Thermal Modeling of the TN-32B Cask for the High Burnup Spent Fuel Data Project

Fort, James A.; Richmond, David; Cuta, Judith M.; Suffield, Sarah

doi:10.2172/1566774

Cited by 9 publications

(12 citation statements)

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“…In reality, the PCT for most systems will be below 325°C. In addition, as shown by the PNNL thermal analyses (Fort et al 2016a(Fort et al , 2016b, only a small fraction of the cladding achieves the PCT, and the overwhelming majority is well below 300°C.…”

Section: Temperaturementioning

confidence: 95%

“…3 (NRC 2003). Based on the PNNL thermal modeling of the Research Project Cask and accounting for expected decay heats, the cladding temperatures range from a low (minimum cladding temperature, MCT) of 138°C to a PCT of 271°C (Fort et al 2016a). PNNL performed a similar analysis on a MAGNASTOR system loaded at the Duke Catawba Nuclear Station.…”

Section: Temperaturementioning

confidence: 99%

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EPRI/DOE High Burnup Fuel Sister Pin Test Plan Simplification and Visualization

Sorenson

Hanson

et al. 2017

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The EPRI/DOE High Burnup Confirmatory Data Project (herein called the "Demo") is a multi-year, multi-entity confirmation demonstration test with the purpose of providing quantitative and qualitative data to show how high-burnup fuel ages in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of highburnup PWR fuel of four common cladding alloys from the North Anna Nuclear Power Plant, drying them according to standard plant procedures, and then storing them in an NRC-licensed TN-32B cask on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and the rods will be examined for signs of aging.

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Section: Temperaturementioning

confidence: 95%

Section: Temperaturementioning

confidence: 99%

EPRI/DOE High Burnup Fuel Sister Pin Test Plan Simplification and Visualization

Sorenson

Hanson

et al. 2017

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“…Temperature models for the HBU cask [Fort, 2019] predicted that stored fuel and adjacent structures would have peak temperatures in the 200-250°C range at initial storage conditions, per Figure 3-1. The temperature of canister materials affects the amount of surface-adsorbed water and could affect the radiolysis and/or back-reactions governing the net H2 generation.…”

Section: Temperaturementioning

confidence: 99%

“…The SNF canister drying process involves a sequence of vacuum and hold conditions with the cask internal volume being brought to an initial 55 Pa (0.41 Torr) vacuum [Bryan et al, 2019a]. The vacuum drying time for the HBU cask was less than 9 hours [Fort et al, 2019]. Drying completion is verified by evacuating and sealing the canister, then monitoring the rise in pressure due to water evaporation or desorption; the pressure must remain below 3 Torr after 30 minutes for the canister to be considered dry [Bryan et al, 2019b].…”

Section: Snf Drying Proceduresmentioning

confidence: 99%

Measurement of radiolytic hydrogen generation from surrogate zr-based materials in a "mini-canister"

