1. Reactor As Neutrino Source
Fission Material
Fission Products
Neutrons
odd-odd nuclei
Structural materials of reactor
Electron capture or + decay ( neglected )
νe Emission
Captured
Mostly rich in neutrons
¯ Decay back to  stable valley
Anti-neutrino Emisson

2. Simplified reactor model
Candidate isotopes:
50Cr in RC , SS & Zr-alloy;
54Fe in RC , SS & Zr-alloy;
58Ni in RC , SS& Zr-alloy;
112Sn in Zr-alloy;
In SS:
50Cr %;
54Fe --4.2%;
58Ni --6.3%;
112Sn --0%.
In RC:
50Cr %;
54Fe --0.1%;
58Ni %;
112Sn --0%.
51Cr + e V + νe
55Fe + e Mn + νe
59Ni + e Co + νe
113Sn + e In + νe
Active mass of RC, SS and Zr-alloy(tons):
4967, 1040, 33

3. Simulation procedure Source neutron sampling; Neutron flux estimation;
Position: from fuel material homogeneously;
Energy: follow Watt fission spectrum;
Neutron flux estimation;
Track length estimation for flux simulation;
Neutron capture probability be tallied;

4. Cross Check K-eff : Total neutron capture probability:
Real value: 0.9~1;
Simulated result on our model: 0.77; (little under estimated)
Total neutron capture probability:
∑P n_capture = 1

5. Neutrino flux calculation
In which:
:ratio of neutron capture probability of the ith isotope in cell j to the total neutron capture probability of all the material in cell j
: nuclei number of the ith kind isotope in cell j;;
: neutron capture cross section of the isotope i in cell j, i=1,2…nj;
: electron capture branch ratio of the (A+1) isotope , which is produced from the ith isotope with atomic mass A via neutron capture;
: neutron capture probability of cell j;
: neutron flux in jth. cell; j=1,2,…M;

6. Simulation Result
Geometry distribution of neutron capture probability
Simulated neutrino emission Spectrum ;
Total neutrino emission rate normalized to per fission is: ~0.075
Correspond to an emission rate of 6 νe per fission on the energy range of < 1MeV ; ~ 10-2;