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Solution of multi-group neutron diffusion equation in 3D hexagonal geometry using nodal Green’s function method

ACCELERATOR, RAY TECHNOLOGY AND APPLICATIONS

Solution of multi-group neutron diffusion equation in 3D hexagonal geometry using nodal Green’s function method

Il-Mun Ho
Kum-Hyok Ok
Chol So
Nuclear Science and TechniquesVol.36, No.9Article number 159Published in print Sep 2025Available online 27 Jun 2025
13000

In this paper, we propose a numerical calculation model of the multi-group neutron diffusion equation in 3D hexagonal geometry using the nodal Green’s function method and verified it. We obtained one-dimensional transverse-integrated equations using the transverse integration procedure over 3D hexagonal geometry and denoted the solutions as a nodal Green’s functions under the Neumann boundary condition. By applying a quadratic polynomial expansion of the transverse averaged quantities, we derived the net neutron current coupling equation, equation for the expansion coefficients of the transverse averaged neutron flux, and formulas for the coefficient matrix of these equations. We formulated the closed system of equations in correspondence with the boundary conditions. The proposed model was tested by comparing it with the benchmark for the VVER-440 reactor, and the numerical results were in good agreement with the reference solutions.

NGFMHexagonal geometryMulti-group neutron diffusion equation
1

Introduction

Among the pressurized water reactors (PWRs) operating worldwide, Russian-type PWRs such as VVER-440 and VVER-1000 comprise hexagonal fuel assemblies. Liquid metal fast breeder reactors and very high-temperature reactors under the Next-Generation Nuclear Plant project also use hexagonal assemblies. In addition, these reactors are employed for broader applications in the nuclear industry, such as plutonium and hydrogen production; therefore, considerable improvements have been achieved in the underlying physics calculations of reactors with hexagonal geometry. In the early years, reactor physics calculations for a hexagonal geometry were performed using the finite difference method [1, 2]. In the 1990s, conformal mapping of a hexagon onto a rectangle were reported and codes based on conformal mapping, such as ANC-H [3] and PANTHER [4], were developed. Moreover, numerical methods based on the finite element method have been proposed to solve the neutron diffusion equation for hexagonal geometries [5-8]. The nodal methods developed in the mid-1970s were introduced to neutron diffusion calculations in hexagonal geometry, and nodal method codes such as DIF3D [9] were developed and applied. Owing to the geometrical complexity, it is difficult to solve nodal equations in hexagonal geometry using Cartesian coordinates; therefore, various approximation methods are applied. In the function expansion nodal method (FENM), the neutron flux within a node is expanded using polynomials or exponential functions [10-12]. This method has been used to solve multigroup neutron space-dependent kinetics equations in some studies [13-17]. In Cartesian geometry, nodal methods apply the transverse integration procedure (TIP) to the 3D neutron diffusion equation, resulting in three one-dimensional diffusion equations called transverse-integrated equations. When TIP is applied to a hexagonal geometry, singular (discontinuous) terms arise in the transverse leakage terms. Wagner [18] ignored these discontinuous terms in their solution and used lower-order polynomials to approximate the average source of a transverse-integrated equation. The use of the nodal Green’s function method (NGFM) allows for more accurate treatment of discontinuous terms that arise from the TIP. However, the accurate treatment of discontinuous terms was not implemented in earlier studies [19, 20]. The method presented in [21] improved the accuracy and fidelity of the code by adding the effect of discontinuous terms to the transverse integrated flux solution; however, the detailed description was limited to a two-dimensional geometry. In this paper, we propose a numerical model to solve the multigroup neutron diffusion equation in 3D hexagonal geometry using NGFM under the Neumann boundary condition. This model considers discontinuous terms in the solution of the 1D transverse integrated equation. By comparing this with the benchmark for the VVER-440 reactor, we demonstrate the accuracy of the proposed model.

2

Description of the method

2.1
The TIP for multigroup neutron diffusion equation in 3D hexagonal geometry

In 3D Cartesian coordinates (Fig. 1), the multigroup neutron diffusion equation within node k can be written as Dgk(2x2+2y2+2z2)Φgk(x,y,z)+Σr,gkΦgk(x,y,z)=Qgk(x,y,z) (1) where: Qgk(x,y,z)=[ggΣggk+χgkeffg=1GνΣf,gk]Φgk(x,y,z) (2) Integrating Eq. (1) over the volume of node k leads to the following nodal balance equation: 13ax=x,u,ν[Jgxk(a)Jgxk(a)]+12azk[Jgzk(azk)Jgzk(azk)]+Σr,gkΦ¯gk=Q¯gk (3) where: Φ¯gk=1Vkazkazkdzaadxys(x)ys(x)Φgk(x,y,z)dy (4) Q¯gk=1Vkazkazkdzaadxys(x)ys(x)Qgk(x,y,z)dyVk=azkazkdzaadxys(x)ys(x)dy=43a2azk (5) ys(x)=13(2a|x|) (6) where Vk is the volume of node k; Φ¯gk and Q¯gk are the neutron flux and neutron source averaged over Vk, respectively; a is the lattice half pitch; and azk is the haft height of node k.

Fig. 1
Hexagonal coordinate system
pic

In Eq. (3), Jgxk(±a)=12azkazkazkdz     [12ys(x)ys(x)ys(x)(DgkxΦgk(x,y,z))dy]x=±a (7) and Jgzk(±azk)=123a2aadx[ys(x)ys(x)(DgkzΦgk(x,y,z))dy]z=±azk (8) are the net neutron currents averaged at both surfaces in the x(x=x,u,ν) and z-directions, respectively.

The transverse integration procedure for Eq. (1) leads to the following equation: Dgkd2dx2Φgxk(x)+Σr,gkΦgxk(x)=Qgxk(x)Lgxk(x),x(x=x,u,ν) (9) where Φgxk(x)=azkazkdzys(x)ys(x)Φgk(x,y,z)dy (10) Qgxk(x)=azkazkdzys(x)ys(x)Qgk(x,y,z)dy (11) and Jgxk(x)=azkazkdzys(x)ys(x)[DgkxΦgk(x,y,z)]dy (12) are the transverse integrated neutron flux, transverse integrated neutron source, and transverse integrated neutron current, respectively. Lgxk is the transverse leakage given by Lgxk(x)=23[Jgnk(x,ys(x))Jgnk(x,ys(x))]+Jgzk(x,azk)Jgzk(x,azk)Dgksgn(x)3ddx[Φgk(x,ys(x))+Φgk(x,ys(x))]Dgk2δ(x)3[Φgk(x,ys(x))+Φgk(x,ys(x))] (13) Here, the following relationship between the transverse integrated flux and transverse integrated current is used: Jgxk(x)=DgkddxΦgxk(x)+Dgkys(x)[Φgk(x,ys(x))+Φgk(x,ys(x))] (14) In Eq. (13) Φgk(x,±ys(x))=azkazkΦgk(x,y,z)|y=±ys(x)dz (15) Jgzk(x,±azk)=ys(x)ys(x)[DgkzΦgk(x,y,z)]z=±azkdy (16) Jgnk(x,±ys(x))=azkazk[DgknΦgk(x,y,z)]y=±ys(x)dz (17) and ys(x)=13sgn(x),ys(x)=23δ(x).

