www.elsevier.com/locate/pnuceneProgress in Nuclear Energy 48 (2006) 559e568

Thermal hydraulic analysis of thethorium-based advanced CANDU Reactor

fuel channel

Jiyang Yu a,*, Wenlong Mao a, Baoshan Jia a,Yanfei Rao b, Yangqiang Ruan b

a Engineering Physics Department, Tsinghua University,

Beijing 100084, Chinab Atomic Energy of Canada Limited, 2251 Speakman Drive,

Mississauga, Ontario, Canada L5K 1B2

Abstract

This paper presents a thermal hydraulic analysis of the thorium-based advanced CANDU Reactor (TACR-1300) fuel channel.The paper first uses CATHENA, a system thermal hydraulics code for CANDU reactors, to determine the boundary conditions ofthe TACR-1300 fuel channel. The boundary conditions are then passed to ASSERT-PV, a subchannel thermal hydraulics code forthe CANDU industry, to obtain detailed thermal hydraulic information on the TACR-1300 fuel channel. Through the subchannelanalysis, it is found that the onset of dry-out power (ODP) for the TACR fresh fuel channel is much less than that of the ACR-700because thorium fuel rods as well as uranium rods are not heated and there is bypass-flow effect in the associated subchannels. Thepaper compares three thorium loading modes, and determines the best one for TACR.� 2006 Elsevier Ltd. All rights reserved.

Keywords: Thorium-based; Thermal hydraulic; Subchannel; Onset of dry-out power

1. Introduction

The proposed thorium-based advanced CANDU Reactor (TACR-1300) utilizing thorium resources is a national nu-clear energy system that is suitable for the resource situation in China and other thorium-rich regions (Jiyang et al., 2004).

TACR includes several important items as follows. TACR takes slightly enriched uranium (SEU) as driver and con-verts 232Th to 233U that is an effective fissile material in nuclear reactor. Similar to the ACR-700 (Hedges, 2002),TACR uses D2O at low pressure as the moderator and reflector and H2O at high pressure as the coolant. The useof light water as the coolant becomes feasible with the SEUethorium fuel system, and this will simplify or even elim-inate the high-pressure D2O systems used in CANDU-6 reactors. The use of H2O coolant can also increase the powerdensity and reduce the cost.

* Corresponding author.

E-mail address: yujiy@tsinghua.edu.cn (J. Yu).

0149-1970/$ - see front matter � 2006 Elsevier Ltd. All rights reserved.

doi:10.1016/j.pnucene.2006.02.002

560 J. Yu et al. / Progress in Nuclear Energy 48 (2006) 559e568

Some general characteristic parameters of the TACR-1300 conceptual design are shown in Table 1.

2. The fuel channel design of TACR

Based on the distribution of thorium rods in the fuel bundle, we consider three candidates, called modes,of fuel bundles, as listed in Tables 2 and 3. Fig. 1 shows the cross sections of TACR fuel bundle modes Aand C.

3. Background of thermal hydraulic analysis tools for TACR

The acronym CATHENA (Beuthe and Hanna, 2004) stands for Canadian Algorithm for THErmalhydraulic Net-work Analysis. The CATHENA code was developed by Atomic Energy of Canada (AECL) and is a two-fluid thermalhydraulic code. CATHENA uses a transient, one-dimensional two-fluid representation of two-phase flow in pipingnetworks. In the thermal hydraulic model, the liquid and vapor phases may have different pressures, velocities,and temperatures. In addition, up to four non-condensable gases may be included in the gas phase. The thermalhydraulic model consists of solving six partial differential equations for the conservation of mass, momentum andenergy for each phase. If non-condensable gases are included in a simulation, an additional mass conservation equa-tion for the mass fraction of each of the non-condensable gas components is solved simultaneously with the two-fluidmodel conservation equations.

ASSERT (Carver et al., 1995; Rao, 2004) (Advanced Solution of Subchannel Equations in Reactor Thermal-hydraulics) is a subchannel thermal hydraulics code developed at Chalk River Laboratories to model steady stateand transient single- and two-phase flow through rod bundles. ASSERT-PV is based on the advanced drift-flux(unequal velocity/unequal temperature) thermal hydraulic model and a drift-flux (relative velocity) correlation spe-cifically formulated to account for transverse flow prevalent in horizontal CANDU-type fuel channels (Carveret al., 1983).

4. Subchannel analysis model of TACR fuel channel

The fuel channel of TACR is similar to that of ACR-700 as shown in Fig. 2. There are 12 bundles in a fuelchannel and the dimensions are shown in Table 4. The relative axial heat flux distribution of TACR is shownin Fig. 3.

Table 1

TACR-1300 conceptual design characteristic parameters

Parameter Value Unit

Total electricity power 1300 MW

Thermal power 3584 MW

Enrichment of driver 235U 2.1 %

The active length of core 6 m

Number of bundles in one channel 12

Pitch of pressure tube 220 mm

Table 2

Fuel materials of different modes

Mode Ring 1 Ring 2 Ring 3 Ring 4

A ThO2 ThO2 SEU SEU

B SEU SEU ThO2 SEU

C ThO2 SEU 7ThO2þ 7SEU SEU

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Table 3

Configuration of the modes

Mode U rods Th rods ThO2 loading (%)

A 35 8 21.2

B 29 14 31.5

C 35 8 18.4

Fig. 1. Cross section of TACR fuel channel of Modes A and C.

