A review of existing SuperCritical Water reactor concepts, safety analysis codes and safety characteristics

Progress in Nuclear Energy 153 (2022) 104409

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Progress in Nuclear Energy
journal homepage: www.elsevier.com/locate/pnucene

Review

A review of existing SuperCritical Water reactor concepts, safety analysis
codes and safety characteristics
Pan Wu a, *, Yanhao Ren a, Min Feng a, Jianqiang Shan a, b, Yanping Huang c, Wen Yang c
a

School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28 Xianning West Road, Xi’an, Shaanxi, China
The State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, Xi’an, 710049, China
c
Nuclear Power Institute of China, Cheng Du, 610000, China
b

A R T I C L E I N F O

A B S T R A C T

Keywords:
Supercritical water reactor
Safety performance
System code development

SuperCritical Water Reactor(SCWR) applies water beyond the thermodynamic critical point as the coolant, which
aims to achieve high efficiency around 45% compared to 33% for existing commercial light water reactors. In

order to raise the reactor operating temperature and reactor criticality, the existing SCWR core designs are quite
different from those of boiling water reactors or pressurized water reactors, which further effect their safety
performance and safety system design. A comprehensive review on existing developed SCWR reactor concepts of
different countries, including pressure-vessel type and pressure-tube type SCWRs, as well as thermal, fast and
mixed spectrum SCWRs, is carried out to deeply explain the core design features of SCWR. The development
methods of safety analysis tool for SCWR are also summarized to shed a light on the key scientific difficulties and
how these problems are solved up to now. All the special techniques applied to enable trans-critical simulations
are still unphysical and lack of validation. Moreover, the safety characteristics of existing SCWR concepts are
discussed. Based on these review work and discussions, the research status of SCWR concepts, safety analysis tool
development and safety characteristics are clearly presented. Safety analysis tool validations and more
comprehensive accident evaluations should be further carried out to better illustrate the safety performance of
these SCWR concepts.

1. Introduction
SuperCritical water-cooled reactor,abbreviated as SCWR, owns
unique core design using water above the critical point as coolant, which
is quite different from other Generation-IV reactor concepts (Schulen­
berg et al., 2011). A large amount of research interests arose from nu­
clear industry and academic community for its huge advantages in high
thermal efficiency, simple system configuration as well as good tech­
nical inheritance from existing commercial power plants. SCWR reactor
system operates under pressure of 25 MPa(374 ◦ C, 22.1 MPa is the water
critical point) and applies water as the coolant and moderator. Coolant
heated by the reactor fuels is directly led to turbine to produce power,
which makes SCWR own a similar system configuration as Boiling Water
Reactor(BWR). At the same time, successful operation of supercritical
and ultra-supercritical thermal power plant (Buongiorno et al., 2003)
facilitates application of supercritical water as heat transfer medium.
Though SCWR has so many advantages, it also has some shortcom­
ings to overcome. Because of demand of large coolant enthalpy rise

through the core, the ratio of mass flowrate over thermal power of SCWR
is much lower than those of pressurized water reactor (PWR) as well as
boiling water reactor (BWR). The ratio of mass flowrate over thermal
power of SCWR is around 1/12 of that of PWR and around 1/10 of that
of BWR, which indicates that SCWR has higher fuel temperature rise
when loss of flow accident(LOFA) happens and has a quicker
depressurizing process when loss of coolant accident(LOCA) happens.
Meanwhile, the specific heat of overheated SCW in upper core region is
quite low. When accidents happen to SCWR, the coolant and the fuel
material will encounter a huge temperature rise, which is negative for
the reactor safety. Many researchers have worked on how to improve
SCWR safety performance under normal operation and accidents
through unique core designs and innovative safety system designs. In
2019, IAEA-TECDOC-1869 (IAEA, 2019) summarized the research sta­
tus of supercritical water reactor, which focused on the contents related
to reactor design, including the existing reactor type, thermal hydraulics
and material and chemistry while little information on the system code
development and SCWR safety performance is included.
In this paper, reactor core and safety system designs developed by

* Corresponding author.
E-mail address: (P. Wu).
/>Received 6 May 2022; Received in revised form 23 August 2022; Accepted 8 September 2022
Available online 22 September 2022
0149-1970/© 2022 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY license ( />

P. Wu et al.

Progress in Nuclear Energy 153 (2022) 104409

Abbreviations

LOFA
Loss Of Flow Accident
LPCI
Low Pressure Core Injection System
MCST
Maximum Cladding Surface Temperature
MOX
Mixed Oxide Fuel
MSIV
Main Steam Isolation Valve
NPIC
Nuclear Power Institute of China
PCCS
Passive Containment Cooling System
PT
Pressure-Tube
PV
Pressure-Vessel
PWR
Pressurized Water Reactor
R&D
Research And Design
RMT
Reactor Make-Up Tank
RPV
Reactor Pressure Vessel
SCP

Supercritical Parameters of Water Coolant
SCWR
Supercritical Water-Cooled Reactor
SCWR-M Mixed Spectrum Supercritical Water-Cooled Reactor
SRV
Safety Release Valve
SUPER FR Supercritical Fast Reactor
SUPER LWR Supercritical Light Water-Cooled Reactor
VVER
Water-Water Energetic Reactor

ACR
Advanced Candu Reactor
ADS
Automatic Depressurization System
AECL
Atomic Energy of Canada Limited
AFS
Auxiliary Feedwater System
ATWS
Anticipated Transient Without Scram
BWR
Boiling Water Reactor
CANDU Canadian Deuterium Uranium Reactor
CANFLEX CANdu Flexible
CGNPC China Guangdong Nuclear Power Corporation
CR
Control Rood
CSR1000 Chinese Supercritical Water-Cooled Reactor
ESBWR Economic Simplified Boiling Water Reactor

