Radiation Resistance of the HRA Elements

The radiation estimation of fuel elements (FA) and structural materials resistance to irradiation damage implemented during tests in reactor IVG-1 [1] and at measure­ments of properties after irradiation. The neutron flux and test temperatures in special reactor loops varied within 1012-1015n/sm2s and 450-2,000K accordingly [2, 3]. The uranium burnout in FA at propulsion mode (PM) of NRER within 1 h attained ~5 ■ 1015 and ~2 ■ 1020flss/sm3 at a power regime during 5,000h. Resource tests were carried out at heat release levels 15-35kW/sm3 during 300-4,000s. 200 FA had been tested in all.

The general regularity of materials’ behavior under irradiation may be regarded in the monography [4] and some information on interstitial phases there is in [5, 6], but the data on irradiation changes of fuel materials properties of the solid solutions with uranium of carbides and nitrides of 4-5 groups of periodic table are absent.

The most important characteristic of the radiation resistance of HREs is their dimensional stability. The swelling of HREs made of compositions based on solid solutions of uranium, zirconium, and niobium carbides depends on the fission density and irradiation temperature in a reactor is nonmonotonic function [3]. An increase in the irradiation dose of UC + ZrC + NbC and UC + ZrC fuel compositions up to 2 x 1019 fissions per cm3 at the irradiation temperature T = 1, 100 K leads to swelling by 5 % and an increase in the electric resistance up to 80 % due to the accumulation of radiation defects (mainly vacancies) (Fig. 5.1).

The bubble swelling change of fuel volume during an irradiation incubatory period is defined by balance between two competitive processes: swelling at the expense of accumulation of radiation-induced defects and irradiation sintering [3]. The swelling during the incubatory period poorly depends on the type of the investigated composi­tions and is defined, substantially, by an irradiation temperature. As concentration of the fission products is still the lowest, the dimensional sample modification is caused by accumulations of the dot defects, which evolution leads to gradual swelling growth under irradiation. Calculation shows that the maximum contribution of solid swelling at the expense of the accumulation of fission products, at B = 2-1020 fiss/sm3 makes 0.85 %., i. e., about 35 % from the total. Thus, dimensional changes of the investigated

A. Lanin, Nuclear Rocket Engine Reactor, Springer Series in Materials Science 170, DOI: 10.1007/978-3-642-32430-7_5, © Springer-Verlag Berlin Heidelberg 2013

compositions irradiated at T = 1,000 K and B from 8.4-1017 to 2-1020fiss/sm3 are defined by the swelling at the expense of the accumulations of radiation-induced defects (predominantly vacancies).

Gas bubble swelling of the fuel compositions UC + ZrC + NbC, UC + ZrC at temperature of irradiation T = 2,100 K, begins at a burnout A = 2-1019fiss/cm3. The volume modification AV/V0 from initial sample porosity p0 irradiated in the limits from 7-1017 to 1.8-1019 is defined by the constitutive equation:

AV/V0 = — C0P0[1 — exp(-B/B0)],

where C0, B—the constants depending basically from temperature irradiation and type of a fuel composition, P0 = initial porosity, B is a burnout.

The rise of irradiation temperature of the same fuel compositions from 1,000 to 2,100 K at constant fission density 2-1020 fiss/sm3 increases swelling to 6%, but reduces the values of lattice constant Aa/a and electrical resistance Ap/p to the initial value without irradiation [8] that is connected with sufficiently high annealing rate of radiation-induced defects (Fig. 5.2).

Noticeable changes of the fuel element electrical resistance A Р/Р are observed in the central HGA with temperature variation of the fuel elements (Fig.5.3).

The microscopic structure is changed simultaneously with swelling. The growing pores are arranged mainly on the grain boundaries and graphite inclusions. The most increment of porosity about 15-20%, is revealed for 3-fold solid solution UC-ZrC — NbC. Origination of pores increases the specific electrical resistance and decreases modulus of elasticity that proves to be true to the Р and E calculations on porosity at known relationships [2].

The research of the carbide fuels, implemented on the basis of the developed express complex techniques [1], confirms the radial fuel inhomogeneity swelling effect on a magnitude and on a stress sign of the fuel surfaces as at high-and low-temperature sections. Dependence of lattice constant change a0 = f(R) along fuel radius is close to the parabolic law [8, 9] (measurements of a0 were car­ried out by X-ray survey on various sections of a strongly oblique fuel section) (Fig. 5.4).

Different concentrations of radiation defects along the HRE radius (higher on the surface and lower at the more heated center) due to annihilation of point defects lead to a parabolic change in the lattice space. This gives rise to compressing stresses in the surface layers resulting in an increase of the strength and thermal strength. UC-ZrC and UC-ZrC-NbC HREs irradiated by the neutron flux J = 1014 — 1015/cm2/s at the fission density B = 1016 — 1018/cm3 and T < 0.4Tm enhances the strength by 30-50% and the thermal strength resistance by 70-80%. In this case, the strength increment decreases upon increasing the fission density above 1018 fissions per cm3 (Fig.5.5) due to the formation of vacancy and gas pores and some amorphization of materials.

The strength loss of carbide graphite (UC-ZrC+5 mas. %) occurs at lower burnout in comparison with double and three-fold solutions. The double solution UC-ZrC at irradiation is stable to fission density not less than 2-1019 fiss/cm3.

Dependence of residual stresses ars acting on a surface at inhomogeneity swelling [8]: can be easily computed by the change of lattice constant a0 on a fuel surface and measured modulus elasticity value E

ars = [A/ (1 — v)] x [(^V/V)] = [E/(1 — v)] x [Ц — 50)]

The isochronous annealing of heated HREs at 1,500 K removes inhomogeneity of swelling and the residual stresses, returning the strength to its initial level before irradiation, whereas the same irradiation doses in the case of structural carbides cause only weak changes in the elastic modulus and strength, along with a noticeable increase in the lattice pitch and electric resistance (Fig.5.6).

Irradiation of structural carbide materials under conditions similar to those of irra­diation of fuel materials, while retaining the volume and shape of the samples, leads

to an appreciable growth of the lattice parameter and growing electrical resistance, accompanied by a minor change of strength and Young modulus [10].

Under the parameters being studied, the Zr carbide density increased, while that of Nb carbide shows virtually no significant change. Lattice parameter and electrical resistance of ZrC increase much more significantly that those of NbC, while micro­hardness of these materials shows inverse behavior. After irradiation, the cracking resistance (as measured by change of the load P on the hardness tester’ indenter) of ZrC drops much more than that of NbC, while their thermal stability grows approx­imately by the same factor 1.7. Equimolar solid solution of ZrC and NbC behaves similarly to ZrC. though p behavior is rather close to that of NbC, while thermal stability demonstrating its own, unique behavior.

It is important to note that an absolute gain of electrical resistance after irradiation is much higher in ZrC and NbC, as compared to that in irradiated metals, which confirms the significant effect of the carbon sublattice on formation of the irradiation- induced defects as measured by electrical resistance change. Difference in ZrC and NbC behavior under irradiation should be attributable to the general differences in physical and chemical properties of IV and V group carbides, due to the electronic structure features [10] (Table 5.1).

It is also typical that, with increasing irradiation temperature to 1,300 K, the dif­ference in ZrC and NbC behavior remains virtually the same as at 450 K, though radiation damage consequences become much smaller. Particularly, noteworthy is also the fact of increased thermal stress resistance, possibly attributable to gen­eration of residual compressive stresses and to the effect of radiation-induced healing.

Radiation-induced fuel rod healing phenomenon has been experimentally con­firmed by declining effect of the introduced surface cracks on UC-ZrC and UC-ZrC — NbC fuel rods after low-power irradiation inside the He-filled ampoules. The degree of crack healing was estimated by monitoring changes in strength a. Electrical resis­tance p and an elastic flexure f of the fuel rods in the middle under consolidated mass impact.

The results of tests show that purely thermal annealing at relatively low (for carbide) temperature of 1,100 K cannot lead even to partial healing of the surface

Table 5.1 Physical-mechanical properties of irradiated samples

Properties

Materials

ZrC

NbC

(Zr, Nb)C

Y (g/cm)2

Initial

6.40

7.4

6.97

Irradiation

150 ° C

6.29

7.4

6.84

1,100 °C

6.27

7.4

6.91

a (A)

Initial

4.692

4.471

4.575

Irradiation

150°C

4.714

4.488

4.598

1,100 °C

4.698

4.472

4.578

p (^^ • cm)

Initial

43

50

68

Irradiation

150°C

250

90

100

1,100 °C

65

50

68

ai (kgf/mm2)

Initial

Irradiation

250

350

380

aimin aimax

150°C 1,100 °C

220-310

320

300-410

350

320-400

280

240-450

260

310-300

390

260-340

340

E •lO-3 (kgf/mm)2

Initial

240-300

41

340-410

49

310-490

46

Irradiation

150°C

41.5

47.5

45.5

1,100 °C

41

50

47

H|x (kg/mm2)

Initial

2,050

1,400

1,600

Irradiation

150°C

2,300

2,500

1,900

1,100 °C

2,200

1,600

2,000

P. g

Initial

100

120

120

Irradiation 150°C

40

100

60

AT(°C)

Initial

45

70

500

Irradiation

150°C

75

130

30

1,100°C

75

135

50

Table 5.2 Average relative values change of fuel elements characteristics (f. p. a) of UC-ZrC-NbC; with created surface cracks after irradiation in the nuclear reactor

Property

Initial

value

After

introduce of cracks

After annealing T = 1100K, t = 290 h

After irradiation T = 1100K, t = 290 h

After additional irradiation at 1870K, t = 1.5 h

f

100

124

125

104

100

p

100

126

125

141

103

a

100

45

47

198

207

cracks (see Table5.2). On the other hand, complete crack healing was observed after irradiation T = 1,100 K or after additional irradiation.

It is particularly apparent that crack-induced halving of fuel rod strength upon irradiation was replaced by 98 % strength rebound, of which 38 % correspond to a irradiation reinforcement observed in crack-free fuel rods, and the balance 60 % are attributable to an additional structural defect healing in rod surface, due to some of the critical surface defects being consumed during crack generation. The observed accelerated defect healing under irradiation is probably attributable to ‘displacement

peaks’ having much smaller size than the thermal cracks. The strength gain did not disappear after additional annealing at temperature above T^, whereas the flexure and electrical resistance returned to their initial values (i. e., before irradiation and crack introduction).

It should also be noted that the results of irradiation researches on the stability of the thermoelectric temperature transducers, for the determination of the possible temperature measurement errors. It was shown [11] that reversible and irreversible changes of the thermal electromotive forces are manifested under act of reactor irradi­ations. The reversible changes caused by additional energy release in small volume of a thermojunction are negligible small, irreversible changes grow appreciably with an increase of fluence neutrons and errors components should be considered necessar­ily as the regular capacity. The regular component of an error measurement ATr for thermocouple tungsten-rhenium VR5/20 is defined by an aspect of dependence [11]:

ATr = av фт +a2Taeaj ■ Tr,

where фТ, Фь—fluences of thermal and fast neutrons and Tirr is an irradiation temperature.

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