Progress in Nuclear Energy, Vol. 32, No. 314, pp. 547-554, 1998 8 1997 Published by Elsevier Science Ltd

Printed in Great Britain 0149-1970/98 $19.00 + 0.00

PII: SO149-1970(97)00042-5

SUPERCRITICAL-PRESSURE, LIGHT-WATER-COOLED REACTORS

FOR ECONOMICAL NUCLEAR POWER PLANTS

S. KOSHIZUKA and Y. OKA

Nuclear Engineering Research Laboratory Faculty of Engineering, The University of Tokyo

2-22 Shirane, Shirakata, Tokai, Naka, Ibaraki 3 19- 11, Japan

ABSTRACT

Design studies of supercritical-pressure light-water-cooled reactors (SCLWRs) have been carried out to pursue drastic improvement of the economy of nuclear power generation. The core is cooled by supercritical water which is superheated without the phase change. The cooling system is a once-through type; the whole core flow is driven by the feedwater pumps and is directly led to the turbine. No recirculation line is necessary. Besides, steam separators and dryers are not needed. Water rods are used to enhance the moderation and to increase the flow velocity around the fuel rods. The radial peaking factor is satisfactorily reduced by controlling uranium enrichment and gadolinia concentration as well as water rods. Flattening of the radial power distribution is important to enhance the thermal efficiency. This can be achieved by the coolant density feedback and the out-in refueling pattern. Orificing is also effective to enhance the thermal efficiency. The thermal efficiency is above 40% with stainless steel cladding. Plant control system and safety system are also designed. The core flow should be directly maintained due to the once-through direct cycle. Plant behaviors of large break LOCAs and loss of offsite power are analyzed. Safety criteria are satisfied in both cases. The feasibility of SCLWR is shown.

Q 1997 Published by Elsevier Science Ltd

INTRODUCTION

A once-through direct cycle is employed in supercritical-pressure light-water-cooled reactors. The once- through direct cycle is widely used in current fossil-fired power plants where the feedwater driven by feedwater pumps flows through the boiler to the turbine without recirculation (Fig. 1). Employing this cycle, the nuclear power plant is markedly simplified; recirculation lines, steam separators and dryers are not

(a) BWR (b) supercritical fossil-fired plant

Fig. 1 Comparison of plant systems

(c) SCLWR

548 S. Koshizuka and Y. Oka

necessary. Besides, both pressure and temperature can be enhanced at the core outlet so that the thermal efficiency increases.

Boiling disappears at supercritical pressure. The critical pressure of water is 22.1MPa. Figure 2 shows the properties of water at 25.OMPa. Density continuously decreases with temperature increase. Specific heat exhibits a peak where density change is large. The temperature at the peak is called pseudo-critical temperature, which appears along the extrapolation of boiling temperature. The water whose temperature is above the pseudo-critical Temperature [“Cl

temperature is called as steam. Fig.2 Properties of supercritical water (25MPa)

Since boiling disappears at supercritical pressure, boiling transition does not occur. Accordingly, superheated steam can be obtained without dryout in the core. However, it is known that heat transfer is deteriorated in supercritical water cooling at high heat fluxes. The heat transfer deterioration is attributed to mixed convection between forced and natural convection. The heat transfer coefficient decreases when the heat transfer deterioration occurs, though this is much milder than dryout in subcritical water cooling.

Design studies of supercritical water cooled reactors are summarized by Oka et al. (1992), Oka and Koshizuka (1993), Oka et al. (1995a) and Oka and Koshizuka (1996). Thermal reactors (SCLWR) are described by Okano et al. (1994), Okano et al. (1996a) and Tanaka et al. (1996). Fast breeder reactors (SCFBR) are described by Jevremovic et al. (1994), Oka et al. (1995b) and Jevremovic et al. (1996). Safety analyses are provided by Koshizuka et al. (1994), Lee et al. (1996), Okano et al. (1996b), Okano et al. (1996~) and Kitoh et al. (1996). Numerical analysis of the heat transfer deterioration is presented by Koshizuka et al. (1995). Recent developments of these studies are summarized in this paper.

