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Alternative Fueling Scheme for the Indonesian Experimental Power Reactor
(10 MWth Pebble-Bed HTGR)
Article in Energy Procedia · December 2017
DOI: 10.1016/j.egypro.2017.09.477
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Hoai-Nam Tran
NAIS Co., Inc.
Phenikaa University
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PT ThorCon Power Indonesia
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Energy Procedia 00 (2017) 000–000
ScienceDirect
ScienceDirect
www.elsevier.com/locate/procedia
Energy Procedia
Procedia 00
131(2017)
(2017)000–000
69–76
Energy
www.elsevier.com/locate/procedia
5th International Symposium on Innovative Nuclear Energy Systems, INES-5, 31 October – 2
November, 2016, Ookayama Campus, Tokyo Institute of Technology, JAPAN
AlternativeThe
Fueling
SchemeSymposium
for the Indonesian
Experimental
15th International
on District Heating
and Cooling Power
Reactor (10 MWth Pebble-Bed HTGR)
Assessing the feasibility of using the heat demand-outdoor
a
b
c
Peng Hong LIEM
, Hoai-Nam
, Tagor Malem
SEMBIRING
, Bakri forecast
ARBIEd,
temperature
function
forTRAN
a long-term
district
heat
demand
d
Iyos SUBKI
a
b
c
I. Andrić
*, A.
Pina Service
, P. Ferrão
J. Fournier
., B.
Lacarrière
, O. Le
Correc
Nippon Advanced
Information
(NAIS Co.,, Inc.),
416 Muramatsu,
Tokaimura,
Ibaraki 319-1112,
Japan
a
a,b,c
a
b
Institute of Research and Development, Duy Tan University, K7/25 Quang Trung, Da Nang, Vietnam
a
IN+ Center for Innovation, Technology and Policy Research - Instituto Superior Técnico, Av. Rovisco Pais 1, 1049-001 Lisbon, Portugal
National Nuclear Energy Agency
of Indonesia (BATAN), Kawasan Puspiptek Gd. No. 80 Serpong, Tangerang Selatan 15310, Indonesia
b
Veolia Recherche & Innovation, 291 Avenue Dreyfous Daniel, 78520 Limay, France
d
PT MOTAB Technology, Kedoya Elok Plaza Blok DA 12, Jl. Panjang, Kebun Jeruk, Jakarta Barat, Indonesia
c
Département Systèmes Énergétiques et Environnement - IMT Atlantique, 4 rue Alfred Kastler, 44300 Nantes, France
c
Abstract
Abstract
The National Nuclear Energy Agency of Indonesia (BATAN) is launching a plan (2014) to build an 10 MWth Experimental
Power Reactor (Reaktor Daya Eksperimental, RDE) in the Agency’s largest Research Center site, i.e. the Puspiptek Complex,
District heating networks are commonly addressed in the literature as one of the most effective solutions for decreasing the
Serpong,
South
a first strategic
milestone
for the
introduction
of large scale
nuclear
power plant
fleets
greenhouse
gasTangerang,
emissions Banten,
from theasbuilding
sector. These
systems
require
high investments
which
are returned
through
theinto
heat
the
country.
The
main
objective
of
the
plan
is
to
demonstrate
safe
and
reliable
electricity
and
process
heat
generation
from
a
sales. Due to the changed climate conditions and building renovation policies, heat demand in the future could decrease,
nuclear reactor. The RDE is a very small sized pebble-bed high temperature gas-cooled reactor (HTGR) with low enriched
prolonging the investment return period.
TRISO
previous
designthe
study
multipass
fueling
scheme, function
a scopingfor
study
fuel
uranium
(LEU)
The main
scopeUO
of 2this
paperfuel.
is to Following
assess the our
feasibility
of using
heatwith
demand
– outdoor
temperature
heat on
demand
composition
parameters,
namely
heavy
metal
(HM)
loading
per
pebble
and
uranium
enrichment
was
conducted
for
the
simpler
forecast. The district of Alvalade, located in Lisbon (Portugal), was used as a case study. The district is consisted of 665
once-through-then-out (OTTO) fueling scheme. No need of burnup measurement devices and refueling mechanism to send back
buildings that vary in both construction period and typology. Three weather scenarios (low, medium, high) and three district
discharged
pebbles
intodeveloped
the core, (shallow,
low power
and powerdeep).
density,
simple
of the heat
RDEdemand
becomevalues
strongest
renovationfuel
scenarios
were
intermediate,
To and
estimate
theoperation
error, obtained
were
motivation
for
the
present
work.
