Numerical Simulation of the Thermal Performance ofa Dry Storage Cask for Spent Nuclear Fuel
Chi-Ming Lai, University Distinguished Professor
Department of Civil Engineering, National Cheng Kung University
1, University Rd., Tainan City, 701, Taiwan R. O. C.
Rong-Horng Chen, University Distinguished Professor
Department of Mechanical and Energy Engineering, National Chiayi University
300, Syuefu Rd., Chiayi City, 600, Taiwan R. O. C.
Chu-Tsen Liao, Assistant Professor
Department of Creative Design and Architecture, National Kaohsiung University
700, Kaohsiung University Rd., Kaohsiung City, 811, Taiwan R. O. C.
Ting-Yi Wang, Undergraduate Student
Department of Civil and Environmental Engineering, National Kaohsiung University
700, Kaohsiung University Rd., Kaohsiung City, 811, Taiwan R. O. C.
Heui-Yung Chang, Associate Professor
Department of Civil and Environmental Engineering, National Kaohsiung University
700, Kaohsiung University Rd., Kaohsiung City, 811, Taiwan R. O. C.

Keywords: dry storage cask; spent nuclear fuel; salt particle collection; thermal performance; heat transfer

Abstract
In this study, the heat flow characteristics of a dry storage cask were investigated and the effect of installing asalt particle collection device on the thermal performance of the dry storage cask was evaluated, via thermal flow
experiments and a computational fluid dynamics (CFD) simulation. The results indicate that there are many
inner circulations in the cask’s flow channel (the channel width is 10 cm). They affect the channel airflow
efficiency, which in turn affects the heat dissipation of the dry storage cask. The daily operating temperatures of
the top concrete lid and the upper locations of the concrete cask are higher than permitted by the design
specification. The installation of the salt particle collection device has no negative impact on the thermal
dissipation performance of the dry storage cask.

1. INTRODUCTION
Taiwan's first and second nuclear power plants use
boiling water reactors (BWRs) made by General
Electric Corporation (United States). In accordancewith the practices abroad, the Taiwan Power
Company (Taipower) built dry storage facilities at its
nuclear power plants. After using the nuclear fuel formany years, Taipower removed the spent fuel fromthe reactor core, cooled it in pools, and then stored
the spent fuel in dry storage facilities to ensure thatthe plants had sufficient space in which to operate. At
present, the dry storage facility at nuclear power plant
No. 1 has been tested. The safety review of the dry
storage facilities of nuclear power plant No. 2 has
also been completed.
There are few studies regarding the heat transfer
performance of dry spent fuel storage casks. Heng etal. (2002) numerically investigated naturalconvection heat transfer in a horizontally placed dry
spent-fuel storage cask. Pugliese et al. (2010)
evaluated the integrity of a spent fuel cask under bothnormal and accident scenario transport conditions,such as impact and rigorous fire events, according to
the International Atomic Energy Agency (IAEA)
accident test requirements. Koga and Tominaga(2008) investigated the characteristic cooling flow in
the annular gap of a concrete cask used to store spent
nuclear fuel via a simplified test model. Li and Liu(2016) studied the thermal performance of a verticaldry storage cask with a welded canister containing
high-burnup fuel. Kim et al. (2014) investigated heattransfer in a concrete cask such as the one used at theintermediate storage facilities for BWR spent fuel.
Thermal modeling of the temperature profiles ofdry casks has been identified as a high-priority item
in a U.S. Department of Energy gap analysis. In this

6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japanwork, a two-dimensional thermal model of a verticaldry cask has been established via a computational
fluid dynamics (CFD) simulation.
2. RESEARCH METHODS
2.1 Dry storage cask
Dry storage casks are used to store spent nuclear
fuel from nuclear power plants. Such casks primarily
consist of a nuclear waste steel canister, a concretecask and a transfer cask. In this paper, theinvestigated model is based on the dry storage caskthat will be employed in Taipower’s nuclear power
plant No. 2. The layout of the cask is shown in Fig. 1;
it is similar to the storage system in NAC's
MAGNASTOR-87. One difference is the concrete
cask’s thickness, which is altered to accommodate theradiation dose limit of the field boundary. Another
difference is that an anti-topple lug is installed tofacilitate transportation of the concrete cask in theplant (Taiwan Power Company, 2011).

