Radial distribution of the burnup and content of plutonium in VVÉR fuel pellets

The results of a study, using electron-probe x-ray-spectral microanalysis, of the radial distribution of the burnup and accumulation of plutonium in the fuel elements of power reactors and the research reactor MIR, irradiated to maximum burnup 70 MW·days/kg, are presented. The influence of the initial enrichment with ²³⁵U and cross-section-averaged burnup on the total accumulation and radial distribution of plutonium is analyzed. The special features of burnup of gadolinium-containing fuel are determined. __________ Translated from Atomnaya énergiya, Vol. 100, No. 1, pp. 3–8, January, 2006.     

Experience with uranium-erbium fuel at the Ignalinsk Atimic Power Plant

Conclusions Experience with uranium-erbium fuel assemblies at the Ignalinsk Atomic Power Plant confirms the possibility of removing additional absorbers and considerably increasing the depth of fuel burnup, while maintaining permissible values of the steam reactivity and other characteristics important for reactor safety. Conversion of the reactor to uranium-erbium fuel permits not only the maintenance of α_ϕ within specified limits but also compensation for its growth on account of replacement of the control rods. The effectiveness of uranium-erbium fuel up to degrees of burnup close to the design value (∼20 MW·day/kg) has been demonstrated. The good agreement of the theoretical predictions and the actual characteristics offers hope that the complete conversion of reactors to uranium-erbium fuel and appropriate removal of additional absorbers will yield high fuel burnup and considerable improvement in the economic performance of atomic power plants. Other means of increasing the fuel burnup in RBMK-1500 reactors might be to increase the fuel enrichment and erbium content, to use zirconium spacers, and to reduce the operational reserve of reactivity. Ignalinsk Atomic Power Plant. Kurchatovskii Institute Russian Scientific Center. Mashinostroitel'nyi Zavod Joint Stock Company. Scientific-Research and Design Institute of Energy Technology. A. A. Bochvar All-Russian Scientific-Research Institute of Nonferrous Materials. Translated from Atomnaya énergiya, Vol. 85, No. 2, pp. 91–97, August, 1998.     

MIR reactor tests of VVER fuel elements in cyclic power variation regimes

Tests of prefabricated VVER fuel elements burnup 50–60 MW·days/kg in regimes with cyclic power variation have been performed in a circuit setup of the MIR research reactor. The testing procedures are described, and the designs of the irradiation setups are presented. Some fuel elements are equipped with sensors for performing in-reactor measurements, which yielded the experimental data on the variation of the fuel element parameters during the tests (gas pressure, fuel temperature, length). Some results of post-reactor materials-engineering investigations are presented. All fuel elements remained airtight. __________ Translated from Atomnaya énergiya, Vol. 104, No. 2, pp. 80–84, February, 2008.     

Comparison of the results of calculations and post-reactor investigations of VVÉR-1000 fuel elements with burnup 49 MW·days/kg

The results of calculations performed with the PINw99, TRANSURANUS (V1M1V03), and TOPRA-2 computer programs are compared with data obtained from post-reactor investigations of fuel elements which operated for four years in the No. 1 unit of the Zaparozh’e nuclear power plant with a VVéR-1000 reactor to burnup ≈ 49 MW·days/kg. The initial data are analyzed, and a comparison is made of the computed and experimental elongation of the fuel elements (49 fuel elements), the yield of gaseous fission products and the subcladding pressure (35 fuel elements), and the decrease of cladding diameter and fuel-cladding gap width. It is shown that these computer programs can be used to calculate VVéR fuel elements. __________ Translated from Atomnaya énergiya, Vol. 101, No. 6, pp. 413–420, December, 2006.     

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.     

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.     

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.     

Optimization of the thorium-uranium regime in CANDU heavy-water reactors

Two operating regimes of a heavy-water power reactor operating in a thorium-uranium fuel cycle are examined: accumulation of the required amount of ²³³U and self-fueling with ²³³U fuel. The parameters of 10 variants of the core-lattice cells of a heavy-water power reactor are calculated. The lattice spacing is kept constant and the variants differ by the number of channels, containing fuel elements or targets, and the ²³³U content in ThO₂. Combined channels containing fuel elements and a target at the same time are examined. Preference is given to a cell variant where a cell contains three channels with fuel elements and one channel with targets. For this variant, the fuel burnup is ∼8 MW-days/kg. A large increase of the burnup, i.e., decrease of the amount of reprocessed targets, can be achieved by decreasing the minimum reactivity excess and also by changeing and increasing the complexity of the technology. For example, a large effect is expected from using combined fuel assemblies. In so doing, provisions must be made for performing complicated operation of disassembling highly active fuel assemblies consisting of fuel elements and targets. __________ Translated from Atomnaya énergiya, Vol. 101, No. 5, pp. 327–336, November, 2006.     

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.     

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.     