Similarly to Eq. (9), the transverse integrated equation along the z-direction is given by Dgkd2dz2Φgzk(z)+Σr,gkΦgzk(z)=Qgzk(z)Lgzk(z) (18) where, Φgzk(z)=aadxys(x)ys(x)Φgk(x,y,z)dy (19) Qgzk(z)=aadxys(x)ys(x)Qgk(x,y,z)dy (20) Lgzk(z)=x=x,u,ν(Jgxk(a,z)Jgxk(a,z)) (21) Jgxk(±a,z)=[ys(x)ys(x)(DgkxΦgk(x,y,z))dy]x=±a (22) We define the transverse averaged neutron flux, transverse averaged neutron current, and transverse averaged neutron source as follows: Φ¯gxk(x)=14ys(x)azkazkazkdzys(x)ys(x)Φgk(x,y,z)dy=14ys(x)azkΦgxk(x) (23) J¯gxk(x)=14ys(x)azkazkazkdzys(x)ys(x)[DgkxΦgk(x,y,z)]              dy=14ys(x)azkJgxk(x) (24) Q¯gxk(x)=14ys(x)azkazkazkdzys(x)ys(x)Qgk(x,y,z)dy=14ys(x)azkQgxk(x) (25) Φ¯gzk(z)=123a2aadxys(x)ys(x)Φgk(x,y,z)dy=123a2Φgzk(z) (26) Q¯gzk(z)=123a2aadxys(x)ys(x)Qgk(x,y,z)dy=123a2Qgzk(z) (27) Substituting Eqs.(23)-(27) into Eqs.(9) and (18), the following equations for the transverse averaged neutron fluxes are obtained Dgkd2dx2Φ¯gxk(x)+Σr,gkΦ¯gxk(x)=Q¯gxk(x)L¯gxk(x),x(x=x,u,ν) (28) Dgkd2dz2Φ¯gzk(z)+Σr,gkΦ¯gzk(z)=Q¯gzk(z)L¯gzk(z) (29) where L¯gxk(x)=13ys(x)[J¯gnk(x,ys(x))J¯gnk(x,ys(x))]+12azk[J¯gzk(x,azk)J¯gzk(x,azk)]Dgksgn(x)23ys(x)ddx[Φ¯gk(x,ys(x))+Φ¯gk(x,ys(x))4Φ¯gxk(x)]Dgksgn(x)23ys(x)[Φ¯gk(x,ys(x))+Φ¯gk(x,ys(x))2Φ¯gxk(x)] (30) L¯gzk(z)=13ax=x,u,ν[J¯gxk(a,z)J¯gxk(a,z)] (31) are the transverse averaged leakages along the x(x=x,u,ν) and z-directions, respectively, and Φ¯gk(x,±ys(x))=12azΦgk(x,±ys(x)),J¯gnk(x,±ys(x))=12azJgnk(x,±ys(x)),J¯gzk(x,±azk)=12ys(x)Jgzk(x,±azk),J¯gxk(±a,z)=[12ys(x)ys(x)ys(x)(DgkxΦgk(x,y,z))dy] (32)

2.2
Solution to the transverse integrated equation using Green’s function under the Neumann boundary condition

Equation (28) can be written as follows (the index g and k and the - symbol are omitted for convenience): Dd2dx2Φ(x)+ΣrΦ(x)=Q(x)L(x),x=x,u,ν (33) The Green’s function is the solution of the following equation, which corresponds to Eq. (33) Dd2dx2G(x,x0)+ΣrG(x,x0)=δ(xx0) (34) Using Neumann boundary conditions, [dG(x,x0)dx]x0=±a and Eqs.(14), (23), and (24), the following equation for the transverse averaged flux is obtained Φ(x)=aa(Q(x0)L(x0))G(x,x0)dx0Jx(a)G(x,a) +Jx(a)G(x,a)D2a[Φ1+Φ22Φx(a)]G(x,a) +D2a[Φ4+Φ52Φx(a)]G(x,a) (35) where Φ1=Φ(a,ys(a)), Φ2=Φ(a,ys(a)), andΦ3=Φ(0,ys(0)), Φ4=Φ(a,ys(a)),Φ5=Φ(a,ys(a)), Φ6=Φ(0,ys(0)) are the corner point flux values of the hexagonal node.

We denote the transverse leakage by the sum of the individual terms as follows: L(x)=LJ,y(x)+LJ,z(x)+Lsgn(x)+Lδ(x) (36) where, LJ,y(x)=13ys(x)[Jn(x,ys(x))Jn(x,ys(x))],LJ,z(x)=12az[Jz(x,az)Jz(x,az)],Lsgn(x)=Dsgn(x)23ys(x)ddx[Φ(x,ys(x))+Φ(x,ys(x))4Φ(x)],Lδ(x)=Dδx3ys(x)[Φ(x,ys(x))+Φ(x,ys(x))2Φ(x)], (37) where Lsgn(x) and Lδ(x) are the singular (discontinuous) terms that appear in the hexagonal geometry.

