Conceptual design of a small-pressurized water reactor using the AP1000 fuel assembly design

 Nuclear Science and Technology, Vol.9, No. 2 (2019), pp. 25-30 
©2019 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute 
Conceptual design of a small-pressurized water reactor using 
the AP1000 fuel assembly design 
Van Khanh Hoang
a, *
, Viet Phu Tran
a
, Van Thin Dinh
b
, Hoai Nam Tran
c
a
Institute for Nuclear Science and Technology, VINATOM, 179 Hoang Quoc Viet, Hanoi, Vietnam 
b
Faculty of Nuclear Engineering, Electric Power University, 235 Hoang Quoc Viet, Hanoi, Vietnam 
c
Institute of Fundamental and Applied Sciences, Duy Tan University, Ho Chi Minh city, Vietnam 
*
E-mail: hvkhanh21@gmail.com 
(Received 01 November 2019, accepted 13 November 2019) 
Abstract: This paper presents the conceptual design of a 300 MWt small modular reactor (SMR) 
using fuel assemblies of the AP1000 reactor. Numerical calculations have been performed to evaluate 
a proper active core size and core loading pattern using the SRAC code system and the JENDL-4.0 
data library. The analysis showed that Doppler, moderator temperature, void, and power reactivity 
coefficients are all negative over the core lifetime. Semi-analytical thermal hydraulics analysis reveals 
acceptable radial and axial fuel element temperature profiles with significant safety margin of fuel and 
clad surface temperature. The minimum departure from nucleate boiling ratio (MDNBR) is also 
calculated. The results indicate that a cycle length of 2.22 years is achievable while satisfying the 
operation and safety-related design criteria with sufficient margins. 
Keywords: small modular reactor, AP1000 reactor, neutronic analysis, thermal hydraulics analysis. 
I. INTRODUCTION 
In recent years, the small modular 
reactors (SMRs), new generation reactors 
designed with an electrical output up to 300 
MWe, have been received increasing attention 
within the nuclear energy community due to a 
number of advantages. Because of the small 
size, the SMRs require less capital investment 
and construction time compared to traditional 
commercialized reactors, so that financial risks 
could be reduced. The SMR designs can adopt 
most of the advanced safety features of the 
current technologies. One of the advantages is 
that the components of the reactor system can 
be fabricated at factories and transported 
modularly to the plant site for installation. It is 
also more flexible to choose locations for the 
SMRs than the traditional reactors, and 
therefore, it would be a suitable solution for a 
wide range of users and applications, for 
instance remote areas with smaller electricity 
demand. The SMRs are being developed for a 
number of principal reactors including 
advanced light water reactors, heavy water 
reactors, and the generation IV reactors such as 
high temperature gas cooled reactors (HTGRs), 
liquid-metal, sodium and gas-cooled fast 
reactors (LMFR, SFR, GFR), and molten salt 
reactors (MSRs). 
About 50 SMR designs are being under 
developed world-wide for both electrical 
generation and non-electrical application 
such as desalination of seawater, district 
heating, hydrogen production and other 
process heat application. Among those, three 
industrial demonstration SMRs are in 
construction including CAREM (integral 
PWR with the output of 150 – 300 MWe) in 
Argentina, HTR-PM (HTGR) in China and 
KLT-40S (compact PWR for a floating 
nuclear power plant) in Russia. The integral 
pressurized water reactor (IPWR) technology 
is one of the major near term SMR designs, 
of which primary components are contained 
CONCEPTUAL DESIGN OF A SMALL-PRESSURIZED WATER REACTOR USING THE AP1000  
26 
in the reactor vessel. CAREM is an example 
of the IPWR technology. CAREM-25, a 
prototype reactor with the power of 27 MWe, 
is under construction. Other designs of 
IPWRs are SMART (Korea), NuScale (US), 
mPower (US), Westinghouse SMR (US). The 
IPWR of the Westinghouse Electric 
Company is based on a partial-height 17x17 
fuel assembly used in the AP1000
reactor. 
