Spent nuclear fuel in dry storage conditions – current trends in fuel performance modeling

Konarski, Piotr; Cozzo, C.; Khvostov, Grigori; Ferroukhi, H.

doi:10.1016/j.jnucmat.2021.153138

Cited by 23 publications

(4 citation statements)

References 103 publications

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“…Since 2001, after 3-5 years of storing, they have been transferred to the spent-fuel dry storage facility (SFDSF) at the site, which can accommodate 9120 FAs in 380 ventilated storage casks of 144 t each. The SFDSF facility was designed for a 50-year storage lifetime, which is a period defined on the basis of the degradation and integrity reduction processes that may affect the fuel which is subjected to high temperature and high tensile stresses, e.g., hydrogen-related effects, creep, fission gas release, and fuel swelling [26,27]. Beyond this period, it would be necessary to provide sufficient evidence that the spent fuel condition would not have deteriorated to an extent that any handling or retrieval would be prevented.…”

Section: Snf Radioactivity Results and Discussionmentioning

confidence: 99%

Assessment of Spent Nuclear Fuel in Ukrainian Storage System: Inventory and Performance

Dolin,

Lo Frano,

Cancemi

2024

Energies

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It is of meaningful importance to evaluate the performance of all the nuclear facilities, and particularly those part of such buildings where spent nuclear fuel (SNF) is stored to assess what kinds of consequences are anomalous/abnormal or to determine what types of accident events may occur. In this preliminary study, the strategies adopted for the management of SNF, and the risk related to them are discussed. The aim of this study is to evaluate the total radioactivity inventory characterising Ukrainian nuclear facilities, including storage facilities. The dataset used to calculate the total activity associated with nuclear fuel is provided and discussed. For the evaluation, it is considered that a SNF pool in VVER-1000 is designed to store 687 fuel assemblies, and 670 are in VVER-440. When it is half full, which is the case for 15 Ukrainian units, it will store about 2200 tU containing up to 1·1019 Bq of 137Cs, 7·1018 Bq of 90Sr, and 1·1019 Bq of TUE. This study focuses particularly on the total activity of the SNF stored at the Zaporozhye plant, the biggest nuclear plant in Europe, and the risk posed by the potential loss that cooling the plant could incur because of pond water level variation. The results of the analysis of the Zaporozhye NPP behaviour suggest that the water flow rate which keeps the SNF pool temperature constant is about 200,000 m3·day−1. Therefore, the water level in the pond should not be lower than 1.5–2 m; otherwise, the plant will need an additional source of water of more than 200,000 m3 per day to guarantee safe storage of SNF.

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Section: Snf Radioactivity Results and Discussionmentioning

confidence: 99%

Assessment of Spent Nuclear Fuel in Ukrainian Storage System: Inventory and Performance

Dolin,

Lo Frano,

Cancemi

2024

Energies

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“…-The temperature evolution depends on fuel enrichment, burnup and cooling performance of the cask and lies in the [20°C, 300°C] range. It has been shown that there is a temperature gradient along the length of the fuel assembly (Hong et al, 2020;Feria and Herranz, 2011;Konarski et al, 2021) but not across the thickness (< 1°C) (EPRI, 2011). -Moreover, fuel assemblies are subjected to a pressure difference between the external and internal surfaces.…”

Section: Introductionmentioning

confidence: 99%

Assessing the fracture toughness of Zircaloy-4 fuel rod cladding tubes: impact of delayed hydride cracking

François,

Petit,

Auzoux

et al. 2024

Int J Fract

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Delayed Hydride Cracking (DHC) is a hydrogen embrittlement phenomenon that may potentially occur in Zircaloy-4 fuel claddings during dry storage conditions. An experimental procedure has been developed to measure the toughness of this material in the presence of DHC by allowing crack propagation through the thickness of a fuel cladding. Notched C-ring specimens, charged with 100 wppm of hydrogen, were used and pre-cracked by brittle fracture of a hydrided zone at the notch root at room temperature. The length of the pre-crack was measured on the fracture surface or cross-sections. Additionally, a finite element model was developed to determine the stress intensity factor as a function of the crack length for a given loading. Two types of tests were conducted independently to determine the fracture toughness with and without DHC, K I DHC and K I C , respectively: (i) constant load tests at 150°C, 200°C, and 250°C; (ii) monotonic tests at 25°C, 200°C, and 250°C. The results indicate the following: (1) there is no temperature influence on the DHC toughness of Zircaloy-4 between 150 and 250°C (K I DHC ∈ [7.2; 9.2] MPa √ m), (2) within this temperature range, the fracture toughness of Zircaloy-4 is halved by DHC (K I C ∈ [16.9; 19.7] MPa √ m), (3) the crack propagation rate decreases with decreasing temperature and (4) the time before crack propagation increases as the temperature and loading decrease.

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“…However, delays in the introduction of permanent geological disposal facilities and in governments' decisions about reprocessing processes have led to the need for storage systems for SNF other than spent fuel pools, which are reaching their maximum capacity. In this framework, dry cask systems have been proposed and accepted worldwide as an interim storage system for SNF for up to 100 years [1][2][3][4][5].…”

Section: Introductionmentioning

confidence: 99%

Uncertainty and Sensitivity Analysis of a Dry Cask for Spent Nuclear Fuel

et al. 2022

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Nuclear safety relies to a good extent on thoroughly validated codes. However, code predictions are affected by uncertainties that need to be quantified for a more accurate evaluation of safety margins. In this regard, the present paper proposes a preliminary uncertainty and sensitivity analysis of the thermal behavior of a concrete-based dry cask for spent nuclear fuel storage, employing the MELCOR code and a series of MATLAB scripts. As thermal behavior is of utmost importance for the fulfillment of United States Nuclear Regulatory Commission (USNRC) safety requirements, the Peak Cladding Temperature (PCT) has been addressed as the key Figure of Merit (FOM). Variables related to the main heat transfer mechanisms have been selected as input parameters for the uncertainty quantification, whereas heat source and heat sink, namely decay power and external air temperature, have been dealt with in a separate sensitivity analysis. The results show that the selected parameters have a weak influence on the PCT, whereas it is strongly related to the decay power and external air temperature values. In any case, PCT stays below the regulatory threshold even under the considered off-normal conditions.

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Spent nuclear fuel in dry storage conditions – current trends in fuel performance modeling

Cited by 23 publications

References 103 publications

Assessment of Spent Nuclear Fuel in Ukrainian Storage System: Inventory and Performance

Assessment of Spent Nuclear Fuel in Ukrainian Storage System: Inventory and Performance

Assessing the fracture toughness of Zircaloy-4 fuel rod cladding tubes: impact of delayed hydride cracking

Uncertainty and Sensitivity Analysis of a Dry Cask for Spent Nuclear Fuel

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