Sensitivity of UO2 Stability in a Reducing Environment on Radiolysis Model Parameters

ABSTRACTResults for a radiolysis model sensitivity study of radiolytically produced H2O2 are presented as they relate to Spent (or Used) Light Water Reactor uranium oxide (UO2) nuclear fuel (UNF) oxidation in a low oxygen environment. The model builds on previous reaction kinetic studies to represent the radiolytic processes occurring at the nuclear fuel surface. Hydrogen peroxide (H2O2) is the dominant oxidant for spent nuclear fuel in an O2-depleted water environment. The most sensitive parameters have been identified with respect to predictions under typical conditions. As compared with the full model with about 100 reactions, it was found that only 30 to 40 of the reactions are required to determine [H2O2] to one part in 10–5 and to preserve most of the predictions for major species. This allows a systematic approach for model simplification and offers guidance in designing experiments for validation.

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A sensitivity study of the factors affecting the risks associated with the disposal of spent nuclear fuel in a geological disposal facility in a clay environment

Progress in Nuclear Energy ◽

10.1016/j.pnucene.2021.103910 ◽

2021 ◽

Vol 140 ◽

pp. 103910

Author(s):

Ashley I. Marsh ◽

Laurence G. Williams OBE, FREng ◽

James A. Lawrence

Keyword(s):

Nuclear Fuel ◽

Spent Nuclear Fuel ◽

Sensitivity Study ◽

Geological Disposal ◽

Factors Affecting ◽

Disposal Facility

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Bayesian Activity Estimation and Uncertainty Quantification of Spent Nuclear Fuel Using Passive Gamma Emission Tomography

Journal of Imaging ◽

10.3390/jimaging7100212 ◽

2021 ◽

Vol 7 (10) ◽

pp. 212

Author(s):

Ahmed Karam Eldaly ◽

Ming Fang ◽

Angela Di Fulvio ◽

Stephen McLaughlin ◽

Mike E. Davies ◽

...

Keyword(s):

Nuclear Fuel ◽

Linear Models ◽

Spent Nuclear Fuel ◽

Emission Tomography ◽

Superior Performance ◽

Model Parameters ◽

Gamma Emission ◽

Unknown Parameters ◽

Noise Models ◽

Activity Estimation

In this paper, we address the problem of activity estimation in passive gamma emission tomography (PGET) of spent nuclear fuel. Two different noise models are considered and compared, namely, the isotropic Gaussian and the Poisson noise models. The problem is formulated within a Bayesian framework as a linear inverse problem and prior distributions are assigned to the unknown model parameters. In particular, a Bernoulli-truncated Gaussian prior model is considered to promote sparse pin configurations. A Markov chain Monte Carlo (MCMC) method, based on a split and augmented Gibbs sampler, is then used to sample the posterior distribution of the unknown parameters. The proposed algorithm is first validated by simulations conducted using synthetic data, generated using the nominal models. We then consider more realistic data simulated using a bespoke simulator, whose forward model is non-linear and not available analytically. In that case, the linear models used are mis-specified and we analyse their robustness for activity estimation. The results demonstrate superior performance of the proposed approach in estimating the pin activities in different assembly patterns, in addition to being able to quantify their uncertainty measures, in comparison with existing methods.

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Selective Cs-removal from highly acidic spent nuclear fuel solutions

Radiochimica Acta ◽

10.1515/ract-2019-3187 ◽

2020 ◽

Vol 108 (8) ◽

pp. 615-626

Author(s):

Mu Lin ◽

Ivan Kajan ◽

Dorothea Schumann ◽

Andreas Türler ◽

Adelheid Fankhauser

Keyword(s):

Nuclear Fuel ◽

Spent Nuclear Fuel ◽

Fission Products ◽

Suitable Condition ◽

Kinetic Studies ◽

High Yield ◽

Exchange Method ◽

Maximum Sorption Capacity ◽

Maximum Sorption ◽

Paul Scherrer

AbstractThirty liters of highly acidic spent nuclear fuel solutions need to be disposed at the “Hot Laboratory (hotlab)” facility in Paul Scherrer Institut (PSI), Switzerland. In order to significantly reduce the γ dose rate before proper disposal treatment, 137Cs must be removed. In the here presented sub-project, the ion-exchange method was evaluated. Two promising sorbents, CLEVASOL® and AMP (ammonium molybdophosphate), and two derived products AMP_PAN (AMP immobilized in polyacrylonitrile) and AMP/SiO2 (AMP immobilized on silica gel) were tested by the batch method using model solutions of important high-yield fission products (Cs, Eu, Zr, Ru, Pd and Ag), as well as U and Pu. The results showed that AMP, AMP/SiO2 and AMP_PAN have higher selectivity for Cs than CLEVASOL® in 0.1–8 M (mol/L) HNO3 solutions. 4 M HNO3 solution was identified as the most suitable condition for Cs-removal with AMP, AMP_PAN and AMP/SiO2 due to the still sufficiently high separation factor of Cs from other metal ions and an acceptable volume increase factor after dilution. The follow-up kinetic studies on AMP, AMP_PAN and AMP/SiO2 indicated that although Cs exchange on AMP and AMP/SiO2 is faster than on AMP_PAN in the first 5 min, they all nearly reach equilibrium after 30 min of contacting time. The isotherm curves of Cs adsorption on AMP, AMP_PAN and AMP/SiO2 in 4 M HNO3 showed that the maximum sorption capacity of Cs is reached asymptotically. The results from Langmuir isotherm modeling agree with results from other publications.

