Trial and life cycle tests of NRE fuel elements in flowing gas-cooled channels of IVG-1 research reactor showed that the state of the irradiated fuel elements to a great extent depends on the specific modes of reactor tests [9-11]. Thus, the fuel elements of (Zr, U)C and (Zr, Nb, U)C after the trial tests in the channels with 8-cell fuel assemblies in the modes, significantly reduced compared to the nominal mode for NRE reactor, were discovered in a quite satisfactory condition: surface cracks and fracture of the fuel rods were few events, the longitudinal beams twisting of fuel elements in HS were small, the increase in the coefficient of gas-dynamic resistance of the cooling duct of fuel assemblies was also small. At the same time a significant post-irradiation increase of strength aS and thermal strength of fuel elements, as well as the presence of compressive residual stresses aR on their surfaces was recorded.

Hardening of fuel elements AaS/aS in three inlet (low-temperature) sections (HS-1, HS-2, and HS-3) reached almost 100%, in four outlet (high-temperature) sections it was small (Fig. 7.5).

The nature of fuel elements thermal strength increase was the same as the char­acter of their hardening. Measurements of fuel elements thermal strength by time — dependent thermal loading (destructive temperature drop ATF or the first criterion of the thermal strength was determined by rapid immersion of the pre-heated fuel rods in water) showed the more than twofold increase in thermal strength of fuel elements of (Zr, U)C in the first three inlet HS and the almost complete no change of

the characteristics of fuel elements of (Zr, U)C and (Zr, Nb, U)C in high-temperature HS. The coincidence of the change of the strength and thermal strength of fuel ele­ments on FA length is due to the fact that the thermal strength of brittle materials products is linearly related to their strength.

The results of X-ray measurements of the magnitude and sign of axial residual stresses or in fuel elements of two channels exposed to different uranium burn-up ND (ND burn-up—the number of uranium fission per a volume unit) are shown in Fig.7.6. In the measurements, we used the developed nondestructive method [12] with photo recording of the stresses in a special X-ray camera. The measurements were made on the fuel elements which had no surface cracks. When comparing the data in Figs. 7.6 and 7.5, we can see the satisfactory correlation of the change in hardening levels of fuel elements along the FA length and the levels of compressive residual stresses in them: the maximal hardening Aos/os and maximum stresses or occur in the fuel elements of the first three HS of both channels.

Furthermore, using only the fuel elements from HS-1 sections of six channels. X-ray study of the uranium burn-up effect on the level of occurred stresses or was conducted out. It turned out that ND increase on an order (from ~2.1-1016 to ~2.1-1017 div/cm3) leads to the more than twofold increase of stresses level or. This has led to express concern for the integrity of the fuel rods involved in a longer state life tests at IVG-1 reactor with the second core. Indeed, a further significant increase of ND burn-up the positive effect of increasing compressive stresses at the surfaces of fuel elements may be replaced by the negative: a significant increase of the tensile axial stresses or inside of fuel elements could lead to their damage into fragments at the time of cooling after start-up (at the time of temporary stresses disappearance ot).

The question on the danger of expected excessive growth of stresses or in the fuel elements was resolved only after the completion of their life cycle tests in the second IVG-1 reactor core, when the opportunity to X-ray measurements of stresses or in fuel elements with ND burn-up in a wide range (from ~2-1016 to ~2-1018 div/cm3). To do this, in addition to data on the fuel elements of (Zr, U)C tested in the six channels of IVG-1 reactor first core, X-ray diffractometry parameter measurements

of the magnitudes and signs of axial residual stresses or on the surfaces of fuel elements of (Zr, U)C+C tested in low-temperature HS of 11 channels of IVG-1 reactor second core were performed. The measurements were carried out on the selected integer fuel elements containing no surface cracks.

The results of measurements or (see Fig.7.7 in which data for the first core fuel elements are marked with square dots) indicated that the continuous growth of the radiation dose up to uranium burn-up about 2-1018 div/cm3 was not accompanied by continuous growth of compressive stresses or on the surface of a fuel element (and tensile or inside) to excessively high values. The increase of stresses or (fuel element material swelling heterogeneity increase), as seen from the figure was stopped after burning of about 3-1017 div/cm3 (indicated by dotted line) before reaching 200MPa, that eliminates the risk of failure of fuel elements from the tensile residual stresses operating in their internal regions.

In contrast to the first core of IVG-1 reactor, the second core fuel elements tested in modes close to the nominal operation of NRE reactor, there was a significant degradation of the initial state, especially noticeable in low-temperature HS-1 and HS-2. In particular, the fuel elements in these sections had numerous cracks and damage, and the longitudinal beams twisting in HS and the numbers of failures of fuel elements were increased with increasing the number of reactor start-ups. For example, after six start-ups the angles of beams twisting of fuel elements reached about ~20°, and the number of broken fuel elements (Fig.7.8) was approximately 80%.

Destroyed fuel elements in HS-1 and HS-2 were small (3-15mm in length) and slightly mixed fragments of rods that significantly—hundreds of percent—increased the coefficient of gas-dynamic resistance at the entrance of FA cooling duct. The fuel elements in the remaining four high-temperature HS remained intact, or were damaged only to large (30 mm length) fragments without mutual displacement, their strength decreased with increasing temperature and duration of exposure due to the degradation of the surface material by erosion and hydration.

It is obvious that the original cause of the negative situation in the inlet HS is the lack of thermal strength of carbide-graphite fuel elements at the temperature of the brittle state. Indeed, the level of thermal strength of these fuel elements is such

Fig. 7.8 View of fuel elements fracture from the second heating section (HS) (integrated flux of fast neu­trons of order 1019sm-2)

that the cracks appear in them even at heat loads qS higher ~5 MW/m2 (see upper dangerous region in Fig.7.9). That is why the fuel elements of the first two sections of 6-cell FA were damaged by cracks and destroyed during tests in IVG-1 reactor second core, and the fuel elements in the first three sections of 8-cell assemblies remained intact during the tests in the first core.

The cracks arising in FE under the influence of thermoelastic stress lead to a final fracture in the first and second HS by a tightness of ten-order microns on a series of the peripheral mainly damaged FE. Generated FE fragments can drop out and increase hydraulic resistance of HS, in a case if they have moving freedom in any direction. It depends on a concrete locating of fragments and also on angular orientation surrounding the yielded FE fragments and forces, capable to move them (Fig.7.10).

Fuel elements in the last HS undergo plasticity at elevated temperatures, forming sometimes through axial channels (Fig. 7.11). The decline of the heat removal, due to change of gas flow in the channel, leads to temperature increase of fuel elements that in the presence of compression loading promotes emersion of plastic deformation.

FE of the last high-temperature HS can also be twisted around axis and lose the stability in the presence of axial compression load under the influence of axial forces of the downthrusted spring and power loads, arising at gas passing in HS (Fig.7.12).

Consequently, for the elimination of a failure of fuel elements in the first two sections of 6-cell assemblies and prevent processes of grinding and mixing of the fragments it is necessary, as seen in Fig.7.9, to increase the initial thermal strength of the fuel elements of (U, Zr)C+C approximately double. As the level of thermal strength of brittle carbide material can be elevated by increasing its strength, in [10] the possible ways of pre-hardening of fuel elements by thermal, radiation, and combination methods were analyzed. Among the possible radiation methods, the way of fuel elements strengthening by residual radiation stresses or, in which the fuel elements of 6-cell FA were previously irradiated to ND burn-up about 3-1017 div/sm3 in a mode of decreased power (at average heat load on FA length qS about 4MW/m2), was pointed to. Pre-irradiation is carried out in the shape of the first start-up (FS) of NRE space facility located in orbit. Double hardening of carbide-graphite fuel elements of low-temperature HS achieved during FS must ensure the absence of cracks and breakage in the subsequent reactor activations.

A comparison of the change of fuel elements strength after reactor tests in only a few (n) working start-ups (nWS) and after the tests in the n working start-ups with prior power start-up (PS+nWS) [10, 11] can serve as an experimental substantiation of the proposed method.

Such positive effect of PS start-up was recorded among the fuel elements tested in the IVG-1 reactor fourth core (see Fig.7.13): hardening of fuel elements in HS-1 and HS-2 after the tests in PS + 2WS and 2WS amounted to ~35 and ~20 %, corre­spondingly.