The
objective
function
for
the
optimization
is
the
fissile
loading
requirement
per
energy
compared with results from a dynamic heat demand model, previously developed and validated by the authors.
generated (kg/GWd). The optimal heavy metal loading was found around 8 g/pebble while the uranium enrichment
The results showed that when only weather change is considered, the margin of error could be acceptable for some applications
corresponding
to the 80
GWd/twas
discharge
burnup
is approximately
13.7 w/o. Compared
burnup calculations
(the error in annual
demand
lower than
20%constraint
for all weather
scenarios considered).
However,with
aftertheintroducing
renovation
results
of
multipass
fueling
scheme,
OTTO
fueling
scheme
burnup
performance
is
slightly
inferior
in
that
slightly
higher
scenarios, the error value increased up to 59.5% (depending on the weather and renovation scenarios combination
considered).
enrichment and fissile loading per pebble are required to achieved the same average discharge burnup.
The value of slope coefficient increased on average within the range of 3.8% up to 8% per decade, that corresponds to the
decrease in the number of heating hours of 22-139h during the heating season (depending on the combination of weather and
©renovation
2017 The Authors.
by Elsevier
Ltd. hand, function intercept increased for 7.8-12.7% per decade (depending on the
scenariosPublished
considered).
On the other
Peer-review
under
responsibility
of
the
organizing
the 5ththe
International
Symposiumfor
onthe
Innovative
Nuclear
Energyand
coupled scenarios). The values suggested could committee
be used toofmodify
function parameters
scenarios
considered,
Systems.
improve the accuracy of heat demand estimations.
Keywords: Indonesian Experimental Power Reactor (RDE), Pebble-Bed Reactor, HTGR, OTTO, Multipass, uranium fuel cycle
© 2017 The Authors. Published by Elsevier Ltd.
Peer-review under responsibility of the Scientific Committee of The 15th International Symposium on District Heating and
Cooling.
* Corresponding author. Tel.: +81-29-2705000; fax: +81-29-2705001. E-mail address: liemph@nais.ne.jp
Keywords: Heat demand; Forecast; Climate change
1876-6102 © 2017 The Authors. Published by Elsevier Ltd.
Peer-review under responsibility of the organizing committee of the 5th International Symposium on Innovative Nuclear Energy Systems.
1876-6102 © 2017 The Authors. Published by Elsevier Ltd.
Peer-review under responsibility of the Scientific Committee of The 15th International Symposium on District Heating and Cooling.
1876-6102 © 2017 The Authors. Published by Elsevier Ltd.
Peer-review under responsibility of the organizing committee of the 5th International Symposium on Innovative Nuclear Energy Systems.
10.1016/j.egypro.2017.09.477
Peng Hong Liem et al. / Energy Procedia 131 (2017) 69–76
Author name / Energy Procedia 00 (2017) 000–000
70
2
1. Introduction
The National Nuclear Energy Agency of Indonesia (BATAN) is launching a plan (2014) to build an Experimental
Power Reactor (Reaktor Daya Eksperimental, RDE) [1] in the Agency’s largest Research Center site, i.e. the
Puspiptek Complex, Serpong, South Tangerang, Banten, as a first strategic milestone for the introduction of large
scale nuclear power plant fleets into the country. The main objective of the plan is to demonstrate safe and reliable
electricity and process heat generation from a nuclear reactor. According to the User Requirement Document (URD)
[1] together with its technical specification documents (TSD) issued by BATAN, the RDE adopted a very small
sized pebble-bed high temperature gas-cooled reactor (HTGR) with a thermal output of 10 MWth, fueled with low
enriched uranium (LEU) UO 2 TRISO fuel under multipass or once-through-then-out (OTTO) fueling scheme. The
thermal heat generated in the core is transferred to steam generator (for electricity generation) and a heat utilization
plant via an intermediate heat exchanger (IHX). The use of IHX is expected to ensure a high degree of plant safety
especially the steam/water ingress events can be avoided.