This reactor utilizes passive safety systems 
and proven components from the AP1000 
plant design. All primary components 
including the steam generator and the 
pressurizer are located inside the reactor 
vessel [3]. 
In the present work, a conceptual 
design of a small PWR core with thermal 
output of 300 MWt based on the AP1000 
fuel assembly has been presented. Neutronics 
analysis has been performed to determine the 
core height, the number of fuel assemblies 
and core configuration. The neutronic 
analysis has been conducted using the SRAC 
code system and the JENDL-4.0 library. 
Thermal hydraulics analysis has been 
conducted to investigate the safety 
parameters of the core. Reactivity 
coefficients regarding the change in 
temperature of fuel, coolant, etc. were also 
investigated to ensure the safety feature of 
the newly designed SMR core. 
II. CODE AND DESIGN PROCEDURE 
The SRAC2006 code system [4] and the 
JENDL-4.0 nuclear data library [5] were used 
for performing the design process of the SMR 
core. For the core burnup calculation process is 
divided into two steps. Firstly, cell burnup 
calculation using the PIJ module was 
conducted to produce few-group burnup 
dependent homogenized macroscopic cross-
sections. Secondly, the COREBN [6] was used 
for three-dimensional core burnup calculation 
based on the macroscopic cross-section 
interpolation. In order to confirm the core 
design criteria, thermal hydraulics analysis for 
the hottest channel in the core was performed 
to establish adequate heat removal capability of 
the design. Analytical models and equations 
are utilized for heat conduction within the fuel 
element and convection to the coolant [7]. 
Calculations of the MDNBR have also been 
performed for the hottest channel using 
empirical correlation and W-3 correlation [8]. 
The design process consists of two steps as 
follows: 
a) Estimation of core size: The objective of 
this step is to determine the active fuel length 
and the number of fuel assemblies. Neutron 
transport calculations have been performed 
for an infinite core by imposing reflective 
boundary condition in the assembly level 
model. From the mass of uranium per 
assembly, the required number of fuel 
assemblies for the core can be determined. In 
this determination of core size, a four-year 
core lifetime, the average discharge burnup of 
40 GWd/t with a single batch refueling 
scheme and a capacity factor of 1.0 were 
assumed. 
b) Determination of core loading pattern: 
Once the required number assemblies are 
achieved, a symmetrical arrangement of the 
fuel assemblies in the active core zone is 
proposed for the initial core configuration. The 
core loading pattern is developed for the core 
design to meet the design criteria. Fuel 
assemblies with higher U-235 enrichment are 
selected with a high priority, and located in the 
peripheral core locations to achieve a uniform 
power distribution and maximum fuel 
utilization. Finally, the core that meets the 
specified performance requirements is selected 
for further analysis. 
HOANG VAN KHANH et al. 
27 
Fig. 1. Locations of fuel rods, guide thimbles (GT) 
and instrumentation thimble (IT) in a fuel assembly. 
III. CORE DESIGN AND 
PERFORMANCE ANALYSIS 
A. Core parameters 
The Robust Fuel Assembly (RFA) 
design is used in the AP1000 reactor. The 
typical fuel assembly, with 17x17 array, 
contain 264 fuel rods, 24 guide thimbles, and 
1 instrumentation thimble as shown in Fig. 1 
[3]. The SMR core is designed with a 
thermal output of 300 MWt. The typical 
uranium oxide fuel rods (UO2) with 2.35, 
3.40, and 4.95 wt.% U-235 enrichments are 
used in the current SMR [9]. The main 
design parameters and targets for the SMR 
are specified in Table 1. In the AP1000 
reactor design, in order to control excess 
reactivity and flatten power distribution, 
burnable absorber rods and integral fuel 
burnable absorbers are also used. In this 
study, only fuel assemblies without burnable 
absorber are considered to design a new core. 
For the control element assemblies (CEAs), 
the standard 24 finger Ag-In-Cd (AIC) rod 
control cluster assembly is exploited. 