GDCS
Gravity Driven Core Cooling System
HEC
High Efficiency Fuel Channel
HPLWR High Performance Light Water Reactor
ICS
Isolation Condenser System
LOCA
Loss Of Coolant Accident
LOECC Loss Of Emergency Core Cooling

different countries and institutions are extensively reviewed. At the
same time, the safety analysis tool development methodologies and
safety performances of different types of SCWR concepts are also deeply
reviewed to provide some insights for further SCWR development.
2. Existing rector concept development
From the aspect of core structure, SCWR concepts can be divided into
pressure tube type and pressure vessel type. Canadian SCWR(Yetisir
et al., 2016) applies pressure tube to contain the high-pressure coolant
and fuel in the reactor core while SCWR concepts developed by other
countries apply pressure vessel to contain the coolant and fuels. From
the aspect of neutron energy level, SCWR concepts could be divided into
fast spectrum type(Oka et al., 2010a), thermal spectrum type(Wu et al.,
2014)(Oka et al., 2010a; Ishiwatari and Oka et al., 2010b; Novog et al.,
2012) and mixed spectrum type SCWR(Xu et al., 2011; Liu et al., 2013).
In this section, SCWR concepts from different countries will be exten­
sively reviewed.
2.1. Chinese SCWR concepts
There are two main types of theoretically mature Chinese SCWR
concepts, which are CSR1000 which is named as Chinese supercritical

water-cooled reactor and the SCWR-M which is named as mixed spec­
trum supercritical water-cooled reactor. Thermal spectrum is used for
CSR1000, whereas mixed neutron spectrum is used for SCWR-M (Zhu
et al., 2012; Liu et al., 2013; Wu and Geffraye et al., 2011).
2.1.1. CSR1000
In 2014, Nuclear Power Institute of China (NPIC) proposed CSR1000
which is a pressure-vessel SCWR. (Wu et al., 2014). CSR1000 applies
thermal spectrum while supercritical water is assumed to cool the core
and moderate the neutrons. Coolant and moderator will be strongly
mixed in the lower plenum. The thermal power and electrical power of
CSR1000 are 2300 MW and 1000 MW respectively. The temperature at
the entrance of the core is 280 ◦ C whereas the temperature at the exit of
the core is 500 ◦ C, as a result, the thermal efficiency of CSR1000 can
reach 45%. In order to have a more even distribution of axial power,
CSR1000 applies a two-pass core design, which is shown in Fig. 1(a). The
difference in coolant temperature in different flow channels are shown
in Fig. 1(b).
There are 177 fuel assemblies in CSR1000 reactor core. The first and

Fig. 1. The flow scheme in CSR1000 and corresponding coolant temperature
variation(Wu et al., 2014).

second-pass cores consist 57 assemblies and 120 assemblies, respec­
tively. The distribution of two pass fuel assemblies is shown in Fig. 2.
Fig. 3 shows the cross section of fuel assembly, consisting of 4 subassemblies and 4 water rods. Water flowing through these two paths
acts as moderator and coolant simultaneously. Additionally, for better
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distribution of fuel assemblies in SCWR-M. The core of SCWR-M is made
up of 284 fuel assemblies while 164 fuel assemblies locates in the outer
zone reacting with thermal spectrum neutrons and 120 fuel assemblies
locates in the inner zone reacting with fast spectrum neutrons (Liu et al.,
2013). For the zone with thermal spectrum, as shown in Fig. 5(a), there
are three layers of fuel assemblies with varying fuel enrichment at
different heights. The fast zone is designed to be short in order to in­
crease neutron leakage so that SCWR-M can obtain a negative void
reactivity feedback, which is shown in Fig. 5(b).
The flow path is shown in Fig. 6. Low-temperature water enters the
pressure vessel and flows upward into the upper chamber, after which it
flows into both the coolant and the moderator channels, with the 25
percent of coolant flowing into the moderator channel in the zone of
thermal. Then coolant flows out of the zone of thermal spectrum into the
lower chamber, and from the lower chamber it flows into the fast zone.
The temperatures at the entrance and the exit of reactor core are 280 ◦ C
and 510 ◦ C respectively.
The average line power of the core is 18 kW/m. The active heights of
the thermal zone and the fast zone are 4.5m and 2.0m respectively.

Fig. 2. Fuel assembly distribution of CSR1000(Wu et al., 2014).

2.2. Japanese SCWR concepts
Japan carried out researches and design works for fast and moder­
ated SCWR concepts simultaneously. Water is used as a moderator and
works in Super LWR in form of water rods. Water rod is a space in the
core filled with light water. Its presence ensures negative void reactivity

and provide additional emergency core cooling injection during acci­
dents. Meanwhile a fast neutron spectrum SCWR named Super FR, is also
under development. In order to ensure a negative void coef­
ficientZirconium hydride layers are used for Super FR. Both of these
different concepts will be introduced to compare respective behaviors
(Oka et al., 2010a).
2.2.1. Super LWR
The University of Tokyo proposed Super LWR for the first time which
includes a once-through coolant cycle without recirculation line, as
shown in Fig. 7. Some of the plant parameters are also included in Fig. 7.
After design updates, the feed water temperature of Super LWR is

Fig. 3. Cross section of CSR1000 fuel assembly(Wu et al., 2014).

control of reactivity, cross-shaped control rods are used.
2.1.2. SCWR-M
Unlike CSR1000, in order to avoid serious problems that may be
encountered in mechanical design and safety analysis (Zhu et al., 2012),
proposed a mixed core design scheme in which the fuel assemblies are
divided into multiple layers, and this design scheme can simultaneously
achieve high core exit temperatures (Zhu et al., 2012; Liu et al., 2013).
The Chinese mixed spectrum reactor(SCWR-M) core is made up of
thermal spectrum core and fast spectrum core. Fig. 4 shows the

Fig. 4. Fuel assemblies distribution in SCWR-M(Xu et al., 2011).

Fig. 5. Struture of fuel assembly of SCWR-M.
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updated to be 290 ◦ C and the reactor exit temperature is 510 ◦ C. The
thermal and electric powers are 4039 MW and 1725 MW respectively. As
shown in Fig. 8, Super LWR applies a two-pass core, in which the fuel
assemblies located at the outer region of reactor core are cooled firstly
by the downward flowing coolant. The coolant flows upward through
the fuel assemblies located at the inner side of the core after mixing in
the lower chamber. The arrangement of fuel assembly in the core is
shown in Fig. 9, with 372 fuel assemblies being divided into three-batch
fueling().
The fuel design of Super LWR uses UO2 for fuel pellets, which is the
same as that of LWRs. The material of the fuel cladding is stainless steel
and nickel-based alloy. The design of fuel assembly, which is the same as
that of LWR, is shown in Fig. 10.
2.2.2. Super-fast reactor
Super FR’s flow circulation is the same as that of the Super LWR
which has already been shown in Fig. 7. Since fast reactors do not
require moderators, the power density of fast reactors is much higher
than that of thermal reactors, which further result in better economy.
The operating conditions of Super FR and Super LWR are totally the
same. The pressure of reactor when it’s normally operated is 25 MPa,
while the temperatures of reactor core’s inlet and outlet are 280 ◦ C and
508 ◦ C respectively.
The principle of mixed-oxide fuel(MOX) design of the Super FR needs
to accommate high Pu content, except which the principle is the same as
that of the Super LWR. A zirconium hydride layer is placed in the blanket
fuel assemblies to make the reactor has a negative coolant void reac­

