Study on Optimization Design for CSR1000 Core

The supercritical water-cooled reactor (SCWR) is one of the generation IV reactor concepts. The economic performance of the SCWR has a core focus, and in this study forms one of the key design goals. A conceptual design of China supercritical water-cooled reactor with the rated electric power of 1000 MW_e (CSR1000) core was proposed with features such as single water rod, combined square fuel assembly (FA) with stainless steel fuel cladding and structure material, two flow-pass of coolant in core, high-leakage reload core with 157 FAs, cycle length of 350 effective full power days (EFPD) [1]. The refueling cycle of the CSR1000 design was about 12 months, which is much shorter than the generation III pressurized water reactor of 18 months. Thus, it is necessary to enhance the economic performance of CSR1000 by prolonging the refueling cycle.

Based on conceptual design of the CSR1000, the FA and the core design are optimized to improve the core performance, and the refueling cycle of CSR1000 is prolonged.

The SCWR FA design depends upon the core neutronics and thermal-hydraulics. The core structure and the feasibility of FA fabrication are also under consideration. A number of SCWR FA designs have been proposed internationally, including square FA with small water rod [2], small FA with single water rod [3], square FA with two row fuel rods [4], square FA with big water rod [1], etc. The square FA with big water rod [5] adopts the simple structure of combined square FA with single water rod, which is well considered for the performance of sufficient moderation and uniform moderation, and with respect to the simple split flow of the moderator and coolant. The square FA with big water rod shows good neutronics and thermal-hydraulics performance and structural implementation ability, and is chosen as the FA of CSR1000 conceptual design. Based on the square FA with big water rod design, this paper studies the FA and the core design optimization, particularly the method to prolong the refueling cycle and reduce the power peaking factor.

The square FA with big water rod consists of four subassemblies. The gap between fuel rods in subassembly is filled with coolant, the water rods in subassembly, and the channel between subassemblies is moderator. The fuel assembly configuration shown in Fig. 1 consists of a 9 × 9 fuel rod array arranged in a square lattice and a single moderator tube at the center displacing 25 fuel rods. The fuel rods have a diameter of 9.5 mm and a pitch of 10.5 mm, and two fuel enrichments were used to obtain a lower power peaking factor for the FA conceptual design. The enrichment of the fuel rods at the four corners is 4.4%, and the enrichment of the rest fuel rods is 5.8%, the average enrichment of the FA is 5.7%. Each fuel rod contains 0.6 wt % Er₂O₃ to suppress the reactivity at the beginning of life [6]. The average discharge burnup of FA is 40,000 MW d/t(U), which meet the requirement of 12 months refueling cycle of the core.

In order to get a longer refueling cycle (at least 18 months), the initial reactivity of FA has to be improved. Therefore, the average fuel enrichment has been increased to 7.8% based on the CSR1000 FA conceptual design, the enrichment of the fuel rods at the four corners, and that of the rest fuel rods has been increase to 5.6% and 7.97%, respectively, and all fuel rods contain 1.5 wt % Er₂O₃ to suppress the initial reactivity.

An improved design of CSR1000 FA is proposed in this study to flatten the power distribution of the FA, which decreases the nonuniform of the power distribution of the core. As shown in Fig. 2, there are three kinds of fuel rods with different enrichments for the optimization design. The enrichment of the fuel rods at the four corners is 5.6%, the two fuel rods next to the corner and the three fuel rods next to the water rod have an enrichment of 7.5%, and the rest fuel rods have an enrichment of 8.26%. The average enrichment of fuel is 7.8%, and all fuel rods contain 1.5 wt % Er₂O₃ to suppress the initial reactivity.

Neutronics performance of the previously mentioned three FA designs is studied, particularly the power peaking factor and the reactivity varying with the FA burnup. The evaluation conditions, which have been determined from the core average, are as follows: the density of the moderator and the coolant is 0.58 g/cm³ and 0.3 g/cm³, respectively, and the average fuel temperature is 830 °C without control rods. The k-inf and power peaking factor of the three FA design varies with burnup is given by Figs. 3 and 4. Figure 3 shows that the expected discharge burnup of the two FA with the average enrichment of 7.8% reaches to 60,000 MW d/t(U) from the viewpoint of core criticality, which is a theoretical burnup without considering the radiation effect and the mechanical performance. This burnup is much higher than that of the conceptual design with 5.7% average enrichment fuel, and satisfies the tentative target of 20 months or even longer refueling cycle. In addition, due to the addition of burnable poison Er₂O₃, the curve of reactivity variation is more flat than that of the conceptual design. Figure 4 shows that the peaking factor of the FA is decreased by using three region fuel rods with different enrichments. The maximum power peaking factor of FA is 1.04, which is much lower than that of the conceptual design (1.11), and the power peaking factor of FA is kept between 1.03 and 1.04 with the variation of burnup. The lower and more flat power peaking factor lays a foundation for the core optimization.