B. Estimation of core size 
In order to determine the active fuel 
length, a three-dimensional infinite core 
model was developed based on an assembly 
level model with reflective boundary 
condition using the SRAC code system. The 
model consists of an active fuel zone and two 
reflector zones at the top and the bottom. The 
AP1000 fuel assembly with the highest U-
235 enrichment, i.e., 4.95 wt.%, was selected 
for the active fuel zone. Reflective boundary 
condition was assumed for peripheral core, 
and extrapolated boundary condition was set 
at the top and the bottom. Fig. 2 shows the 
evolution of the infinite neutron 
multiplication factor (k-inf) at the begin of 
cycle (BOC) and the number of assemblies 
loaded into the core as functions of the active 
fuel length. It can be seen that when the 
active fuel length is less than 190 cm, the k-
inf increases rapidly with the increase of the 
active fuel length. The k-inf value is nearly 
unchanged with the active fuel length greater 
than 190 cm. Therefore, the active fuel 
length of 190 cm was selected, and the 
number of fuel assemblies in the core was 
determined as 45 assemblies. 
Fig. 2. Variation of infinite multiplication factor 
and required number of assemblies with increasing 
active fuel length. 
CONCEPTUAL DESIGN OF A SMALL-PRESSURIZED WATER REACTOR USING THE AP1000  
28 
Table I. Main design parameters and conditions (k-eff: effective neutron multiplication factor). 
Fig. 3. Core configurations of the SMR (yellow, blue, and green blocks represent fuel assembly assemblies 
with 4.95, 3.40 and 2.35 wt.% U-235 enrichment, respectively. White block represents water filled position). 
a) Core model 1 b) Core model 2 c) Core model 3 
D. Analysis of temperature coefficients and 
control element assemblies 
In order to investigate the feedbacks due 
to changes of fuel and coolant temperatures, 
the temperature reactivity coefficients have 
been calculated. A change in temperature of 
fuel material causes a change of neutron cross-
section, which is so-called the Doppler 
broadening effect, resulting in a change in 
reactivity. Meanwhile, a change in coolant 
HOANG VAN KHANH et al. 
29 
temperature results in the change of moderator 
density, which leads to a change in reactivity 
of the core. For safe operation, negative values 
of the temperature reactivity coefficients are 
desirable. In the numerical analysis, the 
temperature of fuel and/or coolant was 
increased by 50K. For the isothermal 
coefficient, both temperature of fuel and 
coolant were increased by 50K. 
The shutdown margins of the control 
rod assemblies were analyzed to ensure that 
the core has sufficient control rod worth. 
Fig. 4 shows the map of the control element 
assemblies. The CEAs are composed of 
shutdown banks (S) and regulating banks 
(R) located in the outer core and the inner 
core, respectively. The reactivity 
coefficients and CEAs performance of the 
selected core model (Core 3) at cold zero 
power (CZP) and hot full power (HFP) 
states are presented in Table 3. It can see 
that all coefficients are negative during the 
core lifetime. The CEAs of the core have 
sufficient shutdown margin. 
Table II. Summary of analysis results of the different core configurations. 
Fig. 4. Locations of the CEAs in the active 
core zone. 
Table III. Temperature coefficients and CEAs 
performance of the final core design. 
CONCEPTUAL DESIGN OF A SMALL-PRESSURIZED WATER REACTOR USING THE AP1000  
30 
IV. CONCLUSIONS 
Conceptual design calculation of a 300 
MWt SMR based on the fuel assemblies of the 
AP1000 reactor has been carried out. The core 
consists of 45 fuel assemblies with an active 
core height of 190 cm could operate up to 2.22 
years without refueling. The Doppler and 
moderator temperature coefficients are 
negative throughout the core lifetime. 
Maximum values of fuel, cladding and coolant 
temperatures are within the design limits. The 
DNBR remains greater than 3.71 for the active 
core region. The analysis results represent that 
the final core design satisfies all the design 
criteria with significant safety margins. 
ACKNOWLEDGEMENTS 
This research is funded by Ministry of 
Science and Technology (MOST) of Vietnam 
under grant number ĐTCB. 02/19/VKHKTHN. 
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