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

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.

Progress in Nuclear Energy 153 (2022) 104409 Contents lists available at ScienceDirect 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 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: wupan2015@mail.xjtu.edu.cn (P Wu) https://doi.org/10.1016/j.pnucene.2022.104409 Received May 2022; Received in revised form 23 August 2022; Accepted 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 (http://creativecommons.org/licenses/by/4.0/) 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 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 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 Fig shows the cross section of fuel assembly, consisting of subassemblies and water rods Water flowing through these two paths acts as moderator and coolant simultaneously Additionally, for better P Wu et al Progress in Nuclear Energy 153 (2022) 104409 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 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 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 Some of the plant parameters are also included in Fig After design updates, the feed water temperature of Super LWR is Fig 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 shows the Fig Fuel assemblies distribution in SCWR-M(Xu et al., 2011) Fig Struture of fuel assembly of SCWR-M P Wu et al Progress in Nuclear Energy 153 (2022) 104409 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 Since fast reactors 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 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 Flow circulation of coolant cycle for Super LWR(Oka et al., 2010a) Fig Schematic diagram of coolant flow inside core(Oka et al., 2010a) Fig The fuel assembly distribution of the core(Oka et al., 2010a) P Wu et al Progress in Nuclear Energy 153 (2022) 104409 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 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 sub-assemblies with total 40 fuel rods in it and an additional moderator box in the P Wu et al Progress in Nuclear Energy 153 (2022) 104409 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 P Wu et al Progress in Nuclear Energy 153 (2022) 104409 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 P Wu et al 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 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) Discussion The basic operating parameters of supercritical water reactors in various countries are summarized in Table 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 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 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 UO2 H2O PV Mixed 25 280 510 3800 44 4.5 UO2/MOX H2O/2 PV Thermal 25 290 560 3794 46 4.2 UO2 H2O PV Fast 25 280 501 1602 44 2.4 MOX -/ZrH PT Thermal 25 350 625 2540 48 Pu–Th(UO2) D2O PV Thermal 25 280 500 2300 43.5 4.2 UO2 H2O PV Fast 24.5 290 540 3830 45 4.07 MOX – PV Thermal 25 280 510 3255 43.68 3.66 UO2 ZrH2 P Wu et al Progress in Nuclear Energy 153 (2022) 104409 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 ¨nninen and Ylijoki, 2008): (Ha Surface tension is assumed to 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 ρw fsp,k = (1.82 log10 (Reb ) − 1.64) ρb L pe L pe hg,sat (p) = hpc (p) + 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 − , if Tb < Tpc < Tw Tpc n= ⎪ ⎪ ( )( ( )) ⎪ ⎪ Tw Tb ⎪ ⎪ ⎪ 0.4 + 0.2 − 1− − , if Tpc < Tb < 1.2Tpc and Tb < Tw ⎪ ⎪ Tpc Tpc ⎩ 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 Thus, there is only transition from supercritical fluid at high temperature to sub-critical steam because of sudden void fraction variation from 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 if bulk temperature is smaller than corresponding pseudo critical temperature and is 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 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 will transition into subcritical vapor region whose void fraction is 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 Discussion A comprehensive summary of existing SCWR codes is shown in Table 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 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 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 P Wu et al 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 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 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 Transients and accidents in safety analysis(Yuki Ishiwatari et al., 2005) Transients Loss of feedwater heating Inadvertent startup of AFS Partial loss of reactor coolant flow Loss of offsite power 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 CR ejection at normal operation CR ejection at hot standby Accidents 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 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 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 Ishiwatari, Y., Oka, Y., et al., 2006 LOCA analysis of super LWR J Nucl Sci Technol 43 (3), 231–241 Jackson, J.D., 2009 Validation of an Extended Heat Transfer Equation for Fluids at Supercritical Pressure Krasnoshchekov, E.A., Protopopov, V.S., 1966 Experimental study of heat exchange in carbon dioxide in the supercritical range at high temperature drops Teplofiz Vysok Temp (3), 389–398 Kurki, J., 2008 Simulation of thermal hydraulics at supercritical pressures with APROS Master’s Thesis, Helsinki University of Technology Espoo, Finland Liu, X.J., Fu, S.W., et al., 2013 LOCA analysis of SCWR-M with passive safety system Nucl Eng Des 259, 187–197 Lou, M., 2016 Loss of Coolant Accident Simulation for the Canadian Supercritical WaterCooled Reactor Using RELAP5/MOD4 McMaster 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accident analysis of SCWR with mixed spectrum core Prog Nucl Energy 54 (1), 150–161 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 This paper is supported by the National Key R&D Program of China (2018YFE0116100) References Allison, C.M., Wagner, R.J., et al., 2016 The development of RELAP/SCDAPSIM/ MOD4.0 for advanced fluid systems design analysis In: The 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety Gyeongju, Korea Bae, K.M., Joo, H.K., et al., 2008 Conceptual Design of a 1400 MWe Supercritical Water Cooled Reactor Core with a Cruciform Type U/Zr Solid Moderator Bae, Y., Jang, J., et al., 2007 Research activities on a supercritical pressure water reactor in Korea Nucl Eng Technol (39) Beuthe, T., Vasic, A., et al., 2020 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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... 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... 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