In our previous design study [2], a scoping study on fuel composition parameters, namely heavy metal (HM)
loading per pebble and uranium enrichment was conducted for the multipass fueling scheme. In the present work, we
extend our scoping study to include the once-through-then-out (OTTO) fueling scheme. The OTTO fueling scheme
has several advantages over the multipass fueling scheme in that no need of burnup measurement devices and
refueling mechanism to send back discharged fuel pebbles into the core. The low power and power density, and
simple operation of the RDE are the strongest motivations for the present work.
As already shown and explored in Ref. [2], the URD and its TSD provide general main specifications of the RDE,
however, many detail design parameters on the core dimension, fuel composition etc. are given in term of ranges of
value which should be further optimized during the conceptual, basic and detail design phases. An optimal fuel
composition in the operation of a pebble-bed HTGR is a very important design parameter since it will directly affect
the fuel cost, new and spent fuel storage capacity as well as other back-end environmental burden.
Similar to our previous work, the present scoping study on the fuel composition parameters, namely heavy metal
(HM) loading per pebble and uranium enrichment is conducted. The main goal of this study is to obtain optimal
ranges of HM loading per pebble and uranium enrichment for the RDE design. HM loading per pebble strongly
affects the neutron moderation (core neutron spectrum) while uranium enrichment is correlated directly with the
achievable discharge burnup. The present work is expected to contribute in providing an optimal fuel composition
under OTTO fueling scheme which can be directly compared with the results of the multipass fueling scheme.
Table 1. Main design parameters and constraints of the 10 MWth RDE.
Design Parameters
Thermal Power (MW)
Core Diameter (m)
Core Height/Diameter Ratio
Ave. Core Power Density
(PD, W/cm3)
Upper Core Void Height (m)
Radial Reflector Thickness (m)
Upper and Lower Reflector Thickness (m)
235
U Enrichment (w/o)
HM Loading (g/pebble)
Ave. Discharge Burnup (GWd/t)
Fueling Scheme
He Inlet/Outlet Temp. (C)
He Inlet Pressure (MPa)
Specifications Issued
by URD/TSD
10
< 2.5
>1.1
2.0 ≤ PD ≤ 3.0
≤ 17
≤ 20
80
Multipass/
OTTO
250/700
3
Present Analysis
10
1.8 [3]
1.1
(Height = 1.97 m)
2.0
(Max. Core Vol.)
0.4 [3]
0.5
1.0
10 – 20 (LEU)
4 – 10
80
OTTO
(10 flow paths)
250/700
3
Peng Hong Liem et al. / Energy Procedia 131 (2017) 69–76
Author name / Energy Procedia 00 (2017) 000–000
71
3
2. RDE Technical Specifications
In Table 1, the main design parameters and constraints of the RDE are listed. The parameters and constraints
shown in the second column of the table were taken from the URD/TSD issued by BATAN. The thermal output,
discharge burnup and the coolant inlet/outlet temperature and pressure are already fixed. As for the core dimension,
three design constraints are imposed, that is, the maximum core diameter, minimum height/diameter ratio and
average core power density. On the other hand, as for the fuel composition, only the maximum uranium enrichment
(17 w/o) and maximum HM loading (20 g/pebble) are imposed. The third column of Table 1 gives the design
parameters taken for the present scoping study. The core diameter is set to 1.8 m (derived from the Chinese 10
MWth HTR-10 [3]) while the core height/diameter ratio and average power density are set to its minimum and
maximum value, i.e. 1.1 and 2 W/cm3, respectively, to obtain good neutron economics. The void cavity at the top of
the core is necessary and its height is set to be approximately 40 cm [3]. As for the fresh fuel composition, the
scoping study is conducted over uranium enrichment in the range of 10 to 20 w/o and HM loading in the range of 4
to 10 g/pebble. Cases of uranium enrichment higher than the maximum design constraint (17 w/o) are also
considered for the sake of completeness since the possible maximum uranium enrichment for LEU is 20 w/o. Cases
with HM loading less than 4 or higher than 10 g/pebble are excluded through preliminary screening calculations.
The URD issued by BATAN allows the RDE to be operated under the once-through-then-out (OTTO) fueling
scheme, however no detail specification on the radial zoning is given. In the present work, the pebble flow is
simulated with 10 flow paths (flow path index of 1 to 10, index=1 is in the core center and index=10 is adjacent to
the radial reflector). Since detail fuel specification and core-reflector structures and dimensions etc. are not presently
available, as in our previous work [2] we assume no boron (equivalent) impurity in the fuel, 50 cm (radial) and 100
cm (axial) thickness of pure graphite reflectors with graphite density of 1.75 g/cm3.
