Thorium-based fuel cycles can improve fuel sustainability within the nuclear power industry. The Canadian supercritical water-cooled reactor (SCWR) concept uses this path to achieve the sustainability requirement of the Gen-IV Forum. The study of thorium dioxide/thoria ThO2-based fuel irradiation behavior is significantly less advanced than that of uranium dioxide (UO2) fuel, although ThO2 possesses superior thermal conductivity, thermal expansion, higher melting temperature, and oxidation resistance that may improve both fuel performance and safety. The fuel and sheath modeling tool (FAST), a fuel performance model for UO2 fuel, was developed at the Royal Military College of Canada (RMCC). FAST capability has been extended to include thoria (ThO2), thorium uranium dioxide (Th,U)O2, and thorium plutonium dioxide (Th,Pu)O2 as fuel pellet materials, to aid in designing and performance assessment of Th-based fuels, including SCWR (Th,Pu)O2 fuel. The development and integration of ThO2 and (Th,U)O2 models into the existing FAST model led to the multipellet material FAST (MPM-FAST). Model development was performed in collaboration between RMCC and Canadian Nuclear Laboratories (CNL). This paper presents an outline of the ThO2 and (Th,U)O2 MPM-FAST model, a comparison between modeling results with postirradiation examination (PIE) data from a test conducted at CNL, and an account of the knowledge gap between our ability to model ThO2 and (Th,U)O2 fuel compared to UO2. Results are encouraging when compared to PIE data.

References

1.
McDonald
,
M. H.
,
Hyland
,
B.
,
Hamilton
,
H.
,
Leung
,
L. K. H.
,
Onder
,
N.
,
Pencer
,
J.
, and
Xu
,
R.
,
2011
, “
Pre-Conceptual Fuel Design Concepts for the Canadian Super Critical Water-Cooled Reactor
,”
Fifth International Symposium on Supercritical Water-Cooled Reactors (ISSCWR-5)
,
Vancouver, BC, Canada
,
Mar. 13–16
, Paper No. P134.
2.
Yetisir
,
M.
,
Diamond
,
W.
,
Leung
,
L. K. H.
,
Martin
,
D.
, and
Duffer
,
R.
,
2011
, “
Conceptual Mechanical Design for a Pressure-Tube Type Supercritical Water-Cooled Reactor
,”
Fifth International Symposium on Supercritical Water-Cooled Reactors (ISSCWR-5)
,
Vancouver, BC, Canada
,
Mar. 13–16
, Paper No. P055.
3.
Kuran
,
S.
,
Hopwood
,
J.
, and
Hastings
,
I. J.
,
2011
, “
Fuel Cycles–A Key to Future CANDU Success
,”
International Conference on Future of Heavy Water Reactors Proceedings
,
Ottawa, ON, Canada
,
Oct. 2–5
, Paper No. 026.
4.
BARC
,
2016
, “
About Us:ANUSHAKTI—Atomic Energy In India: Strategy for Nuclear Energy - BARC
,” Bhabha Atomic Research Centre, Department of Atomic Energy, Government of India, Mumbai, India, accessed Oct. 16,
2016
, http://barc.gov.in/about/anushakti_sne.html
5.
World Nuclear News
,
2016
, “
Canada and China Team up on AFCR
,” World Nuclear News, accessed Oct. 16, 2016, http://www.world-nuclear-news.org/C-Canada-and-China-team-up-on-AFCR-2309164.html
6.
Thorbeck
,
A.
,
2016
, “
Startpage
,”
Thor Energy
, Oslo, Norway, accessed Oct. 16, 2016, http://thorenergy.no/
7.
Chassie
,
G. G.
,
Sim
,
K. S.
,
Wong
,
B.
, and
Papayiannis
,
G.
,
2005
, “
ELESTRES Code Upgrades
,”
Ninth International Conference on CANDU Fuel, ‘Fuelling a Clean Future’
,
Bellville, ON, Canada
,
Sept. 18–21
, Paper No. C3001.
8.
Williams
,
A. F.
,
2005
, “
The ELOCA Fuel Modelling Code: Past, Present and Future
,”
Ninth International Conference on CANDU Fuel, ‘Fuelling a Clean Future’
,
Bellville, ON, Canada
,
Sept. 18–21
, Paper No. B4007.
