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Title: THERMAL ANNEALING OF IRRADIATED PREHYDRIDED ZIRCALOY-4 MATERIALS

Abstract

Hydrogen-assisted irradiation growth is an important issue for reliability and performance of nuclear materials in nuclear fuel cladding and fuel assembly components. The mechanisms of irradiation growth are not entirely clear and the role of hydrogen and hydrides are less so. There is evidence that breakaway irradiation growth is strongly influenced by the formation and dynamics of dislocation loops that vary with alloy composition, heat treatment and hydrogen content. Thermal annealing of eight Zircaloy-4 specimens irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory as part of the joint EPRI Nuclear Fuels Industry Research (NFIR) and Fuel Reliability Program (FRP) AsTeR irradiation growth study were performed by hot-stage transmission electron microscopy (TEM), differential scanning calorimetry (DSC), and dilatometry via NSUF RTE 17-1093. Two fluence levels (6.7 and 12.3 dpa) and three hydrogen contents (~10, 40, 120 ppm) were investigated. TEM with a hot stage was used for the quantification of "a" and "c" dislocation loop population densities and size distributions at a range of temperatures from 25 to 800oC. DSC and dilatometry were performed by dynamically heating and cooling from 30 to 750ºC at rates of 5, 10, and 20 oC/min to determine the behavior of dislocationmore » loop annealing. Dynamic temperature DSC tests were able to detect the temperature ranges and enthalpies of "a" and "c" loop annihilation, hydride dissolution and precipitation, secondary particle precipitation, and a?a + ß phase transformations upon heating and cooling. Dilatometry tests showed that all of the test showed initial linear expansion followed by a decrease in expansion rate from about 400 to 600 ºC associated with annihilation of "a" and "c" loops and showed evidence for the thermal stabilizing effect of hydrogen on dislocation loops. The results are discussed in the context of the larger AsTeR program that will include higher exposure (~20 dpa) and different alloys when they become available in 2021.« less

Authors:
ORCiD logo [1];  [2];  [3]
  1. Idaho National Laboratory
  2. Studsvik International
  3. Electric Power Research Institute
Publication Date:
Research Org.:
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Sponsoring Org.:
USDOE Office of Nuclear Energy (NE)
OSTI Identifier:
1632205
Report Number(s):
INL/CON-19-55848-Rev000
DOE Contract Number:  
DE-AC07-05ID14517
Resource Type:
Conference
Resource Relation:
Conference: ANS Global TopFuel 2019, Seattle, Washington, 09/22/2019 - 09/26/2019
Country of Publication:
United States
Language:
English
Subject:
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS; Zircaloy; calorimetry; dilatometry; transmission electron microscopy; annealing; zirconium hydride

Citation Formats

Middlemas, Scott C, Jadernas, Daniel, and Mader, Erik. THERMAL ANNEALING OF IRRADIATED PREHYDRIDED ZIRCALOY-4 MATERIALS. United States: N. p., 2018. Web.
Middlemas, Scott C, Jadernas, Daniel, & Mader, Erik. THERMAL ANNEALING OF IRRADIATED PREHYDRIDED ZIRCALOY-4 MATERIALS. United States.
Middlemas, Scott C, Jadernas, Daniel, and Mader, Erik. Wed . "THERMAL ANNEALING OF IRRADIATED PREHYDRIDED ZIRCALOY-4 MATERIALS". United States. https://www.osti.gov/servlets/purl/1632205.
@article{osti_1632205,
title = {THERMAL ANNEALING OF IRRADIATED PREHYDRIDED ZIRCALOY-4 MATERIALS},
author = {Middlemas, Scott C and Jadernas, Daniel and Mader, Erik},
abstractNote = {Hydrogen-assisted irradiation growth is an important issue for reliability and performance of nuclear materials in nuclear fuel cladding and fuel assembly components. The mechanisms of irradiation growth are not entirely clear and the role of hydrogen and hydrides are less so. There is evidence that breakaway irradiation growth is strongly influenced by the formation and dynamics of dislocation loops that vary with alloy composition, heat treatment and hydrogen content. Thermal annealing of eight Zircaloy-4 specimens irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory as part of the joint EPRI Nuclear Fuels Industry Research (NFIR) and Fuel Reliability Program (FRP) AsTeR irradiation growth study were performed by hot-stage transmission electron microscopy (TEM), differential scanning calorimetry (DSC), and dilatometry via NSUF RTE 17-1093. Two fluence levels (6.7 and 12.3 dpa) and three hydrogen contents (~10, 40, 120 ppm) were investigated. TEM with a hot stage was used for the quantification of "a" and "c" dislocation loop population densities and size distributions at a range of temperatures from 25 to 800oC. DSC and dilatometry were performed by dynamically heating and cooling from 30 to 750ºC at rates of 5, 10, and 20 oC/min to determine the behavior of dislocation loop annealing. Dynamic temperature DSC tests were able to detect the temperature ranges and enthalpies of "a" and "c" loop annihilation, hydride dissolution and precipitation, secondary particle precipitation, and a?a + ß phase transformations upon heating and cooling. Dilatometry tests showed that all of the test showed initial linear expansion followed by a decrease in expansion rate from about 400 to 600 ºC associated with annihilation of "a" and "c" loops and showed evidence for the thermal stabilizing effect of hydrogen on dislocation loops. The results are discussed in the context of the larger AsTeR program that will include higher exposure (~20 dpa) and different alloys when they become available in 2021.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {2018},
month = {10}
}

Conference:
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