Taipower’s No. 2 nuclear power plant will store
the spent nuclear fuel from the complete GE8x8-2
and ANF8x8-2 rods. The maximum initial averageenrichment concentration of U-235 is 3.25% byweight. The maximum average fuel consumption is
35,000 MWD/MTU. The minimum cooling time is
20 years. The maximum decay heat of each fuelbundle is conservatively estimated to be 0.168 kW.
Each steel canister can load 87 bundles of nuclear
fuel, so the maximum design thermal load of eachsteel canister is 14.6 kW (1.15 kW/m3).
For the specified parameters of the BWR fuelstored in the MAGNASTOR system, the maximum
initial average enrichment concentration is 3.8% byweight. The maximum average fuel consumption is60,000 MWD/MTU. The minimum cooling time is 4
years. The maximum decay heat of each bundle is0.379 kW. The maximum designed thermal load of
the sealed steel cylinder is up to 33.0 kW (2.6 kW/m3)
when 87 groups of fuel bundles are stored. Therefore,
in this study, a thermal load of 1.15 to 2.6 kW/m3 foreach steel canister is employed in the CFD
simulation.
2.2 The investigated modelFig. 2 Schematic diagram of the investigated model(dry storage cask)

Fig. 1 Design diagram of dry storage cask

Fig. 3 Schematic diagram of the investigated SPCD

6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

Fig. 2(a) shows the object of this study. It issimplified from Fig 1. Inside the cask, there is a
nuclear waste canister containing spent nuclear fuel(symbol). Its diameter is 1.8 m, and its height is 5
m. The concrete cask surrounding the steel canister
has a thickness of approximately 1 m (). A flow
channel () of width 10 cm is located between the
nuclear waste canister and concrete cask. The decay
heat of the nuclear fuel heats the air in the channel.
The heated air rises to the channel exit because of thebuoyancy effect () and exits from the dry storage
cask. The external air enters the channel from thebottom (), thus forming a natural ventilation path
(). In addition to the thermal buoyancy ventilation,
the ventilation of the dry storage cask caused byexternal wind is also taken into account. It should be
noted that when the two natural ventilation airflows(airflow induced by buoyancy and wind) merge andconflict, the thermal dissipation performance of thedry storage cask will be compromised. This study
does not discuss this topic. It will be addressed infollow-up studies.
In this study, a salt particle collection device(SPCD, ) was installed at the inlet (), as shown in
Fig. 3. The effect of this device on the heatdissipation performance of the dry storage cask was
observed. The height of the SPCD is 0.3 m (Ho), andthe width is 0.3 m (B). The length of the collection
plate and the length L of the SPCD are 0.5 and 1.0 m,
respectively. The plate number is 15. The salt capture
ratio is approximately 40%.
2.3 CFD simulation
The physical problem under consideration was
numerically simulated using a finite volume method
to solve the governing equations with theaforementioned boundary conditions. The calculation
domain is shown in Fig. 2 (b). A commercial CFD
code, PHOENICS, was used to simulate the airflow
and temperature distributions. The governing
equations solved by PHOENICS include thetwo-dimensional
incompressible
Navier-Stokes
equation and the convection diffusion equation withthe laminar model. These equations can be found in
the PHOENICS user manual (Spalding, 2012) or any
CFD textbook and thus will not be provided here. A
general wall function was employed to bridge thesteep variable dependent gradients near the solid
surface. The iterative calculation was continued until
a prescribed relative convergence of 10-3 wassatisfied for all the field variables in this problem.
The accuracy of the numerical simulation depended
on the resolution of the computational mesh, and afiner grid produced more accurate solutions. A gridwith approximately 195 x 105 cells was used for thenumerical simulation in this study. Increasing the cellnumber increased the solution accuracy at theexpense of significantly increasing the computational
resources required.

2.4 Key-section full-scale experiment
In this study, a full-scale test cell for key sections(as shown in the red dotted area of Fig. 4(a)) of thedry storage cask was built and the heat flow patternsof the test cell was experimentally observed (asshown in Figs. 4 (b) and (c)). The experimental
results were compared with those obtained from the
CFD simulation. The heating power of the spent
nuclear fuel in the nuclear waste canister was set to
be 400 W/m2. The length of the plate in the SPCD
was 0.5 m.
(1) Hot wall with heating flux
The hot wall (shown in Fig. 4 (b)) is formed byattaching a mica-type heater to a copper plate tosimulate the heat generation of spent nuclear fuel.
The output power of the heating plate is 400 W/m2.
The thickness of the copper plate is 1 cm. Its function
is to ensure that the heating wall is uniformly heated.
To control the output power, a power supply isconnected to a power meter (TES WM-01) and theheating plate.