Using the following linear approximations for Φ(x),Φ(x,ys(x)),Φ(x,ys(x)) Φ(x)=Φ(x,ys(x))+Φ(x,ys(x))2 (38) Φ(x,ys(x))={Φ1Φ6ax+Φ6 x>0 Φ5Φ6ax+Φ6 x<0 (39) Φ(x,ys(x))={Φ2Φ3ax+Φ3, x>0Φ3Φ4ax+Φ3, x<0 (40) in Eq. (35), the integral for the discontinuous terms leads to the following equation Φ(x)=aa(Q(x0)LJ(x0))G(x,x0)dx0Jx(a)G(x,a)+Jx(a)G(x,a)D2a[Φ1+Φ22Φx(a)]G(x,a)+D2a[Φ4+Φ52Φx(a)]G(x,a)+D2a[Φ3+Φ62Φx(0)]G(x,0)+D2a(Φ3+Φ6Φ1Φ2)0aG(x,x0)(2ax0)dx0+D2a(Φ3+Φ6Φ4Φ5)a0G(x,x0)(2a+x0)dx0 (41) where LJ(x)=LJ,y(x)+LJ,z(x) denotes the practical leakage term in the y and z-directions.

Substituting Eq. (38), Eq. (39) and (40) again into Eq. (41), the fourth, fifth, and sixth terms in right side of Eq. (41) become zero, and thus, the following equation for the transverse averaged flux is obtained: Φ(x)=aa(Q(x0)LJ(x0))G(x,x0)dx0Jx(a)G(x,a)+Jx(a)G(x,a)+D2a(Φ3+Φ6Φ1Φ2)0aG(x,x0)(2ax0)dx0+D2a(Φ3+Φ6Φ4Φ5)a0G(x,x0)(2a+x0)dx0 (42)

2.3
Constitution of the closed system of equations

The transverse-averaged flux, source, and leakage terms are expanded using orthogonal polynomials: Φ(x)=n=0n=2Φnωn(x)Q(x)=n=0n=2Qnωn(x)LJ(x)=n=0n=2Lnωn(x) (43) where ω0(x)=1,ω1(x)=325xa,ω2(x)=20127(9x2a252) (44) are orthogonal [18]: 123a2aa2ys(x)ωm(x)ωn(x)dx=δmn (45) where aadx;ys(x)ys(x)dy=aa2ys(x)dx=23a2 denotes the area of a hexagon with lattice pitch 2a.

From Eq. (45), the expansion coefficients in Eq. (43) are given by Φn=123a2aa2ys(x)Φ(x)ωn(x)dxQn=123a2aa2ys(x)Q(x)ωn(x)dx (46) Ln=123a2aa2ys(x)L(x)ωn(x)dx (47) where Qn is related to Φn via Eq. (2).

The zeroth order moment of the transverse averaged flux equals the node-averaged flux, that is, Φ0=123a2aa2ys(x)Φ(x)ω0(x)dx=123a2aa2ys(x)Φ(x)dx=Φ¯ (48) By contrast, the zeroth order moments of the leakage terms in the y and z coordinate directions are given by LJ,y0=123a2aa2ys(x)LJ,y(x)dx=13as=u,ν[Js(a)Js(a)]=L¯J,y (49) LJ,z0=123a2aa2ys(x)LJ,z(x)dx=13as=u,ν[Jz(az)Jz(az)]=L¯J,z (50) The first- and second-order moments can be obtained using an extrapolation procedure for two adjacent nodes, as in a square node.

The transverse-averaged flux at the boundary x=±a is as follows: Φ(±a)=aa(Q(x0)LJ(x0)G(x0,±a)dx0Jx(a)G(a,±a)+Jx(a)G(a,±a)+D2a(Φ3+Φ6Φ1Φ2)0aG(x0,±a)2ax0dx0+D2a(Φ3+Φ6Φ4Φ5)a0G(x0,±a)2a+x0dx0 (51) Inserting Eq. (43) for Q(x0) and LJ(x0) into Eq. (51) leads to the following equation Φ(a)=[GI+]Jx(a)G(a,a)+Jx(a)G(a,a)+[GS+] (52) Φ(a)=[GI]Jx(a)G(a,a)+Jx(a)G(a,a)+[GS] (53) where, =[GI±]=n=02(QnLn)[GI±]n (54) [GS±]=D2a(Φ3+Φ6Φ1Φ2)0aG(x,±a)(2ax)dx+D2a(Φ3+Φ6Φ4Φ5)a0G(x,±a)(2a+x)dx=12aκsinh2κa[(Φ3+Φ6Φ1Φ2)0acoshκ(a±x)(2ax)dx+D2a(Φ3+Φ6Φ4Φ5)a0coshκ(a±x)2a+xdx] (55) [GI±]n=aaωn(x)G(x,±a)dx         =1κDsinh2κaaaωn(x)coshκ(a±x)dx,            n=0,1,2 (56) where κ=ΣrD.

