In-reactor investigations of the creep of structural materials

Conclusions 1. A series of in-reactor tests was performed on a sample used to study radiation creep in 00X16H15M3B steel, XHM1 chrome-nickel alloy, the zirconium based alloys é110 and é635, and the vanadium-based alloy BTX8. The radiation creep modulus (in units of Pa⁻¹·(displacements/atom)⁻¹ equals 1.7·10⁻¹¹ for 00X16H15M3B steel, 4.6·10⁻¹¹ for XHM alloy with fluence up to 2.3·10²⁰ cm⁻² and 1.6·10⁻¹¹ for a fluence above 1·10²¹ cm⁻², (4.6–4.9)·10⁻¹¹ for é110 alloy, and 1.8·10⁻¹¹ for é635 alloy. For the alloy BTX8, at stresses below half the yield point and t=450°C, the modulus equals 3.3·10⁻¹² Pa⁻¹·(displacements/atom)⁻¹. At a higher stress, the deformation rate of the alloy increases progressively. 2. In the investigation of the temperature dependence of in-reactor creep of the alloy é110, it was found that at 350–370°C and higher, the thermal creep makes the predominant contribution to deformation. In the experimental range 370–455°C, the thermal activation energy of in-reactor creep was determined to be 36 ± 8 kcal/(g·atom). At temperatures below 350°C the creep of the alloy é110 is a temperature-independent radiation-stimulated process. 3. In the case of tests of zirconium alloys, a previously unobserved phenomenon of periodic rapid deformation of the material against the background of creep at stresses even well below the yield point of the irradiated material was discovered. The effect was manifested at a temperature of about 230°C. As the temperature increases up to 290°C and higher, no plastic movements are observed. Translated from Atomnaya énergiya, Vol. 80, No. 5, pp. 386–391, May, 1996.     

Effect of the composition of radiation on the radiation damage to graphite

A statistical analysis is performed of the results on the determination of the critical neutron fluence in MR, SM-2, and BOR-60 with different irradiation temperature. It is shown that the critical neutron fluence depends not only on the irradiation temperature but also, and to an even greater extent, on the radiation composition factor (ratio of the neutron and γ-ray flux densities). Thus the critical neutron fluence for irradiation at 600°C in MR (radiation composition factor 0.13) is 17·10²¹ cm⁻² and in SM-2 (radiation composition factor 0.1) 11·10²¹ cm⁻² at the same temperature. When the same graphite is irradiated in the region of the outer corner of a working block of RBMK, where the radiation composition factor is 0.55, it is expected that the critical neutron fluence will be 31.7·10²¹ cm⁻². In summary, taking account of the effect of γ-radiation introduces substantial corrections: the experimental results obtained in research reactors are found to be at least a factor of 2 too low. This gives hope of substantiating the substantial increase in the service life of the RBMK graphite masonry. 3 figures, 8 references. Scientific-Research and Design Power-Engineering Institute. State Science Center—Scientific-Research Institute of Nuclear Reactors. Translated from Atomnaya énergiya, Vol. 87, No. 1, pp. 24–28, July, 1999.     

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.     

Inclusion of Radioactive Ion-Exchange Resins in Inorganic Binders

This article is devoted to the inclusion of ion exchange resins in portland, portland blast-furnace, and alumina cements. The degree to which the solidified products are filled with respect to dry resin reaches 7–10, 12, and 18.9–19.7%, respectively, with adequate strength being maintained (at least 5 MPa); the cesium diffusion coefficients are 9.3·10⁻⁴, 1.2·10⁻⁴, and 7.2·10⁻⁵ cm²/day with the normative value 6.7·10⁻⁴ cm²/day. When 10 mass% clay is added to alumina cement, the diffusion coefficient of cesium decreases to 5.1·10⁻⁶ cm²/day, and the volume of the wastes increases by not more than a factor of 1.5 on solidification. __________ Translated from Atomnaya Energiya, Vol. 99, No. 3, pp. 171–177, September, 2005.     

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.     

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.     

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.     

Decrease of Radiation Swelling of Explosion-Worked Graphite

Radiation swelling (change of the unit-cell parameters) of reactor graphite and diamond is measured as a function of the perfection of the crystal lattice. The initial powders are irradiated together with powders which have been exposed to an explosive wave with nominal pressure ∼40 GPa. Such treatment results in up to 100% broadening of the diffraction lines. In addition, ultrasmall-grain diamond is used. Irradiation is conducted in a BOR-60 reactor up to fluence 1·10²² cm⁻² at 390 and 475°C. The investigation shows that the distortion of the crystal lattice and change in the size of crystallites can decrease by factors of 1.6–5 the growth of the unit-cell parameters of graphite and diamond. __________ Translated from Atomnaya Energiya, Vol. 99, No. 1, pp. 43–47, July 2005.     

