The radiation characteristics of the transport packages with vitrified high-level waste

Peculiarities of the radiation formation in dispersed microencapsulated nuclear fuel

“…For this reason, of greater interest is determination of the ratio of the quantity of the neutrons q an formed as a result of (a, n) reactions to the quantity of the neutrons resulting from spontaneous (q sf ) and induced (q f ) fission (in the event of large-size spent nuclear fuel (SNF) blocks) which is of interest both for calculating the neutron radiation dose rate during handling of such fuel Zabrodskaya 2005, Dulin andMatveyenko 2002) and in using the Rossi alpha method to determine the multiplication factor (k eff ) in subcritical systems (Dulin and Matveyenko 2002), such as the НGTRU fuel block (Shamanin et al 2018). This problem was examined rather thoroughly and solved in (Vlaskin et al 2015, Bulanenko 1979, Dulin and Zabrodskaya 2005, Dulin and Matveyenko 2002, Shamanin et al 2017, Bogatov et al 2015) and as part of other studies for fuel in the form of homogeneous dioxide. In the event when dispersed fuel and media containing heterogeneous inclusions of different configurations and sizes are considered, this problem is more difficult to solve and the method to calculate the quantitative and spectral compositions of the neutron source is of interest as such.…”

“…Such studies are conducted with the use of verified codes and dedicated programs. It often turns out during numerical modeling of radiation fields and in development of new procedures and regulations for the SNF handling in a nuclear fuel cycle of a new generation (Shamanin et al 2018, Bogatov et al 2015, Glukhov et al 2016, Spirin et al 2015 that no standard codes can be used without respective scientific rationale and upgrades. For instance, the ORIGEN-S code enables determination of the nuclear fuel nuclide composition and the spectral composition of the SNF neutron and photon radiation sources depending on burn-up and decay time.…”

“…This code is one of the advanced tools used internationally to calculate radiation sources. It is normally employed to calculate the burn-up and radiation sources of standard oxide fuel (UO 2 , (U,Pu)O 2 ) for LWR-type reactors, while the use of ORIGEN-S may lead to incorrect results in the event of other reactor types and for fuel other than of the standard type (Bogatov et al 2015). It should be noted that, apart from ORIGEN-S, there are other neutron source codes, e.g., SOURCES-4С (Wilson et al 2009), NEDIS-2m (Vlaskin and Khomyakov 2017), CHARS (Leniau and Wilson 2014), and others (Spirin et al 2015, Vlaskin and Khomyakov 2017).…”