Fig. 4 Full-scale experimental model (not to scale)Fig. 5 Comparison of simulation results with themeasurement

6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan(2) Temperature, heat flow and channel airflow
velocity measurements
In this study, type K thermocouples were used tomeasure the temperature at multiple locations. The
temperatures measured were the surface temperatureof the heating plate, the air temperatures at the inlet
and outlet of the channels (10 measuring locations
each at the inlet and outlet), and the channel wall
temperature of the concrete cask. A heat flow meter
was placed between the heating plate and the copper
plate
tomeasuretheheating
power.
Micro-anemometers were placed at both the entrance(10 locations) and the exit (5 locations) of the flow
channel.

nuclear waste canister, which in turn affects thetemperature distribution in the dry storage cask andleads to a thermal stress problem. On the other hand,owing to the Coanda effect, the high-temperature
airflow existing in the dry storage cask merges withthe high-temperature airflow from the symmetricalends at the upper-middle section of the dry storage
cask (shown by the symbol ). The high temperature
airflow enclosing the top concrete lid makes heatdissipation at the lid difficult (denoted by the symbol
 in Fig. 5(d)). It even affects the temperature of theconcrete cask (indicated by the symbol  in Fig.
6(d)).

(3) Measurement apparatus
The device and data acquisition system included
Data Acquisition Units (YOKOGAWA MX-100), a
personal computer (PC), heat flux sensors (EKO
Instruments MF-180), anemometers (KANOMAX
Model 1560) and AC/DC power supply units(Gwinstek APS-1102).
Fig. 5 shows that there was no significant
difference between the CFD results and experiment;
thus, the reliability of the simulation results was
confirmed.
3. RESULTS AND DISCUSSION
In this study, the controlling factors included
whether the SPCD was used, the length of the SPCD
plate (50 or 100 cm), the heating power of the sealed
steel canister (1.15 and 2.6 kW/m3), and the ambient
temperature (25, 30, and 35°C). Based on the design
specifications of Taipower (2011), the evaluation
criteria were defined as follows: The maximum
acceptable temperature for the fuel steel basket was
570°C (abnormal and accidental conditions). The
acceptable temperatures for the concrete was either
93.3°C (the average temperature under normal
conditions),
148.8°C(themaximum local
temperature under normal conditions), or 176.6°C(the maximum local temperature under abnormal or
accidental conditions).
3.1 Flow and thermal characteristics of the design
configuration(1) Flow pattern observation
In this section, the flow and thermal characteristicsof the dry storage cask were discussed under thefollowing conditions: a 50-cm SPCD is used, the heatdissipation rate of the sealed steel canister is 2.6
kW/m3, and the ambient temperature is 30°C. Fig. 6(a) shows the flow velocity distribution. Figs. 6(b)-(c)show the flow pattern. It can be observed from thesefigures that after the external cold air flows into thedry storage cask from the bottom, the flow separates
at the flow’s turning locations -. There exist
multiple inner circulations in the channel. This affects
the heat transferred by the channel airflow from theFig. 6 Flow and thermal characteristics of the dry
storage cask (with the 50-cm-SPCD). (The heatdissipation rate of the sealed steel canister is 2.6
W/m3. The ambient temperature is 30°C)(2) Thermal performanceFig. 7 shows the temperature distribution of thedry storage cask at locations of Y = 3, 2, and 1.5 m.

6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

This figure shows that the temperature of the nuclear
waste canister was approximately 334-342°C, whichis less than the evaluation criterion of 570 °C (themaximum acceptable temperature under abnormal or
accidental conditions). The temperature of the top
concrete lid on the concrete box was as high as 232°C(at height Z = 5.6 m), which is greater than the
maximum acceptable temperature (176.6°C, indicatedby the red solid line) of concrete under abnormal or
accidental conditions. Moreover, the temperatures atthe lower and upper halves of the top concrete lid
were both greater than the respective maximum
acceptable temperatures (176.6°C for the lower halfand 148.8°C for the upper half, as indicated by the
red dotted line).
Fig. 7 (b) shows that the maximum temperature of
the concrete cask was 85.2°C (at height Z = 5.53 m),
which satisfies the design requirements. The
temperature of the top concrete lid was greater than
the acceptable average temperature of the concrete(93.3°C, indicated by the red chain line). Fig. 7(c)shows that the maximum temperature of the concretecask was 123°C (at height Z = 5.6 m), which isgreater than the acceptable average temperature of
93.3°C under normal conditions. The temperature of
the lid was also greater than 93.3°C.
Therefore, as a whole, the daily operating
temperatures of the top concrete lid and at highlocations on the concrete cask (indicated by thesymbol  in Fig. 6(d)) were greater than permitted
by the design specifications. Power plant designers
should pay attention to the possible structural damage
caused by this overheating. Possible solutions to thethermal stress problem discussed above include thefollowing:(1) Improve the design of the flow channels. The
current ventilation flow is through structural
spacing. No formal ventilation is designed. If theexisting construction cannot be altered, theauthors suggest that guide vanes be installed tosmooth the airflow, thus increasing the channel
airflow’s heat dissipation efficiency.
(2) Increase the RC’s strength to enhance its
resistance to the thermal stress generated.
(3) Use a thermal dissipation design on the top
concrete lid and high locations of the concretecask. The design may include utilization of
additives to improve the RC’s heat transfer
coefficient and installation of heat pipes, among
other measures, to ensure that the heat gained
inside the RC is smoothly transferred to theoutside.
Fig.7 Temperature changes along the height of thedry storage cask (at Y = 3, 2, and 1.5 m) (with the
50-cm SPCD). (The heat dissipation rate of the sealed
steel canister is 2.6 kW/m3. The ambient temperature
is 30°C)