Kinetics of the leaching of90Sr from fuel particles in soil in the near zone of the chernobyl nuclear power plant

A method is presented for calculating the fraction of⁹⁰Sr included in fuel particles in soil. Data concerning the change in forms of the occurrence of⁹⁰Sr in different soils in the 30-km zone, at different distances from the Chernobyl nuclear power plant, were used to obtain the kinetic characteristics of its leaching: the first-order rate constant and the normalized rate of solution. Depending on the direction and distance from the nuclear power plant, the first-order leaching rate constant varies from 3·10⁻⁵ to 2·10⁻³ days⁻¹ and the normalized rate of solution of the fuel matrix varies from 1·10⁻⁵ to 6.1·10⁻⁴. It was not found possible to clearly identify the influence of the distance from the nuclear power plant on the leaching rate in the northern and western sectors. In contrast, in the southerly and south-easterly directions a clear tendency was observed for the leaching rate to increase with increasing distance from the nuclear power plant. Taifun Scientific Production Enterprise. Translated from Atomnaya énergiya, Vol. 86, No. 2, pp. 129–134, February, 1999.     

Characteristics of heat transfer of models of core and steam generator surfaces with regulation of impurity content in a loop with lead coolant

The results of experimental investigations of the heat transfer by lead coolant in the ring-shaped gaps of a circulation loop during monitored and controlled mass transfer and mass exchange of oxygen and impurity are presented. The investigations were performed in a loop with circulation of lead coolant at temperature of 450–550°C, average velocity 0.1–1.5 m/sec, Peclet number 500–6000, and heat flux 50–160 kW/m². The oxygen content in the loop was varied from the value for thermodynamic activity 10⁻⁵–10⁰ to saturation and above with formation of lead oxide deposits on the heat transfer surface. The processes in a non-isothermal liquid metal loop with heating (core) and cooling (steam generator) experimental sections simulate the dependence of the heat transfer characteristics in the loop on the impurity mass transfer. __________ Translated from Atomnaya énergiya, Vol. 104, No. 2, pp. 74–80, February, 2008.     

Calculation of the effective delayed-neutron fraction and prompt-neutron lifetime in IBR-2M

The main parameters of IBR-2M are presented: the effective delayed-neutron fraction β_eff and the promptneutron lifetime τ, calculated using the DORT two-dimensional multigroup neutron-transport compute code and the SCALE4 code with a system of multigroup nuclear constants. For a regular IBR-2M regime β_eff = 0.00216 ± 0.00007, τ = (6.5 ± 0.5)·10^–8 sec, the delay-neutron value γ = 0.980, the prompt-neutron decay constant in the critical state α = 3.5·104 sec^–1. The calculations showed that the effective delayed-neutron fraction for IBR-2M is identical, within the error limits, to the measured value for IBR-2, the prompt-neutron lifetime is approximately 5% longer (β_eff = 0.00216, τ = (6.2 ± 0.2)·10^–8 sec). It is shown that β_eff and τ increase somewhat as the IBR-2 core size increases in the radial direction.     

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.     

Use of epitaxial gallium arsenide in detectors

The results of investigations of semiconductor detectors on the basis of epitaxial layers of gallium arsenide for detecting x rays and low-energy radiation are examined. It is shown that epitaxial layers ranging in thickness from 60 to 300 μm with current carrier density ≤5·10¹³ cm⁻³ and electron mobility ≥6000 cm²/(V·sec) at 300 K hold promise for such detectors. A new type of photovoltaic x-ray detector based on the epitaxial structures p⁺-n-nⁱ-n⁺ GaAs is described. Such detectors possess high charge collection efficiency with zero bias at room temperature and can operate in two regimes — counting and current integration — and will substantially expand the dynamical range of image formation when used in scanning systems. __________ Translated from Atomnaya énergiya, Vol. 103, No. 5, pp. 322–326, November, 2007.     

Medium-scale experiment with production of prototypical ~100 kg melt and an analysis of its interaction with concrete

The results of a medium-scale experiment in which a prototypical melt is produced by combustion of a chemically active substance in the course of the reaction 2Fe₂O₃ + 3Zr = 3ZrO₂ + 4Fe + 2840 kJ/kg in a ~6·10⁻² m³ concrete container are presented. It is shown that the ~100 kg melt, whose temperature is 2700–3200 K, so obtained produces heat fluxes 100–150 kW/m2 into the walls and bottom of the concrete container for ~10 min. The ablation of the walls of the concrete container was 2.5–3 cm at the completion of the experiment.     