CORE DESIGN

Thermal limits

Three thermal limits are imposed on the design of SCLWR as follows. ( 1) Maximum linear heat rate: The maximum linear heat rate is 40kW/m with enriched UO7 oellets. Release Of’FP gas from the pellets is the same level as the current LWR’s.

-a

(2) Maximum cladding surface temperature: The material of cladding is stainless steel whose surface temperature is required below 450°C. Oxidation of the cladding is expected to be small enough with this criterion, which is much more conservative than the current fossil-fired power plants. (3) Minimum deterioration heat flux (MDHFR) To avoid the heat transfer deterioration in supercritical water cooling, the minimum deterioration heat flux ratio (MDHFR) is required above 1.3 in the steady state operation condition:

MDHFR=& , 9Lx

(1)

q;ier : deterioration heat flux [kW/m*],

4 & : maximum heat flux at the deterioration point [kW/m*]. This criterion corresponds to MCPR in BWRs and the minimum DNBR in PWRs. Correctly, the MDHFR criterion is to be applied to transient analysis, which is shown later. MDHFR is temporally used as a design criterion in this study to simplify the conceptual design process.

Yamagata’s correlation (Yamagata et al., 1972) is used for evaluating the deterioration heat flux:

q;ier = 0.2G’.* , (2) G : mass flowrate [kg/mz].

According to Eq.(2), the deterioration heat flux is in proportion to Cl.*, which means that the deterioration heat flux is sensitive to the mass flow rate. Therefore, the mass flowrate should be kept high enough. The transients accompanied by the decrease of the core flowrate may be severe in SCLWR.

Economical nuclear power plants 549

The heat transfer deterioration, by which the heat transfer coefficient decreases gradually, is much milder than the dryout. The heat transfer coefficient can be estimated numerically based on single-phase flow simulation with a turbulence model (Koshizuka et al., 1995). Thus, the MDHFR criterion can be replaced by another rational criterion which guarantees the integrity of the cladding.

Core confirmration

A fuel assembly of SCLWR with single tube water rods (SCLWR-SWR) is depicted in Fig.3. Seven fuel pins are replaced by a water rod. Part of the water rods contain a guide tube, in which a control rod is inserted. This enhances the reactivity worth due to the water surrounding the control rod. The diameter of the control rod can be larger than that inserted in the fuel rod space. Since SCLWR employs the once-through direct cycle, the core coolant flowrate is smaller than the recirculation type. Thus, a tighter lattice is better to enhance the flow speed around the fuel rods. Water rods are used to enhance the moderation in the tight lattice core. The flow speed in the water rods is slower than that around the fuel rods.

The horizontal cross section of the SCLWR-SWR core is shown in Fig.4. Fresh assemblies are located in the peripheral region to flatten the radial power distribution. The vertical cross section of the reactor pressure vessel is depicted in Fig.5. Since neither steam separators nor dryers are necessary, control rods are inserted from the top of the core. Thus, the. configuration in the pressure vessel-is similar to PWR’s. The coolant lines are onlv two because of the small

Fig.3 Fuel assembly

@ 1st cycle @ 2nd cycle 0 3rd cycle

Fig.4 Cross section of the core

SCLWR with double tube water rods (SCLWR-DWR) has been also designed. In the double tube water rods, the coolant flows upward in the inner tube to the core top, turns down in the outer annular tube, and enters to the fuel channels at the core bottom.

Power distributions

Subcooled water is heated to superheated steam in the core by employing the once-through direct cycle. The coolant density axially changes more than BWR’s as shown in Fig.6. In addition to the water rods, uranium

550 S. Koshizuka and Y. Oka

enrichment and gadolinia concentration are axially divided to three zones to flatten the axial power distribution. Figure 7 shows the axial power distribution calculated in the present study at thle hottest channel. The cosine shape is almost kept in the beginning of cycle (BOC), middle of cycle (MOC) and end of cycle (BOC). The thermal-hydraulic calculation is performed by using the code developed by the authors’ group. The nuclear calculation is performed by using SRAC developed in the Japan Atomic Energy Research Institute (Tsuchihashi et al., 1986).