The CSR1000 core uses rod-type fuel, and the 310S stainless steel is chosen as the fuel cladding and core structure material. Currently, it is widely acknowledged that the core design criteria [7,8] are as follows: ① Under normal condition, maximum cladding surface temperature (MCST) ≤650 °C, maximum linear heat generation rate (MLHGR) ≤39 kW/m; ② When the control rod with the maximum rod worth stuck, the k-eff of the core is no bigger than 0.99. Based on these criteria and the FA mentioned earlier, this paper has studied the optimization core design for a longer refueling cycle.

The optimization core design consists of 157 FAs with assemblies of pitch size 239 mm, and the active core height is 4200 mm. The thermal power of the core is 2300 MW. The average enrichment of the FA is 7.8%, and there are three radial regions with different enrichments (5.6%, 7.5%, and 8.26%) as mentioned earlier. The axial of the core is divided into 20 layers for the axial core design. In order to flatten the power distribution, the axial partition of the burnable poison is proposed: the upper four layers have 1.0 wt % Er₂O₃, and the lower 16 layers have 1.5 wt % Er₂O₃, and the average content of Er₂O₃ is 1.4%. The two-pass coolant flow scheme is adopted [1]. The 57 fuel assemblies in center are the first flow path assemblies, in which the coolant flows from top to bottom. The other 100 fuel assemblies in periphery are the second flow path assemblies, in which the coolant flows from bottom to top.

The core average outlet coolant temperature is 500 °C, and the core mass flow rate is 1189 kg/s correspondingly. The design target is getting a higher and uniform outlet coolant temperature of the flow pass II. Based on the two-pass flow scheme, the flow scheme is optimized as shown in Fig. 5. 10% of the inlet water flows down the downcomer, 10% flows as the first flow pass moderator, 50% flows as the first flow pass coolant, and 30% flows as the second flow pass moderator. All of the coolant and moderator fully mixes in the downcomer, and then flows up as the second flow pass coolant.

The fuel management is optimized to obtain a lower leakage, flat power distribution, and a higher average discharge burnup. A low leakage fuel management scheme shown in Fig. 6 is adopted with three batch cycles: 52 FAs are fresh, 52 FAs have burned one cycle, 52 FAs have burned two cycles, and one FA in the center of the core has burned three cycles. Two-cycle burned FAs are loaded in core peripheral regions to decrease the leakage of neutrons, the rest two-cycle burned FAs and one-cycle burned FAs are loaded in core inner regions alternatively to decrease the power share of the core center. The fresh FAs are loaded in core middle regions to balance the neutron leakage and the power peaking factor.

In contrast to pressurized water reactor, the CSR1000 only uses control rods and burnable poisons to control its reactivity, and the control rod management is the key factor for reactivity control. The control rod management has to meet the following requirements: ① The reactivity control ability is enough and the rod worth distribution is appropriate. ② It is good for flattening the power distribution to decrease the power peaking factor and MCST. ③ It weakens the disturbance on power distribution to make sure the stability of neutronics and thermal-hydraulic coupling.

In order to obtain a longer refueling cycle, an average enrichment of 7.8% fuel has been used. The 1.4 wt % burnable poison Er₂O₃ has been added to the fuel to balance the large excess reactivity at begin of cycle. In addition, the control rod arrangement has been proposed to suppress the initial reactivity as shown in Fig. 7. There are in total 124 control rods in the core, which are divided into 17 groups; group Q with 28 rods is the safety rod. When the reactor experiences shutdown, all control rods insert into the core. Under normal condition, multiple groups of control rods move together to avoid a large disturbance on the power distribution. Figure 8 shows the position of the 17 groups control rods over the cycle.

