skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: Milestone report - M4FT-14OR0302102b - Evaluation of Tritium Content and Release from Surry-2 Fuel Cladding

Abstract

To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified.To gain a better understanding of how tritium in cladding will behave during processing, scoping tests are being performed to determine the tritium content in the cladding pre- and post-tritium pretreatment. A sample of Surry-2 pressurized water reactor (PWR) cladding was heated to 1100–1200°C to oxidize the zirconium and release all of the tritium in the cladding sample. The tritium content was measured to be ~240 µCi/g. Cladding samples were heated to 500ºC, which is within the temperature range (480 - 600ºC) expected for standard air tritium pretreatment systems, and to a slightly higher temperature (700ºC) to determine the impact of tritium pretreatment on tritium release from the cladding. Heating at 500°C for 24 hr removes ~0.2% of the tritium from the cladding, and heating at 700°C for 24 hr removes ~9%. Thus, a significant fraction of the tritium remains bound in the cladding and must be considered in operations involving cladding recycle.

Authors:
 [1];  [1];  [1];  [1]
  1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Publication Date:
Research Org.:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org.:
USDOE Office of Nuclear Energy (NE)
OSTI Identifier:
1210128
Report Number(s):
ORNL/LTR-2014/394
AF5805000; NEAF357
DOE Contract Number:
AC05-00OR22725
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English

Citation Formats

Robinson, Sharon M., Chattin, Marc Rhea, Giaquinto, Joseph M., and Jubin, Robert Thomas. Milestone report - M4FT-14OR0302102b - Evaluation of Tritium Content and Release from Surry-2 Fuel Cladding. United States: N. p., 2014. Web. doi:10.2172/1210128.
Robinson, Sharon M., Chattin, Marc Rhea, Giaquinto, Joseph M., & Jubin, Robert Thomas. Milestone report - M4FT-14OR0302102b - Evaluation of Tritium Content and Release from Surry-2 Fuel Cladding. United States. doi:10.2172/1210128.
Robinson, Sharon M., Chattin, Marc Rhea, Giaquinto, Joseph M., and Jubin, Robert Thomas. Mon . "Milestone report - M4FT-14OR0302102b - Evaluation of Tritium Content and Release from Surry-2 Fuel Cladding". United States. doi:10.2172/1210128. https://www.osti.gov/servlets/purl/1210128.
@article{osti_1210128,
title = {Milestone report - M4FT-14OR0302102b - Evaluation of Tritium Content and Release from Surry-2 Fuel Cladding},
author = {Robinson, Sharon M. and Chattin, Marc Rhea and Giaquinto, Joseph M. and Jubin, Robert Thomas},
abstractNote = {To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified.To gain a better understanding of how tritium in cladding will behave during processing, scoping tests are being performed to determine the tritium content in the cladding pre- and post-tritium pretreatment. A sample of Surry-2 pressurized water reactor (PWR) cladding was heated to 1100–1200°C to oxidize the zirconium and release all of the tritium in the cladding sample. The tritium content was measured to be ~240 µCi/g. Cladding samples were heated to 500ºC, which is within the temperature range (480 - 600ºC) expected for standard air tritium pretreatment systems, and to a slightly higher temperature (700ºC) to determine the impact of tritium pretreatment on tritium release from the cladding. Heating at 500°C for 24 hr removes ~0.2% of the tritium from the cladding, and heating at 700°C for 24 hr removes ~9%. Thus, a significant fraction of the tritium remains bound in the cladding and must be considered in operations involving cladding recycle.},
doi = {10.2172/1210128},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon Sep 01 00:00:00 EDT 2014},
month = {Mon Sep 01 00:00:00 EDT 2014}
}

Technical Report:

Save / Share: