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

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Nuclear Science and Technology, Vol.9, No. 2 (2019), pp. 25-30

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

^aInstitute for Nuclear Science and Technology, VINATOM, 179 Hoang Quoc Viet, Hanoi, Vietnam

^bFaculty of Nuclear Engineering, Electric Power University, 235 Hoang Quoc Viet, Hanoi, Vietnam

^cInstitute of Fundamental and Applied Sciences, Duy Tan University, Ho Chi Minh city, Vietnam

^*E-mail: [email protected]

(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.

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).

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

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 …

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].

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.

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

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.

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.

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

HOANG VAN KHANH et al.

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

Fig. 1. Locations of fuel rods, guide thimbles (GT)

and instrumentation thimble (IT) in a fuel assembly.

evolution

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.

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.

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 …

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

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

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

HOANG VAN KHANH et al.

temperature results in the change of moderator

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

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.

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.

The shutdown margins of the control

rod assemblies were analyzed to ensure that

the core has sufficient control rod worth.

Table II. Summary of analysis results of the different core configurations.

Fig. 4. Locations of the CEAs in the active

Table III. Temperature coefficients and CEAs

core zone.

performance of the final core design.

CONCEPTUAL DESIGN OF A SMALL-PRESSURIZED WATER REACTOR USING THE AP1000 …

Status, Technical Feasibility and Economics of

Small Nuclear Reactors”, June 2011.

IV. CONCLUSIONS

Conceptual design calculation of a 300

[3]. Jun Liao, et al.. “Preliminary LOCA Analysis

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.

of the Westinghouse Small Modular Reactor

Using the WCOBRA/TRAC-TF2 Thermal-

Hydraulics Code”, Proceedings of ICAPP'12

Chicago, USA, June 24-28, 2012.

[4]. Keisuke Okumura, Teruhiko Kugo, Kunio

Kaneko

and

Keichiro

Tsuchihashi.

“SRAC2006: A Comprehensive Neutronics

Calculation Code System”, JAEA-Data/Code

2007-004, 2007.

[5]. Keiichi Shibata, et al.. “JENDL-4.0: A New

Library for Nuclear Science and Engineering,

Journal of Nuclear Science and Technology”,

vol. 48, No. 1, p. 1-30, 2011.

[6]. Keisuke Okumura. “COREBN: A Core Burn-

up Calculation Module for SRAC2006”,

JAEA-Data/Code 2007-003, 2007.

ACKNOWLEDGEMENTS

This research is funded by Ministry of

Science and Technology (MOST) of Vietnam

under grant number ĐTCB. 02/19/VKHKTHN.

[7]. James J. Duderstadt Louis J. Hamilton.

"Nuclear Reactor Analysis", John Wiley &

Sons, Inc., 1976.

REFERENCES

[8]. Tong L.S. “Prediction of Departure from

Nucleate Boiling for an Axially Non-Uniform

Heat Flux Distribution”. Journal of Nuclear

Energy, vol. 21, p. 241-248, 1967.

[1]. International Atomic Energy. “Advances in

Small

Modular

Reactor

Technology

Developments,

Supplement to: IAEA

Advanced Reactors Information System

(ARIS)”, 2018.

[9]. U.S. Nuclear Regulatory Commission (NRC),

“Westinghouse AP1000 Design Control

Documentation (DCD)”, Rev. 19, Chapter 4,

Westinghouse Electric Company, 2011.

[2]. Organisation for Economic Co-operation and

Development, Nuclear Energy Agency. “Current

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