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.     

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.     

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.     

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.     

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.     

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.     

50 years after the start-up of BR-2 — The first fast reactor in Europe

February 14, 2006 is the 50th anniversary of the start-up of the first fast reactor in Europe — the BR-2. This was an important step in the development of fast reactors for nucler power production. Experience in operating a core under conditions with cramped bundles of fuel elements and a liquid-metal coolant with heat flows through the surface of the claddings of about 1.2·10⁶ kJ/(m²·h) was gained for the first time on the BR-2 reactor. Investigations performed on the BR-2 of the breeding ratio showed that BR = 1.8 ± 0.3. The first experience of physical start-up of a fast reactor, determining the efficiency of the reactivity control organs and their calibration, was gained. __________ Translated from Atomnaya énergiya, Vol. 100, No. 3, pp. 178–183, March, 2006.     

Temperature and burnup reactivities and operational lifetime for the submersion-subcritical, safe space (S4) reactor

The He–Xe gas-cooled, S4 reactor has a sectored, Mo–14%Re solid core for avoidance of single point failures in reactor cooling and Closed Brayton Cycle (CBC) energy conversion. The reactor core is loaded with UN fuel and each of its three sectors is thermal-hydraulically coupled to a separate CBC loop and radiator panels. The solid core minimizes voids, and the BeO reflectors are designed to easily disassemble upon impact, ensuring that the bare S4 reactor is sufficiently subcriticial when submerged in wet sand or seawater and flooded with seawater, following a launch abort accident. Spectral shift absorber (SSA) additives in the core and thin SSA coatings on the outer surface of the core can also be used to ensure subcriticality in such an accident. This paper investigates the effects of various SSAs (Re, Ir, Eu-151, B-10 and Gd-155) on the temperature and burnup reactivity coefficients and the operating lifetime of the S4 reactor at a steady thermal power of 550 kW. The calculations of the burnup, reactivity feedback coefficient used a mixture of the top 10 light and top 10 heavy fission products plus Sm-149 and are performed for isothermal reactor core and reflector temperatures of 1200 and 900 K. In this fast spectrum space reactor, SSAs markedly increase fuel enrichment and decrease the burnup reactivity coefficient, but only slightly decrease the temperature, reactivity feedback coefficient. With no SSAs, the UN fuel enrichment is lowest (58.5 wt.%), the temperature and burnup reactivity coefficients are the highest (−0.2709 ¢/K and −1.3470 $/at.%), and the estimated operating lifetime is the shortest (7.6 years). The temperature and burnup reactivity coefficients decrease to −0.2649 ¢/K and −1.0230 $/at.%, and the operating lifetime increases to 8.3 years when rhenium additives are used. With europium-151 and gadolinium-155 additions, fuel enrichment (91.5 and 94 wt.%) and operating lifetime (9.9 and 9.8 years) are the highest and both the temperature reactivity feedback coefficient (−0.2382 and −0.2447 ¢/K) and the burnup reactivity coefficient (−0.9073 and −0.8502 $/at.%) are the lowest.     

Radiation damage in beryllium at 70–440°C and neutron fluence (0.3–18)·1022 cm−2 (E n > 0.1 MeV)

The results of investigations of the damage to samples of five brands of beryllium (TV-56, TV-400, TV-30, TIP, and DIP), prepared using the hot-pressing and extrusion technology and also hot isostatic pressing, are presented. The beryllium samples were irradiated in the channels of the core of an SM reactor and as part of the photoneutron source of the BOR-60 fast reactor at 70–440°C and neutron fluence (0.3–18)·10²² cm⁻² (En > 0.1 MeV). The experimental program included mechanical tests for stretching and compression, measurements of the microhardness, swelling and specific thermal conductivity, as well as a study of the microstructure. The experimental results are used to construct the dose dependences of the change of the physicomechanical characteristics of beryllium at different irradiation temperatures and the trends in these dependences are analyzed. __________ Translated from Atomnaya énergiya, Vol. 101, No. 4, pp. 289–296, October, 2006.     

<Superscript>59</Superscript>Ni Accumulation in Corrosion-Resistant Steel under the Action of Neutron Flux

A γ-ray line with energy E_γ = 11.3 MeV was detected during an experiment, performed on a nuclear reactor, investigating the characteristics of the energy spectrum of γ-rays. The most likely source of this line is radiative capture of thermal neutrons by ⁵⁹Ni nuclei, which accumulated in the corrosion-resistance steel as a result of the more than 20 years of irradiation in the reactor, via the reaction ⁵⁸Ni(n, γ)⁵⁹Ni. It was found that for thermal-neutron fluence 10²¹ cm⁻² the ⁵⁹Ni concentration is 0.47% of the ⁵⁸Ni concentration. __________ Translated from Atomnaya Energiya, Vol. 99, No. 4, pp. 268–272, October, 2005.