Front. Energy Power Eng. China 2007, 1(2): 141–149 DOI 10.1007/s11708-007-0018-6

RESEARCH ARTICLE

CHENG Xu

Studies on advanced water-cooled reactors beyond generation III for power generation

© Higher Education Press and Springer-Verlag 2007

Abstract China’s ambitious nuclear power program motivates the country’s nuclear community to develop advanced reactor concepts beyond generation III to ensure a long-term, stable, and sustainable development of nuclear power. The paper discusses some main criteria for the selection of future water-cooled reactors by considering the specific Chinese situation. Based on the suggested selection criteria, two new types of water-cooled reactors are recommended for future Chinese nuclear power generation. The high conversion pressurized water reactor utilizes the present PWR technology to a large extent. With a conversion ratio of about 0.95, the fuel utilization is increased about 5 times. This significantly improves the sustainability of fuel resources. The supercritical water-cooled reactor has favorable features in economics, sustainability and technology availability. It is a logical extension of the generation III PWR technology in China. The status of international R&D work is reviewed. A new supercritical water-cooled reactor (SCWR) core structure (the mixed reactor core) and a new fuel assembly design (two-rows FA) are proposed. The preliminary analysis using a coupled neutron-physics/thermal-hydraulics method is carried out. It shows good feasibility for the new design proposal. Keywords GEN-IV nuclear reactor, high conversion PWR, supercritical water cooled reactor, mixed reactor core

1

Introduction

The operating experience of more than five decades of nuclear energy for civil use shows significant advantages with respect to environment protection, economic competitiveness and power supply stability. Nowadays, nuclear

Received January 10, 2007; accepted March 2, 2007 CHENG Xu ( ) School of Nuclear Science and Engineering, Shanghai Jiaotong University, Shanghai 200240, China E-mail: [email protected]

power plays an important role in power generation and produces about 16% of the total electricity worldwide. The development of nuclear power technology can be classified into four generations. The first generation includes demonstration plants with small power capacity. Based on experience gathered from the first generation, a large number of standardized concepts of nuclear power plants (NPP) were proposed and the second-generation (GEN-II) NPP was born. Most NPPs operating nowadays belong to the second generation. After the accidents of TMI and Chernobyl intensive efforts were made to improve the safety features of secondgeneration NPP, and the third generation of nuclear power technology was developed. Compared to the second generation, the third generation has a much higher safety level. The core damage frequency (CDF) is lower than 10−5 per reactoryear. It is expected that in the next 2–3 decades, newly constructed NPPs will mainly use reactors of GEN-III. In spite of the high safety level, GEN-III NPPs show some shortcomings related to the requirements of long-term nuclear power development. 1 Economics: continuous improvement in safety features makes the system more complicated and expensive. Economic competitiveness is strongly affected. 2 Sustainability: nearly all NPPs operating today use reactors with thermal neutron spectrum. The conversion ratio is low, e.g. in a conventional PWR the conversion ratio is about 0.6. This low conversion ratio restricts fuel utilization, which is less than 1%. For countries with limited uranium resources as China, fuel utilization is a key factor affecting the long-term nuclear power development. In addition, low fuel utilization leads to high production of nuclear waste. This would result in a more challenging task for nuclear waste management. Recently, China issued a very ambitious long-term program for nuclear power development [1]. The main reactor concepts for the next 2–3 decades have been determined. They are water-cooled reactors of generation II+ or generation III. However, the selection of reactor designs beyond GEN-III for the future nuclear power program is still open. In this paper, some criteria are suggested for the selection of future reactors. Two possible types of water cooled reactors beyond GEN-III are introduced for power generation that

142 includes features like enhanced fuel utilization and improved economics.

