Lead reactor of small power with metallic fuel*

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 (natPb) 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 208Pb, a low neutron-moderating 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 (UO2) substituted by metallic uranium-plutonium fuel (U-Pu-Zr), 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 one-group fission cross-section of 241Am increases from 0.359 to 0.536 barn, while the probability of the 241Am fission increases from 22 to 39%. It is proposed that power-grade 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 long-term storage conditions for high-level nuclear waste in dedicated devices.


Introduction
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 high-level waste (HLW) is also undesirable due to large quantities of heat it releases and its high volatility.
In one of the scenarios of a two-component (VVER+BN) nuclear power system (Troyanov 2016) in Russia, as it is known, fast sodium cooled reactors (BN) are expected to have the role as producers of plutonium for the MOX fuel of thermal reactors.And BN reactors will be fueled with power-grade plutonium obtained by regeneration of irradiated fuel from VVER reactors.Low-fission MAs in spent nuclear fuel (SNF) are expected to be converted into fission products.However, the neutron spectrum in fast sodium and lead cooled reactor cores appears to be not hard enough for the effective MA transmutation as the average core neutron energy does not exceed 0.5 MeV (Khorasanov and Blokhin 2013), this limiting the probability of the 241 Am fission to about 15%.As a result, some of the MAs remain unburnt or are converted into long-lived isotopes, and the equilibrium content of MAs in fast reactors may reach around 1 % (Lopatkin 2013).These MAs withdrawn from the SNF of BN reactors shall be either disposed or reburnt in a hard spectrum burner reactor in which the MA fission probability exceeds 15%.
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 (Samokhin et al. 2015) with uranium oxide fuel and BRUTs-M2 reactor (Khorasanov and Samokhin 2017) with metallic uranium-plutonium fuel (Vaganov et al. 2000, Aitkalieva 2016) have been considered as innovative reactors.The fuel composition calculations at this stage used the isotopic composition of power-gra-de plutonium produced from SNF of light water thermal reactors.Additional calculations will be needed to obtain data on the isotopic vector of the plutonium withdrawn from BN reactors with MOX fuel.

BRUTs and BRUTs-M2 reactors
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 (BRUTs-M2) with the reactor power increased and the uranium oxide fuel substituted for zirconium doped uranium-plutonium fuel.The parameters of the BRUTs and BRUTs-M2 reactors are presented in Table 1.

Calculation procedure
The neutron fluxes in the BRUTs and BRUTs-M2 core centers were calculated at the Institute of Physics and Power Engineering (IPPE) with a 28-group neutron spectrum approximation by Monte Carlo method using the MCNP/4B code (Briesmeister 1997) with a library of cross-sections based on the ENDF/B-VII.1 evaluated nuclear data files.The following neutronic parameters were calculated based on the obtained neutron spectra and using the same nuclear constants: one-group neutron energy in the core center (energy averaged upon the core center neutron spectrum); fraction of hard (E n > 0.8 MeV) neutrons in the core center neutron spectrum; one-group fission cross-sections for the 235, 238 U, 238,  239, 240, 241, 242 Pu, and 241 Am isotopes; cross-sections of the radiative neutron capture by these nuclei; probabilities of fission for these nuclei.The probability of the 241 Am fission, P f Am241 , was calculated from the relation P f Am241 = <σ fisAm241 > / (<σ fisAm241 > + <σ capAm241 >), where <σ fisAm241 > and <σ capAm241 > are one-group cross-sections of the 241 Am nucleus fission and cross-sections of the radiative neutron capture by the 241 Am nucleus respectively.The same procedure was used to calculate the probabilities of the U and Pu isotope nuclei fission.

Calculation results
Table 2 presents the results of calculating the neutron characteristics of the BRUTs and BRUTs-M2 reactor cores and one-group nuclear cross-sections of actinides in the calculated neutron spectra of the reactor cores.
It follows from Table 2 that the use of the U-Pu-Zr metallic fuel instead of uranium oxide fuel and of heavy nat Pb coolant in a small core reactor leads to an increase in: -the average energy of neutrons in the core center (by 30%); -the fraction of hard neutrons, E n > 0.8 MeV, in the core center neutron spectrum (by 57%); -the one-group cross-section of the 238 U nucleus fission (by 35%) and the probability of its fission (by 65%); -the one-group cross-sections of the 240, 242 Pu nuclei fission (by 40 to 50%) and the probabilities of their fission (by 30 to 37%); -the one-group cross-section of the 241 Am nuclei fission (by 49%) and the probability of its fission (by 78%).

Conclusion
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 BRUTs-M2 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 one-group cross-section of the 241 Am nuclei fission is 0.536 barn, which is also 2 to 2.5 times as high as the cross-section 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 two-component (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.

Table 2 .
Neutronic parameters of the BRUTs and BRUTs-M2 reactor cores and the actinide isotope range.
* OCNRC -one-group cross-section of nuclear radiative capture