Determination of the neutron flux and <Emphasis Type="Italic">γ</Emphasis> radiation in the core of operating and shut-down reactors using quartz glasses and element monitors

The neutron fluxes and the intensity of γ radiation are measured in 26 channels of a VVR-SM reactor and its thermal column. The fast neutron fluxes in the channels are determined using Ni, Fe, Co, Au, and Mn element monitors with different threshold energies, together with a theoretical calculation using the MCNP-4C program. The energy distribution of the neutron flux inside the fuel assembly is obtained for selected channels around the core. The flux of neutrons with energies >1 MeV is in the range (0.5–43)·10¹² cm⁻²sec⁻¹, depending on the location of the channel. A linear correlation is discovered between the induced optical absorption at the 215 nm line (E′ center) of SiO₂–BaO glass and the fast neutron flux in the channels. The γ-ray intensity in the thermal channel is estimated for the reactor during operation (∼38.4 Gy/sec) and 24 hours after it is shut down (∼24.7 Gy/sec) using the E′ centers induced in pure quartz glasses. The observed difference in the efficiency with which oxygen defects are formed during dry and wet irradiation of glass owing to the radiolysis of water must be taken into account when developing radiation technology and during the burial of radioactive waste. Translated from Atomnaya énergiya, Vol. 105, No. 3, pp. 160–164, September, 2008.     

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.     

241Am excretion with the urine for personnel at a radiochemical plant

Conclusions We have obtained for the first time the coefficients of excretion of²⁴¹Am after chronic inhalation in personnel working at a radiochemical plant. We found that the level of excretion in the urine is affected by different diseases, which result in the appearance of pronounced morphological changes in the liver tissue, as established from pathological-anatomical data. The levels of excretion of²⁴¹Am long after the start or termination of contact vary depending on the seriousness of the pathological processes in the liver over the range (0.63–6.2)·10⁻⁵ day⁻¹. To obtain dosimetric estimates of the²⁴¹Am content in the body according to the level of excretion in the urine, the coefficient 1.2·10⁻⁵ day⁻¹, established for essentially healthy people, must be used. We thank A. P. Nifatov for performing the morphological investigations of the organ tissues. Affiliate No. 1 of the Institute of Biophysics, Ministry of Health of the Russian Federation. Translated from Atomnaya énergiya, Vol. 77, No. 1, pp. 69–72, July, 1994.     

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.     

Erosion of Pyrolytic Graphite and Ti-doped Graphite due to High Flux Irradiation

The erosion of pyrolytic graphite and titanium doped graphite RG-Ti above 1,780K was investigated by 5keV Ar beam irradiation with the flux from 4x10¹⁹ to 1x10²¹ m^?2·s^?1. The total erosion yields were significantly reduced with the flux. This reduction would be attributed to the reduction of RES (radiation enhanced sublimation) yield, which was observed in the case of isotropic graphite with the flux dependence of RES yield of φ^?0.26 (φ: flux) obtained in our previous work. The yield of pyrolytic graphite was roughly 30% higher than that of isotropic graphite below the flux of 10²⁰ m^?2·s^?1 whereas each yield approached to very close value at the highest flux of 1x10²¹ m^?2·s^?1. This result indicated that the effect of graphite structure on the RES yield, which was apparent in the low flux region, would disappear in the high flux region probably due to the disordering of crystal structure.

In the case of irradiation to RG-Ti at 1,780K, the surface undulations evolved with a mean height of about 3μm at 1.2×10²⁰ m^?2·s^?1, while at higher flux of 8.0×10²⁰ m^?2·s^?1 they were unrecognizable. These phenomena can be explained by the reduction of RES of graphite parts excluding Tic grains.     

Correcting the isotopic composition of regenerated uranium with respect to <Superscript>232</Superscript>U by a centrifuge method with introduction of a carrier gas

Theoretical and experimental studies dealing with correcting the isotopic composition of regenerated uranium with respect to ²³²U by a centrifuge method with introduction of a carrier gas are reported. In order to increase the efficiency of separating ²³²U from the spent uranium and reduce the loss of ²³⁵U, the use of a carrier gas is proposed – the gaseous compound ¹²C₈H₃F₁₃, which is inert to uranium hexafluoride, and whose molecular weight, M_c = 346 amu, matches that of ²³²UF₆. Freon, C₈H₃F₁₃, is shown not to decompose during operation in the rotor of a centrifuge or to interact with the centrifuge material. The measured absorption parameters of freon on sodium fluoride NaF confirm the feasibility of efficient separation of a mixture of uranium hexafluoride and freon with return of the freon to the separation process. It is shown that introducing a carrier gas into the centrifuge technology can yield some new results: lowering the radioactivity of the commercial product, normalizing the overall radiation situation during production, increasing the recovered ²³⁵U in the commercial product, and reducing the volume of radioactive waste. The recovery of ²³⁵U in the commercial product can be increased to 99% or more. Then the ²³²U content in the commercial product is ∼2·10⁻⁸% or a factor of 10 less than the maximum allowable content of 2·10⁻⁷%. Translated from Atomnaya énergiya, Vol. 105, No. 3, pp. 150–156, September, 2008.