Using the continuity conditions of the transverse-averaged flux and the neutron current at the interfaces between adjacent nodes, f+kΦk(a)=fk+1Φk+1(a) (57) Jk(a)=Jk1(a),Jk+1(a)=Jk(a), (58) the net neutron current coupling equation can be rewritten as [GI±]n=aaωn(x)G(x,±a)dx         =1κDsinh2κaaaωn(x)coshκ(a±x)dx,            n=0,1,2 (59) where f+k and fk+1 are the discontinuity factors on the right boundary of node k and left boundary of node k+1, respectively. Moreover, Rk=Gk(a,a),Tk=Gk(a,a) are called the reflective and transmission factors, respectively.

We substitute Eq. (43) into Eq. (42) and multiply both sides of Eq. (42) by ωm(x). Using 2ys(x)23a2 as the weighting function and integrating over the interval [a,a], the following equation for the expansion coefficients of the transverse-averaged flux is obtained: Φ=[G](QL)+[G+]J(a)+[G]J(a)+D2a(Φ3+Φ6Φ1Φ2)[GIS+]+D2a(Φ3+Φ6Φ4Φ5)[GIS] (60) where [G] denotes a 3 × 3 matrix with the following elements: [G]mn=123a2aa2ys(x)ωm(x)dxaaωn(x0)G(x,x0)dx0=13a2aa(2a|x|)ωm(x)dx[axωn(x0)coshκ(ax)coshκ(a+x0)κDsinh2κadx0+xaωn(x0)coshκ(a+x)coshκ(ax0)κDsinh2κadx0],m,n=0,1,2 (61) The elements of the column vectors [G±] and [GIS±] are as follows: [G±]n=123a2aa2ys(x)ωn(x)G(x,±a)dx=13a2κDsinh2κaaa(2a|x|)ωn(x)coshκ(a±x)dx,n=0,1,2 (62) [GIS+]n=123a2aa2ys(x)ωn(x)0aG(x,x0)2ax0dx0=13a2κDsinh2κa0a12ax0dx0[coshκ(ax0)a0(2a+x)ωn(x)coshκ(a+x)dx+coshκ(ax0)0x0(2ax)ωn(x)coshκ(a+x)dx+coshκ(a+x0)x0a(2ax)ωn(x)coshκ(ax)dx],n=0,1,2 (63) [GIS]n=123a2aa2ys(x)ωn(x)a0G(x,x0)2a+x0dx0=13a2κDsinh2κaa012a+x0dx0[coshκ(ax0)ax0(2a+x)ωn(x)coshκ(a+x)dx+coshκ(a+x0)x00(2a+x)ωn(x)coshκ(ax)dx+coshκ(a+x0)0a(2ax)ωn(x)coshκ(ax)dx],n=0,1,2 (64) By contrast, the corner point flux values of a hexagonal node are included in Eqs. (55) and (60), which are also unknown quantities. For the interior nodes away from the reactor core boundaries, we approximate the corner point flux in terms of the node-averaged fluxes, surface-averaged fluxes, and diffusion coefficients in three adjacent nodes, as follows [21]: Φ0=[2(D1+D2)Φ1+2(D2+D3)Φ2+2(D3+D1)Φ3i=13DiΦi]/3(D1+D2+D3) (65) where Φi and Φi(i=1,2,3) are the node- and surface-averaged fluxes, respectively.

Note that the corner-point numbers in Eqs.(55) and (60) correspond to the calculation along the x-direction. The corner point numbers in the u- and ν-directions are smaller than one and two, respectively, compared to those in the x-direction; that is, uaxis direction;Φ1Φ6,Φ2Φ1,Φ3Φ2,Φ4Φ3,Φ5Φ4,Φ6Φ5νaxis direction;Φ1Φ5,Φ2Φ6,Φ3Φ1,Φ4Φ2,Φ5Φ3,Φ6Φ4 Next, we consider the equation for the transverse averaged flux in the z-direction Dd2dz2Φ(z)+ΣrΦ(z)=Q(z)L(z) (66) In this case, there are no discontinuous terms in the transverse leakage; therefore, the procedure is similar to that for a square node [22-24]. In addition, Legendre polynomials are used instead of Eq. (44). Then, the zeroth-order moment of the transverse leakage term becomes Lz0=1azazazL(z)dz=13as=x,u,ν12azazaz{[12ys(x)ys(x)ys(x)(DxΦ(x,y,z))dy]x=adz[12ys(x)ys(x)ys(x)(DxΦ(x,y,z))dy]x=adz}=13as=x,u,ν[Js(a)Js(a)]=L¯z (67) The first- and second-order moments of the transverse leakage term are the same as those of the square node. The net neutron current coupling equation, equation for the expansion coefficients of the transverse averaged flux, and matrix elements included in these equations also have the same forms as those in a square node.

