire grid with 3.5 × 3.5 mm cell size), a tube heated with a separate furnace for the fuel-element simulator setup, a tube with coaxially installed rod for pushing the pellets out with a massive weight secured at the end, and an apparatus for signalling that

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The Containment of Fission Product Iodine in the Reprocessing of Lmfbr Fuels by Pyrochemical Reactions

Cited by 3 publications

References 3 publications

Rapid determination of iodide ion content in chloride molten salt by ascorbic acid reduction

Rapid determination of iodide ion content in chloride molten salt by ascorbic acid reduction

Pyrochemical Processing of LWR and L M F B R Fuels by the Salt Transport Method

Extraction of Uranium Nitride Pellets During Extraction of Lead from Brest Reactor Fuel-Element Simulators by Melting

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