0

SCLWR

Fig.6 Axial distribution of coolant density

0 1 2 3 ‘I Height [m]

Fig.7 Axial power distribution

In the once-through direct cycle, the whole coolant is mixed at the top of the core. In order to keep the average outlet temperature high, the flowrates in low power assemblies are limited by orificing. Thge core is axially divided into four orifice regions as shown in Fig% The radial power distribution is analyzed with considering the feedback between the coolant density and power distributions. A code is developed by combining the thermal- hydraulic and nuclear calculation codes. The calculation result is shown in Fig.9. Figure 10 provides the core outlet temperature averaged in each orifice region. The outlet temperature in the peripheral region is low to satisfy the thermal limits. The total core average outlet temperature is 397”C, which is only 12’C lower than the hottest channel. The temperature decrease is not so large because the coolant density feedback effectively flattens the radial power distribution.

If the orificing is not used, the core average outlet temperature decreases 3°C more. Nevertheless, the total electric power increases when the channel boxes are replaced by fuel rods. The cross flow must be evaluated to assess whether the channel boxes CiUl be really removed.

I --__ Region 1 ,/’ Region 2,~~Region 31 Region 4 /

Fig.8 Orifice regions

0 0 50 100 150

Radius[cm]

Fig.9 Radial power distribution

5 410 e e 405 a 2 400

B E 395 2 d 390 4 0 385 _ _- _......... _-.._.: e

6 380 I 2 3 ‘I

Radial region

Fig. 10 Core outlet temperatures from orifice regions

Economical nuclear power plants 551

PLANT DESIGN

A steam cycle is designed for SCLWR-SWR. It is a three stage turbine system with 8 feedwater heaters like supercritical fossil-fired power plants. Main steam is used for reheating like current LWRs. Figure 11 shows the designed steam cycle with the SCLWR-SWR steam conditions, 25MPa and 397°C. The thermal efficiency is 40.7%. Compared with the fossil-fired power plants, nuclear plants have higher efficiency with the same steam conditions. This is due to the exhaust loss of the boiler in the fossil-fired power plants.

Principal characteristics of the designed SCLWR-SWR are provided in Table 1. SCLWR-DWR and SCLWR with zirconium hydride rods are also involved. Although the atomic density of hydrogen is the highest with the zirconium hydride rods, uranium enrichment is the highest since the neutron absorption of zirconium cannot be neglected. The core with double tube water rods, whose structure is more complicated, has higher core outlet coolant temperature and lower uranium enrichment due to higher coolant density in the water rods.

moderation improvement thermal I electric power [MWI thermal ef!iciency [%I pressure [MPal

Fig. 11 Steam cycle

Table 1 Principal characteristics of SCLWRs

SCLWR-SWR SCLWR-DWR SCLWR (%i) single rube water mds double tube wnler rods zirconium hydride rods single tube w~ter rods

^ . ..*. . ,,I., 19C&,l IQ4 ,<<“,ln46 3.293 / I.100 ‘“YUJ~;‘” 1, &,O_l”I 41.5 L,.“.s _,___. 41.0 ..- .-

25.0 25.0 25.0 33.4 6.9

fuJgd=W 0.80SS0.95

ss Zircaloy fuel rod diameter1 pitch km1 0.80 / 0.95 o.*os/s.95 I.231 1.47

number of fuel reds I gadolinis rods 258 I30 ‘:::P

735’:;;) 62218

number of water rods I conlral rods 3019 2l-

upper I middle / lower *35U enrichment tw!%l 6.41 15.22 14.66 5.65 15.30 14.55 7.2 16.2 I 5.5 3.4 (average)

upper I middle I lower gadolinia concentration [wt%l 2.1 13.1 14.3 3.0 / 4.0 I5.0 2.0 13.0 14.1 3.0 or 4.0 / 4.0

