64 
 

Computational and Experimental Research  
in Materials and Renewable Energy (CERiMRE)         
Volume 2, Issue 2, page 64-71 
eISSN : 2747-173X 
 
 

Submitted : July 8, 2019 
Accepted  : September 8, 2019 
Online  : November 24, 2019 
doi : 10.19184/cerimre.v2i2.27368 

Study of Neptunium, Americium and Protactinium Addition for 
300MWth GFR with Uranium Carbide Fuel  

 

Ratna Dewi Syarifah
1,a

 and Alvi Nur Sabrina
2
 

1
Department of Physics, Faculty of Mathematics and Natural Sciences, Universitas Jember, 

Jalan Kalimantan No. 37 Jember 68121  

a
rdsyarifah.fmipa@unej.ac.id 

 

Abstract. A study of Neptunium, Americium, and Protactinium addition for GFR 300MWth with Uranium 
Carbide fuel has been performed. The purpose of this study was to determine the characteristics of 
addition Neptunium, Americium, and Protactinium in a 300MWth Gas-Cooled Fast Reactor. Neutronics 
calculation was design by using Standard Reactor Analysis Code (SRAC) version 2006 with data nuclides 
from JENDL-4.0. Neutronics calculations were initiated by calculating the fuel cell calculation (PIJ 
calculation) and continued with the reactor core calculation (CITATION calculation). The reactor core 
calculation used two-reactor core configurations, namely the homogeneous core configuration and 
heterogeneous core configuration. The Neptunium, Americium, and Protactinium additions were 
performed after obtaining the optimal condition from heterogeneous core configuration. The addition of 
Neptunium and Americium which are Spent Nuclear Fuel (SNF) from LWR fuels, aims to reduce the 
amount of Neptunium and Americium in the world and also to reduce the effective multiplication factor (k-
eff) value from the reactor. The results obtained that the addition of Neptunium and Americium causes the 
k-eff value was decreased at the beginning of burn-up time, but increase at the end of burn-up time. It 
was because Neptunium and Americium absorb neutrons at the beginning of burn-up time and turns into 
fissile material at the end of burn-up time. The addition of protactinium in the reactor causes the k-eff 
value to be decreased both at the beginning of the burn-up time and at the end of the burn-up time. It 
happens because Protactinium absorbs neutrons both at the beginning of the burn-up time and at the end 
of the burn-up time. Therefore protactinium is often called a burnable poison. 

Keywords: GFR, Uranium Carbide, Neptunium, Americium, Protactinium 

Introduction 

Nuclear reactor generation IV are reactors formed from the development of generation III
+
 

reactors. Generation IV reactor consists of six reactor types, i.e. Gas-Cooled Fast Reactor 
(GFR), Lead Cooled Fast Reactor (LFR), Molten Salt Reactor (MSR), Sodium Cooled Fast 
Reactor (SFR), Supercritical Water-Cooled Reactor (SCWR), and Very High-Temperature 
Reactor (VHTR). There are several advantages of generation IV reactors, i.e. inherent safety, 
sustainability, non-proliferation resistances, and economic competitiveness [1-2]. A Gas-Cooled 
Fast Reactor (GFR) is one of the gas reactors that emerged as the Generation IV reactors. GFR 
is one candidate for generation IV nuclear power reactor which is estimated to be used in 2030. 
Neutronic calculation about GFR with Uranium Plutonium Nitride and Thorium Nitride as a fuel 
has been calculated before with varies of power, geometry, and actinide minor addition [3-12]. 
In this study, the neutronics calculation of GFR used Uranium Carbide as fuel and added 
Neptunium, Americium, and Protactinium as spent nuclear fuel. The purpose of this study was 
to determine the characteristics of the addition of Neptunium, Americium, and Protactinium on 
Uranium Carbide fuel. The addition of minor actinides on the fuel used in the modular Gas-



  

 
 
 
 
 

65 
 

Computational and Experimental Research  
in Materials and Renewable Energy (CERiMRE)         
Volume 2, Issue 2, page 64-71 
eISSN : 2747-173X 
 
 

Submitted : July 8, 2019 
Accepted  : September 8, 2019 
Online  : November 24, 2019 
doi : 10.19184/cerimre.v2i2.27368 

Cooled Fast Reactor is one of the right actions to reduce the waste of residual LWR fuel in the 
world. 
 