9.
Geelhood
,
K. J.
,
Luscher
,
W. G.
, and
Beyer
,
C. E.
,
2011
, “
FRAPCON-3.4: A Computer Code for the Calculation of Steady-State, Thermal-Mechanical Behavior of Oxide Fuel Rods for High Burnup (NUREG/CR-7022, Volume 1, PNNL-19418, Volume 1)
,” Nuclear Regulatory Commission, Richland, WA, Report No. PNNL-19418.
10.
Luscher
,
W. G.
, and
Geelhood
,
K. J.
,
2011
, “
Material Property Correlations: Comparisons Between FRAPCON-3.4, FRAPTRAN 1.4, and MATPRO
,” Nuclear Regulatory Commission, Richland, WA, Report No. PNNL-19417.
11.
Lyon
,
W.
,
Montgomery
,
R.
,
Zangari
,
A.
,
Sunderland
,
D.
,
Rashid
,
Y.
, and
Dunham
,
R.
,
2002
, “
Fuel Performance Analysis Capability in Falcon
,”
Nucl. Eng. Des.
,
295
, pp.
910
921
.
12.
Lassmann
,
K.
,
1992
, “
TRANSURANUS: A Fuel Rod Analysis Code Ready for Use
,”
J. Nucl. Mater.
,
188
, pp.
295
302
.
13.
Baurens
,
B.
,
Sercombe
,
J.
,
Riglet-Martial
,
C.
,
Desgranges
,
L.
,
Trotignon
,
L.
, and
Maugis
,
P.
,
2014
, “
3D Thermo-Chemical–Mechanical Simulation of Power Ramps With ALCYONE Fuel Code
,”
J. Nucl. Mater.
,
452
(
1–3
), pp.
578
594
.
14.
Newman
,
C.
,
Hansen
,
G.
, and
Gaston
,
D.
,
2009
, “
Three Dimensional Coupled Simulation of Thermomechanics, Heat, and Oxygen Diffusion in Nuclear Fuel Rods
,”
J. Nucl. Mater.
,
392
(
1
), pp.
6
15
.
15.
Bell
,
J. S.
,
2017
, “
Thorium-Based Nuclear Fuel Performance Modelling With Multi-Pellet Material Fuel and Sheath Modelling Tool
,”
Ph.D. thesis
, Royal Military College of Canada, Kingston, ON, Canadahttps://espace.rmc.ca/handle/11264/1262.
16.
Floyd
,
M. R.
,
Bromley
,
B.
, and
Pencer
,
J.
,
2016
, “
A Canadian Perspective on Progress in Thoria Fuel Science and Technology
,”
CNL Nucl. Rev.
,
6
(1), pp.
1
17
.
17.
Prudil
,
A.
,
2013
, “
FAST: A Fuel and Sheath Modeling Tool for CANDU Reactor Fuel
,” Ph.D. thesis, Royal Military College of Canada, Kingston, ON, Canada.
18.
Corbett
,
S.
,
Floyd
,
M. R.
,
Livingstone
,
S. J.
,
Hamilton
,
H.
, and
Harrison
,
N. F.
,
2013
, “
DME-221 Thoria Fuel: Fabrication, Irradiation Testing, and Post-Irradiation Examination
,”
12th International CANDU Fuel Conference Proceedings
,
Kingston, ON, Canada
,
Sept. 15–18
, pp. 1–14.
19.
Morgan
,
D.
,
2007
, “
A Thermomechanical Model of CANDU Fuel
,” M.S. thesis, Royal Military College of Canada, Kingston, ON, Canada.
20.
Shaheen
,
K.
,
2011
, “
A Mechanistic Code for Intact and Defective Nuclear Fuel Element Performance
,” Ph.D. thesis, Royal Military College of Canada, Kingston, ON, Canada.
21.
Prudil
,
A.
,
Lewis
,
B. J.
,
Chan
,
P. K.
, and
Baschuk
,
J. J.
,
2015
, “
Development and Testing of the FAST Fuel Performance Code: Normal Operating Conditions—Part 1
,”
Nucl. Eng. Des.
,
282
, pp.