number of fuel assembly 163 I87 764

discharge bumup IGWtl 45

253.,&l.3*l

:z 39.5

~~~~ _. maximum / average power density [MWlm’l

^,_ _. .^,.I 103.31 Luo ’ 299 INS.6 l20/50

coolant mass flow rate [kg/secl 2.314 1,943 Il.640

coolant inlet I oudet tempemture [“Cl 324 I397 sod I408 3201417 279 I286 3.71 14.75 core height /core diameter [ml *r.n,*04 ,.,“I &o-l 3.70 13.12 3.70/3.14

axial / radial I lccal peaking f*ClOr l.56/ 1.26/ I.16 I.51 I l.30/ I.19 1.51/1.981-

*I The total peaking factor 2.5 is assumed.

552

Plant control svstem

S. Koshizuka and Y. Oka

The turbine-following-reactor operation is employed for SCLWR. The plant control system consists of the following three components (Fig. 12). (1) The reactor pressure is controlled by the turbine control valves. (2) The reactivity is controlled by the control rods to keep the core power constant. (3) The feedwater flowrate is controlled by the main feedwater pumps to keep the core outlet temperature constant. Both pressure and temperature should be controlled in SCLWR, though temperature is not independent in a saturation system like BWRs. On the other hand, water level should be. controlled in BWRs.

Auxiliarv and safetv systems

The safety system of SCLWR is required to maintain the core flow directly, since the plant system is the once-through direct cycle. This is different from the current LWRs in which natural circulation is expected in the recirculation lines or the primary loops. To keep the core flow in SCLWR, the following two conditions are required: (1) the core inlet flow supplied and (2) the core outlet open. To backup the first requirement, a turbine-driven high-pressure auxiliary feedwater system (AFS) and a low-pressure core injection system (LPCI) are provided (Fig.13). The main feedwater pumps are designed to be used as AFS; the main feedwater pumps are equipped with an auxiliary turbine for main steam and an auxiliary coolant supply line from the condensate water storage tank. Only turbine-driven pumps are used as the high-.pressure. safety system because fast start-up without electric power is necessary to keep the core flow in initial transients. Long term feedwater supply is not expected at high pressure because the system should be depressurized when the main steam lines are filled with water. Emergency diesel power is much reduced without a high-pressure motor-driven safety system. To backup the second requirement, turbine bypass valves and safety relief valves (SRV) are provided. SRV has a function of automatic depressurization system (ADS) whose steam outlet is the suppression chamber pool in the containment vessel. Measiurement of the feedwater flowrate is linked to actuation levels of the feedwater supply systems. Measurement of the core pressure is linked to actuation levels of the outlet valves. The relations between the requirements and the measurements are direct, which is good for the operators’ understanding of the plant behavior. AFS does not need to be a “safety system” because the plant safety is guaranteed by LPCI+ADS.

VW Main steam isolatiq valves

6 Main steam Turbine , stop valves control valves I

2 * / Turbine bypass valves

Main steam /

HP 3

heaters &

A w u I

Main feedwater pumps

Fig. 12 Plant control system

Economical nuclear power plants 553

accumulators

main steam contml valves

safety relief valves turbine bypass valves

Fig .13 Safety system

SAFETY ANALYSIS

A supercritical-pressure light water reactor LOCA analysis code (SCRELA) is developed (Lee et al., 1996). SCRELA consists of blowdown and reflood estimation modules. The homogeneous equilibrium model is employed for the two-phase flow. In the reflood estimation module, relative velocity between water and steam is considered.

Large break LOCAs at the cold-leg of SCLWR-DWR are analyzed with LPCI and ADS. Two lines of LPCI are assumed to be available with 30 second time delay considering the diesel start-up. The total capacity is 1,6 lOkg/sec. ADS is assumed to actuate at 30sec. Accumulators are not considered in this analysis, though the capacity of LPCI can be reduced by involving them. Figure 14 shows the behaviors of cladding surface temperatures. The maximum temperature, about l,lOO”C, appears at the core middle in the case of 70% cold-leg break; this is lower than the limit of stainless steel, 1260°C, which was obtained by USNRC (Coffman, Jr., 1976). The peak cladding temperature is much lower in the cases of hot-leg break since the core flow is accelerated due to the blowdown.