Fig. 2. Fuel channel of TACR.

Table 4

Dimensions of TACR fuel channel

Items Value (cm)

Pressure tube (PT) inside diameter 10.338

Bundle diameter (including bearing pads) 10.250

Bundle total length 49.53

Bundle heated length 48.03

Whole channel length 594.36

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There are 70 subchannels and 43 rods in the subchannel analysis model as shown in Fig. 4. In Fig. 4, rods 1e8 arethorium rods and 9e43 are uranium rods.

5. Simulation results

The paper simulates two different cases to analyze thermal hydraulic phenomena in the fuel channel of differentmodes. The first one is at 6.65 MW channel power, and the second one is at 9.24 MW channel power. For both casesthe same boundary conditions shown in Table 5 are used.

Fig. 5 shows the void distribution in various subchannels of Mode A at 6.65 MW channel power. Figs. 6 and 7 showthose of Modes B and C, respectively. The ACR-700 case was also simulated for comparison of the result with those ofTACR; the void distribution is shown in Fig. 8.

We can see that the void-fraction differences between different subchannels of Mode A are much larger thanthose of the other two modes. The reason is that in Mode A, much more coolant is bypassed through the inner(ring 1 and ring 2) subchannels surrounded by unheated thorium rods. In Fig. 9 for Mode A at the channelpower of 6.65 MW, we can see that the subchannel 49 has the lowest coolant mass flux among the subchannels

Fig. 3. Relative axial heat flux.

Fig. 4. Subchannel analysis model of TACR fuel channel.

563J. Yu et al. / Progress in Nuclear Energy 48 (2006) 559e568

selected, and more so at the channel power of 9.24 MW, as shown in Fig. 10. The results for the othertwo modes and the ACR-700 are shown in Figs. 11e16. We can see that Mode C has a comparable perfor-mance to the ACR-700, but the thorium loading of Mode C is less than those of Modes A and B as shownin Table 3.

Besides void and mass flux distribution, the paper has also calculated the onset of dry-out power and dry-out locations of different modes and compared them with the ACR-700 case. The results are summarized inTable 6, where ODP is shown as the percentage change over that of ACR-700. We can see that the dry-outlocation is in Bundle 9 (from the channel inlet), and near the rods 34 and 35. Compared to ACR-700,the TACR fresh fuels (Modes AeC) have lowered ODPs. This is because the thorium fuel rods are notheated as much as the uranium rods and there is the bypass-flow effect in the associated subchannels.Among the three modes considered, Mode C is the best candidate in ODP point of view, and Mode A theworst.

6. Conclusions

Thermal hydraulic analysis has been performed of the proposed TACR fuels for three different distributions ofthorium rods in the fuel bundle (three fuel modes). The subchannel code ASSERT-PV was used to obtain thedetailed thermal hydraulic information on the TACR-1300 fuel channel. The results showed that the onset ofdry-out power (ODP) of TACR fresh fuels is considerably lower than that of ACR-700, with the best mode havinga CCP 10% lower than the ACR-700. This is considered to be due to the fact that the thorium fuel rods are notheated as much as the uranium rods and thus there exists the bypass-flow effect in the associated subchannels thatnegatively affects the thermal hydraulic performance. This result will be used in further study of the concept de-sign of TACR.

Table 5

Boundary conditions of subchannel analysis

Channel inlet mass flux 6.0 Mg/m2s

Channel outlet pressure 12.5 MPa

Channel inlet temperature 268 �C

Fig. 5. Void distribution in different subchannels. (Mode A at 6.65 MW channel power.)

564 J. Yu et al. / Progress in Nuclear Energy 48 (2006) 559e568

Fig. 6. Void distribution in different subchannels. (Mode B at 6.65 MW channel power.)

Fig. 7. Void distribution in different subchannels. (Mode C at 6.65 MW channel power.)

Fig. 8. Void distribution in different subchannels. (ACR-700 at 6.65 MW channel power.)

565J. Yu et al. / Progress in Nuclear Energy 48 (2006) 559e568

Fig. 9. Mass flux distribution in different subchannels. (Mode A at 6.65 MW channel power.)

Fig. 10. Mass flux distribution in different subchannels. (Mode A at 9.24 MW channel power.)

Fig. 11. Mass flux distribution in different subchannels. (Mode B at 6.65 MW channel power.)

566 J. Yu et al. / Progress in Nuclear Energy 48 (2006) 559e568

Fig. 12. Mass flux distribution in different subchannels. (Mode C at 6.65 MW channel power.)

Fig. 13. Mass flux distribution in different subchannels. (ACR-700 at 6.65 MW channel power.)

Fig. 14. Mass flux distribution in different subchannels. (Mode B at 9.24 MW channel power.)

567J. Yu et al. / Progress in Nuclear Energy 48 (2006) 559e568

Table 6

Onset of dry-out power (ODP) and dry-out position

Mode CCP pct diff. Dry-out position

Bundle Rod Subchannel

ACR e 10 34 36

A �23% 10 35 36

B �16% 10 34 61

C �10% 10 34 36

Fig. 15. Mass flux distribution in different subchannels. (Mode C at 9.24 MW channel power.)

Fig. 16. Mass flux distribution in different subchannels. (ACR-700 at 9.24 MW channel power.)

568 J. Yu et al. / Progress in Nuclear Energy 48 (2006) 559e568

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