158
168
.
22.
Prudil
,
A.
,
Lewis
,
B. J.
,
Chan
,
P. K.
,
Baschuk
,
J. J.
, and
Wowk
,
D.
,
2015
, “
Development and Testing of the FAST Fuel Performance Code: Transient Conditions—Part 2
,”
Nucl. Eng. Des.
,
282
, pp.
169
177
.
23.
Belle
,
J.
, and
Berman
,
R. M.
,
1984
, “
Thorium Dioxide: Properties and Nuclear Applications
,” USDOE Assistant Secretary for Nuclear Energy, Naval Reactors Office United States Department of Energy, Pittsburgh, PA, Report No. DOE/NE–0060.
24.
Long
,
Y. Y. Y.
,
Kazimi
,
M. S.
,
Ballinger
,
R. G.
, and
Pilat
,
E. E.
,
2002
, “
A Fission Gas Release Model for High-Burnup LWR ThO2-UO2 Fuel
,”
Nucl. Technol.
,
138
(
3
), pp.
260
272
.
25.
Bakker
,
K.
,
Cordfunke
,
E. H. P.
,
Konings
,
R. J. M.
, and
Schram
,
R. P. C.
,
1997
, “
Critical Evaluation of the Thermal Properties of ThO2 and Th1−yUyO2 and a Survey of the Literature Data on Th1−yPuyO2
,”
J. Nucl. Mater.
,
250
(
1
), pp.
1
12
.
26.
Lucuta
,
P. G.
,
Matzke
,
H.
, and
Hastings
,
I. J.
,
1996
, “
A Pragmatic Approach to Modelling Thermal Conductivity of Irradiated UO2 Fuel: Review and Recommendations
,”
J. Nucl. Mater.
,
232
(
2–3
), pp.
166
180
.
27.
Hastings
,
I. J.
, and
Evans
,
L. E.
,
1979
, “
Densification Algorithm for Irradiated UO2 Fuel
,”
J. Am. Ceram. Soc.
,
62
(
3–4
), pp.
217
218
.
28.
Olander
,
D. R.
,
1976
, “
Fundamental Aspects of Nuclear Reactor Fuel Elements
,” California University, Berkeley, CA, Report No.
TID-26711-P1
https://www.osti.gov/biblio/7343826-fundamental-aspects-nuclear-reactor-fuel-elements.
29.
Matzke
,
H.
,
1980
, “
Hertzian Indentation of Thorium Dioxide, ThO2
,”
J. Mater. Sci.
,
15
(
3
), pp.
739
746
.
30.
Booth
,
A. H.
,
1957
, “
A Method of Calculating Fission Gas Release From UO2 Fuel and Its Application to the X-2-F Loop Test
,” Atomic Energy of Canada Limited, Chalk River, ON, Canada, Report No. AECL-496.
31.
Turnbull
,
J. A.
,
Friskney
,
C. A.
,
Johnson
,
F. A.
,
Walter
,
A. J.
, and
Findlay
,
J. R.
,
1977
, “
The Release of Radioactive Gases From Uranium Dioxide During Irradiation
,”
J. Nucl. Mater.
,
67
(
3
), pp.
301
306
.
32.
Turnbull
,
J. A.
,
Friskney
,
C. A.
,
Findlay
,
J. R.
,
Johnson
,
F. A.
, and
Walter
,
A. J.
,
1982
, “
The Diffusion Coefficients of Gaseous and Volatile Species During the Irradiation of Uranium Dioxide
,”
J. Nucl. Mater.
,
107
(
2–3
), pp.
168
184
.
33.
Friskney
,
C. A.
,
Turnbull
,
J. A.
,
Johnson
,
F. A.
,
Walter
,
A. J.
, and
Findlay
,
J. R.
,
1977
, “
The Characteristics of Fission Gas Release From Monocrystalline Uranium Dioxide During Irradiation
,”
J. Nucl. Mater.
,
68
(
2
), pp.
186
192
.
34.
White
,
R. J.
, and
Tucker
,
M. O.
,
1983
, “
A New Fission-Gas Release Model
,”
J. Nucl. Mater.
,
118
(
1
), pp.
1
38
.
35.
