Corresponding author: Georgiy Khorasanov ( khorasanow@yandex.ru ) Academic editor: Yury Korovin
© 2018 Georgiy Khorasanov, Dmitriy Samokhin, Aleksandr Zevyakin, Yevgeniy Zemskov, Anatoliy Blokhin.
This is an open access article distributed under the terms of the Creative Commons Attribution License (CC BY 4.0), which permits unrestricted use, distribution, and reproduction in any medium, provided the original author and source are credited.
Citation:
Khorasanov G, Samokhin D, Zevyakin A, Zemskov Y, Blokhin A (2018) Lead reactor of small power with metallic fuel. Nuclear Energy and Technology 4(2): 99102. https://doi.org/10.3897/nucet.4.30527

The possibility for obtaining a hard neutron spectrum in small reactor cores has been considered. A harder spectrum than spectra in known fast sodium cooled and molten salt reactors has been obtained thanks to the selection of relatively small core dimensions and the use of metallic fuel and natural lead (^{nat}Pb) coolant. The calculations for these compositions achieve an increased average neutron energy and a large fraction of hard neutrons in the spectrum (with energies greater than 0.8 MeV) caused by a minor inelastic interaction of neutrons with the fuel with no light chemical elements and with the coolant containing 52.3% of ^{208}Pb, a low neutronmoderating isotope.
An interest in creating reactors with a hard neutron spectrum is explained by the fact that such reactors can be practically used as special burners of minor actinides (MA), and as isotope production and research reactors with new consumer properties. With uranium oxide fuel (UO_{2}) substituted by metallic uraniumplutonium fuel (UPuZr), the reactors under consideration have the average energy of neutrons and the fraction of hard neutrons increasing from 0.554 to 0.724 MeV and from 18 to 28% respectively. At the same time, the onegroup fission crosssection of ^{241}Am increases from 0.359 to 0.536 barn, while the probability of the ^{241}Am fission increases from 22 to 39%. It is proposed that powergrade plutonium resulting from regeneration of irradiated fuel from fast sodium cooled power reactors be used as part of the fuel for future burner reactors. It contains unburnt plutonium isotopes and some 1% of MAs which transmutate into fission products in the process of being reburnt in a harder spectrum. This will make it possible to reduce the MA content in the burner reactor spent fuel and to facilitate so the longterm storage conditions for highlevel nuclear waste in dedicated devices.
Fast reactor, hard neutron spectrum, metallic uraniumplutonium fuel, natural lead coolant, americium241
Currently, the issues of the MA transmutation into the fission products of these nuclei receive a great deal of attention in literature. The content of ^{241}Am, e.g. in the MOX fuel of thermal reactors, needs to be minimized both for the safe handling of fuel in the process of its fabrication and for safe reactor control. The presence of large amounts of ^{241}Am in disposable highlevel waste (HLW) is also undesirable due to large quantities of heat it releases and its high volatility.
In one of the scenarios of a twocomponent (VVER+BN) nuclear power system (
This paper considers the feasibility of creating such reactor with a harder neutron spectrum using innovative fuel compositions and innovative heavy liquid metal coolant.
The purpose of the study is to show numerically the possibility of achieving a high probability of the ^{241}Am fission (over 15 %) in innovative hard neutron spectrum reactors.
BRUTs reactor (
The BRUTs reactor was proposed by the Obninsk Institute for Nuclear Power Engineering, NRNU MEPhI, as a training reactor. It was upgraded for operation as a burner reactor (BRUTsM2) with the reactor power increased and the uranium oxide fuel substituted for zirconium doped uraniumplutonium fuel. The parameters of the BRUTs and BRUTsM2 reactors are presented in Table
BRUTs and BRUTsM2 reactor parameters.
Parameter  BRUTs value  BRUTsM2 value 
Thermal power, MW  0.5  15 
Equivalent core diameter, mm  618  460 
Core height, mm  620  500 
Number of FAs in core  7  7 
Number of pins in FA  252  125 
FA flattoflat dimension, cm  17  17 
FA spacing, cm  17.2  17.2 
Outer diameter of pin, mm  12.7  9 
Fuel cladding thickness, mm  0.5  0.4 
Fuel pellet diameter, mm  11.5  8.0 
Fuel spacing, mm  14  14 
Fuel  UO_{2}  U 53 (^{235}U19.7%) + Pu 30 + Zr 17 
Fuel density, g/cm^{3}  10.5  13.37 
Coolant  ^{nat}Pb  ^{nat}Pb 
Core inlet/outlet coolant temperature, °C  460 / 500  450 / 500 
Fuel cladding surface temperature, °C  550  550 
Incore coolant/fuel/structural material volume fraction, %  26 / 67 /7  63 / 30 / 7 
Core fuel load weight, kg  1176  293 
Loaded weight of fissile nuclides  205 kg ^{235}U  88 kg Pu and 31 kg ^{235}U 
Reactor fueling K_{eff}  1.00721±0.00082  1.00018±0.00086 
Core center neutron flux, 1/cm^{2}·s  1.6∙10^{13}  1.4∙10^{15} 
The neutron fluxes in the BRUTs and BRUTsM2 core centers were calculated at the Institute of Physics and Power Engineering (IPPE) with a 28group neutron spectrum approximation by Monte Carlo method using the MCNP/4B code (
Table
Neutronic parameters of the BRUTs and BRUTsM2 reactor cores and the actinide isotope range.