3. Analytical Code and Group Constants
The pebble fuel movement, burnup, core criticality calculations and core equilibrium search are carried out by an
upgraded version of BATAN-MPASS [4], a general in-core fuel management code for pebble-bed HTGRs, featured
with many automatic equilibrium and criticality searching options as well as thermal-hydraulic calculation capability.
The code has been validated with the German HTR-Module design, the validation results have also been used as a
comparative solution for other code [5]. The TRISO coated particle fuel and pebble fuel element specifications are
shown in Table 2. The microscopic cross-sections (4 energy groups, Table 2) and their self-shielding factors as a
function of temperature and composition were prepared using several parts of the VSOP code system [6]: ZUT-DGL,
THERMOS and GAM. In the upgraded version, amongst other additional options, users may specify a different
fissile enrichment for each flow path.
In Table 3, the heavy metals, fission products, moderators and poisons nuclides used in the burnup chain of
BATAN-MPASS are listed. Group constants for the cavity at the top of the core are determined according to the
method developed by Gerwin and Scherer [7]. Using their method, different axial and radial diffusion coefficients
can be obtained. The detail discussion of the in-core thermal-hydraulic model used in the BATAN-MPASS code was
already given by Liem and Sekimoto [8]. By using the core equilibrium search option, all calculation cases
converged to an equilibrium core with effective multiplication factor (keff) equals to 1.0 (critical).
4. Results and Discussion
The scoping analysis results are shown in Figures 1 and 2. Figure 1 (a) shows the fissile loading requirement as a
function of HM loading for 4 values of uranium enrichment. The results for multipass fueling scheme are also shown
for comparison. For both fueling schemes, the fissile loading requirement per unit energy generated (kg/GWd)
decreases as the uranium enrichment increases. For all values of uranium enrichment, one can find an optimal value
of HM loading per pebble, which is around 8 g/pebble for both fueling schemes. These optimal compositions also
indicate the optimal moderation for the reactor.
Figure 1 (b) shows the pebble fuel element discharge burnup as a function of HM loading for 4 values of uranium
enrichment. The results for multipass fueling scheme are also shown for comparison. From the figure, it can be
Peng Hong Liem et al. / Energy Procedia 131 (2017) 69–76
Author name / Energy Procedia 00 (2017) 000–000
72
4
observed an almost linear proportionality of discharge burnup to uranium enrichment for both fueling schemes.
Similar to Figure 1(a), for a particular value of uranium enrichment one can find an optimal value of HM loading per
pebble, which is also around 8 g/pebble. If the design constraint of 80 GWd/t discharge burnup is imposed then for
the optimal HM loading of 8 g/pebble case, the uranium enrichment is found to be around 13.7 w/o and 13.1 w/o for
OTTO and multipass fueling scheme, respectively. For the case of optimal HM loading and uranium enrichment of
13.7 w/o, the pebble fuel residence time in the core is around 4.6 years (cf. Figure 2(a)).
Table 2. Standard data for pebble fuel element cell calculation.
TRISO coated particle fuel
Kernel
material
diameter (μm)
density (g/cm3)
material
thickness (μm)
density (g/cm3)
Coatings
(from inner)
(235U/238U)O 2
500
10.9
PyC/PyC/SiC/PyC
90/40/35/35
0.9/1.85/3.2/1.85
Pebble fuel element
Fuel matrix
material
diameter (cm)
material
thickness (cm)
diameter (cm)
Outer layer
Fuel element
graphite
5
graphite
0.5
6
Few group structure
Upper bound (eV)
107
1.11×105
2.90×101
2.38
Lower bound (eV)
1.11×105
2.90×101
2.38
～
Range
fast fission
slowing down
resonance
thermal
Table 3. Nuclides used in BATAN-MPASS code
Heavy metals
Fission products
Moderators
Poisons
Th,233Pa,233U,234U,235U,236U,237Np, 239Np,
Pu,240Pu,241Pu,242Pu
83
Kr,95Zr,95Mo,97Mo,99Tc,101Ru, 103Ru,
103
Rh,105Rh,105Pd,106Pd, 109Ag, 113Cd,131I,
131
Xe,133Xe,135Xe,136Xe,133Cs,134Cs,141Pr,
143
Pr,143Nd,145Nd,146Nd,147Pm,148mPm,
148g
Pm,147Sm,148Sm,149Sm,150Sm,151Sm,
152
Sm,153Eu,154Eu,155Eu,155Gd,156Gd,
157
Gd, Non-saturating FP
12 16
C, O
B, Impurity in C
232
239
Figure 2(b) shows the maximum power density as a function of HM loading. This parameter is sensitive to both
uranium enrichment and heavy metal loading per pebble. It can be observed that abrupt changes in the maximum
power density curves for uranium enrichment values of 15.0, 17.5 and 20.0 w/o. These trends can be explained from
the fact that higher HM loading will result in a longer pebble fuel residence time and the axial power density profile
will produce higher peak toward the core top. Nevertheless, even for 10 g/pebble HM loading cases the maximum
power density is still lower than 4 W/cm3 which is still smaller than the maximum allowable value.