Livingstone
,
S. J.
, and
Floyd
,
M. R.
,
2013
, “
Thoria Irradiation and Post-Irradiation Examination Experience at AECL
,”
12th International CANDU Fuel Conference
,
Kingston, ON, Canada
,
Sept. 15–18
, pp. 1–11.
36.
Kim
,
H.
,
Park
,
K.
,
Kim
,
B. G.
,
Choo
,
Y. S.
,
Kim
,
K. S.
,
Song
,
K. W.
,
Hong
,
K. P.
,
Kang
,
Y. H.
, and
Ho
,
K.
,
2004
, “
Xenon Diffusivity in Thoria-Urania Fuel
,”
Nucl. Technol.
,
147
(
1
), pp.
149
156
.
37.
Lee
,
C. B.
,
Yang
,
Y. S.
,
Kim
,
Y. M.
,
Kim
,
D. H.
, and
Jung
,
Y. H.
,
2004
, “
Irradiation Behavior of Thoria-Urania Fuel in a PWR
,”
Nucl. Technol.
,
147
(
1
), pp.
140
148
.
38.
McCauley
,
J. E.
,
1969
, “
Observations on the Irradiation Behavior of a Zircaloy-4 Clad Rod Containing Low Density ThO2—5.3 W/O UO2 Pellets (LWBR Development Program)
,” Bettis Atomic Lab, Pittsburgh, PA, Technical Report No. WAPD-TM-664.
39.
Nichols
,
F. A.
,
1968
, “
Further Comments on the Theory of Grain Growth in Porous Compacts
,”
J. Am. Ceram. Soc.
,
51
(
8
), pp.
468
468
.
40.
Goldberg
,
I.
,
Waldman
,
L. A.
,
Giovengo
,
J. F.
, and
Campbell
,
W. R.
,
1979
, “
Fission Gas Release and Grain Growth in ThO2-UO2 Fuel Irradiated at High Temperature
,” Bettis Atomic Lab, Pittsburgh, PA, Technical Report No. WAPDTM-1350.
41.
Cognet
,
G.
,
Efanov
,
A.
,
Fortov
,
V.
,
Fink
,
J. K.
,
Froment
,
K.
,
Gromov
,
G.
,
Hwang
,
I. S.
,
Jaroma-Weiland
,
G.
,
Jeong
,
K. J.
,
Jiang
,
Y.
,
Kim
,
Y. S.
,
Mares
,
R.
,
Matthew
,
P. M.
,
Petoukhov
,
V.
,
Piluso
,
P.
,
Sengupta
,
A. K.
,
Venugopal
,
V.
, and
Vinogradov
,
V.
,
2006
,
Thermophysical Properties Database of Materials for Light Water Reactors and Heavy Water Reactors
,
International Atomic Energy Agency
,
Vienna, Austria
.
42.
Konings
,
R.
,
2012
,
Comprehensive Nuclear Materials
,
Elsevier
, Amsterdam, The Netherlands.
43.
Long
,
Y.
,
Siefken
,
L. J.
,
Hejzlar
,
P.
,
Loewen
,
E. P.
,
Hohorst
,
J. K.
,
MacDonald
,
P. E.
, and
Kazimi
,
M. S.
,
2004
, “
The Behavior of ThO2-Based Fuel Rods During Normal Operation and Transient Events in LWRs
,”
Nucl. Technol.
,
147
(
1
), pp.
120
–1
39
.
44.
Insulander Björk
,
K.
, and
Kekkonen
,
L.
,
2015
, “
Thermal–Mechanical Performance Modeling of Thorium–Plutonium Oxide Fuel and Comparison With On-Line Irradiation Data
,”
J. Nucl. Mater.
,
467
(
Pt. 2
), pp.
876
885
.
45.
Boer
,
B.
,
Lemehov
,
S.
,
Wéber
,
M.
,
Parthoens
,
Y.
,
Gysemans
,
M.
,
McGinley
,
J.
,
Somers
,
J.
, and
Verwerft
,
M.
,
2016
, “
Irradiation Performance of (Th,Pu)O2 Fuel Under Pressurized Water Reactor Conditions
,”
J. Nucl. Mater.
,
471
, pp.
97
109
.
You do not currently have access to this content.