tivity, which is shown in Fig. 11. The arrangement of the reactor core is
shown in Fig. 12(Oka et al., 2010a).
Super FR also applies two-pass flow, as shown in Fig. 13 (Oka et al.,
2010a). The coolant flows downward through the blanket assemblies
and part of the seed assemblies. The coolant gathered in the lower
plenum flows through the rest part of seed assemblies. Two-pass flow
core is helpful to increase the reactor operating temperature while
satisfy the cladding temperature design limits.

Fig. 6. Flow paths in the core(Xu et al., 2011).

2.3. Canadian SCWR
Canadian SCWR is only pressure-tube type SCWR concept. It was
updated from the mature CANDU in several aspects. The main features
of CANDU have been preserved, such as modular fuel channels

Fig. 7. Flow circulation of coolant cycle for Super LWR(Oka et al., 2010a).

Fig. 8. Schematic diagram of coolant flow inside core(Oka et al., 2010a).

Fig. 9. The fuel assembly distribution of the core(Oka et al., 2010a).
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Fig. 10. Design of fuel assembly for Super LWR.

Fig. 13. Two-pass flow core of Super FR.

Fig. 11. Fuel assemblies of Super FR(Oka et al., 2010a).

Fig. 14. Schematic map of conceptual Canadian SCWR core design.

flows into the center channel downward first and then flows upward to
carry away heat.(Khartabil, 2008).
The design of the fuel assembly has experienced a series of upgrades,
as shown in Fig. 15. Each channel includes a single fuel assembly while a
stainless-steel fuel cladding remains in direct contact with the fuel
pellets.
Fig. 12. Fuel assembly distribution of the core for Super FR(Oka et al., 2010a).

2.4. European SCWR-HPLWR

configuration and selecting heavy water as the moderator. Canadian
SCWR’s operating pressure is 25 MPa, while the temperatures of core’s
inlet and outlet are 350 ◦ C. and 625 ◦ C respectively. The thermal power
and the electric power are 2540 MW and 1200 MW with a thermal ef­
ficiency of 48%(Novog et al., 2012).
‘No-core-melt’ concept is proposed for Canadian SCWR. The radia­
tion heat exchange between fuel rods inside the pressure tube and the
low temperature heavy water moderator outside the pressure tube can
carry away the decay heat of the fuel under extreme operating condi­
tions, which greatly reduces the probability of core meltdown in the
reactor.
There are 336 fuel channels in the core of Canadian SCWR. The
average channel power is 7.5 MW(t). The schematic map of the design of
Canadian SCWR is shown in Fig. 14. Coolant entering each pressure tube

SCWR developed by the European Union is a pressure vessel type
reactor, which is called High Performance Light Water Reactor
(HPLWR). The operating pressure of HPLWR is 25 MPa and the tem­
perature of core exit is 500 ◦ C. The thermal power and the electric
powers are 2300 MW and 1000 MW respectively. The structure of
pressure vessel of HPLWR is shown in Fig. 16(Allison et al., 2016).
The unique feature of HPLWR core design is that it applies a threepath flow scheme, which is shown in Fig. 17(a). Since there are three
processes of coolant flow, the heating process of coolant is also divided
into 3 stages. Meanwhile, the core of HPLWR is separated into three
parts, which is shown in Fig. 17(b).
Each fuel assembly has an assembly box which has 9 sub-assemblies
with total 40 fuel rods in it and an additional moderator box in the
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Fig. 15. The fuel assembly design of the Canadian SCWR.

Fig. 16. The structure of pressure vessel of HPLWR(Allison et al., 2016).

center. Cross section of HPLWR fuel assembly is shown in Fig. 18
(Starflinger et al., 2010).

to VVER, PWR and BWR reactors. The fuel material for VVER-SCP could
be uranium dioxide, MOX fuel, and other kinds of fuel. The temperatures
of the core’s inlet and outlet are 280 ◦ C and 540 ◦ C respectively. The

efficiency of VVER-SCP increases from 40% to 44–45%. Unusually,
VVER-SCP applies single-pass coolant flow scheme. The coolant flow
scheme of single-pass core is shown in Fig. 19. Compared with the

2.5. Russian SCWR concept-VVER-SCP
The Russian SCWR concept(VVER-SCP) is developed with reference
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Fig. 19. Coolant flow scheme of single-pass core(Kalyakin and Kirillov
et al., 2014).

2.6. Korean SCWR concept -SCWR-R
A 1400MWe SCWR concept named SCWR-R, is developed by Korea
Atomic Energy Research Institute. Different from other thermal type
SCWR concepts, SCWR-R applies cruciform-type solid moderator(U/
ZrH2), instead of water or heavy water, to provide additional modera­
tion and simplify the core structure, which is helpful for decreasing the
power peak. The design of fuel assembly is shown in Fig. 20. There are
300 fuel rods in each fuel assembly, whereas 25 cruciform-type solid
moderator pins, and 16 single solid moderator pins. The core of SCWR-R
includes 193 fuel assemblies using a typical four-batch fuel-loading
pattern. Meanwhile, there are jet pumps installed in the downcomer to
enable coolant recirculation, as shown in Fig. 21, which is helpful to
decrease the rise of enthalpy in the core. The core flowrate of SCWR-R is
6441 kg/s and the flowrate and temperature of the feedwater is around

2518 kg/s and 280 ◦ C. Internal circulation results in the increase of core
inlet temperature from 280 to 350 ◦ C(Bae et al., 2008), which is beyond
the pseudo critical temperature. High core inlet temperature helps avoid

Fig. 17. Design of thermal care.