Figure 2(c) shows the conversion ratio as a function of HM loading. As expected, lower uranium enrichment
gives better conversion ratio. Higher HM loading per pebble also significantly increases the conversion ratio. With
higher HM loading, the pebble fuel resides longer in the core and the fertile to fissile conversion is enhanced.
Peng Hong Liem et al. / Energy Procedia 131 (2017) 69–76
Author name / Energy Procedia 00 (2017) 000–000
3.00
1.8E +05
2.75
12.5 (O)
15.0 (O)
17.5 (O)
20.0 (O)
12.5 (M )
15.0 (M )
17.5 (M )
20.0 (M )
1.6E +05
1.4E +05
D ischarg e Burnup (M W d/t)
2.50
F issile Loading (kg /G W d)
73
5
2.25
2.00
1.75
1.50
1.2E +05
1.0E +05
8.0E +04
6.0E +04
4.0E +04
1.25
2.0E +04
1.00
12.5 (O)
15.0 (O)
17.5 (O)
20.0 (O)
12.5 (M )
15.0 (M )
17.5 (M )
20.0 (M )
0.0E +00
4
5
6
7
8
9
10
4
H ea vy M eta l Loading (g /pebble)
5
6
7
8
9
10
H ea vy M eta l Loading (g /pebble)
(a)
(b)
Fig. 1. (a) Fissile loading requirement; (b) Average discharge burnup as function of heavy metal loading and uranium enrichment (O=OTTO,
M=multipass).
For the optimal HM loading and uranium enrichment of 13.7 w/o, the fissile nuclide distributions are shown in
Figures 3 and 4, for 235U, 239Pu, and 241Pu, 237Np, respectively. From Figure 3(a), it can be observed that 235U is
depleted more in the outer pebble flow paths than the inner ones since the pebble flow velocities in the outer paths
are slower. Consequently, the discharged burnup levels are also higher for pebbles discharged from the outer flow
paths.
4500
4.00
0.50
12.5
4000
15.0
0.45
17.5
17.5
3.50
M a x. P ower Density (W /cm3)
20.0
3000
2500
2000
1500
1000
C onversion R a tio (-)
3500
R esidence Tim e (days)
12.5
3.75
15.0
3.25
3.00
2.75
20.0
0.40
0.35
0.30
12.5
2.50
15.0
500
2.25
0
2.00
0.25
17.5
20.0
4
5
6
7
8
9
H ea vy M eta l Loading (g /pebble)
(a)
10
0.20
4
5
6
7
8
9
H ea vy M eta l Loading (g /pebble)
(b)
10
4
5
6
7
8
9
10
H ea vy M eta l Loading (g /pebble)
(c)
Fig. 2. (a) Pebble fuel average residence time; (b) Maximum power density; (c) Conversion ratio as function of heavy metal loading and uranium
enrichment.
Peng Hong Liem et al. / Energy Procedia 131 (2017) 69–76
Author name / Energy Procedia 00 (2017) 000–000
74
6
Figures 3(b) and 4(a) show that fissile plutonium is produced and accumulated along the pebble flow paths,
however, at core lower regions the consumption rates are greater than the production rates so that the fissile
plutonium nuclide densities decrease. Figure 4(b) shows the 237Np nuclide densities as function of pebble flow path
and axial positions. For pebble flow path index 1 to 8, 237Np is produced and accumulated along the pebble flow
paths, however, for flow path index 9 and 10 near the radial reflector where the pebble flow velocities are slower, the
consumption rates are greater than the production rates at the core lower regions and their trends are similar with the
fissile plutonium shown in Figures 3(b) and 4(a).
