Effect of the nonuniformity of graphite samples on their shape change under irradiation

It is shown that the use of the initial characteristics of samples makes it possible to decrease substantially the variance and the bias in the determination of the critical neutron fluence for GR-280 RBMK graphite. It is recommended that a graphite model that includes neutron fluence and the initial value of the elastic modulus be used as an approximation for dimensional changes of graphite samples. 2 figures, 2 tables, 3 references. Russian Science Center “Kurchatov Insitute”. Translated from Atomnaya énergiya, Vol. 87, No. 1, pp. 28–33, July, 1999.     

Radiation Resistance of Reactor Graphite with Enhanced Operating Properties

Neutron fluence dependences of the dimensional changes, the thermal expansion coefficient, the dynamic elastic modulus, and the maximum strength of samples of high-strength structural graphite GR-1 and GSP-50 after irradiation in BOR-60 at 360–400°C up to fluence 6.5·10²⁶ m^–2 (E > 0.18 MeV) were determined. The shape change of graphite reached the secondary-swelling stage, and the state of the material is characterized by disseminated fracturing.The influence of the initial density in the range 1.58–1.9 g/cm³ is determined. It is shown that GSP-50 graphite, based on pyrolytic carbon matrix, possesses a higher radiation resistance than GR-1 graphite based on a composite filler and granular binder.     

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.     

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.     

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.     

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.     

Radiation resistance of high-density uranium—molybdenum dispersion fuel for nuclear research reactors

The results of radiation tests are discussed and the character of the failure of fuel compositions and the operability of fuel elements is analyzed as a function of the type of fuel and the irradiation conditions. The intense interaction of the fuel with the matrix material is considered as the main factor limiting the operability of fuel elements in power-dense high-flux nuclear reasearch reactors. It is concluded that low-enrichment high-density uranium—molybdenym fuel can provide reliable operation of dispersion fuel elements in low-and medium-power research reactors. Such fuel can be used in power-dense high-flux research reactors if the fuel load is decreased below the maximum admissible amount, the compatibility of the uranium—molybdenum alloy with an aluminum matrix is radically improved, or fuel elements with a different construction, for example, monolithic, are used. __________ Translated from Atomnaya énergiya, Vol. 100, No. 1, pp. 35–44, January, 2005.     

Thermophysical properties of irradiated uranium-zirconium fuel

The dependence of the thermophysical properties of metallic nuclear fuel — the alloy Zr-40U — in a wide temperature range on the amount of fission products accumulated is presented. Non-irradiated and irradiated samples with different degree of accumulation of fission products — 0.4, 0.6, and 0.9 g/cm³ — are investigated. The specific heat is measured in the range 50–1000°C, the temperature diffusivity is measured in the range 300–1000°C, and the variation of the dimensions and density of the samples on heating is also investigated. The thermal conductivity in the range 50–1000°C is calculated on the basis of the experimental data. __________ Translated from Atomnaya énergiya, Vol. 108, No. 1, pp. 6–9, January, 2008.     

Raman spectroscopy of carbon dust samples from NSTX

The Raman spectrum of dust particles exposed to the NSTX plasma is different from the spectrum of unexposed particles scraped from an unused graphite tile. For the unexposed particles, the high energy G-mode peak (Raman shift ∼ 1580 cm⁻¹) is much stronger than the defect-induced D-mode peak (Raman shift ∼ 1350 cm⁻¹), a pattern that is consistent with Raman spectrum for commercial graphite materials. For dust particles exposed to the plasma, the ratio of G-mode to D-mode peaks is lower and becomes even less than 1. The Raman measurements indicate that the production of carbon dust particles in NSTX involves modifications of the physical and chemical structure of the original graphite material. These modifications are shown to be similar to those measured for carbon deposits from atmospheric pressure helium arc discharge with an ablating anode electrode made from a graphite tile material. We also demonstrate experimentally that heating to 2000-2700 K alone cannot explain the observed structural modifications indicating that they must be due to higher temperatures needed for graphite vaporization, which is followed either by condensation or some plasma-induced processes leading to the formation of more disordered forms of carbon material than the original graphite.     