Fig. 18. Design of HPLWR fuel assembly (Allison et al., 2016).

multiple-pass flow scheme applied by other SCWR core designs, singlepass core has advantages in system simplification while it may also lead
to non-uniform power distribution along the axial direction with large
hot channel factor(Kalyakin and Kirillov et al., 2014).
Fig. 20. Fuel assembly design.
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Progress in Nuclear Energy 153 (2022) 104409

concepts applied water as moderator, which also act as coolant as well.
Some SCWR concepts don’t need moderator, such as VVER-SCP. Cana­
dian SCWR applies heavy water as moderators while Korean SCWR-R
applies cruciform-type solid moderator pin.
Beyond these key differences, the above design concepts have many
similar challenges which provides the possibility for existing SCWR re­
searchers to collaborate with each other, for example, materials selec­
tion for fuel cladding and reactor internal components, water chemistry
study applicable for all SCWR concepts. As well as thermal-hydraulics
and safety analysis. For the aspect of thermal-hydraulics and safety,
there are huge gaps in SCWR’s heat transfer and critical flow database

for SCWR concept development. Data from the SCWR prototype pile are
needed. The unique thermohydraulic behavior and sharp property
changes with water around the critical point needs to be investigated
more deeply. A test reactor needs to be designed and built to provide
verification and reference for the reactor design and fuel design.
4. Safety analysis tool development for SCWR
Safety analysis code is an essential analysis tool for SCWR safety
evaluation and safety system design. As SCWR apply water as coolant, its
safety analysis code has many similarities with those used for PWR or
BWR. Many researchers take advantage of mature PWR commercial
safety analysis codes’ predicting ability under subcritical pressure and
expand these codes’ application range to supercritical pressure. Through
this method, a large amount of pressurized accidents and depressurized
accidents with slow depressurization rate can be evaluated. However,
there are still problems when coolant system pressure decreases quickly
to subcritical pressure, which is a typical process in loss of coolant ac­
cident(LOCA) scenario. Water passing through critical point experiences
sharp property change, which makes system codes hard to converge.
Three basic methods are developed to overcome this problem.

Fig. 21. SCWR-R design concept.

risks of flow instability and heat transfer deterioration inside the core
(Bae et al., 2007).
3. Discussion
The basic operating parameters of supercritical water reactors in
various countries are summarized in Table 1. As can be seen from the
table, SCWR can achieve high cycle efficiency because of its higher
outlet temperature. SCWR concepts can be designed as thermal, fast or
mixed neutron spectrum type. Most of SCWR designers apply multiple

passes to make up the core, which aims to reduce the coolant outlet
temperature of each pass-through coolant pre-mixing in the lower
plenum or upper plenum. This is because that SCWR has a very large
coolant enthalpy rise between core inlet and outlet, which is eight times
of that of existing PWR reactors. Thus, a hot channel factor of 2 would
result in coolant outlet temperature of 1200 ◦ C in single-pass core
configuration, which exceed the coolant temperature limit by a large
degree(Schulenberg et al., 2011). Another method to decrease the
enthalpy rise inside the core is achieved through adding an internal
circulation, which is adopted by Korean designers.
On the aspects of SCWR applied moderator, most of the SCWR

4.1. Separate code applied for simulation under supercritical and
subcritical pressure
Researchers from Japan apply a series of codes, named SPRAT, to
carry out safety analysis for their SCWR concepts(Super LWR and Super
FR). Code SPRAT applies homogenous model and fully implicit nu­
merical method to solve conservation equations. Code SPRAT-DOWN is
developed based on SPRAT, which could only simulate transients under
supercritical pressure(Yuki Ishiwatari, 2005). A separate code named
SPRAT-DOWN-DP is developed to simulate quickly depressurization
process for SCWR. When the coolant system depressurizes to equilib­
rium pressure between coolant system and containment, another spe­
cific code SCRELA is applied to simulate the core reflood process
(Ishiwatari et al., 2006). Code SCRELA has detailed constitutive models
to evaluate reflood process and it’s the only code focusing on validating

Table 1
Key parameters of SCWR.
Name

Type
Spectrum
Pressure (MPa)
Inlet Temp. (◦ C)
Outlet Temp. (◦ C)
Thermal Power (MW)
Efficiency (%)
Active Core Height (m)
Fuel
Moderator
No. of Flow Passes

China

Japan

Canada

EU

Russian
Federation

Korea

CSR1000

SCWR-M

Super LWR

Super FR

Canadian
SCWR

HPLWR

VVER-SCP

SCWR-R

PV
Thermal
25
280
500
2300
43
3
UO2
H2O
2

PV
Mixed
25
280
510

3800
44
4.5
UO2/MOX
H2O/2

PV
Thermal
25
290
560
3794
46
4.2
UO2
H2O
2

PV
Fast
25
280
501
1602
44
2.4
MOX
-/ZrH
2

PT
Thermal
25
350
625
2540
48
5
Pu–Th(UO2)
D2O
1

PV
Thermal
25
280
500
2300
43.5
4.2
UO2
H2O
3

PV
Fast
24.5
290
540
3830

45
4.07
MOX
–
1

PV
Thermal
25
280
510
3255
43.68
3.66
UO2
ZrH2
1

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its code prediction ability on reflood process for tight lattice bundles of
SCWR. Comprehensive safety analyses for super LWR require the above
codes to cooperate together to finish simulating accident like LOCA.
4.2. Pseudo two-phase region development under supercritical pressure
region

For researchers who try to upgrade mature safety analysis code for
PWR or BWR based on two-fluid model to supercritical pressure,
developing pseudo two-phase region for supercritical pressure condition
is a solution. Pseudo two-phase region development for supercritical
water tries to regard supercritical water as subcooled supercritical
water, overheated supercritical water and two-phase supercritical
water, which is consistent with phase state definition for water under
subcritical pressure. In this way, water blowing down from supercritical
pressure to subcritical pressure will experience a continuous phase
change, which is helpful for code numerical convergence. Many codes
ănninen and Kurki,
for SCWR apply this method, such as APROS(Ha
2008), ATHLET(Zhou et al., 2012), CATHARE(Geffraye et al., 2011;
IAEA, 2014) and so on.

Fig. 22. Pseudo two phase method applied in ATHLET-SC.