Parameter  BRUTs value  BRUTsM2 value  Variation of BRUTsM2 value against BRUTs value, % 
Core center average neutron energy, <E_{n}>, MeV  0.554  0.724  + 30.69 
Hard neutron fraction, E_{n} > 0.8 MeV, %  18.11  28.45  + 57.10 
Onegroup ^{235}U fission crosssection, barn  1.550  1.338  –13.68 
OCNRC ^{*)} for ^{235}U, barn  0.362  0.238  –34.25 
Probability of ^{235}U fission, P_{f U235}, %  81.07  84.90  + 4.72 
Onegroup ^{238}U fission crosssection, barn  0.059  0.080  +35.59 
OCNRC for ^{238}U, barn  0.210  0.140  –33.33 
Probability of ^{238}U fission, P_{f U238,} %  21.93  36.36  +65.80 
Onegroup ^{238}Pu fission crosssection, barn  1.166  1.369  +17.41 
OCNRC for ^{238}Pu, barn  0.499  0.341  –31.66 
Probability of ^{238}Pu fission, P_{f Pu238,} %  70.03  80.06  +14.32 
Onegroup ^{239}Pu fission crosssection, barn  1.649  1.647  –0.12 
OCNRC for ^{239}Pu, barn  0.275  0.154  –44.00 
Probability of ^{239}Pu fission, P_{f Pu239,} %  85.71  91.46  + 6.71 
Onegroup ^{240}Pu fission crosssection, barn  0.471  0.667  + 41.61 
OCNRC for ^{240}Pu, barn  0.335  0.206  –38.51 
Probability of ^{240}Pu fission, P_{f Pu240,} %  58.44  76.40  +30.73 
Onegroup ^{241}Pu fission crosssection, barn  2.062  1.795  –12.95 
OCNRC for ^{241}Pu, barn  0.294  0.196  –33.33 
Probability of ^{241}Pu fission, P_{f Pu241,} %  87.52  90.16  + 3.02 
Onegroup ^{242}Pu fission crosssection, barn  0.344  0.517  + 50.29 
OCNRC for ^{242}Pu, barn  0.289  0.178  –38.41 
Probability of ^{242}Pu fission, P_{f Pu242,} %  54.34  74.43  + 36.97 
Onegroup fission crosssection, ^{241}Am, barn  0.359  0.536  + 49.30 
OCNRC for ^{241}Am, barn  1.281  0.835  –34.82 
Probability of ^{241}Am fission, P_{f Am241,} %  21.89  39.10  + 78.62 
It follows from Table
It has been shown that rather a hard spectrum of neutrons with the average neutron energy of <E_{n}> = 0.724 MeV in the core center and a large fraction (28%) of neutrons with an energy greater than 0.8 MeV is achieved in a lead fast reactor with a small sized core of D × H = 0.46 × 0.5 m^{2} and metallic fuel (U53wt%+Pu30wt%+Zr17wt%).
The calculated probability of the ^{241}Am fission in the hard neutron spectrum of the BRUTsM2 fast lead cooled reactor has a value of around 39%, which is 2 to 2.5 times as high as the probability value of this isotope fission in sodium fast reactors. At the same time, the onegroup crosssection of the ^{241}Am nuclei fission is 0.536 barn, which is also 2 to 2.5 times as high as the crosssection value of this isotope in sodium fast reactors.
The proposed method to increase the MA nuclei fissionability in the cores of lead reactors with metallic fuel can be used to reburn equilibrium MA residues in SNF of sodium fast reactors as part of the twocomponent (VVER+BN) nuclear power system in Russia.
Apart from its key function as a burner reactor, the hard neutron spectrum reactor can be used for production of medical isotopes through the (n, p) and (n, a) reactions, which are not readily achievable in currently effective isotope production reactors.