Study of hydrogen generation and radionuclide release during wet damaged oxide spent fuel storage

Gaiazov, Artem Zuferovich; Komarov, Sergey V.; Leshchenko, Leshchenko A.Yu.; Revenko, Ksenia Evgenievna; Smirnov, Smirnov V.P.; Zvir, E. A.; Ilyin, Pavel A.; Teplov, Vladimir Gennadievich

doi:10.26583/npe.2018.3.11

Cited by 4 publications

(5 citation statements)

References 74 publications

(98 reference statements)

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“…It contains 1% to 20% of 235 U and consists of either uranium oxide or uranium-molybdenum alloy in a magnesium matrix ((U-9% Mo)+Mg) or uranium carbide in a calcium matrix (UC+Ca). Taking into account the release of hydrogen and fission products from the spent fuel oxides described in (Gayazov et al 2018), the experiments used fragments of the irradiated (U-9% Mo)+Mg and UC+Ca nuclear fuel from the AM reactor, a prototype of the AMB spent fuel.…”

Section: Selection and Characterization Of Spent Nuclear Fuel For Exp...mentioning

confidence: 99%

“…The experimental equipment included a model canister, heating, steam pipeline, filtering, condensation, evacuation, gas filling and sampling systems, an EMG-20-6 mass spectrometer, an IRG BDGB-02P detector, and an automated control system (ACS). A more detailed description of the test rig and the measuring equipment is given in (Gayazov et al 2018).…”

Section: Experimental Equipment and Proceduresmentioning

confidence: 99%

“…In general, the generation rate for combustible gases depends on many reasons, such as the nature and concentrations of substances, pressure, temperature, the presence and concentrations of catalysts and impurities, the SNF damage, the equipment geometry, and the radiation dose rate. But even a simple case of the VVER spent fuel oxides gives us no chance to analytically take into account the effect of all components on the hydrogen generation (Gayazov et al 2018). It is even more the case when describing the release of combustible gases and radionuclides from the multicomponent spent fuel, such as the spent fuel fragments from the AMB reactor ("Large Peaceful Atom"; Russian: Атом Мирный Большой) in water or a humid environment.…”

Section: Introductionmentioning

confidence: 99%

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Study on combustible gas generation and radionuclide release during underwater handling of the AM reactor spent fuel

Gayazov¹,

Leshchenko²,

Smirnov³

et al. 2021

NUCET

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Introduction. The paper addresses studies on the accumulation of combustible gases during underwater handling simulations for the leaky spent nuclear fuel from the AM reactor. Two fuel compositions were studied- uranium-molybdenum dispersed in magnesium and uranium carbide dispersed in calcium. Methods. The 137Cs release rate was measured during underwater storage of the uranium-molybdenum fuel. The kinetics of hydrogen release for both fuels and methane release for the carbide SNF were obtained. The kinetics approximate most with exponential dependences that formally correspond to first-order chemical reactions. A contribution of radiolytic hydrogen to the gases generated during the experiments was estimated. It was demonstrated that the determining source of the gases is the chemical interaction between the spent fuel and the water. The experiment with the uranium-molybdenum fuel demonstrated a pronounced passivation effect of the chemical processes on the fuel surface due to insoluble corrosion products. For the carbide SNF, an incubation period of about 20 hours was observed followed by an intensive release of hydrogen and methane. Results. The obtained results were subject to a comparative analysis against publications on the behavior of the fuel components in water. Conclusion. The findings can be applied to justify fire and explosion safety of underwater handling techniques for the damaged spent nuclear fuel with the considered fuel compositions (the spent fuel from reactors AM, AMB, EGP-6, etc.), e.g., to justify underwater preparations of the AMB spent fuel for reprocessing.

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Section: Selection and Characterization Of Spent Nuclear Fuel For Exp...mentioning

confidence: 99%

Section: Experimental Equipment and Proceduresmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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Study on combustible gas generation and radionuclide release during underwater handling of the AM reactor spent fuel

Gayazov¹,

Leshchenko²,

Smirnov³

et al. 2021

NUCET

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“…The same applies to the fuel elements of the EGP-6 reactor at the Bilibino NPP, the storage life of which is approaching 50 years. Issues related to the safety of nuclear power plants and the state of fuel rods after longterm storage have been the subject of numerous studies (Sidorenko 1994;Solonin et al 2000;Volkova et al 2003;INES 2010;Golosov et al 2013a;Golosov et al 2013b;Mukhamadeev et al 2018;Parafilo et al 2018).…”