Theoretical Background 
A nuclear reactor is a controlled fission chain reaction. Based on the neutron energy used to 
maintain the fission reaction, there are two types of reactor, i.e. thermal reactors and fast 
reactors. GFR is a reactor that uses a fast neutron spectrum and uses a coolant in the form of 
Helium (He). The high temperature generated in the GFR, allows this reactor to be used as a 
power plant in the future. The addition of minor actinides, which are recycled fuels from LWR 
waste, has the advantage of reducing radioactive waste in the world. The Neptunium and 
Americium have a large percentage of LWR used fuel and have a half-life of hundreds to 
millions of years [13]. Neutronics calculation in the nuclear reactors can be performed using the 
following programs, i.e. RELAP, ORIGIN-8, AutoCAD 20108, LabView, ArcGIS-10, Understand 
2.6, MATLAB, SRAC, etc. [14]. Neutronics calculation in this study was performed by using the 
Standard Reactor Analysis Code (SRAC) which is a program of code analysis and design for 
nuclear reactors. SRAC has a library or data set on neutron cross-section, additional codes 
used in the analysis of nuclear reactor design, routines or movements of the neutron spectrum, 
neutron transport, one group – multi-group diffusion, etc. [15]. The database was used in the 
SRAC code, i.e. JENDL, ENDF/B, and JEF which have more than 300 nuclide data. This study 
using SRAC ver 2006 with the JENDL-4.0 database which is the latest version of nuclide data 
from JENDL. 
 
Materials and Methods 
Table 1 shows the reactor design specifications. The specifications include the specification of 
the core and the fuel that had been used. The fuel cell pin used in this study is the hexagonal 
cell. The hexagonal cell geometry in this study is divided into six regions, the first three are fuel 
regions, the next two regions are cladding, and the next two regions are coolant. The division of 
the region can be seen in Figure 1 and the sub-region from hexagonal cell geometry can be 
seen in Figure 2. 

Table 1. Reactor design specifications  
No. Parameter Specification 

1. Fuel /Cladding /Coolant UC/ SiC/ He 
2. Fuel volume fraction 60% 
3. Cladding volume fraction 10% 
4. Coolant volume fraction 40% 
5. Minor actinide Neptunium-237 and Americium 
6. Pin pitch 1.45 cm 
7. Burnable poison Protactinium-231 
8. Percentage U-235 5% - 15% 
9. Percentage Protactinium-231 0.5% - 4% 
10. Percentage Neptunium-237 0.5% - 5% 
11. Percentage Americium 0.5% - 5% 
12. Power 300 MWth 
13. Burn-up time >20 years 
14. Core geometry Cylinder pancake 
15. Fuel pin cell Hexagonal cell 
16. Height active core 100 cm 
17. Diameter active core 240 cm 
18. Reflector 50 cm 



  

 
 
 
 
 

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Computational and Experimental Research  
in Materials and Renewable Energy (CERiMRE)         
Volume 2, Issue 2, page 64-71 
eISSN : 2747-173X 
 
 

Submitted : July 8, 2019 
Accepted  : September 8, 2019 
Online  : November 24, 2019 
doi : 10.19184/cerimre.v2i2.27368 

 

RX 

FUEL CLADDING COOLANT 

      

 

Figure 1. Regional division in fuel pin calculation 

 
Figure 2. Sub-region hexagonal cell geometries  

 
The calculation method uses SRAC 2006 with the JENDL-4.0 database. The first step of 
calculation was the calculation of fuel pin cell (PIJ calculation) and the second step was reactor 
core calculation (CITATION calculation). The calculation for PIJ and CITATION used two core 
configurations, i.e. homogeneous and heterogeneous core configurations. The homogeneous 
core configuration used one type of percentage of Uranium-235. And the heterogeneous core 
configuration used three types of the percentage of Uranium-235 in the reactor core. The 
Uranium-235 is used as fissile material in the reactor core.  
 
Results and Discussion 
The neutronics calculation began with the calculation of the homogeneous core configurations. 
The homogeneous core configurations used percentage 5% - 15% of Uranium-235. The results 
obtained indicate that the fuel variations with the average U-235 value of 12% showed the most 
critical condition. Based on the results of the homogeneous core configuration i.e. percentage 
12% of U-235, it became referred to as calculate heterogeneous core configuration. Table 2 
shows fuel variations used in the calculations of heterogeneous core configurations. 