3 High conversion pressurized water reactors

2

The water-cooled reactors of GEN-II and GEN-III operating or being built in China have a thermal neutron spectrum and a conversion ratio of about 0.6. The theoretical relationship between the conversion ratio and the fuel utilization in a closed fuel cycle is expressed as

Main criteria of selecting future WCRs

In China, nuclear power is still in the development stage. In the Mainland, nuclear power makes less than 2% of total electricity production [1]. However, according to the mediumterm plan, in the year 2020 nuclear power capacity will reach over 40 GWe and will contribute 4% of total electricity production in the country. In the Chinese nuclear community, it is commonly agreed that in the middle of this century nuclear power will contribute about 15% of total electricity production and provide an electrical capacity of about 250 GWe. Water-cooled reactors have been selected as the main reactor type for Chinese nuclear power. The government has issued several programs to develop technology for advanced large-scale pressurized water reactors. Although GEN-III technology is attracting a lot of attention of Chinese nuclear scientists and engineers, the following question has to be raised for discussion: what are the advanced water-cooled reactors beyond GEN-III for the Chinese Nuclear Power generation? The technology development of a new type of NPP requires a relatively long period, which could take even several decades. Therefore, the author strongly suggests putting attention to the selection of future reactors beyond GEN-III and proposes the following criteria. (1) Safety: safety remains one of the most important criteria. It has been widely accepted by various countries that the Core Damage Frequency has to be lower than 10−5 per reactor-year. The existing reactor concepts of GEN-III do fulfill this safety requirement. For future nuclear power plants this safety goal has to be kept. The need for further improvement in safety is still open and shall be discussed by combining with other aspects, e.g. economics. (2) Sustainability: the uranium resource is relatively poor in China. According to the OECD/NEA [2], from the technical and economic point of view, the inland exploitable uranium may cover the need of the next 2–3 decades. Therefore, with the selection of future WCRs, attention should be paid to both fuel utilization and the capability of burning other fuels, e.g. thorium. (3) Technology and experience availability: vendors and utilities will pay significant attention to the availability and the continuation of technology and experience in design, construction and operation of NPPs and related fuel cycle procedure. In addition, the time point for the availability of the new technology has to obey the requirement of the nuclear power development program. According to the criteria mentioned above, the author recommends two candidates for water-cooled reactors beyond GEN-III to be used for future Chinese nuclear power plant; a high conversion pressurized water reactor (HCPWR) and a supercritical water-cooled reactor (SCWR).

Fu =

En5 1-C

(1)

And in a once-through fuel cycle ⎛ E ⎞ Fu = En5 ⎜1- d5 ⎟ E5 ⎠ ⎝

(2)

where, Fu stands for fuel utilization, En5 is the content of uranium-235 in the natural uranium ore, which is about 0.72%, Ed5 is the uranium-235 content of the discharged fuel, E5 is the uranium-235 content of the fresh fuel and C is the conversion ratio. Considering the separation efficiency, the fuel utilization is schematically illustrated in Fig. 1. As we can see, the fuel utilization is significantly enhanced (about 10 times) and increases from 1% to about 10%, if the conversion ratio is increased from the present PWR value (0.6) to a value close to 1.

Fig. 1 Relationship between fuel utilization and conversion ratio

In the 1980s, extensive research activities were made to develop water-cooled reactors with high conversion ratio, i.e. a conversion ratio larger than 0.9 [3–5]. Due to the unfavorable political situation caused by the Chernobyl accident, the R&D activities of high conversion water-cooled reactors were terminated in Western countries. In Japan, however, activities are still ongoing to develop high conversion boiling water reactors (BWR) [6]. Considering the specific situation in China, high conversion PWR (HCPWR) might be a favorable solution to improve the fuel utilization and to enhance resource sustainability, which is crucial for the Chinese long-term nuclear development program.

143 A high conversion ratio can be realized by increasing the neutron energy. This can be achieved by modifying the reactor core of existing PWR, whereas other components of the PWR NPP are kept unchanged. The key feature of HCPWR is a tight arrangement of the fuel rods. This way, the water volume, and subsequently the moderation capacity is reduced. Figure 2 compares schematically the fuel rods arrangement of an HCPWR fuel assembly with that of a PWR fuel assembly. In this example, the fuel rod diameter is the same for both concepts. However, the pitch is reduced from 12.6 mm in PWR down to 10.5 mm in HCPWR. Figure 3 shows the relative fuel utilization (normalized by the value of PWR) on the dependence on the water-to-fuel mass ratio in the fuel assembly. These results were obtained using a detailed neutron-physical analysis [7]. The mass ratio in PWR is approximately 0.15, at which the conversion ratio is about 0.6. By decreasing the mass ratio from 0.15 to 0.04, the conversion ratio increases to 0.93 and the relative fuel utilization increases by a factor of 5. A strong increase in fuel utilization can be achieved by a further reduction in the moderatorto-fuel mass ratio. However, this requires an extremely tight lattice, which may prove a challenging task in the design and construction of fuel assemblies. Another approach leading to a further reduction of the moderator-to-fuel mass ratio is to reduce the density of water. This will be discussed in section 4. A tight lattice is usually arranged in a hexagonal structure, which allows for a larger pitch-to-diameter ratio at the same moderator-to-fuel mass ratio. The linear power ratio in HCPWR is similar to that of PWR. Thus, the reactor core power density is nearly doubled and the active reactor volume is reduced to about 50%. By using the same pressure vessel, the effective core diameter is also kept unchanged. The active core height is reduced from about 4 m in PWR to about 2 m in HCPWR (see Fig. 4). Because the HCPWR takes the existing PWR technology and devices, except the reactor core design, it has excellent technology availability and can be realized in a short term. The technology and experience achieved in the frame of

the PWR development can be directly applied to HCPWR, including design, manufacture and operation. Nevertheless, related to the reactor core design, a more detailed study on neutron-physics and thermal-hydraulics are required. Usually, a harder neutron spectrum would reduce the temperature reactivity coefficient of the coolant and cause additional safety issues. Therefore, design optimization is required considering the fuel utilization and reactivity coefficients. Existing computer codes have to be verified for tight lattice PWR with epi-thermal spectrum. Concerning thermal-hydraulics there is no sufficient knowledge about the flow and heat transfer behavior in tight lattices, e.g. critical heat flux. The non-uniformity of local heat transfer in a tight lattice could cause a serious problem and needs further study [8].

4

Supercritical water-cooled reactors

The Generation IV International Forum (GIF), of which China is one of the 13 members, has recommended six generation IV (GEN-IV) reactor concepts [9]. The SCWR is the only one using water as coolant. One of the main objectives of introducing SCWR is to enhance the economics of NPP. Despite the more than five decades of development of watercooled reactors, the thermal efficiency of NPP has not been significantly improved and is still about 35%. The main reason for the low thermal efficiency is the low operating pressure of the steam cycle. A significant higher thermal efficiency can only be achieved by increasing the operating pressure beyond the critical value (22.1 MPa). Figure 5 shows the sketch of a SCWR. It is similar to a coal-fired power plant, just replacing the boiler by a reactor. Compared to a conventional PWR, an SCWR is a one-loop system; this eliminates the need for a steam generator and pressurizer and simplifies the system structure. An SCWR operates at a pressure of about 25 MPa, higher than the critical pressure of water (22.1 MPa). Water exiting the reactor has a temperature higher than 500 °C and goes directly to the turbine. Due to the low density of water at high temperatures, the SCWR has

Fig. 2 Examples of fuel rods arrangement in fuel assemblies (a) HCPER [5]; (b) PWR

144 4.1

Existing reactor core designs

Extensive R&D activities, including conceptual design, feasibility study and basic technology development, have been carried out since the 1990s [10]. Several pre-conceptual designs of the SCWR reactor cores have been proposed in the open literature. Some of their parameters are summarized in Table 1. Table 1 Parameters of some existing SCWR core designs

Fig. 3 Relative fuel utilization in dependence on moderatorto-fuel mass ratio [6]

Fig. 4 Sketch of pressure vessel and core of PWR and HCPWR [5]

Fig. 5 Sketch of an SCWR [9]

a good possibility to be designed with a much faster neutron spectrum (in the present paper, such spectrum is called “fast spectrum”, although it is more epi-thermal).

Country Spectrum El. Power/MW qp/(kW · m−1) Efficiency/% Pressure/MPa Tinlet/°C Texit/°C Core H/m Core D/m Cladding No. of FA No. of FR FR D/mm Pitch/mm

[11]

[12]

[13]

[10]

[14]

Europe Thermal 1 000 24 44.0 25 280 508 4.2 ― SS 121 216 8.0 9.5

USA Thermal 1 600 19.2 44.8 25 280 500 4.3 3.9 Ni 145 300 10.2 11.2

Japan Thermal 1 217 18 44.4 25 280 530 4.2 3.7 Ni 121 300 10.2 11.2

Japan Fast 1 728 ― 44.4 25 280 526 3.2 3.3 ― 419 ― 7.60 8.66

China Mixed 1 500 16 44.0 25 280 510 4.0, 2.0 3.5 TBD 269 256, 324 8.0 9.6, 10.7

Most of the concepts are based on thermal spectrum. Efforts were also made to design SCWR reactor cores with fast neutron spectrum. In the fast spectrum design of Oka [10], fuel pins are tightly arranged inside fuel assemblies in a hexagonal lattice, similar to that in an HCPWR indicated in Fig. 2. Compared to the fast core design, the fuel assembly structure of a thermal reactor core is much more complicated. This is due to the introduction of an additional moderator into the fuel assemblies. This also results in a large number of design options proposed in the open literature, which can be divided into two classes, i.e. PWR-type and BWR-type. The PWR-type fuel assembly stems from PWR similar fuel assembly and has a similar size as a PWR fuel assembly, as shown in Fig. 6. The flow channels inside the fuel assembly are divided into two groups, i.e. coolant channels and moderator channels. Inside the moderator channels high density (low temperature) water flows. Usually, count-current flow mode is taken (see Fig. 6(c)), to achieve high water temperature at the core exit. Water entering the pressure vessel flows into two paths. One part goes through the down-comer to the lower plenum, whereas the other part flows upward to the upper part of the pressure vessel, enters the moderator channels and flows downwards to the lower plenum, where it merges with the water from the down-comer and flows upwards through the cooling channels. Compared to the PWR-type fuel assembly, the BWR-type fuel assembly is small. Due to insufficient moderation capability inside the BWR-type fuel assembly, the additional water gaps between fuel assemblies are required. In the fuel

145

Fig. 6 Example of thermal SCWR fuel assemblies (a) PWR type [15]; (b) BWR type [16]; (c) Count-current flow mode in pressure vessel

assembly design [16], two groups of moderator channels are identified, i.e. moderator channels inside FA and assembly gaps. Water flowing through both the moderator channels and the assembly gaps should exit at the lower plenum of the vessel and then flow through the cooling sub-channels, to ensure a high exit temperature of the coolant. This would yield an extremely challenging task in the mechanical design of the reactor core. Reactors with thermal spectrum have advantages in using lower enrichment and have a high inventory of water in the reactor core. This is important for the dynamic behavior in the case of transient scenarios. However, an SCWR reactor with thermal spectrum requires additional moderator channels. The count-current flow mode inside the moderator channel and the cooling channel yields an extremely challenging design task. Moreover, reactors with thermal spectrum are unfavorable with respect to the fuel utilization and resource sustainability. An SCWR with a fast spectrum will enhance fuel utilization and resource sustainability, and simplify the mechanical

design of the fuel assemblies and the reactor core. A fast spectrum also provides the possibility of nuclear waste transmutation. However, reactors with fast spectrum use UO2 and PuO2 mixed fuel (MOX fuel) with high enrichments. This requires a new fuel cycle technology. Moreover, an SCWR with fast spectrum has less water inventory, which will have negative impact on the dynamic behavior in the case of transient scenarios. To combine the major merits and to minimize shortages of both options, the author proposes an SCWR reactor core with a mixed neutron spectrum (mixed core). 4.2

Mixed reactor core

Figure 7 shows schematically the geometrical arrangement of the proposed mixed core. The basic idea is to divide the reactor core into two zones with different neutron spectra. In one zone (e.g. the outer zone in Fig. 7, it could also be located in the inner zone) the neutron energy spectrum is similar to that of a thermal reactor. In this zone the fuel assembly has

Fig. 7 Scheme of the mixed SCWR core (a) Flow paths in core; (b) Fuel assembly arrangement in core

146 a PWR-type structure, but with the co-current flow mode. The cold water entering the pressure vessel goes upward to the upper dome and into both the moderator channels and the cooling channels of the thermal zone. It exits the thermal zone in the lower plenum; from there it enters the fast zone of the reactor core (e.g. the inner zone in Fig. 7). Table 2 summarizes some main parameters of the proposed mixed core. Table 2 Parameters of the mixed core Thermal zone Fast zone Entire core Thermal power/MW 2 350 Inlet temperature/°C 280 Outlet temperature/°C 400 Active height/m 4.0 FA box size/mm 173.2 No. of fuel assemblies 180 No. of fuel pins 180 Fuel pin diameter/mm 8.0 Pitch-to-diameter ratio 1.20 190 Ave. linear power/(W · cm−1) Power density/(MW · m−3) 114 Relative moderation capacity 1.53 Outer diameter/m 3.30 Mass flux/(kg · m−2 · s−1) 922 Maximum fluid velocity/(m · s−1) 5.5 Pressure drop/kPa 25.0 Maximum coolant temperature/°C 550.5 Maximum cladding temperature/°C 588.4 Fuel UO2 or MOX Enrichment/% 5–6

1 150 400 510 2.0 137.2 100 289 8.0 1.27 190 92 0.15 2.0 1 145 13.1 98.0 526.9 616.7 MOX ∼20

3 500 280 510 ― ― 280 ― ― ― ― 102 ― 3.30 ― ― 123.0 550.5 616.7 ― ―

The water temperature at the inlet of the pressure vessel is 280°C and at the pressure vessel exit is 510 °C. It is assumed that the water through the thermal zone is heated up to 400 °C. The average linear power rate is 190 W/cm for both the thermal zone and the fast zone. The active height is 4 m of the thermal zone and 2 m for the fast zone. There is a one meter blanket (breeding material) in both the lower part and the upper part of the fuel rods in the fast zone. In the thermal zone, 30% of the total mass flow rate goes through the

moderator channels. To make both zones geometrically compatible, the fuel assemblies of both thermal zone and fast zone are arranged in a square lattice. The size of the fuel assembly box is the same for both zones. However, the fuel pin size or the pitch-to-diameter ratio of both zones is different from each other. In the thermal zone, the PWR-type fuel assemblies are applied, as indicated in Fig. 8(a). Two rows of fuel pins are arranged between the moderator channels. Each moderator channel takes the position of 4x4 fuel pins. Inside each fuel assembly 3x3 moderator channels are placed. This gives a total number of fuel pins of 180. The fuel pin has a diameter of 8.0 mm and the pitch-to-diameter ratio is 1.20. The distance between the fuel pin and the moderator channel, and between the fuel pins and the fuel assembly box is 1.0 mm respectively. This gives the fuel assembly box a span distance of 173.2 mm. The square-shaped fuel assembly of the fast zone shall have the same box size. The fuel assembly of the fast zone has the same structure as that of a conventional PWR. The diameter of the fuel pin is the same as that in the thermal zone (8.0 mm). A larger pitch-to-diameter ratio of 1.27 is taken to reduce the circumferential non-uniformity of local heat transfer [8]. In each fuel assembly 17x17 fuels pins are arranged, as indicated in Fig. 8(b). The uranium oxide or MOX of low enrichment is used in the thermal zone, whereas in the fast zone MOX fuel is applied with an enrichment of about 20%. The plutonium composition could be similar to that of the discharged fuel from PWR. For the performance assessment of the thermal zone, coupled neutron-physical and thermal-hydraulic analysis is carried out [17]. This coupled method combines the SKETCHN code for neutronic analysis and COBRA code for thermalhydraulic calculations. Due to the good homogeneous arrangement in the fast zone, the single channel analysis method is applied by assuming a hot channel factor of 1.10. The axial power distribution is taken from Ref. [17]. Some of the results are summarized in Table 2. The effective core diameter of the proposed mixed core is 3.5 m, which gives an average power density of 90 MW/m3. The inner fast zone has an effective diameter of 2.0 m, which makes 37% of