Irradiation induced dimensional changes in graphite: The influence of sample size

Dimensional changes in irradiated anisotropic polycrystalline GR-280 graphite samples as measured in the parallel and perpendicular directions of extrusion revealed a mismatch between volume changes measured directly and those calculated using the generally accepted methodology based on length change measurements only. To explain this observation a model is proposed based on polycrystalline substructural elements – domains. Those domains are anisotropic, have different amplitudes of shape-changes with respect to the sample as a whole and are randomly orientated relative to the sample axes of symmetry. This domain model can explain the mismatch observed in experimental data. It is shown that the disoriented domain structure leads to the development of irradiation-induced stresses and to the dependence of the dimensional changes on the sizes of graphite samples chosen for the irradiation experiment. The authors derive the relationship between shape-changes in the finite size samples and the actual shape-changes observable on the macro-scale in irradiated graphite.     

Some aspects of nuclear and radiological safety of operating a spent nuclear fuel repository

A model of an irregular situation in a spent nuclear fuel repository with the introduction of excess reactivity into the system, consisting of containers with spent fuel assemblies and water, is examined. The neutron kinetics of a critical system is calculated taking account of the thermohydraulics of the system. The character of the flow of a short-time self-sustained chain reaction — “neutron burst” — is described. It is found that an excursion of the system in the range of reactivity introduction rates examined will result in heating of the system and self-quenching of the chain reaction by negative reactivity effects with respect to fuel temperature. Intense fluxes of fission neutrons and prompt gamma rays, accompanying a self-sustained chain reaction, are formed in the excursion process. A mixed neutron and gamma ray field near the system considered is investigated. __________ Translated from Atomnaya énergiya,Vol. 104, No. 3, pp. 141–147, March, 2008.     

Longitudinal instability of a continuous beam in a proton synchrotron

The results of a calculation of the threshold for longitudinal coherent instability of a continuous beam circulating in a proton synchrotron and constant energy are presented. A beam in an intermediate state — not bunched into bunches but not uniform over the azimuth — is said to be continuous. Such a beam moves outside (possibly, near) empty rf separatrices of the longitudinal sinusoidal electric field and has a ribbon portrait in the longitudinal phase plane. The computational method is used for the U-70 synchrotron at the Institute of High Energy Physics, where in the course of stochastic slow extraction of a circulating beam is continuous in the sense indicated. It bends around empty separatrices of the 200-MHz accelerating field and for several seconds interacts with the electromagnetic fields of the working oscillations of the disconnected resonators of the main accelerating system with frequency 5.5–6 MHz, which can result in a loss of longitudinal stability. __________ Translated from Atomnaya énergiya, Vol. 104, No. 1, pp. 26–33, January, 2008.     

Simulation of high-temperature swelling of uranium dioxide and deformation behavior of a fuel element

Models and computer codes, developed based on them, for simulating the swelling of uranium dioxide (BARS) and the stress-deformation state of a fuel element (SDS) under high-temperature irradiation are presented. It is shown that when developing a design for high-temperature fuel elements and validating their serviceability the quantitative indicator required for the swelling of uranium dioxide in the range ≥1400°C is the change in the external dimensions of the fuel caused by constant formation and growth of bubbles containing gaseous fission products during irradiation. The results of computational investigations using the models indicated are examined. These results eliminate the inconsistency of the data on the effect of the main operating parameters — the temperature and burnup — on the radiation characteristics and service life behavior of a fuel element. It is shown that the central channel in the fuel kernel and strengthening of the cladding improve the dimensional stability fuel elements. __________ Translated from Atomnaya énergiya, Vol. 103, No. 3, pp. 172–179, September, 2007.     