4.2.1. APROS
For the upgrated APROS, the latent heat of condensation or vapor­
ization at supercritical pressure is assumed to be constant. The pseudo
saturation enthalpies are achieved through the following equations
ănninen and Ylijoki, 2008):
(Ha

Surface tension is assumed to 0 under supercritical pressure.
Under supercritical pressure, velocities of pseudo gas and liquid are
assumed to be the same. Thus, a very large number is assigned to the
interfacial friction under supercritical pressure.
Additionally, correlation of Kirillov is applied(Pioro et al., 2004) to
calculate the wall friction under supercritical pressure:

( )0.4
1
ρw
fsp,k =
2
(1.82 log10 (Reb ) − 1.64) ρb

L pe
2
L pe
hg,sat (p) = hpc (p) +
2

h1,sat (p) = hpc (p) −

Besides the pseudo two phase region development for physical
property calculation, the empirical correlations for interfacial heat
transfer under supercritical pressure should also be implemented into
APROS code. A very large interfacial heat transfer coefficient is assumed
at pseudo two phase region at supercritical pressure, which could make
the void fraction vary almost instantly when coolant go through the
pseudo two phase region.
Under subcritical pressure, the heat transfer calculation regimes of
ănninen and Kurki, 2008; Kurk, 2008) are represented by
APROS(Ha
wetted wall regime,dry wall regime and a transition regime. Critical
heat flux (CHF), minimum film boiling temperature(MFB), wall tem­
perature as well as coolant temperature are used to define different heat
transfer regimes. However, critical heat flux, minimum film boiling
temperature don’t exist under supercritical pressure. Upgrated APROS

code applies Jackson correlation to evaluate the heat transfer co­
efficients under supercritical pressure(Hall et al., 1967). Different values
for variable n in the following equation are defined by comparing wall
temperature with coolant temperature and pseudo critical temperature.

(
0.5
Nub = 0.0183Re0.82
b Prb

⎧
⎪
⎪
⎪
⎪
⎪
⎪
⎪
⎪
⎪
⎨

ρw
ρb

)0.3 (

cp
cp,b

4.2.2. ATHLET-SC
Pseudo two phase region is set for ATHLET-SC where pressure is over
22.05 MPa, instead of critical pressure(22.1 MPa). The latent heat at
22.05 MPa is set as the latent heat over all supercritical pressure region,
as shown in Fig. 22. The effects of width of the pseudo two phase zone
are studied in order to avoid convergence problem of the modified code
and large deviation from reality. Bishop et al.(Bishop et al., 1964)、
Krasnoshchekov and Protopopov(Krasnoshchekov and Protopopov,
1966) and Yamagata et al.(Yamagata et al., 1972), Jackson(Jackson,
2009), Cheng et al.(Cheng et al., 2009) are incorporated into the code to
simulate the heat transfer under supercritical pressure.
The velocities of pseudo gas and liquid under supercritical pressure
are assumed to be the same. Interfacial heat transfer is mainly made up
of heat conduction. The flow type is assumed to be annular flow(under
unheated condition) or inverted annular flow(under heated condition).
Thus, the interfacial heat transfer and interfacial area can be calculated
by following equations(Zhou et al., 2012):

)n

0.4,
if Tb < Tw < Tpc or 1.2Tpc < Tb < Tw
(
)
Tw
0.4 + 0.2
− 1 ,
if Tb < Tpc < Tw
Tpc
n=

⎪
⎪
(
)(
(
))
⎪
⎪
Tw
Tb
⎪
⎪
⎪
0.4 + 0.2
− 1 1− 5
− 1 ,
if Tpc < Tb < 1.2Tpc and Tb < Tw
⎪
⎪
Tpc
Tpc
⎩

9


P. Wu et al.

Progress in Nuclear Energy 153 (2022) 104409

Fig. 23. Regions of liquid and vapor for CATHARE (IAEA (2014)).

2λ

L
̅
αL = √̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅

(1 − x)D2h

2λ

V
̅
αV = √̅̅̅̅̅̅̅

xD2h

Fig. 24. Strategy for supercritical/subcritical transitions in CATHENA(Beuthe
et al., 2020).

√̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅
Ai,annual flow = π (1 − x)D2h ⋅l
Ai,inverted annual flow = π

2013).

√̅̅̅̅̅̅̅̅
xD2h ⋅l

4.3.2. CATHENA
CATHENA is a safety analysis code developed by Atomic Energy of
Canada Limited(AECL) for CANDU reactors. It applies one-dimensional
two-fluid model to simulate water flowing in pipes. In order to apply
CATHENA to carry out safety analysis for Canadian SCWR (Beuthe et al.,
2020), implements a novel method to enable CATHENA simulating ac­
cidents and startup transients for SCWR. Instead of introducing pseudo
two phase regime (Beuthe et al., 2020), treats the supercritical regime as
a single uninterrupted phase, whose void fraction is always 0. Thus,
there is only transition from supercritical fluid at high temperature to
sub-critical steam because of sudden void fraction variation from 0 to 1,
as shown in Fig. 24. A special strategy is developed to split the super­
critical phase into vapor and liquid phases under subcritical pressures
and an initial value of void fraction is estimated when trans-critical
process from supercritical to subcritical pressure occurs. Additionally,
the sharp physical property change near the vicinity of the critical point
is manually mitigate to avoid convergence problem of numerical simu­
lation. Supercritical blowdown process is simulated by modified
CATHENA to verify its ability to simulate trans-critical process.

λ, thermal conductivity; Dh, hydraulic diameter; Ai, interface area; l,
length of volume; x, void fraction; α, interphase heat transfer coefficient.
Subscript L denotes the liquid phase while V denotes the vapor phase.
4.2.3. CATHARE
Unlike APROS and ATHLET-SC, there is only pseudo liquid or vapor
region while no pseudo two phase region exists under supercritical
pressure for CATHARE, as shown in Fig. 23. The void fraction of su­
percritical water is 0 if bulk temperature is smaller than corresponding
pseudo critical temperature and is 1 if bulk temperature is larger than
corresponding pseudo critical temperature. No buffer zone exists be­

tween pseudo liquid and vapor region. CATHARE version for SCWR is
reported to be numerically robust to carry out fast trans-critical pressure
simulation according to (Geffraye et al., 2011). However, it’s a pity the
author doesn’t find any paper describes how CATHARE code make it
since sudden void fraction change when bulk temperature moves cross
the pseudo critical temperature is apt to cause non-convergence during
numerical calculation.