Fig. 8 Fuel assembly structures of the thermal and the fast zone (a) FA in thermal zone; (b) FA in fast zone

147 the entire reactor core area. The relative moderation capacity in the thermal zone is higher than that of a conventional PWR, whereas the moderation capacity in the fast zone is about oneseventh of a PWR, still 50% lower than that of an HCPWR. Therefore, the conversion ratio close to 1 is expected in the fast zone. Due to the comparable flow area, the mass flux in the fast zone is similar to that in the thermal zone. The maximum velocity in the fast zone is about 13 m/s. This should still be well acceptable in relation to pressure drop and erosion issues. The maximum cladding temperature in the thermal zone is 588 °C, lower than that in the fast zone (617 °C). However, both values are still well below the design limit of 650 °C, which is well accepted by the international community. This is one of the key criteria for the SCWR design. Table 3 summarizes some calculated parameters of the two-row type FA, which is proposed by the author and is indicated in Fig. 8(a). For comparison the results obtained for the one-row type FA (Fig. 6(a)) are also listed in the table. Table 3 Some calculated parameters for both fuel assemblies [17]

Radial power factor Axial power factor Hot channel factor Max. cladding temperature/°C Max. fuel temperature/°C Thermal neutron fraction/% Fuel Doppler temperature reactivity coefficients/(10−5 · K−1) Moderator temperature reactivity coefficient/(10−5 · K−1) Coolant temperature reactivity coefficient/(10−5 · K−1)

One-row FA

Tow-row FA

1.135 1.740 1.137 684.4 1 890.2 8.7 −1.85

1.086 1.940 1.053 588.4 1 558.5 14.4 −1.78

−54.3

−36.2

−20.1

−29.3

A more uniform radial power distribution is obtained in the two-row FA. In the one-row FA each interior fuel rod is surrounded by two moderator channels, whereas the fuel rod near the assembly wall has only one moderator channel nearby. This gives stronger moderation, and subsequently higher power density in the interior region. Thus, it gives a strong non-uniform distribution of the radial power density. The power density difference between rods could be as high as 40% [17], and the radial power factor is 1.14. In the tworow FA each fuel rod is facing only one moderator channel. Thus, a more uniform moderation over the entire cross section is achieved. The radial power factor is 1.09. The smaller radial power factor is one of the key factors responsible for the smaller hot channel factor, i.e. more uniform distribution of coolant enthalpy rise in various cooling channels. In spite of the higher axial power factor, the maximum cladding temperature and the maximum fuel temperature in the two-row FA are much lower. Figure 9 compares the axial distribution of power and cladding temperature. The one-row FA design gives two power peaks in the lower and the upper part of the core, respectively. The two-row FA has only one peak in the lower part. Because of the effective heat transfer in the lower

Fig. 9 Axial distribution of power and cladding temperature [11]

part, the cladding temperature in the lower part is kept low, in spite of the high power density released in the fuel. In the upper part, the heat transfer from the cladding surface becomes worse due to low density and small heat capacity. The cladding temperature increases strongly. The second peak of the power in the one-row FA leads to a high peak of the cladding temperature in the upper part. The neutron spectrum in the two-row FA is softer. The thermal neutron fraction is defined as the fraction of the neutron flux with energy lower than 0.625 eV to the total neutron flux. This leads to a larger negative value of the temperature reactivity coefficient of the coolant. Due to the smaller size of the moderator channel, a more effective utilization of the moderator is expected in the one-row fuel assembly. Therefore, the temperature reactivity coefficient of the moderator in the one-row FA is more negative than that in the two-row FA. 4.3

Basic R&D requirements for SCWR development

In addition to the reactor core and FA design, as mentioned above, some other key technical challenges are briefly summarized below, which need extensive study. (1) Cladding and structural materials: the structure materials, especially the cladding material, used in PWR cannot be applied to SCWR any more due to the high operating conditions. Cladding material candidates proposed up to now are either austenitic stainless steels or Ni-base super-alloys, whereas ferritic-martensitic steels are envisaged for outof-core components. Both kinds of stainless steels have already been used in boilers of fossil-fired power plants, without irradiation, or as claddings of fast test reactors, without being exposed to the aggressive supercritical water environment. A high corrosion rate is expected for stainless steels. The Ni-base super-alloys, however, show high neutron capture and possibly serious swelling problems under irradiation conditions. (2) Thermal-hydraulics: there is still a large deficiency in understanding the thermal-hydraulic phenomena and in corresponding test data in the FA geometry under SCWR

