Skip to Main Content
Skip Nav Destination
ASTM Selected Technical Papers
Zirconium in the Nuclear Industry: 20th International Symposium
Editor
Suresh K. Yagnik
Suresh K. Yagnik
Symposium Chairperson and STP Editor
1
Electric Power Research Institute (EPRI)
,
Palo Alto, CA,
US
Search for other works by this author on:
Michael Preuss
Michael Preuss
Symposium Chair and STP Editor
2
The University of Manchester Manchester
,
GB
;
Monash University
,
Clayton/Melbourne,
AU
Search for other works by this author on:
ISBN:
978-0-8031-7737-6
No. of Pages:
928
Publisher:
ASTM International
Publication date:
2023

Steam oxidation tests under steam-starved and non-steam-starved conditions were conducted up to 1,573 K using a prototypic BWR fuel assembly (four fuel pins and fuel channel box) with a length of approximately 750 mm. Significant suppression of oxide layer growth and enhancement of hydrogen uptake were found at the downstream positions under the steam-starved conditions. To understand the results obtained in the tests using the prototypic BWR fuel assembly, three separate-effects tests were conducted to obtain a fundamental understanding of the mechanism of oxygen and hydrogen uptake and its axial variations and evaluation of hydrogen solubility in oxygen-dissolved Zircaloy-2. The fuel channel box was found to contribute to the axial variations of oxide layer growth and hydrogen uptake of the fuel pins by acting as a source of hydrogen and a sink of oxygen. The evaluation of hydrogen uptake and release requires a detailed estimation of steam oxidation with time at each elevation.

1.
Inter-Ministerial Council for Contaminated Water and Decommissioning Issues, “
Mid- and Long-Term Roadmap towards the Decommissioning of TEPCO's Fukushima Daiichi Nuclear Power Station
,”
METI
,
2019
, https://web.archive.org/web/20220301192234/https://www.meti.go.jp/english/earthquake/nuclear/decommissioning/pdf/20191227_3.pdf
2.
TEPCO
, “
The Development of and Lessons from the Fukushima Daiichi Nuclear Accident
,”
TEPCO
,
2013
, https://web.archive.org/web/20220309030816/https://www.tepco.co.jp/en/decommision/accident/images/outline01.pdf
3.
Madokoro
M.
and
Sato
I.
, “
Estimation of the Core Degradation and Relocation at the Fukushima Daiichi Nuclear Power Station Unit 2 Based on RELAP/SCDAPSIM Analysis
,”
Nuclear Engineering and Design
376
(
2021
): 111123,
4.
Uetsuka
H.
and
Otomo
T.
, “
High Temperature Oxidation of Zircaloy-4 in Diluted Steam
,”
Journal of Nuclear Science and Technology
26
, no.
2
(
1989
): 240–248.
5.
Wu
X.
and
Shirvan
K.
, “
System Code Evaluation of Near-Term Accident Tolerant Claddings during Boiling Water Reactor Short-Term and Long-Term Station Blackout Accidents
,”
Nuclear Engineering and Design
356
(
2020
): 110362,
6.
Guo
Z.
,
Dailey
R.
,
Zhou
Y.
,
Sun
Z.
,
Wan
J.
, and
Corradini
M. L.
, “
Effect of ATF Cr-Coated-Zircaloy on BWR In-Vessel Accident Progression during a Station Blackout
,”
Nuclear Engineering and Design
372
(
2021
): 110979,
7.
Miyashita
T.
,
Nakae
N.
,
Ogata
K.
,
Baba
T.
,
Kamimura
K.
,
Matsumoto
T.
, and
Kakiuchi
K.
, “
Corrosion and Hydrogen Pick-Up Behavior of Cladding and Structural Components in BWR High Burnup 9x9 Lead Use Assemblies
(paper presentation, International LWR Fuel Performance Meeting,
San Francisco, CA
, September 30–October 3,
2007
).
8.
Hirano
Y.
,
Mozumi
Y.
,
Kamimura
K.
, and
Tsukuta
Y.
, “
Irradiation Characteristics of BWR High Burnup 9x9 Lead Use Assemblies
” (paper presentation, Water Reactor Fuel Performance Meeting,
Kyoto, Japan
, October 2–6,
2005
).
9.
Hatano
Y.
,
Shi
J.
,
Yoshida
N.
,
Futagami
N.
,
Oya
Y.
, and
Nakamura
H.
, “
Measurement of Deuterium and Helium by Glow-Discharge Optical Emission Spectroscopy for Plasma–Surface Interaction Studies
,”
Fusion Engineering and Design
87
, no.
7-8
(
2012
): 1091–1094.
10.
Grosse
M.
,
Steinbrueck
M.
,
Schillinger
B.
, and
Kaestner
A.
, “
In Situ Investigations of the Hydrogen Uptake of Zirconium Alloys during Steam Oxidation
,” in
Zirconium in the Nuclear Industry: 18th International Symposium
, ed.
Comstock
R.
and
Motta
A.
(
West Conshohocken, PA
:
ASTM International
,
2018
): 1114–1135,
11.
Yamanaka
S.
,
Tanaka
T.
, and
Miyake
M.
, “
Effect of Oxygen on Hydrogen Solubility in Zirconium
,”
Journal of Nuclear Materials
167
(
1989
): 231–237.
12.
Steinbruck
M.
, “
Hydrogen Absorption by Zirconium Alloys at High Temperatures
,”
Journal of Nuclear Materials
334
(
2004
): 58–64.
13.
Moalem
M.
and
Olander
D. R.
, “
The High-Temperature Solubility of Hydrogen in Pure and Oxygen-Containing Zircaloy
,”
Journal of Nuclear Materials
178
(
1991
): 61–72.
14.
Veshchunov
M. S.
and
Shestak
V. E.
, “
Models for Hydrogen Uptake and Release Kinetics by Zirconium Alloys at High Temperatures
,”
Nuclear Engineering and Design
252
(
2012
): 96–107.
15.
Stuckert
J.
and
Veshchunov
M. S.
, Behaviour of Oxide Layer of Zirconium-Based Fuel Rod Cladding under Steam Starvation Conditions, FZKA 7373 (Karlsruhe, Germany:
FZKA
,
2008
).
This content is only available via PDF.
You do not currently have access to this chapter.
Close Modal

or Create an Account

Close Modal
Close Modal