Resolution of the Generalized Eigenvalue Problem in the Neutron Diffusion Equation Discretized by the Finite Volume Method

Table 6

Biblis reactor cross sections.

Material	𝐷1 (cm)	𝐷2 (cm)	Σ𝑎1 (cm^−1)	Σ𝑎2 (cm^−1)	Σ12 (cm^−1)	𝜐Σ𝑓1 (cm^−1)	𝜐Σ𝑓2 (cm^−1)
1	1.4360	0.3635	0.0095042	0.0750580	0.017754	0.0058708	0.0960670
2	1.4366	0.3636	0.0096785	0.0784360	0.017621	0.0061908	0.1035800
3	1.3200	0.2772	0.0026562	0.0715960	0.023106	0.0	0.0
4	1.4389	0.3638	0.0103630	0.0914080	0.017101	0.0074527	0.1323600
5	1.4381	0.3665	0.0100030	0.0848280	0.017290	0.0061908	0.1035800
6	1.4385	0.3665	0.0101320	0.0873140	0.017192	0.0064285	0.1091100
7	1.4389	0.3679	0.0101650	0.0880240	0.017125	0.0061908	0.1035800
8	1.4393	0.3680	0.0102940	0.0905100	0.017027	0.0064285	0.1091100