4.3.3. RELAP5 series codes
Several modifications have been made for RELAP5-3D to enable its
ability to simulate slow and fast blowdown process from supercritical to
sub-critical pressure(Rassame et al., 2017). Different versions of RELAP5

4.3. Physical property modification around critical point
Another method applied by SCWR system codes is to process the
physical property around critical point unphysically to avoid sharp
property change. These codes include SCTRAN(Wu et al., 2015),
CATHENA(Beuthe et al., 2020), RELAP5/MOD4(Allison et al., 2016;
Lou, 2016) or RELAP5-3D(Riemke et al., 2003).
4.3.1. SCTRAN
SCTRAN is developed by Xi’an Jiaotong University (Wu et al., 2015)
to provide a safety analysis tool for SCWR, which applies homogenous
model to predict the flow characteristics of coolant. Homogenous model
is suitable for supercritical water as there is no phase change and the
fluid can be treated as single phase. What should be done to enable
SCTRAN to simulate trans-critical process is that the physical properties
under supercritical pressure and near critical pressure should be
continuous and their derivatives near the critical point should be
decreased artificially. Through calculation of fluid temperature, specific
volume, specific heat and saturated enthalpy through fitted property

correlations, physical properties could change smoothly through critical
pressure. SCTRAN code has been used to carry out safety evaluations for
accidents including LOCA and non-LOCA type for pressure vessel type
SCWR, such as CGNPC SCWR(Wu et al., 2015), CSR1000(Wu et al.,
2014), and pressure tube type SCWR, such as Canadian SCWR(Wu et al.,

Fig. 25. Trans-critical transition mechanisms in RELAP5/MOD4.
10


P. Wu et al.

Progress in Nuclear Energy 153 (2022) 104409

applied property tables to calculate physical property. The specific heat,
volume expansibility and isothermal compressibility at critical point
was reset to avoid sharp change. At the same time, more pressure and
temperature points are set near the critical point to rebuild the property
table. The interpolation of specific volume and isothermal compress­
ibility changed from cubic to linear expressions. After these modifica­
tion, RELAP5-3D could be used to simulate slow transients above the
critical pressure or under trans-critical process.
(Hu and Wilson, 2014) expands RELAP5/MOD3.3’s application
range to supercritical pressure through similar method as (Riemke et al.,
2003). More data points are set for the high-pressure and
high-temperature range in the physical table to ensure better physical
property calculation accuracy under supercritical pressure. The modi­
fied RELAP5/MOD3.3 is designed to couple with a neutron code named
PARCS. However, this version of RELAP5/MOD3.3 can’t carry out
simulation for trans-critical process.

RELAP5/MOD4 also treats the supercritical water as single-phase
liquid and the vapor void fraction above critical pressure is always 0.
When volume experiences pressure change from supercritical to
subcritical pressure, the trans-critical transition mechanism can be
found in Fig. 25. It’s obvious there is one case that supercritical water
whose void fraction is 0 will transition into subcritical vapor region
whose void fraction is 1. For these cases, the fluid density as well as fluid

enthalpy at last time step under supercritical pressure will be used to
present the subcritical vapor properties at current time step. For the
liquid properties, saturated liquid properties under current subcrtical
pressure are applied. The velocity of the subcritical vapor phase equals
to supercritical velocity at last time step. There is also a case that the
volume condition will change from supercritical state to saturated
mixture state under subcritical pressure, as indicated in Fig. 25. In this
case, the liquid and vapor densities equal to the saturated densities
under subcritical pressure, and the velocity of both phases are the same
as the supercritical velocity from last time step, while the void fraction is
determined by the enthalpy under subcritical pressure. In this way,
sudden physical change from supercritical to subcritical pressure can be
avoid(Allison et al., 2016; Lou, 2016). The prediction accuracy of vis­
cosity and thermal conductivity is improved by (Lou, 2016) through
refitting data obtained from NIST, which results in steady-state MCST of
Canadian SCWR increasing from 990K to 1060K.
5. Discussion
A comprehensive summary of existing SCWR codes is shown in
Table 2. These codes apply different basic conservation equations,
coolant thermal properties, heat transfer correlations under subcritical
and supercritical pressure, as well as critical flow model. The key

Table 2
Summary of existing SCWR codes.
Code

Serial codes of SPRAT

ATHLET-SC

APROS

RELAP5-3D

RELAP5 MOD3.3

RELAP5 MOD4

SCTRAN

Developing
organization

The University of Tokyo

Shanghai Jiaotong
University
Two-fluid model

US Nuclear
Regulatory
Commission

Two-fluid model

Xian Jiaotong
University

Homogeneous flow

Idaho National
Engineering
Laboratory
Two-fluid model

Innovative
System Software

Basic model

Two-fluid model

Fluid property
model for
subcritical
pressure

IAPWS-IF97

Physical property
table

Physical

property table

Physical property
table

Heat transfer
model for
subcritical
pressure

D-B correlation, Thom
correlation, SchrockGrossman correlation,
Groeneveld film-boiling
look-up table,
McDonough, Milich and
King correlation,
Groeneveld CHF look-up
table
Unknown

Water-vapor physical
property package,
Pressure-specific
enthalpy physical
property package
Series of heat transfer
correlations

VTT Technical
Research Centre

of Finland
Two-fluid
model
IAPWS-IF97

Wetted wall:DB correlation,
Thom
correlation;
Dry wall:fitting
correlation;
Transition
region: linear
interpolation
IAPWS-IF97

Multiple heat
transfer models

Multiple heat
transfer models

Multiple heat
transfer models

Homogeneous
flow
Independently
developed
polynomial
physical property

correlation
Correlation of
different heat
exchange forms

Add data points

Add data points

Add data points

Jackson and
Hall correlation

Bishop correlation,
Oka-Koshizuka
correlation,
Jackson correlation

D-B correlation

No special
consideration

Bishop
correlation, OkaKoshizuka
correlation,
Jackson
correlation
No special

consideration

No special
consideration

Moody
homogeneous
equilibrium model

Increase the
physical
properties near
the critical point

Calculate
separately for a
single control
volume or
adjacent control
volumes
D-B correlation

Linearization of
physical
properties in
transcritical
process

Fluid property
model for

supercritial
pressure

IAPWS-IF97

Heat transfer
model for
supercritical
pressure

Oka-Koshizuka
correlation, D-B
correlation

Critical flow
model

Moody homogeneous
equilibrium model

Bishop correlation,
Yamagata correlation,
Jackson correlation,
Cheng correlation,
Krasnoshchekov
correlation
Modified Y. Z. Chen
critical flow model

