The flow chart of the lattice physics code KYLIN V2.0 is shown in Figure 1; the program is mainly used for neutron simulation calculation and analysis of two-dimensional problems and provides assembly-homogenized parameters for three-dimensional core simulation software. The KYLIN V2.0 program can select the 45-/190-group structure of multi-group cross-section libraries for calculation. It adopts an advanced subgroup method (Nikilaev et al., 1970; Hebert, 2004) to solve the effective resonance cross-section of resonance nuclides, adopts an MOC method (Hong and Cho, 1998; Kosaka and Takeda, 2004) to calculate the neutron transport of complex structure geometry, and adopts a generalized coarse mesh finite difference (GCMFD) acceleration method to accelerate neutron transport calculation. The depletion calculation is carried out based on the improved predictive-corrected critical-fuel consumption iterative method (Yamamoto et al., 2009), and the burnup equation is solved by the Chebyshev method. At the same time, for the convenience of users, the program also has a graphical modeling tool for components with complex structure geometry and a post-processing display tool.

Specifically, KYLIN V2.0 software mainly has the following characteristics:

1) It can accurately describe geometric shape, mesh division, material distribution of various grids, or components with different geometric structures (including the bar bundle type and plate type) and provide users with clear and convenient input methods.

The PACFAC code adopts the interpolation/fitting method to parameterize the few-group assembly-homogenized parameters generated by KYLIN V2.0 and provides the polynomial parameters to CORCA-3D to calculate the few-group constants which can be determined by the core state (Li et al., 2016). The main functions of PACFAC contain the few-group constant parameterization model of few-group constants and the inverse calculation model of cross-section parameters. The former provides the polynomial parameters of the interpolation/fitting relationship between the few-group constants (few-group cross-section, shape factor, and so on) and core state parameters. The inverse calculation model is mainly used to determine the few-group constants in the core simulator, which is closely related to polynomial parameters in the interpolation/fitting relation. The parameterization formulation of PACFA is shown in the following equation: ∑ a c t u a l = ∑ i = 1 n N i a c t u a l σ i ( D M , B u , F T , ... ) + ∑ p i s ( D M , ... ) + Δ Σ r o d ( D M , ... ) . Here, ∑ p i s means the contribution of pseudo-nuclides to the macroscopic cross section, which is caused by the difference between the species of nuclides in the burnup chain of CORCA-3D and burnup chain of KYLIN V2.0. Δ Σ r o d means the influence component of the control rod on the cross section.

The main calculation flow chart of CORCA-3D is shown in Figure 2, which is mainly used for the simulation of three-dimensional steady-state core, and is one of the most important calculation software for core fuel management. CORCA-3D can carry out the diffusion calculation based on the few-group constants calculated via the homogenized cross-section parameter library, the burnup calculation of important actinides, fission products, and combustible poisons. CORCA-3D can perform thermal-hydraulic feedback calculation, pin-power reconstruction calculation, and is able to calculate the equilibrium concentration of iodine and xenon. CORCA-3D can read multiple-cycle databases for refueling core modeling calculation. At the same time, it has the basic functions of core design, such as coefficient calculation, variable parameter calculation, boron concentration/control rod position critical search calculation, and control rod differential/integral value calculation.

The CORCA-3D code mainly has the following characteristics:

1) According to the current core information and historical information of the component (including burnup depth, boron concentration, effective fuel temperature, moderator density, xenon concentration, and control rod status), the few-group cross-section parameters of each segment in the core can be calculated from the multi-parameter cross-section library provided by the linking code PACFAC.

Compared parameters of the reactor startup physics test include critical boron concentration (CBC), control rod integral value, boron differential value, and isothermal temperature coefficient.

The results of critical boron concentration are shown in Figure 3. The absolute errors of the critical boron concentration in the all-rod-out (ARO) state are within ±50 ppm and those of the rod-inserted state agreed with the rod-inserted criterion. The formula of the rod-inserted criterion is as follows: ( CBC ) R M = ( CBC ) R P + [ ( CBC ) ARO M − ( CBC ) R 0 M ] ± f ( ( CBC ) R 0 M , ( CBC ) R M , ( CBC ) R 0 P , ( CBC ) R P ) , where M means the measured value, P means the predicted value, R₀/R means the initial/final rod position at the time of measurement, and f is the value of the uncertainty of the measurement, including chemical analysis uncertainty.

The results indicate that the accuracy of ARO calculation is better than that of rod-inserted calculation with the exception of a very few calculations. It is because when the control rods are inside the core, the effect and shadow effect would increase the difficulty for the core simulation.

The results of control rod integral value, boron differential value, and isothermal temperature coefficient are shown in Figure 4, Figure 5, and Figure 6, respectively. The relative errors of control rod integral value are within 10%, the absolute errors of boron differential value are within 1.0 pcm/ppm, and the absolute errors of isothermal temperature coefficient are within 3.6 pcm/°C. The large errors appearing in the results belong to the subsequent cycles. It is caused by the core historical effect. In CORCA-3D, the micro burnup method is applied to handle this problem. The models which are based on the two-step calculation scheme adopted in the software cannot fully account for the historical processes. Anyway, all of them agreed with the acceptance criteria. It is indicated that TORCH V2.0 software has enough accuracy in predicting the reactor startup physics test parameters.

Compared parameters of the NPP operation include critical boron concentration, assembly-wise power distribution, hot spot factor, and nuclear enthalpy rise factor. The partial results of critical boron concentration are shown in Figure 7. The absolute errors of critical boron concentration of all the measured points shown in Figure 8 are within ±50 ppm except for very few points. At the beginning and end of the core cycle life, the absolute errors are almost worse than those of other times. The changes in xenon concentration and distribution at the beginning of the core cycle life and the core historical effect enhanced with depletion lead to increased errors.

The partial results of maximum assembly-wise power relative error are shown in Tables 2–4, summary of the maximum assembly-wise power relative errors of all burnup steps is shown in Figure 9 and Figure 10, and the maximum relative errors are within 5% when the relative assembly power is greater than 0.9 and within 8% when it is less than 0.9, except for very few points. The assembly-wise power distributions of Daya Bay NPP (Unit 1 Cycle 11) at the beginning, middle, and end of the core life are shown in Figures 11–13 respectively. It can be found that the calculation accuracy does not get worse with the increase in depletion.

The hot spot factors of Daya Bay NPP (Unit 1 Cycle 11) are shown in Figure 14. The relative errors of hot spot factors of all cycles are shown in Figure 15; the relative errors of hot spot factor are within 8%. The nuclear enthalpy rise factors of Daya Bay NPP (Unit 1 Cycle 11) are shown in Figure 16. All the relative error results of nuclear enthalpy rise factor are shown in Figure 17; the relative errors are within 5%.

All the calculated parameters are in good accordance with the measured values, which are in agreement with the acceptance criteria. For each core cycle of the NPP used to perform the validation, the average computer processing time of calculations is several minutes. Before the core simulation, the assembly-homogenized few-group constants and the databank of core front-order loops should be ready. The results show that TORCH V2.0 software has high calculation accuracy and efficiency for each cycle operation data of each power plant.