 
Table 2. Variation of U-235 in heterogeneous core calculation 
Number of case Composition (%) 

Case 1 F1=10%, F2=12%, F3=14% 
Case 2 F1=9%, F2=12%, F3=15% 
Case 3 F1=11%, F2=12%, F3=13% 
Case 4 F1=10.5%, F2=12%, F3=13.5% 
Case 5 F1=9.5%, F2=12%, F3=14.5% 

 
Figure 3 shows the result of heterogeneous core configuration. The result shows that variations 
in case 2 (F1=9%, F2=12%, F3=15%) are the variation with the most stable k-eff value from the 

Center 2 1 3 4 5 6 



  

 
 
 
 
 

67 
 

Computational and Experimental Research  
in Materials and Renewable Energy (CERiMRE)         
Volume 2, Issue 2, page 64-71 
eISSN : 2747-173X 
 
 

Submitted : July 8, 2019 
Accepted  : September 8, 2019 
Online  : November 24, 2019 
doi : 10.19184/cerimre.v2i2.27368 

beginning till the end of burn-up time. Maximum excess reactivity obtained on average 6% 
∆k/k%. 
 
Figure 4 shows the comparison of average power density in homogeneous and heterogeneous 
core configurations. The average power density in homogeneous core configuration has a 
peaking power in the center of the reactor core, while in heterogeneous core configuration has a 
flat value of average power density in the center of the reactor core. Based on the results 
obtained, the average power density in the heterogeneous core configuration indicates a safer 
condition to using in the operation of the reactor. Therefore, the calculation in the next step uses 
a core with a heterogeneous configuration. The comparison of average power density value and 
maximum power density value in homogeneous and heterogeneous core configurations are 
shown in Table 3. 
 

 
Figure 3. Effective multiplication factor (k-eff) value of heterogeneous core configuration  

 

0 5 10 15 20

1.058

1.060

1.062

1.064

1.066

1.068

E
ff

e
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u
lt
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li
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a

ti
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 F
a

c
to

r 
(k

-e
ff

)

Burn-up Time (year)

 F1=10%,F2=12%,F3=14%

 F1=9%,F2=12%,F3=15%

 F1=11%,F2=12%,F3=13%

 F1=10.5%,F2=12%,F3=13.5%

 F1=9.5%,F2=12%,F3=14.5%



  

 
 
 
 
 

68 
 

Computational and Experimental Research  
in Materials and Renewable Energy (CERiMRE)         
Volume 2, Issue 2, page 64-71 
eISSN : 2747-173X 
 
 

Submitted : July 8, 2019 
Accepted  : September 8, 2019 
Online  : November 24, 2019 
doi : 10.19184/cerimre.v2i2.27368 

 
Figure 4. Comparison of average power density value in homogeneous core configuration and 

heterogeneous core configuration 
 

Table 3. Comparison of average power density and maximum power density in homogeneous 
and heterogeneous core configuration 

Core configuration Case 
Average power 

density (watt/cc) 
Maximum power 
density (watt/cc) 

Homogeneous  U-235 12% 87.9 175 

Heterogeneous  
F1=9%, F2=12%, 

F3=15% 
78.6 97 

After calculated heterogeneous core configuration, it was necessary to add Neptunium-237 and 
Americium in the reactor. Neptunium-237 and Americium are one of minor actinide nuclide. 
Minor actinides are residual fuel obtained from light water reactor (LWR) waste. Minor actinides 
are commonly referred to as spent nuclear fuel (SNF) which has a relatively long half-life and a 
high level of toxicity. This study used the minor actinides of the nuclides Neptunium-237 and 
Americium because it was the most abundant LWR fuel waste. Figure 5 and 6 shows the 
effective multiplication factor (k-eff) value of Neptunium and Americium addition in the UC fuel.  
 