D'Entremont

VERST

Sindelar

2022

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This report describes a test campaign to investigate radiolytic hydrogen generation from surrogate materials resembling the (Zr-based) cladding of commercial spent nuclear fuel (SNF) with an inventory of residual water post-dryout. The radiolytic hydrogen generation rates from residual waters in the SNF-incanister system are important for predicting the in-canister environment for sealed SNF dry storage canisters over time and the impact to continued safe dry storage.Under gamma radiation, radiolytic breakdown of water remaining in a canister (free, physisorbed, or chemisorbed) causes the generation of hydrogen gas (H2). SNF assemblies, clad in zirconium alloys, provide a large surface area to host surface-adsorbed water (physisorbed water). Other components in the cask, including stainless steel and aluminum structural components and aluminum neutron absorbers, also contribute to the total water inventory and radiolytic H2 generation in an SNF canister.Water adsorbed to the ZrO2 on the cladding surface may experience accelerated radiolysis compared to free water due to energy exchange with the oxide. In addition, surface water decomposed by radiolysis may be replenished by water vapor in the canister gas, providing a mechanism for accelerated radiolysis of the free water as well. The previous simplified analysis [d'Entremont et al., 2020] predicted that radiolysis of physisorbed water on the Zr-alloy components would dominate the short-term H2 generation due to these factors. Therefore, the yield associated with water radiolysis on the Zr surfaces (physisorbed water, with potential replenishment from water vapor) is the focus of the current testing.The lab-scale testing is conducted on non-radioactive surrogates, namely assemblies of zirconium coupons, irradiated inside a miniature steel canister ("mini-canister") designed to allow intermittent, insitu sampling of the canister gas during irradiation. The mini-canister approach for irradiation with in-situ monitoring was used in recent and on-going studies [Verst and d'Entremont, 2021] to measure radiolytic hydrogen generation from aluminum SNF cladding surrogate materials. This proposed testing forms part of the overall materials performance evaluation of the commercial SNF-in-canister system and is part of the technical bases for their continued safe dry storage.The test plan draws on the previous radiolysis testing on aluminum as well as the previous analysis, "Evaluation of Hydrogen Generation in High Burnup Demonstration Dry Storage Cask" [d'Entremont et al., 2020], which provided a best-estimate evaluation of residual water content (post-dryout) in the High Burnup (HBU) LWR Spent Fuel Demonstration project TN-32 cask and evaluated the expected radiolysis of the residual water, including free, physisorbed, and chemisorbed water in the sealed cask, based on literature data and models.The experiments focus on the contribution associated with the Zr-alloy components, i.e., surface-adsorbed water as well as interactions with free water vapor present in the gas p...

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“…As illustrated in Figure 2, the local temperature of the fuel rod cladding varies with axial elevation and based on the dry storage system; however, a flat 400°C was imposed at all axial locations for the FHT. The expected axial temperature distribution in the Research Project Cask (RPC) is also provided in Figure 2 for information [6]. The hydrogen content in the fuel rod cladding is expected to vary axially as well, with the highest cladding hydrogen concentrations located in the upper elevations where the cladding and coolant were hottest during reactor operation.…”

Section: Full-length Fuel Rod Heat Treatments (Fht)mentioning

confidence: 99%

Sister Rod Destructive Test Results (FY19)

Montgomery

Morris

Ilgner

et al. 2019

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This report documents work performed under the Spent Fuel and Waste Disposition, Spent Fuel and Waste Science and Technology program for the US Department of Energy (DOE) Office of Nuclear Energy (NE). This work was performed to fulfill Level 2 Milestone M2SF-19ORO010201026 within work package SF-19OR01020102 and is an update to the work reported in M2SF-19OR010201028 and M3SF-19OR010201027. The document incorporates comments received with respect to the previous reports and incorporates additional test results. As a part of the US Department of Energy (DOE) Office of Nuclear Energy (NE) High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 gigawatt days per metric ton uranium) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called the "sister rods," are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin Zircaloy-4 (LT Zirc-4), ZIRLO ® , and M5 ®. The DEs are being conducted to obtain a baseline of the HBU rod's condition prior to dry storage and are focused on understanding overall SNF rod strength and durability. Both composite fuel and empty cladding will be tested to derive material properties. While the data generated can be used for multiple purposes, a primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities by (1) addressing identified knowledge gaps and (2) enhancing the technical basis for post-storage transportation, handling, and consolidation activities. This report documents the status of the ORNL Phase I DEs of 8 sister rods and outlines the DE tasks performed and the data collected to date, as guided by the sister rod test plans. The DEs are performed using a phased approach, and the Phase 1 DEs being performed at ORNL include: • full-length rod heat treatments (FHT) of 3 selected sister rods to examine the effects of temperatures reached during dry storage preparation, Sister Rod Destructive Examinations

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Thermal Modeling of the TN-32B Cask for the High Burnup Spent Fuel Data Project

Cited by 9 publications

References 0 publications

EPRI/DOE High Burnup Fuel Sister Pin Test Plan Simplification and Visualization

EPRI/DOE High Burnup Fuel Sister Pin Test Plan Simplification and Visualization

Measurement of radiolytic hydrogen generation from surrogate zr-based materials in a "mini-canister"

Sister Rod Destructive Test Results (FY19)

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