## 3 Research Results and Discussion
& & & & \\ et al., 2020 & steel fiber cement mixture & 300 years of operation & 20.0 kGy/h & Higher compressive strength and stable flexural strength limit \\
[127] & & (accelerated testing & (energy–1.25 MeV) & values \\  & & conducted using isotope Cs\({}^{137}\)) & & & \\  & & & & \\  & & & & \\  & mixed Portland cement SPC & 0.8–1.8 MGy & 1.6 kGy/h & strength—by 26.3\%, \\ Hlav  
The above data (Table 1) suggest that recently published sources of research and engineering information have contradictory information about the effect of gamma radiation on strength if the effect of gamma radiation on concrete and cement mortar is addressed. Authors of [75; 126; 128] emphasize that this process is largely associated with radiolysis and carbonization, and tests of control specimens irradiated in conditions of a radioactive waste storage facility do not show any changes in strength characteristics [126]. Indeed, Table 1 has information on standard structural monolithic reinforced concrete or cement mortar. There are no test data for higher-strength concrete or reinforced concrete structures produced offsite using a steam curing chamber. Authors of [59] propose four types of interaction between a gamma photon and atoms in a cement hydrate, namely, dislocation of atoms (with or without the rupture of chemical bonds) in the solid structure of the cement phase, hydrolysis of water, causing the formation of reactive particles, and thermal shrinkage with dehydration. In turn, the above trend makes it obvious that the type of interaction between gamma radiation and concrete composite is largely determined by the energy potential of the acting gamma quanta, the rate, and the accumulated radiation dose. In this case, it is necessary to establish parameters of the operation of protective walls and ceilings of buildings and structures where radioactive materials are handled. As for high-activity waste storage facilities, the predicted accumulated dose is estimated at up to 9 MGy, with a maximum rate of 23 Gy/h for the concept of a super-container designed for the final isolation of high-activity radioactive waste packages during their anticipated operation for 300 years [74]. Although [PERSON] et al. [60] further specify a figure of 120 MGy for the design service life of elements of a high-activity waste container. At the same time, they consider a container for the final isolation of spent nuclear fuel. [PERSON] et al. [126] emphasize that the accumulated dose corresponds to 35.8 MGy at a rate of 134.4 Gy/h for concrete specimens in a high-activity waste storage facility designed for 30+ years of storage. For example, if conditioned and cemented waste containing isotope Cs\({}^{137}\) (3.7\(\cdot\)10\({}^{10}\) Bq/kg) is stored [129], the absorbed dose reaches \(\sim\)1 MGy for the cement stone (not even packaging) in case of a nearly complete decay of the isotope activity of \(\sim\)2 MGy during the half-life of the radionuclide (30,17 years). The protective structure of the Shelter facility will accumulate up to 10.0 MGr over the course of 300 years of operation [127]. Since the purpose of this work is to consider protective structures of buildings and facilities where radioactive materials and waste are handled, the effect on them should be expected to be even smaller than the one specified above, since the above effects mainly refer to the packaging structures and waste cementation materials. For example, the design requirements applied to the dose rate for structures of unattended rooms of NPPs (located in the internal subshell space of the reactor island containment) are set at around 1.008 Gy/h [130], which corresponds to \(\sim\)0.9 MGy for 100 years of their operation. The average energy of the gamma spectrum in the concrete shield of NPPs was measured; it reached 1.93 MeV for WWER 1000 [75]. Authors [PERSON] and [PERSON]. [30] emphasize that if a service life of an NPP is 60 years, the calculated accumulated dose should not be expected to exceed 10\({}^{9}\) Gy.