A coupled three-dimensional neutronics/thermal-hydraulics code SNTA (SCWR coupled neutronics/thermal-hydraulics analysis code) has been developed for SCWR core steady-state analysis by modular coupling the improved neutronics nodal methodological code NGFMN_S and SCWR thermal-hydraulic code advanced thermal-hydraulics analysis subchannel [9]. The in-core fuel management code NGFMN_S is developed for SCWR core design with functions of cross sections readout, burn-up calculation, critical control rod position search, and equilibrium cycle search. The Nodal Green's function method is adopted for solving three-dimensional neutron diffusion equations to improve the coupled neutronics/thermal-hydraulics calculating efficiency. The code SNTA enhances convergence, stability, and efficiency of coupled N/T calculation by using outer iteration coupling method and self-adaptive relaxation iterative method. By comparing the existing steady-state code for SNTA preliminarily verification, the numeric results show that SNTA is accurate and efficient to SCWR core steady-state analysis [9]. The code SNTA also has a function of matching the coolant flow distribution with the power distribution of the assemblies in pass II to decrease the MCST and increase the outlet coolant temperature of pass II.

Main parameters of the improved CSR1000 core are listed in Table 1. The refueling cycle of the CSR1000 optimized design is 580 EFPD, which is an increase of 66%, compared to the preliminary conceptual design of 350 EFPD. The average fuel assembly discharge burnup is 59,124 MW d/t(U), and the maximum fuel assembly discharge burnup is 68,255 MW d/t(U). The maximum MCST is 647.6 °C, which appears at 300 EFPD. The maximum MLHGR is 36.9 kW/m, which appears at 580 EFPD. The shutdown margin at cold state is 1140 pcm. The maximum MCST, the maximum MLHGR, and the shutdown margin satisfy the safety criteria.

Figure 9 shows the core axial power distribution at 2 EFPD (begin of cycle), 300 EFPD (middle of cycle), and 580 EFPD (end of cycle (EOC)). Since the core adopts a two-pass flow scheme, there is power peaking at the top half and the bottom half of the core. The control rods continually draw out of the core with the increase of the burnup, and the power portion of the top half core increases accordingly. The power peaking factor of the core is 2.366, which appears at the core top half at EOC.

The maximum radial power peaking factor of the core is 1.448, which appears at 110 EFPD and the corresponding core radial power distribution is shown in Fig. 10. In order to decrease the MCST and increase the outlet coolant temperature, it is necessary to match the coolant flow distribution with the power distribution of the assemblies in pass II. The coolant flow distribution in pass II is calculated by the code SNTA and shown in Fig. 11.

Figure 12 shows the core MCST distribution at 300 EFPD. The inner core is for flow pass I, where the coolant temperature is low and with a good cooling capability to get a lower MCST. The peripheral core is for flow pass II, where the cooling capability decreases because of the high coolant temperature. The maximum MCST is 647.6 °C, which appears at the highest power peaking fuel assembly in the flow pass II.

The core burnup distribution of EOC is given in Fig. 13, which shows that the average and the maximum fuel assembly discharge burnup is 59,124 MW d/t(U) and 68,255 MW d/t(U), respectively. The maximum discharge burnup fuel assembly is at the center of the core, which has burned four fuel cycles.

Based on conceptual design of CSR1000, the FA design is optimized, and an optimized conceptual design of CSR1000 core is proposed. Steady-state performance of the proposed core is then studied with the code SNTA. The optimized FA design uses fuel with three different enrichments and the fuel average enrichment is 7.8%. The optimized core design adopts a two-pass flow scheme, three-batch fuel management with low-leakage fuel loading patterns, and a reactivity control method via 124 control rods and burnable poison Er₂O₃. This study shows that the optimization design satisfies all of the design criteria, and the refueling cycle of CSR1000 core can be extended effectively to 580 EFPD.

ABWR =

advanced boiling water reactor

CSR1000 =

China supercritical water-cooled reactor with the rated electric power of 1000 MW_e

EFPD =

effective full power days

EOC =

end of cycle

FA =

fuel assembly

k-eff =

effective multiplication factor

k-inf =

infinite multiplication factor

MCST =

maximum cladding surface temperature

MLHGR =

maximum linear heat generation rate

NGFMN_S =

Nodal Green's function method based on Neumann boundary condition for SCWR

N/T =

neutronics/thermal-hydraulics

SCWR =

supercritical water-cooled reactor