Calculation

method of
transcritical

The coupling of the two
programs

Pseudo two phase
region

Zuber-Griffith
correlation for
low flow rate,
Biasi
correlation
Pseudo two
phase region

Heat transfer
model in
transcritical
process

D-B correlation,

Cheng correlation,
Jackson correlation

Modified Thom
correlation

11

Increase the
physical properties
near the critical
point, Change cubic
interpolation to
linear interpolation
No special
consideration

No special
consideration

Independently
developed
polynomial
physical property
correlation
Bishop
correlation,
Jackson
correlation

Zahlan
transcritical lookup table


P. Wu et al.

Progress in Nuclear Energy 153 (2022) 104409

difference of these codes is the method to solve the trans-critical nu­
merical process, which is also the common difficulty for SCWR code
development. Both of the two methods described in section 3.2 and 3.3
are designed to decrease the physical property’s partial derivatives
unphysically, which may result in numerical convergence problem.
Though they are not the best methods, these codes can somehow predict
representative SCWR transient behaviors including under supercritical
and trans-critical pressure process. These methods’ effects on the pre­
diction accuracy are not evaluated and fully understood yet. Most of the
above system codes applied code to code comparison or separate-effect
experiment to carry out validations, which are not enough to quantify
the prediction error caused by the special trans-critical techniques
applied by all the SCWR system codes. A new method which takes the
effect of sharp physical property near critical point into consideration is
still needed to be developed in the future. Additionally, the validation
work for SCWR system code needs further effort and more system level
experiments should be carried out to support system code validation.

automatic depressurization system (ADS), the passive containment
cooling system (PCCS), and the gravity driven core cooling system
(GDCS), which are shown in Fig. 26.
Functions of each safety system are as follows:
● RMT: After accident, the pressure difference between the hot and
cold pipe, as well as the gravitational pressure head between the tank
and the core, drives the cold coolant to flow into the core within a
short period of time.
● ICS: The ICS relies on the natural circulation drive for uninterrupted
cooling of the core and ensures that the reactor is cooled by the safety

system in the late stages of an accident.
● ADS: In response to emergency situations, the automatic depressur­
ization system provides an automatic and effective means of
relieving pressure. Safety relief valves are used to provide over­
pressure protection for the reactor while it’s also used to depres­
surized the system in LOCA type accident.
● GDCS: Function of GDCS is to automatically provide emergency core
cooling in the event of any accident that may affect the reactor
coolant charge. Once the reactor is depressurized to containment
pressure via the ADS system, the GDCS tank has the capability to
automatically fill the pressure vessel with large volumes of water by
gravity effect.
● PCCS: PCCS consists of a condenser and dryer and is used to ensure
that the pressure and temperature inside the containment are kept
below the design limits in the event of a design basis accident, such as
LOCA.

6. Typical safety system design and safety characteristics
The operation of SCWR can be divided into three types, which are
normal operation, abnormal transients, and accidents. SCWR behaves
differently from the existing water-cooled reactors in normal operation,
since it has neither the circulation loop of BWR nor the primary circuit of
PWR. In existing SCWR designs, supercritical fluid flows directly into the
turbine from the exit of the reactor pressure vessel, so there is also no
dryer or separator in the system. Therefore, SCWR’s safety guidelines
requires that there is a certain coolant flow rate in the core, instead of
requiring a sufficient coolant charge (Oka et al., 2010a). Safety analysis
is important in SCWR R&D. According to existing research, the
maximum cladding surface temperature (MCST) of SCWR needs to be
ensured that is not higher than 850 ◦ C under abnormal transient con­

ditions and no higher than 1206 ◦ C in the event of an accident that could
result in core damage.
Existing safety system design for different types of SCWR and pre­
liminary safety performance evaluations are reviewed in this part to
shed a light on the safety characteristics of both types of SCWR concepts.

Initial security evaluation of CSR1000 which adopts this set of pas­
sive safety system is carried out by Wu et al. (Wu et al., 2014) applying
code SCTRAN. Safety performance of CSR1000 under different types of
accidents are analyzed.
● LOCA type: loss of coolant accident(LOCA);
● Non-LOCA type: uncontrolled CR withdrawal, partial loss of main
coolant flow, loss of offsite power, Pump seizure, main coolant
control system failure, loss of feedwater heating, loss of coolant
flowrate accident(LOFA), main steam line valve closure.

6.1. Pressure-vessel type SCWR

The corresponding MCST is shown in Fig. 27. From the results we can
find that this set of passive safety system is effective to protect core from
overheating. The largest MCSTs under transients and accidents are

6.1.1. CSR1000’s safety system design and safety characteristics
Wu et al.(Wu et al., 2014) designs a completely passive safety system
for the CSR1000 referring to safety system of AP1000(Schulz, 2006),
CPR1000(Wang et al., 2013, 2014), ESBWR(Rassame et al., 2017) as
well as innovative integral inherent safety light water reactor(I2S-LWR)
(Wang et al., 2020), which aims to provide in-time safety injection for
SCWR. The safety system of CSR1000 is made up of the isolation
condenser system (ICS), high pressure reactor make-up tank (RMT), the

Fig. 26. Passive safety system of CSR1000.

Fig. 27. MCST of CSR1000 under different transients and accidents.
12


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Progress in Nuclear Energy 153 (2022) 104409

(IAEA, 2019; Yuki Ishiwatari, 2005; Ishiwatari and Oka et al., 2010a;
Oka and Koshizuka et al., 2010) have carried out most comprehensive
safety analysis for Super LWR, including fifteen transients and accident,
as well as LOCA type accident and ATWS, part of which are shown in
Table 3.
Fig. 29 shows the safety performances of different accidents. The
maximum cladding temperature increase of transients and accidents are
50 ◦ C and 250 ◦ C respectively while the maximum pressure increase is
2.5 MPa for transients and 2 MPa for accidents, which satisfy the safety
criterion. Water rod designed in Super LWR concept play an important
role in providing coolant under transients and accidents, which could
enable the auxiliary feedwater system being actuated with a 30s delay
(Yuki Ishiwatari, 2005). For LOCA type accidents, ADS actuation will
help decrease the core pressure and increase the core flowrate, which is
benefit to carry away the core decay heat. The LPCI system is responsible
for reflooding the overheated core and provide coolant for longer
cooling in the late stage of LOCA type accidents(Ishiwatari et al., 2006).

e

Fig. 28. Sketch of the Super LWR ′ s safety system (Oka et al., 2010a).