Section: Introductionmentioning

confidence: 99%

“…In the event of an emergency depressurization of the canisters containing the irradiated assemblies of the EGP-6 reactor, water from the storage pool (SP) will come into contact with the fuel elements, causing their corrosion and subsequent water pollution. In Gaiazov et al 2018;Ivanov et al 2019;Gaiazov et al 2021, the results of corrosion tests in the aqueous medium of the spent fuel elements of the AM and VVER-440 reactors after long-term storage are presented, which are necessary for predictive assessments of the release of radioactive products from leaky fuel elements of the EGP-6 reactor into the aqueous medium of the pool when the canisters are depressurized.…”

Section: Introductionmentioning

confidence: 99%

Examination of Fuel Elements Irradiated in the Reactor of the World’s First NPP After Long-Term Storage

Ivanov

Porollo²,

Viktorovich³

et al. 2022

YadEn

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Examinations of fuel elements with two different fuel compositions, U-Mo+Mg and UO 2 +Mg, irradiated in the AM reactor after their long-term storage do not reveal any visible defects on the surface of their outer claddings. However, in the fuel elements with U-Mo fuel, an increase in the diameter of the outer cladding is observed. This is most noticeable in the upper part of the fuel element. Storage of the fuel elements with UO 2 fuel for 15-22 years does not lead to a change in their diameter within the measurement accuracy. At the same time, metallographic studies have shown that on the external surface of the outer cladding and the internal surface of the inner cladding of the fuel elements with U-Mo+Mg and UO 2 +Mg fuel compositions, after long-term storage, defects are observed in the form of intergranular and irregular frontal corrosion, pits and pittings up to 20 µm deep. No interaction is found at the points of contact between the fuel claddings and the fuel composition of the layers. There is no noticeable decrease in the thickness of the outer and inner claddings of the fuel elements after long-term storage, nor does the thickness of the claddings at the locations of defects go beyond its minimum initial value, taking into account the technological tolerance for variations in thickness. It is noteworthy, however, that cracks are found in both types of fuel elements both in the fuel grains and in the magnesium matrix. As a result of long-term storage of the fuel elements with U-Mo fuel for 45-55 years, the mechanical properties of their outer claddings gradually degrade, due to which the plasticity of the cladding is significantly reduced.

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Study of flammable gases generation and radionuclide release during and after drying of damaged VVER SNF

Gaiazov,

Leshchenko,

Smirnov

et al. 2024

NUCET

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The paper presents an experimental evaluation of drying the damaged VVER SNF after the wet storage. The aim of the experiments was to justify the fire safety of air-tight canisters with the dried SNF during transportation, storage, and reprocessing. Keeping the SNF in boric acid during the experiments determined the concentration of uranium, plutonium, and specific activity of fission products in the solution inside the model canisters with the SNF. The canisters with the SNF underwent the thermal vacuum drying to determine the process parameters (temperature, residual pressure, drying rate), a hydrogen release, 85Kr activity during the drying, and an airborne radionuclide release onto the filters of the SNF thermal vacuum drying system. The results of weighing the baskets of the canisters before soaking in the boric acid and after drying at 160 °С to the residual pressure below 6 mbar showed that the difference in weight is almost the same as the calculated difference in weight of H3BO3 loaded into the canister and the weight after the complete decomposition into B2O3. Thus, we have discovered that the boric acid undergoes complete decomposition into boric anhydride, and the SNF corrosion products generate hydrogen that accumulates in the air-tight canister. We also considered the hydrogen that accumulated in the air-tight model canister with the dried SNF as a result of the radiolysis of hydrated corrosion products of the spent oxide nuclear fuel, as well as a release of gaseous fission products (GFP) inside the canister. We calculated the quantity of the hydrated corrosion products of uranium oxide after the wet storage and determined their chemical composition after the thermal-vacuum drying. The equation with an exponential decay function describes the hydrogen accumulation. The calculated maximum volume of hydrogen and uranium in the SNF corrosion products suggests that the SNF corrosion product is schoepite with the number of gyrated water molecules corresponding to the formulas from UO3·0.95H2O to UO3·1.5H2O. The received data allows for the fire safety justification of drying the wet damaged SNF and handling the dried SNF during transportation and storage.

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Study of hydrogen generation and radionuclide release during wet damaged oxide spent fuel storage

Cited by 4 publications

References 74 publications

Study on combustible gas generation and radionuclide release during underwater handling of the AM reactor spent fuel

Study on combustible gas generation and radionuclide release during underwater handling of the AM reactor spent fuel

Examination of Fuel Elements Irradiated in the Reactor of the World’s First NPP After Long-Term Storage

Study of flammable gases generation and radionuclide release during and after drying of damaged VVER SNF

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