1.50E+19
1.00E+18
1.25E+19
209.0
255.9
7.50E+18
306.0
356.2
5.00E+18
419.7
Nuclide Density (1/cm3)
Nuclide Density (1/cm3)
7.50E+17
1.00E+19
209.0
255.9
5.00E+17
306.0
356.2
419.7
2.50E+17
2.50E+18
0.00E+00
1
2
3
4
5
6
Pebble Flow Index
7
8
9
0.00E+00
10
1
2
3
4
(a) 235U
5
6
Pebble Flow Index
7
8
9
10
(b) 239Pu
Fig. 3. Nuclide densities as function of pebble flow index and axial position (core top = 209.0 cm, core bottom = 419.7 cm).
1.20E+17
4.00E+17
1.00E+17
209.0
255.9
2.00E+17
306.0
356.2
419.7
Nuclide Density (1/cm3)
Nuclide Density (1/cm3)
3.00E+17
8.00E+16
209.0
255.9
6.00E+16
306.0
356.2
4.00E+16
419.7
1.00E+17
2.00E+16
0.00E+00
1
2
3
4
5
6
Pebble Flow Index
7
8
9
(a) 241Pu
10
0.00E+00
1
2
3
4
5
6
Pebble Flow Index
7
8
9
10
(b) 237Np
Fig. 4. Nuclide densities as function of pebble flow index and axial position (core top = 209.0 cm, core bottom = 419.7 cm)
For the optimal HM loading and uranium enrichment of 13.7 w/o, the group neutron flux distributions are shown
in Figure 5, and the power density distributions are shown in Figure 6. High thermal neutron peaks appear at the
radial and axial reflector regions adjacent to the core. From Figure 6, one can observed that typical OTTO power
density distributions where the power peaks are located in the upper regions of the core.
5. Concluding Remarks and Future Works
A scoping study on the optimal fuel composition (heavy metal loading per pebble and uranium enrichment) for
the 10 MWth RDE, with LEU UO 2 TRISO fuel under OTTO fueling scheme, was conducted using an upgraded
Peng Hong Liem et al. / Energy Procedia 131 (2017) 69–76
Author name / Energy Procedia 00 (2017) 000–000
75
7
version of BATAN-MPASS code. The objective function for the optimization is the fissile loading requirement per
energy generated (kg/GWd). The optimal heavy metal loading was found around 8 g/pebble while the uranium
enrichment corresponding to the 80 GWd/t discharge burnup constraint is approximately 13.7 w/o. Compared with
the burnup calculations results of multipass fueling scheme [2], OTTO fueling scheme burnup performance is
slightly inferior in that slightly higher enrichment and fissile loading per pebble are required to achieved the same
average discharge burnup. However the optimal fuel compositions for the two fueling scheme are close to each other,
i.e. identical 8 g/pebble and uranium enrichment of 13.7 and 13.1 w/o for OTTO and multipass, respectively.
