BOR-60 Reactor Research Articles

Radiation resistance of isotropic structural graphites

Experimental data are presented on the effect of neutron irradiation (360–400°C, neutron fluence up to 3 × 1022 n/cm2 , reflector of a BOR-60 reactor) on the dimensions and properties of isotropic, fine-grained structural graphites. The shrinkage rate and irradiation-induced size changes are shown to depend on the original characteristics of graphite.

High dose neutron irradiation damage in beryllium as blanket material

The paper presents the investigation results of beryllium products that operated in the SM and BOR-60 reactors up to neutron doses of 2.8×10 22 and 8.0×10 22 cm −2 ( E>1 MeV), respectively. The calculated and experimental data are given on helium and tritium accumulation, swelling, micro-hardness and thermal conductivity. The microstructural investigation results of irradiated beryllium are also presented. It is shown that the rate of helium and tritium accumulation in beryllium in the SM and BOR-60 reactors is high enough, which is of interest from the viewpoint of modeling the working conditions of the DEMO fusion reactor. Swelling of beryllium at irradiation temperature of 70–150 °C and neutron fluence of 2.8×10 22 cm −2 ( E>1 MeV) makes up 0.8–1.5%, at 400 °C and fluence of 8×10 22 cm −2 ( E>1 MeV) – 3.2–5.0%. Irradiation hardening and decrease of thermal conductivity strongly depend on the irradiation temperature and are more significant at reduced temperatures. All results presented in the paper were analyzed with due account of the supposed working parameters of the DEMO fusion reactor blanket.

Effect of the saturation conditions and structure on the retention of helium in structural materials

A comparison between the kinetics of helium desorption upon linear heating of samples saturated using various regimes is performed, and the effect of dislocations on the retention of helium in materials is estimated. In order to investigate the effect of the conditions of saturation of materials with helium on its retention, samples of austenitic stainless steel 0Kh16N15M3B saturated using various methods were studied, namely, helium irradiation in a cyclotron, in a magnetic mass-separation setup, inside IRT-2000 and BOR-60 reactors, and using the so-called “tritium trick” technique. The investigations show that when saturation of the samples with helium is accompanied by the introduction of radiation defects (in wide limits of helium concentrations and radiation damage), the kinetics of helium evolution from samples of this type is adequate to the kinetics of its evolution from samples irradiated in a reactor. The investigation of the kinetics of helium evolution from the samples of 0Kh16N15M3B steel both after a preliminary deformation and in the process of deformation showed that, in the process of heating, the helium atoms can migrate along dislocation pipes, resulting in a significant effect on the release of helium and its redistribution in the volume of the material. The activation energy for helium pipe diffusion in austenitic steel 0Kh16N15M3B is about 0.7 eV. Mobile dislocations favor the ejection of helium onto the surface of the material, to grain boundaries, interphase interfaces, etc.

Some experimental work performed on the BOR-60 reactor

The experimental work performed on a BOR-60 reactor over a period of 30 years of reactor operation is briefly reviewed. The results of investigations of the neutron-physical, thermohydraulic, and dynamical characteristics and the safety parameters of the reactor are presented. The investigations performed and the analysis of transient and emergency regimes made it possible to improve the standard shutdown and cooldown systems in order to soften the temperature conditions on the reactor components. The result of a series of experiments on the safety of fast sodium reactors, among which the introduction of gas into the core, sodium boiling, blocking of the flow in the experimental fuel assembly with destruction of fuel elements, interloop leakage in the steam generators, and so on, are discussed. A complex of diagnostics systems has been developed and tested on the basis of the safety investigations. Analysis of the radiation parameters and characteristics of the reactor made it possible to develop methods and means for monitoring and improving the radiation conditions and the safety of the reactor. Experimental irradiation of various initial materials, using threshold and other reactions, enabled the serial production of radionuclides for medical purposes.

Investigations of BOR-60 Structural Materials and Prospects for Further Work

Since its startup in December in 1969, the BOR-60 reactor has been used effectively for irradiation of structural and fuel materials in a wide range of dose–temperature parameters. Analysis of the actual computational-experimental parameters (irradiation temperature, damage rates) shows that the irradiation conditions are highly reproducible and can be maintained accurately.

Features of radiation damage of vanadium and its alloys at a temperature of 330–340°C

Microstructural changes of vanadium alloys after irradiation at 340°C to 12 dpa in the BOR-60 reactor in 7Li environment is analyzed. Materials are vanadium and its alloys V–3Ti, V–3Fe, V–6Cr, V–4Cr–4Ti, V–5Cr–10Ti, V–6Cr–1Zr–0.1C. Void formation was observed in the binary alloys V–3Fe, V–3Ti and V–6Cr. It is shown that three–four-fold increase in V–4Cr–4Ti yield stress is produced by the formation of dislocation loops (DLs) and fine radiation-induced precipitates (RIPs) with a density of 1.7×1017cm−3. It is expected that embrittlement of the welds will be worse because density of DLs and RIPs is 1.4–1.6 times higher. Besides, invisible coherent or semi-coherent RIPs are formed in the fusion zone. Elemental maps of the rupture surface of irradiated V–4Cr–4Ti are presented.

Radionuclide production using a fast flux reactor

The production of 89Sr, 32,33P, 35S via the ( n, p)-reaction, and 117mSn, 153Gd via the ( n, γ)-reaction using the BOR-60 fast flux reactor was experimentally studied. Test samples were irradiated in the active core of the BOR-60 reactor with fast neutron flux of 1 × 10 15 cm −2s −1 for 40–100 effective days. Gadolinium-153 was produced in a radial blanket cell, characterized by a modified (“softened”) neutron spectrum. Data on target materials, procedure of irradiated target reprocessing, radionuclide yield, and specific activity are summarized in the report. The results of the experiment showed that large-scale production of the radionuclides listed above is possible using a fast-flux reactor.

Mechanical properties and microstructure of advanced ferritic–martensitic steels used under high dose neutron irradiation

Some results of the study of mechanical properties and structure of ferritic–martensitic chromium steels with 13% and 9% chromium, irradiated in the BOR-60 reactor up to different damage doses are presented in this report. Results concerning the behaviour of commercial steels, containing to molybdenum, vanadium and niobium, and developed for the use in fusion reactors, are compared to low-activation steels in which W and Ta replaced Mo and Nb. It is shown that after irradiation to the dose of ∼10 dpa at 400°C 0.1C–9Cr–1W, V, Ta steels are prone to lower embrittlement as deduced from fracture surface observations of tensile specimens. Peculiarities of fine structure and fracture mode, composition and precipitation reactions in steels during irradiation are discussed.

Experience in irradiation testing of low-activation structural materials in fast reactor BOR-60

A Russia/US collaborative irradiation experiment on low-activation structural materials, mainly vanadium alloys, was recently completed in the BOR-60 reactor at the Research Institute of Atomic Reactors (RIAR) in Russia. The experiment, designated Fusion-1, was designed to investigate the effects of neutron-induced displacement damage on the microstructure and mechanical properties of these materials at low temperatures, 300–360°C. For purpose of heat transfer and impurity control, the specimens were lithium-bonded to the capsule. After reaching the goal damage dose of 20 displacements per atom (dpa) in approximately 1 year of irradiation, the capsule was discharged and disassembled in a hot cell at RIAR. All specimens were retrieved from the capsule. This paper summarizes the experiment, covering capsule construction, lithium purification and filling, irradiation, disassembly and specimen retrieval, and calculated results of specimen temperatures and damage dose.

Tensile properties and fracture behaviour of V–Cr–Ti alloys after neutron irradiation at 330°C

Results of irradiation effects in the fast neutron spectrum of the BOR-60 reactor on the mechanical properties and fracture behaviour of vanadium as well as on binary and ternary alloys (including V–4Cr–4Ti and V–5Cr–10Ti) are presented. The irradiation was carried out at 330°C in a 7Li isotope environment to a damage dose of 18 dpa. It is shown that all alloys without exception experienced severe radiation embrittlement. Most specimens failed without discernible traces of ductile strain at room temperature. At a test temperature of 350°C, ductility was observed but the total elongation, as a rule, did not exceed 1%. The fracture was mixed: transgranular brittle cleavage and ductile dimple rupture. As the test temperature was increased from 20°C to 350°C, the brittle fracture fraction decreased. At 600°C only ductile fracture was observed. The behaviour of V–4Cr–4Ti alloys of American and Russian production is analyzed in the 275–530°C temperature range, where a shift in behaviour from brittle to ductile failure is observed.