6.2. Pressure-tube SCWR
As the only pressure tube SCWR concepts in the world, Canadian
SCWR’s safety characteristics are quite different from previously
mentioned pressure-vessel type SCWR. A goal of “No-Core-Melt” is
achieved by radiative heat transfer between pressure tube and fuel pins,
as well as a specifically designed passive moderator cooling system. The
heat transfer path under LOCA without emergency core cooling(LOCA/
LOECC) is shown in Fig. 30. The proposed moderator cooling system
allows for completely passive heat removal from the fuel during accident
conditions. A passive moderator heat rejection is possible through the
use of high efficiency fuel channel (HEC), which maintains direct con­
tact with the surrounding moderator. At accident conditions, fuel
channels allow passive decay heat removal which prevents fuel melting.
Under the worst case scenario of total loss of coolant, fuel temperature
increases to a point that decay heat can be transferred by radiation to the
insulator which allows sufficient transfer of heat from fuel assembly to
the pressure tube and after that transfer to the moderator to prevent
melting of the cladding (Novog et al., 2012).
(Wu et al., 2013) applies SCTRAN incorporating a special radiation
heat transfer model, to simulate the LOCA/LOECC accidents. From
Fig. 31-b) we can see that with coolant draining out of the coolant
channel, radiation heat transfer starts to play a very important role in
removing fuel heat after around 30s. The surface temperature variation
in Fig. 31-c) shows there are two peak temperature values during the
accidents. The first peak is caused by mismatch between core heat and
core mass flowrate. The second peak is due to fact that the radiation heat
transfer is still not big enough to carry away all the decay heat. With

accident progresses and the decay heat decreases, the radiation heat
transfer and decay heat gradually achieve balance. In this process the
highest cladding temperature(1278 ◦ C) is lower than 1400 ◦ C which is

Table 3
Transients and accidents in safety analysis(Yuki Ishiwatari et al., 2005).
Transients
1
2
3
4
5
6

Loss of feedwater heating
Inadvertent startup of AFS
Partial loss of reactor coolant
flow
Loss of offsite power

7
8
9

MSIV closure
CR withdrawal at normal operation
CR withdraw at startup

10

Loss of load with turbine bypass
Loss of load without turbine
bypass

11

Main coolant flow control system
failure
Pressure control system failure

3
4

CR ejection at normal operation
CR ejection at hot standby

Accidents
1
2

Total loss of reactor coolant flow
Reactor coolant pump seizure

780 ◦ C and 850 ◦ C, which are far below the corresponding safety
criterion.
6.1.2. Japanese SCWR safety system design and evaluation
Super LWR mainly apply active safety system because the designer
thinks that the design of water rods in the core can act as an in-vessel
accumulator and it could provide buffer time for active system to
launch, such as auxiliary feedwater system. Fig. 28 shows the Super

LWR’s safety system, which is made up of automatic depressurization
system(ADS), main steam isolation valves(MSIVs), low pressure core
injection system(LPCI), auxiliary feedwater system(AFS) and Suppres­
sion chamber(Oka et al., 2010a).

Fig. 29. Transient and accident result summary for Super LWR.
13


P. Wu et al.

Progress in Nuclear Energy 153 (2022) 104409

Fig. 30. Proposed active and passive moderator cooling systems.

Fig. 31. LOCA/LOECC accident simulation using SCTRAN.

the modified stainless steel SS310’s melting temperature. Increase the
heat exchange of radiation heat exchange of fuel cladding and moder­
ator channel is benifical to decrease the maximum surface temperature
under LOCA/LOECC.
Another unique feature of Canadian SCWR is that there is a power
excursion due to void reactivity feedback in postulated accident such as
LOFA and LOCA, which has not been reported for pressure-vessel type
SCWRs (Hummel and Novog, 2016). applies coupled 3D neutron
transport code DRAGON and system code CATHENA to investigate this
phenomenon. The power pulse could be 160% full power in a transient
caused by core inlet pressure decrease. A fast reactor shut-down system
is suggested to avoid these kinds of power pulse in accidents.

8. Conclusion
Existing SCWR concepts, system code development, safety system
and safety characteristics are comprehensively reviewed in this paper.
Through the review, we can find that design of multi-flow pass in SCWR
core is the main method to reduce hot spot. Low coolant inventory and
ratio of coolant mass flowrate over thermal power are distinguish fea­
tures of SCWR, compared to PWR and BWR, which could result in
reduced safety performance. Passive safety system, design of water rod
in the core, or application of innovative heat transfer method such as
radiation heat transfer, are all good ways to improve safety performance
for SCWR. All the existing safety evaluation paper or reports indicate
that safety is not a problem for SCWR. However, the analysis tools
applied to carry out safety analyses still needs further development and
validation. The trans-critical techniques applied in existing SCWR codes
all try to avoid the sharp thermal property changes or decrease the peak
value of thermal properties without any physical principles. The effects
caused by these techniques on overall safety performance evaluation is
not clarified up to now. Additionally, existing SCWR analysis codes still
lack of validation to prove their ability in predicting methods. Therefore,
large-scale and reliable experiments are needed to prove the accuracy of
existing codes near the critical pressure and to develop scientific and
wide-ranging theoretical models.

7. Discussion
SCWR has the problems of low water inventory and small specific
heat of core coolant. Safety performance of reactors could be improved
in different ways. Passive safety system is a good way to provide in-time
emergency core cooling for SCWR. Design of water rod in the core is also
useful to provide in-core cooling water, which has been proved to be
useful in safety evaluation of Super LWR. Canadian SCWR has its unique

safety feature, which applies radiation heat transfer mechanism to
remove decay heat. Existing safety analyses shows that both pressurevessel and pressure-type SCWR concepts can satisfy the safety crite­
rion in assumed accidents.

14


Progress in Nuclear Energy 153 (2022) 104409

P. Wu et al.

Declaration of competing interest

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The authors declare that they have no known competing financial
interests or personal relationships that could have appeared to influence
the work reported in this paper.
Data availability
No data was used for the research described in the article.
Acknowledgements
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