Z
Z
X
Y
2.20E+13
2.09E+13
1.98E+13
1.87E+13
1.76E+13
1.64E+13
1.53E+13
1.42E+13
1.31E+13
1.20E+13
1.09E+13
9.78E+12
8.67E+12
7.56E+12
6.44E+12
5.33E+12
4.22E+12
3.11E+12
2.00E+12
1.50E+13
1.00E+13
5.00E+12
500
400
300
Z(C
M)
200
100
00
50
0.00E+00
150
100
)
R(
Second Group Neutron Flux (n/cm2.s)
Fast Neutron Flux (n/cm2.s)
2.00E+13
4.00E+13
2.00E+13
1.00E+13
500
400
300
Z(C
M)
200
100
CM
00
50
0.00E+00
150
100
)
R(
CM
(Second Group; Slowing Down)
Z
Z
Y
X
8.00E+12
6.00E+12
4.00E+12
2.00E+12
400
300
M)
200
100
00
50
0.00E+00
150
100
)
R(
CM
(Third Group; Resonance)
9.00E+12
8.50E+12
8.00E+12
7.50E+12
7.00E+12
6.50E+12
6.00E+12
5.50E+12
5.00E+12
4.50E+12
4.00E+12
3.50E+12
3.00E+12
2.50E+12
2.00E+12
1.50E+12
1.00E+12
5.00E+11
Y
Thermal Neutron Flux (n/cm2.s)
Third Group Neutron Flux (n/cm2.s)
1.00E+13
Z(C
3.60E+13
3.40E+13
3.20E+13
3.00E+13
2.80E+13
2.60E+13
2.40E+13
2.20E+13
2.00E+13
1.80E+13
1.60E+13
1.40E+13
1.20E+13
1.00E+13
8.00E+12
6.00E+12
4.00E+12
2.00E+12
3.00E+13
(First Group; Fast)
500
X
Y
X
4.00E+13
3.00E+13
2.00E+13
1.00E+13
500
400
300
Z(C
M)
200
100
00
50
0.00E+00
150
100
)
R(
3.80E+13
3.60E+13
3.40E+13
3.20E+13
3.00E+13
2.80E+13
2.60E+13
2.40E+13
2.20E+13
2.00E+13
1.80E+13
1.60E+13
1.40E+13
1.20E+13
1.00E+13
8.00E+12
6.00E+12
4.00E+12
2.00E+12
CM
(Fourth Group; Thermal)
Fig. 5. Group neutron flux distribution for the optimal fuel composition case
The results of the present scoping study can be improved in the future when other detail design parameters
become available in the basic and detail design phases. These include the detail core-reflector structures and
dimensions, control rod and reserved shutdown absorber in the radial reflectors, impurities in fuel and graphite etc.
Peng Hong Liem et al. / Energy Procedia 131 (2017) 69–76
Author name / Energy Procedia 00 (2017) 000–000
76
8
The results of safety evaluation in the future may provide feedback to the present scoping study and the optimal
composition of the fuel may slightly shifts from the above mentioned values.
Z
Y
X
180
200
Top
Power Density (W/cm3)
3.00
2.90
2.79
2.69
2.59
2.48
2.38
2.27
2.17
2.07
1.96
1.86
1.76
1.65
1.55
1.44
1.34
1.24
1.13
1.03
0.93
0.82
0.72
0.61
0.51
0.41
0.30
0.20
220
2.00E+00
1.00E+00
500
240
2.60
2.40
2.20
2.00
1.80
1.60
1.40
1.20
1.00
0.80
0.60
0.40
0.20
260
280
Z(CM)
Power Density (W/cm3)
3.00E+00
300
320
340
360
400
300
Z(C
M)
200
100
00
50
0.00E+00
150
100
)
R(
CM
380
400
420
Bottom
50
(a)
100
150
R(CM)
200
250
(b)
Fig. 6. (a) Power density spatial distribution; (b) Power density along the pebble flow path for the optimal fuel composition case
Acknowledgements
The authors are very grateful to Ms. Suzuko Ikegami for preparing and checking carefully the tables and figures.
References
[1] User Requirement Document – Reaktor Daya Eksperimental –. BATAN, 2014.
[2] Liem PH, Sembiring TM, Arbie B, Subki I. Analysis of the Optimal Fuel Composition for the Indonesian Experimental Power Reactor (10
MWth Pebble-Bed HTGR). To be published in Kerntechnik 2016.
[3] Xu Y, Zuo K. Overview of the 10 MW High Temperature Gas Cooled Reactor – Test Module Project. Nuclear Engineering and Design 2002;
218: 13-23.
[4] Liem PH. BATAN-MPASS: A general fuel management code for pebble-bed high-temperature reactors. Annals of Nuclear Energy 1994; 21
(5): 281-290.
[5] Tery WK, Gougar HD, Ougouag AM. Direct deterministic method for neutronics analysis and computation of asymptotic burnup distribution
in a recirculating pebble-bed reactor. Annals of Nuclear Energy 2002; 29: 1345-1364.
[6] Teuchert E et al. VSOP – Computer Code System for Reactor Physics and Fuel Cycle Simulation. Jul-1649 1980.
[7] Gerwin H, Scherer W. Treatment of the Upper Cavity in a Pebble-Bed High Temperature Gas-Cooled Reactor by Diffusion Theory. Nuclear
Science and Engineering 1987; 97: 9-19.
[8] Liem PH, Sekimoto H. Neutronic and Thermal Hydraulic Design of the Graphite Moderated Helium-Cooled High Flux Reactor. Nuclear
Engineering and Design 1993; 139 (2): 221-233.
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