Loss of offsite vower

o-----L--“.““““““““““” 0 50 100 150 200 250 300

Time (SIX)

Fig. 14 Large break LOCA analysis

Loss of offsite power, which is an anticipated transient, is analyzed in SCLWR-DWR with AFS. In SCLWR, loss of feedwater directly causes loss of core flow. The SCLWR plant transient analysis code is developed (Okano et al., 1996b; Okano et al., 1996~; Kitoh et al., 1996). The reactor scram signal is assumed to release at the same time as the transient starts, and the scram is completed from lsec to 4sec. The turbine-driven AFS is assumed to start to supply 320kg/sec water with 3sec time delay. The main feedwater pumps coast-down starts at 6sec. The coast-down time is assumed 10sec. Fly-wheels are necessary to realize this coast-down time.

554 S. Koshizuka and Y. Oka

The result is shown in Fig. 15. MDHFR decreases from 1.6 at Osec to 1.25 at 13sec because the deterioration heat flux decreases due to the coast- down of the core flowrate. The criterion MDHFbl.0 at anticipated transients is satisfied. If the MDHFR criterion is replaced by another rational criterion, the fly-wheels are not necessary any more.

CONCLUSIONS

The present study shows the feasibility of suuercritical-uressure light-water-cooled reactors (SCLWR) through the core design, plant design and safetv analvses. The current LWR svstems have bee; little’changed for many years,* while fossil-fired power plants have been developed steadily, for example, by improving the steam conditions (Fie.16). The thermal efficiencv is approaching‘fiO% in the advanced combined cycle (ACC). Cost competitiveness is the first requirement for the commercial power plants. The concept of SCLWR is aiming drastic cost reduction by employing the simple once-through direct cycle. Breeding is also possible by tightening the core lattice with the same plant system.

REFERENCES

Coffman, Jr. F. D. (1976). NUREG-006.5 Jevremovic T., Oka Y. and Koshizuka S. (1994).

Nucl. Technol. 108,24 Jevremovic T., Oka Y. and Koshizuka S.

Fig. 15 Loss of external power analysis

Km-

50° - 1952~Bonus?~6~ (FBR)

400 %lk River (1962) ~Palhfinder (1963) SCLWbR, SCFBR

Tokai (GCR)

Kashiwazaki-Kariwa-6 (ABWR)

loo -

0 IO 20 30 40

Pressure [MPa]

Fig. I6 Steam conditions of fossil-fired and nuclear power plants

(1996). Nucl. Technol 114,273 Kitoh K., Koshizuka S. and Oka Y.. (1996). Proc. 4th Int. Conf. Nucl. Eng. (ICONE-4) Vo1.2, p.537 Koshizuka S., Shimamura K. and Oka Y. (1994). Ann. Nucl. Enea, 177 Koshizuka S., Takano N. and Oka Y. (1995). Int Heat Mass Transfer 38,3077 Lee J. H., Koshizuka S. and Oka Y. (1996). l?roc:ith Int. Conf. Nucl. Eng. (ICONE-4) Vol. l-B, p.533 Oka Y. and Jevremovic T. (1996). Ann. Nucl. Enerev 23, 1105 Oka Y. and Koshizuka S. (1993). Nucl. Technol. 103, 195 Oka Y. and Koshizuka S. (1996). Proc. 4th Int. Conf. Nucl E g. (ICONE-4) Vo1.2, p. 191 Oka Y., Koshizuka S. and Yamasaki T. (1992). J. Nucl. SC;. Tnechnol. 29,585 Oka Y., Koshizuka S., Jevremovic T. and Okano Y. (1995a). mess in Nucl. Energy_.Z% 43 1 Oka Y., Koshizuka S., Jevremovic T. and Okano Y. (1995b). Nucl. Technol. 109, 1 Okano Y., Koshizuka S. and Oka Y. (1994). AM. Nucl. Energy 21,601 Okano Y., Koshizuka S. and Oka Y. (1996a). J. Nucl. Sci. Technol. 33,365 Okano Y., Koshizuka S., Kitoh K. and Oka Y. (1996b). J. Nucl. Sci. Technol. 33,307 Okano Y., Koshizuka S. and Oka Y. (1996c). Proc. 4th Int. Conf Nucl. Eng (ICONE-4) Vol.l-B, p.‘77 1 Tanaka S., Shirai Y., Mori M., Takekuro I., Komano Y., Nunokawa K., Otonari J., Kataoka K., Kataoka Y

and Moriya K. (1996). Proc. 4th Int. Conf. Nucl. Ene. (ICONE-4) Vol.2 p. 199 Tsuchihashi K., Ishiguro Y., Kaneko K. and Ido M. (1986). JAERI-13Q2 Yarnagata K., Nishikawa K., Hasegawa S. and Yoshida S. (1972). Int. J. Heat Mass Transfer 15,2575