In addition to the heat dissipation problem
associated with a single dry storage cask, if several
dry storage casks are placed together or external air
affects the overall heat dissipation (such as the

6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japanconflict created by thermal buoyancy-induced airflow
and wind, as mentioned in paragraph 1 of Section2.2), the risk of structural damage caused by poor
heat dissipation will be even greater.

3.2 The impact of the SPCD
Compared to devices without SPCDs, utilizing a
50-cm SPCD can increase the heat transferred by thechannel airflow by approximately 7.6%. Utilizing a
100-cm SPCD increases the heat transferred via thechannel airflow by approximately 7.1% (in both cases,
the heating power of the sealed steel canister ranges
from 1.15 to 2.6 kW/m3). Since the SPCD has aplate-like configuration, when the 50-cm SPCD unit
is installed in the dry storage cask, its
rectification-like effect (Fig. 2  and Fig. 3)
increases the average rate of the channel airflow(compared with cases without SPCDs). However,when the plate length is increased from 50 to 100 cm,
the average flow rate is slightly reduced owing to theincrease in the flow friction. Compared to devices
without SPCDs, the average flow rate of the dry
storage cask with a 100-cm SPCD is greater, and thus,
more heat is transferred by the channel airflow.

REFERENCE
Kim H.M., No H.C., Bang K.S., Seo K.S., Lee
S.H., Development of scaling laws of heat removal
and CFD assessment in concrete cask air path,
Nuclear Engineering and Design 278 (2014) 7-16.
Koga T., Tominaga Y., Heat removal
characteristics of a concrete cask by a simplified test
model, Nuclear Engineering and Design 238 (2008)
1189-1195.
Li J., Liu Y.Y., Thermal modeling of a vertical dry
storage cask for used nuclear fuel, Nuclear
Engineering and Design 301 (2016) 74-88.
Pugliese G., Lo Frano R., Forasassi G., Spent fueltransport cask thermal evaluation under normal andaccident conditions, Nuclear Engineering and Design
240 (2010) 1699-1706.
Spalding DB. The PHOENICS Encyclopedia.
London: CHAM Ltd.; 2012.
Taiwan Power Company: Analysis Report on the
Safety of a Dry Storage System in the No.2 Nuclear
Power Plant, December 2011. (in Chinese)
Xie H., Gao Z., Zhou Z., A numerical investigationof natural convection heat transfer in horizontal
spent-fuel storage cask, Nuclear Engineering and
Design 213 (2002) 59-65.

4. CONCLUSION
In this study, based on a CFD simulation and heatflow experiments, the heat flux characteristics of adry storage cask were investigated and the effect ofan SPCD on the heat dissipation performance of thedry storage cask was evaluated. The object studied
was a dry storage cask with nuclear waste canisters(with a diameter of 1.8 m and a height of 5 m)
containing spent nuclear fuel and a surrounding
1-m-thick concrete cask. The flow channel between
the nuclear waste canister and concrete casks has aspacing of approximately 10 cm. The results
indicated that there are multiple inner circulations in
the airflow channel. This affects the heat transferredby the channel airflow from the nuclear waste
canister, which in turn affects the temperature
distribution of the dry storage cask and results inthermal stress. Overall, the daily operating
temperatures in the top concrete cover lid and theupper locations of the concrete cask were greater thanpermitted by the design specification. Power plantdesigners should pay attention to the possible
structural damage caused by this overheating. The
SPCD does not adversely affect the heat dissipation
performance of the dry storage cask.

ACKNOWLEDGEMENT
The authors gratefully acknowledge the financial
support from the Ministry of Science and Technology,
R.O.C. under Grant No. 106-NU-E-390-001-NU. The
authors also thank the Fuel Cycle and Materials
Administration, Atomic Energy Council, R.O.C. forproviding their insight and expertise that greatly
assisted the research.

6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan