Radiation intensity and expansion saturation of the crystal lattice of irradiated diamond

It is shown that the saturation of the expansion of the diamond crystal lattice during irradiation in a reactor depends not only on the temperature but also on the intensity of the irradiation. Expansion saturation decreases with decreasing irradiation intensity. The temperature influences saturation expansion indirectly via an increase of the effectiveness of the annealing of defects, and the intensity of irradiation stimulates additional annealing by increasing its duration. The same level of the saturation of expansion of the diamond lattice, specifically, 2.75%, can be attained at irradiation temperatures 80 and 300°C if the diamond is irradiated with neutron flux density 10¹² and 10¹⁴ sec^–1·cm^–2, respectively.     

Study of the behavior of the graphite-sodium system for the central rotating column in a BN-600 reactor

The characteristics of sodium permeation through graphite and the accompanying swelling of the graphite are examined for the central rotating column of a BN-600 reactor. The sodium transport parameters when sodium comes into contact with graphite at 350–500°C for up to 400 h are determined experimentally. Under these conditions, the permeation parameter is (0.13–1.3)·10⁻¹¹ m²/sec, which corresponds to an effective diffusion coefficient (0.2–2)·10⁻¹¹ m²/sec. The ratio of the increment to the graphite volume and the sodium mass there is ∼0.85. __________ Translated from Atomnaya énergiya, Vol. 101, No. 6, pp. 431–437, December, 2006.     

State of helium in irradiated beryllium

Experimental data are presented on the change of the pycnometric density and unit-cell volume of beryllium irradiated in an SM reactor to neutron fluence 14·10²² cm^–2. The behavior of the unit-cell volume as a function of the neutron fluence is similar to the data for other metals. This and data on the volume change in beryllium oxide, where a large amount of gas is also formed during irradiation, make it possible to evaluate critically the assertion that helium forms a solid substitution solution. It is proposed that helium is confined in small accumulations.     

Properties of the final product of thermochemical reprocessing of irradiated reactor graphite wastes in the Al-TiO<Subscript>2</Subscript>-C system

Laboratory investigations of the strength and chemical resistance of the final product of thermochemical reprocessing of reactor graphite wastes in the Al-TiO₂-C system are presented. The ¹³⁷Cs and ⁹⁰Sr leaching rate, which is determined for samples synthesized from a charge with real irradiated graphite from an AM research reactor, does not exceed 10⁻⁶ g/(cm²·day) at the 28th day. __________ Translated from Atomnaya énergiya, Vol. 104, No. 4, pp. 224–227, April, 2008.     

60Co content in the spent graphite masonry of commercial reactors at the siberian chemical combine

Some results of comprehensive investigations of the radioactive contamination of graphite masonry from shutdown commercial uranium-graphite reactors at the Siberian Chemical Combine are reported. The objective of the investigations was to study the distribution of radionuclides and to determine the contamination level. In the present paper information about⁶⁰Co in the gaphite of the I-1 and él-2 reactors is reported. Its content in the samples was measured by γ-spectrometry. There were about 250 graphite samples from the I-1 reactor and 200 from él-2. According to the data obtained, the surface contamination level of the blocks can be taken as the same for the entire core within the limits of the errors presented. The average⁶⁰Co contamination of the graphite in the surface of blocks from the I-1 core is 5600 ₋₅₀₀ ⁺⁵⁵⁰ Bq/g and 8400 ₋₁₀₀₀ ⁺¹²⁰⁰ Bq/g for the él-2 core. The⁶⁰Co content in the interior volume of the graphite blocks of a I-1 reactor is now 1100 ₋₁₆₀ ⁺²⁰⁰ Bq/g and 2000 ₋₃₀₀ ⁺¹³⁵⁰ Bq/g in éI-2. The⁶⁰Co activity in all blocks from the I-1 core is 1.22·10¹² Bq, and for éI-1 the figure is 2.16·10¹² Bq. 4 figures, 3 tables, 7 references. Translated from Atomnaya énergiya, Vol. 86, No. 3, pp. 183–188, March, 1999.     

The radiation change in the coefficient of linear thermal expansion of carbon materials

Conclusion Hence, the thermal coefficient of volume expansion of graphite is related exponentially to the height of the crystals and the density of the material and depends on the specific surface of the structure and micropores. The coefficient of linear thermal expansion of graphite is inversely proportional to the dynamic modulus of elasticity. The negative change in α of graphite on neutron irradiation: changes nonmonotonically with the neutron fluence and the radiation temperature — initially it increases, reaches a maximum, then falls and again increases; is inversely proportional to the power 1/3 of its initial value, to the rate of steady radiation creep and the neutron fluence; is determined by the degree of perfection of the crystal structure and the concentration of spherolites (carboids) of the elements of the microstructure. Their increase facilitates a fall in α below its initial value; it does not recover completely on thermal annealing to 2300 K. The relative change in α of carbon-carbon composition materials when irradiated to a neutron fluence of 3·10²⁰ cm⁻² and a temperature from 320 K to 2100 K does not exceed 10%. The complex nature of the radiation change makes it difficult to calculate the value of α, and hence it has to be determined in experiments up to the resource dose. Graphite Scientific Research Institute. Translated from Atomnaya énergiya, Vol. 82, No. 6, pp. 417–424, June, 1997.     

Monitoring of the neutron fluence on the structural components of a VVR-M reactor

Experimental and computational methods for monitoring the fluence of fast neutrons on the most critical structural components of the VVR-M reactor are presented. The dynamics of the accumulation of the fluence at the bases of the experimental channels and the bearing lattice of the core over the last 10 years of reactor operation is presented. A method of preirradiation of samples of the main structural alloy CAB-1 under real conditions in the VVR-M core was developed. This made it possible to reach a fluence up to 2.5·10²² cm⁻² on the samples. Over 40 years of reactor operation the maximum fluence on the structural components reached ∼1.7·10²² cm⁻². The study of the mechanical properties of forcibly irradiated samples will make it possible to draw conclusions about the remaining period of safe operation of the reactor. This is important for practical applications and is of economic value. 2 figures, 1 table, 14 references. Deceased. B.P. Konstantinov St. Petersburg Institute of Nuclear Physics. Translated from Atomnaya énergiya, Vol. 86, No. 3, pp. 175–178, March, 1999.     

Investigation of radiation creep of reactor graphite

The results of investigations of the radiation creep of GR-280 graphite under a high compression load (about 15 MPa) after irradiation in a BOR-60 reactor at 520°C to fast-neutron fluence 1.2·10²² cm⁻² are presented. It is shown that the fluence dependence of the creep deformation, calculated using the standard relation as the difference of the change in the dimensions of loaded and control samples, is anomalous. The linear thermal expansion coefficients of loaded and control samples are found as functions of the neutron fluence under the same conditions. It is noted that the linear thermal expansion coefficient of the samples irradiated under a load is much higher than that of the control samples. Simmons' theorem is used to take account of the effect of a load on the linear thermal expansion coefficient, and the dimensional changes of graphite exposed to radiation and the dependence of the true creep deformation on the neutron fluence are calculated. It is shown that these dependences are close to linear in the experimental fluence range (0.4–1.2)·10²² cm⁻². Translated from Atomnaya énergiya, Vol. 105, No. 2, pp. 83–87, August, 2008.     

Results of SM reactor tests of dysprosium hafnate

Models of absorbing elements with a promising material for the control organs of nuclear reactors have been tested in the SM reactor – pelleted and powder kernels with different composition based on dysprosium hafnate in a mixture with boron carbide. The neutron fluence with energy >0.1 MeV averaged over a kernel volume was (0.9–1.3)·10²² cm^–2 at the moment the tests were completed for different samples. The temperature at the center of the kernels of the absorber element models during irradiation was 620–1100°C in channel No. 4 and 400–500°C in channel No. 9. The results of the materials science studies show that on the whole the serviceability of the absorbing elements based on pellets and powders of dysprosium hafnate is high.     

Investigation of radioactive contamination of graphite samples from the AM reactor

The world’s first nuclear power plant operated for almost 48 years. Over this period of time, the neutron fluence on the graphite masonry reached ∼10²² cm⁻², which resulted in activation of the impurities present in the graphite. During operation, incidents occurred with loss of seal and sometimes loss of integrity of the fuel-element claddings in some cells and particles of the fuel and steam-water mixture entered the graphite masonry. This resulted in radiation contamination with a complex radionuclide composition. Experimental information about the content and distribution of radionuclides in the spent nuclear graphite is needed in order to plan methods and periods of time for disassembly and salvaging of the graphite masonry of the stopped reactor taking account of the dose loads on the workers and the ecological safety norms. The problems which can be solved on the basis of the present work included the determination of the ¹⁴C and ³H contents by liquid-scintillation β spectrometry, analysis of the actinide content by direct γ spectrometry, and neutron-activation analysis followed by γ spectrometry. These investigation yielded new data on the content of fission products and activation impurities in graphite. __________ Translated from Atomnaya énergiya, Vol. 101, No. 5, pp. 358–364, November, 2006.     

Measurement and analysis of noise spectra – an effective diagnostics method for nuclear reactors

The basic principles of measuring and analyzing nuclear-reactor noise are described. The results obtained for the IBR-2 reactor by the noise method are presented. It is shown that analysis of the noise spectra of the power and the main reactor parameters makes it possible to find deviations from normal reactor operation at the level 10^–6 ∆k/ k of the change in reactivity.     

Determining the irradiation temperature of witness samples of VVER-1000 AND -440 reactor vessel steel

The most convenient methods for measuring the temperature in nuclear reactors are enumerated in this paper, specifically thermocouples, fusible metal inserts, and diamond probes. Their advantages and disadvantages are listed and data are presented on their use for measuring the irradiation temperatures of containers with witness samples of steel in VVER-1000 and -440 reactor vessels. It is found that the sample temperature in these assemblies in the VVER-1000 is 300 ± 2°C, as measured by fusible inserts, while thermocouples in chains in the VVER-440 indicated 270°C. The advantage of the method employing diamond probes is their small size, but the range of possible direct measurements for this technique is restricted to a maximum neutron fluence of 10¹⁸ cm⁻² (for neutron energies above 0.5 MeV). Translated from Atomnaya énergiya, Vol. 105, No. 3, pp. 145–150, September, 2008.     

Reconstruction of metal embrittlement in a reactor vessel of the atomic icebreaker Lenin

Data from a study of radiation damage to the vessel of a reactor from the retired atomic icebreaker Lenin are used to determine the radiation embrittlement characteristics of the metal. Irradiation by a low neutron flux of 10¹⁰–10¹¹ cm⁻²sec⁻¹ at the beginning of operation is found to correspond to more intense embrittlement of the metal. Then, apparently, as harmful elements are depleted in the matrix of the metal, embrittlement is reduced until there is a change in sign relative to the standard curve obtained for neutron fluxes above 10¹³ cm⁻²sec⁻¹. It is proposed that, because of irradiation by low fluxes of neutrons in the peripheral zones of reactor vessels during some stages of operation, these zones may be damaged to a greater extent than those lying closer to the core. The irradiating neutron flux is a factor that influences the embrittlement of reactor vessel materials, so there is some interest in studying how material is damaged in the vessels of power reactors with low radiation loads which are under development. This is also needed in order to evaluate the efficacy of measures undertaken to reduce the effect of neutron irradiation on reactor vessels. Translated from Atomnaya énergiya, Vol. 105, No. 4, pp. 201–205, October, 2008.     

Swelling and radiation creep of austenitic corrosion-resistant steel irradiated by neutrons in wide dose and temperature ranges

The results of a study of the swelling and in-reactor creep of EI-847, EP-172, and ChS-68 austenitic steel after irradiation in materials science assemblies in the range 330–700°C and damaging dose 20–96 dpa are presented. The temperature dependences of the volume change of steel were obtained from measurements of the diameter of unloaded ampuls. It is shown that the swelling of the steel increases linearly with increasing tangential stress. The modulus of in-reactor creep in the interval 410–630°C for the steel investigated in the cold-deformed state varies in the range (0.5–3)·10^–6 MPa^–1·dpa^–1. For lower and higher temperatures, the creep modulus increases to (5–8)·10^–6 MPa^–1·dpa^–1.     

An intense channeling radiation source

A non-conventional X-ray source which is based on the production of electron channeling radiation in a diamond crystal has been installed at the radiation source ELBE. The brilliant electron beam with an average current of up to 200 μA allows to reach photon rates of quasi-monochromatic channeling radiation between 10¹⁰ s⁻¹ and 10¹¹ s⁻¹ per 10% bandwidth. The photon energy can be tuned by variation of the beam energy. On-line X-ray monitoring was realised at high beam currents using a Compton spectrometer. Monochromisation of channeling radiation and bremsstrahlung background reduction has been investigated applying X-ray diffraction at a highly ordered pyrolytic graphite crystal.     

The absorption of deuterium by carbon-based facing materials on components in contact with the plasma in a thermonuclear reactor

Conclusions The conditions have been proposed for performing modeling experiments making it possible to predict the accumulation of hydrogen isotopes in carbon materials which are in contact with a tokamak plasma acting as a source of particles having a flux density of between 3×10¹⁶ and 3×10¹⁹ cm⁻²·sec⁻¹. By analyzing the reemission fluxes formed in the stopping zone of the particles implanted from the plasma it is suggested that the action of the plasma as regards the sorption of hydrogen is identical to that of annealing the material in an atmosphere of hydrogen isotopes at a pressure of 1–10³ Pa and a temperature of 1200–1700 K. The quantity of absorbed deuterium in POCO, UAM, RGT-B, and USB increases as the temperature is lowered and the pressure is raised (1500 K, 0.66 Pa→1200 K, 133 Pa). As regards their sorption of deuterium, POCO, UAM, and RGT behave similarly. There is a tendency for the sorption capacity of materials doped with boron to be reduced. In a class of itself is the isotropic material USB, whose sorption capacity is a factor of 10–100 lower than that of undoped graphite. The introduction into these materials of radiation-induced defects (T=300 K) by means of ion irradiation in the range 0.1–1 dpa results in a continuous rise in the deuterium sorption capacity by a factor of 10–100 (up to 10⁻² atomic fraction). The USB graphite demonstrates record low increments in the sorption capacity. In the fluence range identical to 1–10 dpa the sorption capacity of carbon materials for hydrogen is almost constant. The process of the sorption of hydrogen isotopes can be described as the filling of two ensembles of traps, deep traps which are difficult to access and readily accessible Langmuir traps. In the RGT-B materials containing 0.1% of boron, the traps introduced by irradiation with 300-keV neon ions vanish on annealing in a vacuum (T=1800 K, t=1 min). Institute of Physical Chemistry, Russian Academy of Sciences. SINTEZ Scientific and Technical Center, Scientific-Research Institute of Electrophysical Apparatus. Graphite Scientific-Research Institute. National Scientific Center, Kharkov Physicotechnical Institute. Translated from Atomnaya énergiya, Vol. 82, No. 6, pp. 448–464, June, 1997.     

Thermal,structural, and radiological properties of irradiated graphite from the ASTRA research reactor – Implications for disposal

The release of Wigner energy from the graphite of the inner thermal column of the ASTRA research reactor has been studied by differential scanning calorimetry and simultaneous differential scanning calorimetry/synchrotron powder X-ray diffraction between 25 °C and 725 °C at a heating rate of 10 °C min⁻¹. The graphite, having been subject to a fast-neutron fluence from ∼10¹⁷ to ∼10²⁰ n cm⁻² over the life time of the reactor at temperatures not exceeding 100 °C, exhibits Wigner energies ranging from 25 to 572 J g⁻¹ and a Wigner energy accumulation rate of ∼7 × 10⁻¹⁷ J g⁻¹/n cm⁻². The shape of the rate-of-heat-release curves, e.g., maximum at ca. 200 °C and a fine structure at higher temperatures, varies with sample position within the inner thermal column, i.e., the distance from the reactor core. Crystal structure of samples closest to the reactor core (fast-neutron fluence >1.5−5.0 × 10¹⁹ n cm⁻²) is destroyed while that of samples farther from the reactor core (fast-neutron fluence <1.5−5.0 × 10¹⁹ n cm⁻²) is intact, with marked swelling along the c-axis. The dependence of the c lattice parameter on temperature between 25 °C and 200 °C as determined by Rietveld refinement for the non-amorphous samples leads to the expected microscopic thermal expansion coefficient along the c-axis of ∼ 26 × 10⁻⁶ °C⁻¹. However, at 200 °C, coinciding with the maximum in the rate-of-heat-release curves, the rate of thermal expansion abruptly decreases indicating a crystal lattice relaxation. The ¹⁴C activity in the inner thermal column graphite ranges from 6 to 467 kBq g⁻¹. The graphite of the inner thermal column of the ASTRA research reactor has been treated by heating to 400 °C for 24 h in a hot-cell facility prior to interim storage.     

Fuel burnup related change of the reactivity effect in IBR-2 from 1982 to 2006

The dependence of the change of reactivity on energy production is obtained from an analysis of IBR-2 operation during the period 1982–2006. It is shown that at the start of reactor operation, aside from the pure effect of burnup, additional positive effects which are most likely associated with fuel densification and structural change of the core material operate. These effects decrease with time and go to zero. After 40000 MW·h only the effect of pure burnup remains, and from this moment the reactivity decreases linearly with coefficient k_b = −4.3·10⁻⁵%/(MW·h). A formula is obtained for calculating the coefficient of energy release at any moment of operation of the reactor. __________ Translated from Atomnaya énergiya, No. 104, No. 3, pp. 147–152, March, 2008.     

Fission vacuum chambers and secondary-emission detectors as effective tools for measuring the characteristics of <Emphasis Type="Italic">γ</Emphasis> and neutron radiation from intense sources

The results of measurements of the flux of fast neutrons in the density range 2·10⁸–2·10¹⁹ sec^–1·cm^–2 and γ-ray dose rate in the range 2·10^–3–1·10⁹ Gy/sec in different operating regimes of pulsed nuclear reactors and accelerators are presented. The parameters of the delayed photon radiation are presented.     

Effect of the variation of properties on the working capacity of GR-280 reactor graphite

Data on the variance of the properties of GR-280 block reactor graphite are presented. The effect of the variance on the radiation-induced change in the properties, including on the characteristics of the shape change, is presented. In view of this, the working capacity of graphite is analyzed for masonry blocks of a uranium-graphite reactor, 2 figures, 4 tables, 20 references. Translated from Atomnaya énergiya, Vol. 88, No. 2, pp. 119–125, February, 2000.