Nuclear central heating — prospects and solutions

Large-scale development of nuclear central heating — a radical expansion of a sphere of application, large increase of cost-effectiveness and self-financing of the construction of nuclear sources of energy, increase of their fraction in the base part of the load schedule, and large-scale displacement of fossil fuel — is validated. Suggestions for a program for developing nuclear heat and power plants are examined. It is shown that the power generating units of nuclear heat and power plants must satisfy specific requirements, which requires developing specialized reactor systems. The main technical and economic characteristics of an innovative simplified boiling water reactor VK-300, specially designed for central heating power generating units, the parameters of a central heating power generating unit with VK-300, and the results of validation of investments in the construction of the VK-300 nuclear heat and power plant in Arkhangel’sk are presented. __________ Translated from Atomnaya énergiya, Vol. 103, No. 1, pp. 36–40, July, 2007.     

Experience with60Co production in BN-600

The basic results of test of two methods of producing ⁶⁰ Co in BN-600 reactors—in fuel assembly type experimental irradiation setups in the side breeding zone and in experimental compensating control rods in the core—are reported. It is confirmed that the production of ⁶⁰ Co with specific activity at the level 100 kCi/g without disruption of the normal operating conditions of the power-generating unit is possible in the core and side breeding zone of a high-power fast-neutron reactor. 1 table. I. V. Kurchatov Beloyarsk Nuclear Power Plant. Translated from Atomnaya énergiya, Vol. 86, No. 3, pp. 216–219, March 1999.     

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.     

Investigation of fuel structure and composition after irradiation in VVÉR-440 to burnup 70 MW·days/kg

The results of structural investigations performed on fuel and fission products — neodymium, xenon, and cesium — along the radius of a fuel kernel after irradiation in VVéR-440 to burnup 70.2 MW·days/kg are presented. The radial distribution of neodymium is used to calculate the radial distribution of burnup and the accumulation of xenon and cesium. It is shown that a decrease of the xenon content in the fuel matrix as compared with the amount formed over the irradiation time is observed over the entire cross section of the pellet and is due to complete or partial fuel recrystallization occurring predominately along the boundaries of the initial grains and characterized by the formation of a fine-grain structure together with submicron and micron pores. __________ Translated from Atomnaya énergiya, Vol. 101, No. 4, pp. 286–289, October, 2006.     

From laboratory no. 2 of the USSR Academy of sciences to the Russian Science Center “Kurchatov Institute”

A detailed exposition of the history of the creation and development and the present status of one of the leading nuclear centers in the world—the Kurchatov Institute—is presented. The Institute was created in April 1943 during the height of the Second World War to solve a military problem —the development of nuclear weapons in our country. Having solved the basic problem in a relatively short time, the Kurchatov Institute switched to utilizing its powerful atomic potential to solve a wide range of scientific, technological and industrial problems in various fields of science and technology. The development of nuclear power and the utilization of atomic energy in various media—air, oceans, and space—as well as the development of thermonuclear studies and very important fundamental studies in various fields of modern science are all happening here. Using its unique scientific potential, the Kurchatov Institute has been able to establish the starting points everywhere. Russian Science Center “Kurchatov Institute.” Translated from Atomnaya énergiya, Vol. 86, No. 4, 247–260, April, 1999.     

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.     

Investigation of the radiation characteristics of the masonry in the ai uranium–graphite reactor

The results of investigations of the nuclear safety of the masonry in the AI uranium–graphite (Industrial Association Mayak) reactor are presented. It is concluded on the basis of these results that the masonry is nuclear safe and a radiation certificate is composed. The radiation examination made it possible to determine the level, composition, and distribution of the radioactive contamination of the masonry as well as the level and distribution of the neutron and γ radiation, and to construct a forecast of the change in the activity of radionuclides in graphite as a function of the holding time. These data are necessary for safety analysis and for making decisions about the subsequent stages of decommissioning of the reactor. Translated from Atomnaya énergiya, Vol. 105, No. 5, pp. 266–269, November, 2008.