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Title: Capture of Tritium Released from Cladding in the Zirconium Recycle Process

Abstract

This report is issued as the first revision to FCRD-MRWFD-2016-000297. Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using non-radioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrationsmore » on radioactive used cladding.« less

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), Fuel Cycle Technologies (NE-5)
OSTI Identifier:
1328328
Report Number(s):
ORNL/TM-2016/531
AF5805010; NEEAF315
DOE Contract Number:
AC05-00OR22725
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS

Citation Formats

Spencer, Barry B., Walker, T. B., Bruffey, Stephanie H., and DelCul, Guillermo Daniel. Capture of Tritium Released from Cladding in the Zirconium Recycle Process. United States: N. p., 2016. Web. doi:10.2172/1328328.
Spencer, Barry B., Walker, T. B., Bruffey, Stephanie H., & DelCul, Guillermo Daniel. Capture of Tritium Released from Cladding in the Zirconium Recycle Process. United States. doi:10.2172/1328328.
Spencer, Barry B., Walker, T. B., Bruffey, Stephanie H., and DelCul, Guillermo Daniel. 2016. "Capture of Tritium Released from Cladding in the Zirconium Recycle Process". United States. doi:10.2172/1328328. https://www.osti.gov/servlets/purl/1328328.
@article{osti_1328328,
title = {Capture of Tritium Released from Cladding in the Zirconium Recycle Process},
author = {Spencer, Barry B. and Walker, T. B. and Bruffey, Stephanie H. and DelCul, Guillermo Daniel},
abstractNote = {This report is issued as the first revision to FCRD-MRWFD-2016-000297. Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using non-radioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.},
doi = {10.2172/1328328},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2016,
month = 8
}

Technical Report:

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  • Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when themore » solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using nonradioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.« less
  • DCART (Doses from Chronic Atmospheric Releases of Tritium) is a spreadsheet model developed at Lawrence Livermore National Laboratory (LLNL) that calculates doses from inhalation of tritiated hydrogen gas (HT), inhalation and skin absorption of tritiated water (HTO), and ingestion of HTO and organically bound tritium (OBT) to adult, child (age 10), and infant (age 6 months to 1 year) from routine atmospheric releases of HT and HTO. DCART is a deterministic model that, when coupled to the risk assessment software Crystal Ball{reg_sign}, predicts doses with a 95% confidence interval. The equations used by DCART are described and all distributions onmore » parameter values are presented. DCART has been tested against the results of other models and several sets of observations in the Tritium Working Groups of the International Atomic Energy Agency's programs, Biosphere Modeling and Assessment and Environmental Modeling for Radiation Safety. The version of DCART described here has been modified to include parameter values and distributions specific to conditions at LLNL. In future work, DCART will be used to reconstruct dose to the hypothetical maximally exposed individual from annual routine releases of HTO and HT from all LLNL facilities and from the Sandia National Laboratory's Tritium Research Laboratory over the last fifty years.« less
  • No abstract prepared.
  • Using accurate apparatus and a drop'' method, the heat content (enthalpy) of Zr, five Zr hydrides (NH from 1.34 to 4.14), and stainless steel type 316 were measured over the range from 0 to 900 deg C. Using the values for the stainless steel and those from the literature for Mo and Nb, the heat capacities of typical clad samples of the hydrides may be computed additively. Thermal hysteresis of the hydrides was investigated in several cases. Corrections were applied for the impurities in the samples measured, but the two sets of hydrides gave heat values believed to be somewhatmore » inconsistent in the range 550 to 800 deg C through systematic differences in phase compositions. The heat content data were extensively correlated with certain published equilibrium data for the Zr-H in order to extend knowledge of the heats of hydriding, equilibrium H/sub 2/ pressures, and limits of solid solubility to wider ranges of temperature and composition than those covered by direct measurements. The results are discussed critically, and several structural implications are pointed out. (auth)« less