5 10 15 20 25 30

0.0

0.5

1.0

1.5

2.0

A
v
e
ra

g
e
 P

o
w

e
r 

D
e
n
s
it
y
 A

lo
n
g
 R

o
w

 (
w

a
tt
/c

c
)

Radial Mesh (cm)

 Homogen

 Heterogen



  

 
 
 
 
 

69 
 

Computational and Experimental Research  
in Materials and Renewable Energy (CERiMRE)         
Volume 2, Issue 2, page 64-71 
eISSN : 2747-173X 
 
 

Submitted : July 8, 2019 
Accepted  : September 8, 2019 
Online  : November 24, 2019 
doi : 10.19184/cerimre.v2i2.27368 

 
Figure 5. Effective multiplication factor (k-eff) value of Neptunium addition in the UC fuel 

 
Figure 5 and 6 shows that Neptunium-237 and Americium decreased the k-eff values at the 
beginning of the burn-up time and increased the k-eff value at the end of burn-up time. The 
greater variation of the Neptunium-237 and Americium given, the greater decreased of k-eff 
value at the beginning of burn-up time. It happened because Neptunium-237 and Americium 
has cross-section absorption greater than fissile material. Therefore, the k-eff value decreased 
at the beginning of burn-up time. After absorbing the neutrons, the Neptunium-237 and 
Americium would become fissile material (Pu-239), so at the end of burn-up time the k-eff value 
increase. 

 
Figure 6. Effective multiplication factor (k-eff) value of Americium addition in the UC fuel 

0 5 10 15 20

1.025

1.030

1.035

1.040

1.045

1.050

1.055

1.060

1.065

1.070

E
ff

e
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u
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li
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a

ti
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 F
a

c
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r 
(k

-e
ff

)

Burn-up Time (year)

 Neptunium 0.5%

 Neptunium 1%

 Neptunium 1.5%

 Neptunium 2%

 Neptunium 2.5%

 Neptunium 3%

 Neptunium 3.5%

 Neptunium 4%

 Neptunium 4.5%

 Neptunium 5%

0 5 10 15 20

1.01

1.02

1.03

1.04

1.05

1.06

E
ff
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 F

a
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(k

-e
ff
)

Burn-up Time (year)

 Americium 0.5%

 Americium 1%

 Americium 1.5%

 Americium 2%

 Americium 2.5%

 Americium 3%

 Americium 3.5%

 Americium 4%

 Americium 4.5%

 Americium 5%



  

 
 
 
 
 

70 
 

Computational and Experimental Research  
in Materials and Renewable Energy (CERiMRE)         
Volume 2, Issue 2, page 64-71 
eISSN : 2747-173X 
 
 

Submitted : July 8, 2019 
Accepted  : September 8, 2019 
Online  : November 24, 2019 
doi : 10.19184/cerimre.v2i2.27368 

 
After that, it was necessary to add burnable poison in the core reactor to decrease k-eff value to 
become critical. The burnable poison used in this study is Protactinium (Pa-231). Protactinium 
has high cross-section neutron absorption. Pa-231 absorbed neutrons at the beginning of the 
burn-up time, causing the k-eff value was decreased. Figure 7 shows the result of 
heterogeneous core configuration with the addition of Pa-231. The variation of Pa-231 used 
between 0.5% up to 4% with an interval 0.5. The addition of Pa-231 3.5% had the optimum k-eff 
value (k-eff ~ 1) with a maximum excess reactivity obtained 1.04% ∆k/k%.  

 

 
Figure 7. Effective multiplication factor (k-eff) value of addition Protactinium-231 

 

Conclusions 

Neptunium-237 and Americium decreased the k-eff value at the beginning of burn-up time and 
increased the k-eff value at the end of burn-up time. At the beginning of burn-up time, the 
Neptunium-237 and Americium absorbed the neutron and become fissile material (Pu-239 and 
Pu-241). Therefore it increased the burn-up time at the end of burn-up. Protactinium is called 
burnable poison. It decreased the k-eff value both at the beginning and the end of burn-up time.  

References 

[1]  GIF (The Generation IV International Forum) and the OECD Nuclear Energy Agency, 2014, 
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[2]  GIF (The Generation IV International Forum) and the U.S DOE Nuclear Energy Advisory 
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[3]  R. D. Syarifah, Y. Yulianto, Z. Su’ud, K. Basar, and D. Irwanto, 2016, Neutronic Analysis of 
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0 5 10 15 20

1.00

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E
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 Pa 1%

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 Pa 3%

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71 
 

Computational and Experimental Research  
in Materials and Renewable Energy (CERiMRE)         
Volume 2, Issue 2, page 64-71 
eISSN : 2747-173X 
 
 

Submitted : July 8, 2019 
Accepted  : September 8, 2019 
Online  : November 24, 2019 
doi : 10.19184/cerimre.v2i2.27368 

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