The nodal balance equation, net neutron current coupling equation in each coordinate direction, its boundary conditions, and equation for the expansion coefficients of the transverse averaged flux constitute a closed system of equations.

3

Numerical results

The three-dimensional VVER-440 benchmark [3] was used to test the proposed numerical method. The material layout of the core of the problem is shown in Figs. 2 and 3. The group constants of these materials are listed in Table 1. The numerical results of this problem are summarized in Table 2, where Δkeff is the error of the effective multiplication factor and ΔPmax and ΔPavg are the maximum and average errors of the nodal power, respectively. For comparison, the numerical results of the other methods are also presented in this table. NGFM is the code based on the numerical method proposed in this study, and FENM, AFEN, and ANC-H are the codes based on the flux expansion nodal method, analytic function expansion nodal method, and conformal mapping, respectively, [12]. The accuracy of our numerical results was similar to that of other numerical methods. The effective multiplication factor obtained from our calculation was 1.01161, and the difference from the reference value (1.01132) was 0.029%. The maximum and average percentage errors of the nodal power were 2.51% and 0.63%, respectively. In the calculations, the axial mesh spacing was set to 25 cm. The reference solution was obtained from DIF3D-FD [25] runs with 216 and 294 triangle/hexagon subdivisions and a 2.5-cm axial mesh spacing. Figure 4 shows the assembly-wise normalized power distribution. The maximum error in the assembly-wise normalized power was 0.89%. These results show that the proposed numerical method is accurate for solving neutron diffusion equations in hexagonal-z geometry.

Fig. 2
The material layout of the twelfth core in the VVER-440 problem
pic
Fig. 3
The material layout on the vertical section of the core
pic
Table 1
Group constants of the materials in the VVER-440 problem
Material D1 (cm) D2 (cm) Σr1 (cm-1) Σa2 (cm-1) νΣf1 (cm-1) νΣf2 (cm-1) Σ1→2 (cm-1)
1 1.34660 0.37169 0.025255 0.064277 0.0044488 0.073753 0.016893
2 1.33770 0.36918 0.024709 0.079361 0.0055337 0.105810 0.015912
3 1.33220 0.36502 0.024350 0.10010 0.0070391 0.149640 0.014888
4 1.19530 0.19313 0.035636 0.13498 0.0 0.0 0.022264
5 1.44850 0.25176 0.033184 0.032839 0.0 0.0 0.032262
6 1.34130 0.24871 0.029301 0.064655 0.0 0.0 0.027148
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Table 2
Numerical results of the VVER-440 problem
Method Δkeff (%) ΔPmax (%) ΔPavg (%)
NGFM 0.029 2.51 0.63
FENM 0.026 2.35 0.58
AFEN 0.030 3.20 0.73
ANC-H 0.025 1.28 -
Show more
Fig. 4
Assembly wise normalized power distribution of the twelfth core in the VVER-440 problem
pic
4

Conclusion

A numerical model was proposed to solve the neutron diffusion equation in hexagonal-z geometry, and the discontinuous terms in the transverse leakage were explicitly considered in this model. The solution of 1D transverse integrated equation was expressed using the nodal Green’s function under the Neumann boundary condition. Using the quadratic polynomial expansion of the transverse averaged quantities, the net neutron current coupling equation and equation for the expansion coefficients of the transverse averaged neutron flux were obtained. We constructed a closed system of equations by deriving the equations corresponding to the boundary conditions. The numerical model proposed in this study was tested by comparison with the benchmark for the VVER-440 reactor, and the numerical results were in good agreement with the reference solutions.

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Footnote

The authors declare that they have no competing interests.