148 conditions. The uncertainty is even more pronounced, in case the pressure is close to the critical value. Some of the main phenomena requiring further studies are heat transfer, critical flow and flow stability. (3) Design analysis computer codes: existing computer codes require improvements to cover the needs of SCWR design studies. With respect to thermal-hydraulic analyses, the transition between supercritical and sub-critical region has to be taken into account. This requires the extension of models and numerical methods used in the present codes. In addition, large density variation of water in the reactor core requires the coupling approach, which combines the thermalhydraulic codes with the neutronic codes. Both the separate phenomena codes and the coupling codes need experimental validation.

5

Conclusions

Nuclear power is one of the key energy sources for power generation. The ambitious nuclear program motivates the Chinese nuclear community to pay attention not only to the existing and the short-term deployable nuclear power plants, but also to the medium-term and long-term reactor program, in order to ensure long-term stable and sustainable development of nuclear power. Form the technologic point of view, the Chinese nuclear power would surely keep water-cooled reactors as the main reactor types, even in the far future. However, up to now less attention has been paid in China to the future NPP with advanced water-cooled reactors beyond generation III. The paper discusses some criteria for the selection of the future water-cooled reactors by considering the specific Chinese situation. Based on the selection criteria suggested, i.e. safety, economics, fuel resource sustainability, and technology availability, two new types of water-cooled reactors are recommended for the future of Chinese nuclear power generation. They are HCPWR and SCWR. High conversion pressurized water reactor utilizes the present PWR technology to the largest extent. The system structure and the components used in a PWR nuclear power plant are kept unchanged, except the reactor core arrangement. With a tight hexagonal lattice, a moderator-to-fuel mass ratio of 0.04, and subsequently a high conversion ratio of about 0.95 can be realized. This improves the fuel utilization by about 5 times. Prior to the engineering realization, further studies are required on the following topics: core optimization considering the neutron-physical and thermal-hydraulic performance; detailed thermal-hydraulic behavior in tight latticed FA, e.g. critical heat flux and local heat transfer non-uniformity; dynamic behavior of the coolant system in the case of transient scenario, e.g. LOCA; and the related MOX fuel issues, including the fuel cycle technology. The supercritical water-cooled reactor is the only one reactor type cooled by water among the six reactor concepts recommended by the GEN-IV International Forum. It should

be the logical extension of the generation III PWR technology in China. In spite of the significant advantages of SCWR, there exists a large number of challenging tasks, especially in the FA design, and development of advanced structural materials, numerical analysis tools, and other basic technologies. According to the roadmap proposed by the international community the engineering design of a SCWR will be available by 2030 or earlier. The author proposes a new SCWR core design consisting of two zones, one with thermal spectrum and the other with fast (or epi-thermal) spectrum. Compared to a thermal SCWR core, the mixed core simplifies the FA design and enhances the fuel utilization. Compared to the fast SCWR, it would improve the safety behavior. The FA structure (two-row type FA) proposed in this paper differs from those suggested by the international researchers (one-row type FA). The preliminary analysis using a coupled neutron-physics/thermal-hydraulics approach shows clearly a more favorable performance of the two-row type FA. It gives a much more uniform radial power distribution, lower hot-channel factor, lower cladding temperature and lower fuel temperature. The neutron spectrum is softer, which leads to a lower enrichment required and more negative temperature reactivity coefficient of coolant. Based on the summarized information and preliminary analysis, the author strongly recommends both HCPWR and SCWR as the future water-cooled reactor beyond GEN-III in China. Corresponding R&D activities need to be started as soon as possible, to ensure a smooth transition from the GENIII PWR technology to the water-cooled reactor technology beyond GEN-III.

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