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The Limitations of an Air-Oxidation Breakaway Model to Predict a Zirconium Fire in a Spent Nuclear Fuel Pool Accident

Sustainability ◽

10.3390/su11226364 ◽

2019 ◽

Vol 11 (22) ◽

pp. 6364

Author(s):

Sanggil Park ◽

Min Bum Park

Keyword(s):

Nuclear Fuel ◽

Spent Nuclear Fuel ◽

Critical Phenomenon ◽

Fission Products ◽

Model Parameters ◽

Spent Fuel ◽

Air Oxidation ◽

Loss Of Coolant Accident ◽

Spent Nuclear Fuel Pool ◽

Spent Fuel Pool

The OECD/NEA Spent Fuel Pool (SFP) project was conducted to investigate consequences of spent nuclear fuel pool accident scenarios. From the project, it was observed that cladding temperature could abruptly increase at a certain point and the cladding was completely oxidized. This phenomenon was called a “zirconium fire”. This zirconium fire is one of the crucial concerns for spent fuel pool safety under a postulated loss of coolant accident scenario, since it would lead to an uncontrolled mass release of fission products into the environment. To capture this critical phenomenon, an air-oxidation breakaway model has been implemented in the MELCOR code. This study examines this air-oxidation breakaway model by comparing the SFP project test data with a series of MELCOR code sensitivity calculation results. The air-oxidation model parameters are slightly altered to investigate their sensitivities on the occurrence of the zirconium fire. Through such sensitivity analysis, limitations of the air-oxidation breakaway model are identified, and needs for model improvement is recommended.

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Development of Equivalent Beam Model of High Burnup Spent Nuclear Fuel Rods under Lateral Impact Loading

Metals ◽

10.3390/met10040470 ◽

2020 ◽

Vol 10 (4) ◽

pp. 470

Author(s):

Sanghoon Lee ◽

Seyeon Kim

Keyword(s):

Nuclear Fuel ◽

Structural Integrity ◽

Spent Nuclear Fuel ◽

Model Parameters ◽

Beam Model ◽

Spent Fuel ◽

Analysis Model ◽

Fuel Rods ◽

Integrity Assessment ◽

Fuel Rod

Spent nuclear fuel (SNF) is nuclear fuel that has been irradiated and discharged from nuclear reactors. During the whole management stages of SNF before it is, in the end, disposed in a deep geological repository, the structural integrity of fuel rods and the assemblies should be maintained for safety and economic reasons. In licensing applications for the SNF storage and transportation, the integrity of SNF needs to be evaluated considering various loading conditions. However, this is a challenging task due to the complexity of the geometry and properties of SNF. In this paper, a simple and equivalent analysis model for SNF rods is developed using model calibration based on optimization and process integration. The spent fuel rod is simplified into a hollow beam with a homogenous isotropic material, and the model parameters thus found are not dependent on the length of the reference fuel rod segment that is considered. Two distinct models with different interfacial conditions between the fuel pellets and cladding are used in the calibration to account for the effect of PCMI (Pellet-Clad Mechanical Interaction). The feasibility of the models in dynamic impact simulations is examined, and it is expected that the developed models can be utilized in the analysis of assembly-level analyses for the SNF integrity assessment during transportation and storage.

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Investigation of Groundwater Composition in Relation to Spent Nuclear Fuel Disposal

MRS Proceedings ◽

10.1557/proc-26-209 ◽

1983 ◽

Vol 26 ◽

Author(s):

Fred Karlsson

Keyword(s):

Nuclear Fuel ◽

Spent Nuclear Fuel ◽

Carbonate Content ◽

Field Equipment ◽

Carbonate Concentration ◽

Groundwater Environment ◽

Suitable Site ◽

High Level ◽

High Degree ◽

Sensitive Parameters

ABSTRACTGroundwater from boreholes in granitic rock at six different sites in Sweden has been sampled and analyzed. Sensitive parameters such as redox potential, pH sulphide– and oxygen content have been measured with a field equipment. This is an integrated part of a program of geological, geophysical, geochemical and hydrogeological investigations with the final aim to select a suitable site for a high-level radioactive waste repository. According to present results deep granitic groundwaters are reducing due to the presence of iron(II) ions. The pH is normally ranging from 7 to 9. The total sulphide content is generally less than 0.5 mg/l. The normal alkalinity range is 90-275 mg/l. Copper, which has been suggested as canister material in the present concept for spent nuclear fuel disposal, is stable in this groundwater environment, except for a very slow sulphide corrosion and a limited initial attack of oxygen from the emplacement operations. The dissolution of the spent nuclear fuel matrix, UO2, is ultimately controlled by the total carbonate concentration. Carbonate content, pH and redox conditions will also be decisive for the mobilities of actinides and technetium. The conditions are generally favouring a high degree of retention for these species in the undisturbed deep groundwater rock environment.

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