The effect of rhenium on the radiation damage resistivity of Mo–Re alloys

The results of investigations of alloys in the Mo–Re system are presented with Re varying from 0.5 to 47%. Samples were irradiated in the SM-2 reactor up to doses of 1–5.5×10 25 n/m 2 ( E>0.1 MeV) at 100°C, 210°C, 260°C and 330°C and in the BOR-60 reactor to doses of 0.8–2.0×10 26 n/m 2 ( E>0.1 MeV) at 480°C, 550°C, 760°C and 800°C. Irradiation to high doses in the SM-2 and BOR-60 reactors results in pronounced embrittlement of Mo–Re alloys in the testing temperature range of 300–600°C. The study of the electrical conductivity of the alloys has shown that irradiation in SM-2 results in a linear increase of the electrical resistivity with increasing Re concentration. Based on the analysis of available data the tentative conclusion has been made, that during irradiation in a mixed-spectrum reactor the resonance reaction proceeds on Re giving rise to the accumulation of transmutants (Os) reducing the electrical conductivity of the alloys.

Effect of neutron dose and spectra, He/dpa ratio and Ni and Zn accumulation on irradiation damage of pure copper and PH and DS copper alloys

The present analysis is based mainly on the results of recently undertaken investigations of radiation resistance of copper alloys irradiated in the SM-2 and BOR-60 reactors, as well as on the results of the just accomplished collaborative spectrally tailored experiment in the SM-2 reactor. Low-temperature radiation embrittlement at T irr∼100°C presents considerable problems for dispersion strengthened (DS) and precipitation-hardened (PH) copper alloys, as their total elongation at T test= T irr=100°C drops to ∼1% and uniform elongation to ∼0.1%. At irradiation temperatures above 300°C, the most significant mechanical property change for PH copper alloys is the softening effect due to radiation enhanced precipitate coarsening and dislocation recovery and recrystallization processes. The softening effect is sensitive to the neutron spectrum, and the threshold temperature of its manifestation in a mixed-spectrum reactor is apparently ∼100°C less than in a fast neutron reactor. The results of copper alloys irradiation in mixed-spectrum reactors serve as the basis for an analysis of the neutron spectrum effect and the problem of fission–fusion correlation. At irradiation temperatures greater than 350°C, helium embrittlement associated with the high helium accumulation rate for copper alloys exposed to fusion neutrons will be one of the main problems limiting the lifetime. At increased irradiation temperatures, creep will present a special problem. It is apparent that the DS copper alloys have some advantage in their radiation resistance compared to PH alloys such as CuCrZr, this being attributable to the high-temperature stability of their hardening structure.

Properties of dispersion strengthened copper made by high energy milling : B.F.Kieback et al. (Technical University of Dresden, Germany.)

The effects of neutron irradiation on the electrical resistivity of precipitation hardened (PH) and dispersion strengthened (DS) copper alloys are discussed. The analysis is based on the experimental study of radiation damage of PH and DS copper alloys, irradiated in the fast neutron reactor BOR-60 with doses of 8–16 × 1025 n/m2 and in the mixed spectrum neutron reactor SM-2 with doses of 3.7–5.5 × 1025 n/m2. The experimental data on the change Δϱ in electrical resistivity of DS-type copper alloys irradiated in the BOR-60 reactor show that irradiation to 7–10 dpa at T = 340–450°C causes a drop in electrical conductivity by not more than 20%. The obtained results show that in mixed-spectrum reactors the rate of Δϱ normalized to the dpa is about 20 times as high as in fast neutron reactors. The conclusion is made that the calculations performed for ITER must take into account the presence of appreciable fluxes of thermal neutrons in certain components of the reactor. The latter will play a decisive role in the drop in thermal conductivity of copper alloys in these components.

The effect of neutron irradiation on the electrical resistivity of high-strength copper alloys

The effects of neutron irradiation on the electrical resistivity of precipitation hardened (PH) and dispersion strengthened (DS) copper alloys are discussed. The analysis is based on the experimental study of radiation damage of PH and DS copper alloys, irradiated in the fast neutron reactor BOR-60 with doses of 8–16 × 10 25 n/m 2 and in the mixed spectrum neutron reactor SM-2 with doses of 3.7–5.5 × 10 25 n/m 2. The experimental data on the change Δϱ in electrical resistivity of DS-type copper alloys irradiated in the BOR-60 reactor show that irradiation to 7–10 dpa at T = 340–450°C causes a drop in electrical conductivity by not more than 20%. The obtained results show that in mixed-spectrum reactors the rate of Δϱ normalized to the dpa is about 20 times as high as in fast neutron reactors. The conclusion is made that the calculations performed for ITER must take into account the presence of appreciable fluxes of thermal neutrons in certain components of the reactor. The latter will play a decisive role in the drop in thermal conductivity of copper alloys in these components.

Neutron spectrum and transmutation effects on the radiation damage of copper alloys

Experimental results on the neutron spectrum effect on strength and ductility properties, transmutation rate and change in electric conductivity of copper-based alloys are presented in the paper. The analysis is based on the experimental study of radiation damage of pure copper, precipitation-hardened (PH) and dispersion-strengthened (DS) alloys, irradiated in the mixed spectrum reactor SM-2 and in the fast neutron spectrum reactor BOR-60 with the doses of (1–16) × 10 21 neutrons cm −2. It is shown that accumulation rate of the transmutants (Ni, Zn and He) in the SM-2 reactor is higher than in the BOR-60 reactor, which can be attributed to a high power flux of thermal neutrons in the SM-2 reactor. This causes a significant reduction in electric resistivity and mechanical properties of copper-based alloys. The effect of irradiation dose and temperature upon DS and PH copper alloy characteristics is analyzed. A threshold softening temperature for PH alloys is shown to be 300°C at a dose of 5 × 10 21 neutrons cm −2. The mechanism of the spectrum effect on the shift of the threshold softening temperature of PH copper alloys is also considered.

Influence of radiation transformations on the dielectric properties of electrically insulating ceramic during irradiation with a high neutron fluence

A great deal of experimental data has now been accumulated on the action of radiation on electrically insulating materials and dielectrics. As a rule, the effect of radiation transformations on the properties of solids has been neglected in this material [1, 2]. However, in the case of irradiation with a high neutron fluence it is inadmissable to neglect the accumulation of the elements which are formed in transmutational transitions. Our objective in the present paper is to analyze the possible influence of transmutants on the properties of ceramic dielectrics after the dielectrics are irradiated with a flux of neutrons with a wide energy spectrum (up to 14 MeV) to a fluence of about 102-2 cm -2. Irradiation of samples of I 1 brands of ceramics was conducted in the VI~K-8 channel (fluence 3.7.1021 cm -2) and VO-7 channel (fluence 2.3-1022. cm -2) of a BOR-60 reactor. The computational results suggest that in studying the properties of ceramic materials (UF-46, GB-7, MG-2, M-23, L-24, SNTs, SK-1, and others), transmutational processes must be taken into account since the observed changes are large [3]. In the case of the newly formed atoms (B, Si, Li, P, and others), the chemical bonds with the main crystal lattice are broken and these atoms can be treated as weakly bound ions. A typical example is the reaction l~ whose cross section is large, so that the reaction is characterized by a maximum yield of the daughter elements Li and He. The contribution of the transmutants to the specific dielectric losses was calculated using a simplified formula neglecting conductivity losses [4] according to the theory of thermal ionic polarization with a different frequency of the external electric field, temperature, and activation energy. The interatomic distance was taken as the distance between two neighboring positions of equilibrium of the weakly bound ions. The frequency dependence of the specific dielectric losses in the range 50-107 Hz after the ceramic was irradiated in the VI~K-8 channel was investigated. The energy losses P grow almost in proportion to the increase in the frequency. This corresponds to the region of relaxational polarization (o,,8 < < 1). At 300 K and with an activation energy of 0.1 eV the energy losses increase from 1.49.10 -9 up to 61.9 W/m 3 for MG-2, from 1.52" 10 -9 up to 63.4 W/m 3 for GB-7, and from 1.04.10 -9 up to 43.5 W/m 3 for microlite. Comparing with the experimental data for the unirradiated ceramic shows that the influence of the transmutants becomes substantial when the frequency is high and the temperature is relatively low (Fig. 1). The temperature does not appear explicitly in the formula for the dielectric losses. However, the dielectric losses depend strongly on the temperature, since as the temperature varies, the time constant O changes, decreasing with increasing temperature, as does the initial absorption current g, which increases with increasing temperature [3]. Investigations of the function P = F(T) showed that the function has no critical points in the region o~0 < < 1. The character of the temperature dependence was calculated for the temperature interval 523-1373 K. For example, at frequency 50 Gz and activation energy 0.1 eV the specific dielectric losses of the ceramics irradiated in the VO-7 channel decrease with increasing temperature from 4.17.10 -12 to 4.02.10 -12 W/m 3 for GB-7, from 4.56"10 t l to 4.41"10 -12 W/m 3 for MG-2, from 3.05" 10-I t to 2.95.10-12 W/m 3 for uralite. For this value of the activation energy the influence of the transmutants is found to be very small, whereas at 1 eV the yield of transmutants can strongly influence the dielectric losses. For example, after irradiation in the VI~K-8 channel the specific dielectric losses decrease from 4.31'10 -2 to 1.82.10 -8 W/m 3 for GB-7, from 1.98.10 -2 to 8.34.10 -9 W/m 3 for UF-46, and from 2.31-10 -3 to 9.75.10 l ~ W/m 3 for SPK-2 (Fig. 2). The relative contribution of transmutants in the case of an activation energy of 0.44 eV, frequency 50 Hz, and temperature 300 K equals 0.01% for M-23, 0.2% for UF-46, 2.8% for MG-2, and 9.2% for GB-7. Therefore the

Microstructural study of an irradiated high-nickel alloy by X-ray line profile analysis and TEM observations

The microstructure was investigated in previously deformed specimens of a high-nickel alloy with four levels of initial dislocation density, both before and after irradiation in BOR-60 reactor at 370–400°C and three displacement rates. The network dislocation density, dislocation loop radius and loop number density were determined by both X-ray line profile analysis and TEM observations. The dependence of dislocation structure parameters on irradiation conditions and initial network density was obtained.

Radiation tests of ITER diagnostic system materials in the BOR-60 reactor

An in-pile experiment was conducted up to a fluence of 9.9 × 10 25 m 2 ( E ≫ 0.1 MeV) which investigated the following electrophysical characteristics of a cable with mineral insulation and nickel conductor: insulation resistance, radiation-induced current and EMF. Irradiation was also performed in the BOR-60 reactor up to a fluence 10 23 m −2 on six crystal types for the monochromator: mica, LiF, multilayered mirrors Fe/C, W/Si, Cr/C and Mo/Si. Change of the reflectivity, width and shape of diffraction reflections were investigated.

High-temperature beryllium embrittlement

The neutron irradiation effect on the mechanical properties, swelling and fracture surface structure of various beryllium grades was studied in the BOR-60 reactor at 340 to 350°C up to a fluence of 7.2 × 10 21 n/cm 2. At a mechanical testing temperature of 400°C there was observed a strong anisotropy of plastic beryllium deformation depending on the direction of sample cutting relative to the pressing direction. An increase of the testing temperature up to 700°C resulted in an abrupt embrittlement of all irradiated samples. In the most part of the surface structure the intercrystallite fracture along the grain boundaries was covered entirely with large pores, 1 to 4 μm in size. It was suggested that the increase rate of pore formation along the grain boundaries resulted from a high-temperature embrittlement under irradiation.

Radiation embrittlement of low-alloyed Mo alloys

The influence of irradiation in the BOR-60 reactor at 250–1020°C up to fluences 0.013–7.6 × 10 26 n/m 2 ( E > 0.1 MeV) on the mechanical properties of a number of low-alloyed Mo alloys has been studied. Zr, Ti, Ru, B, C were used as alloying elements. Significant radiation strengthening and embrittlement of all the alloys investigated took place during the course of irradiation. According to tension test results, the brittle—ductile transition temperature, T k, was 600–700°C. Under comparable irradiation conditions, alloying had no substantial effect on the extent of irradiation embrittlement, since it did not prevent the radiation defect formation causing strengthening and, therefore, embrittlement. It was assumed that improvement of the Mo radiation resistance could be only obtained by high alloying with such elements as rhenium.