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.     

Incorporation of bottoms from nuclear power plants into a matrix based on portland cement and silicic additives

The solidification of partially evaporated bottoms of RBMK and VVER with salt concentration 500–650 g/liter by compositional binders consisting of Portland cement and silicic additives – aerosil, microsilica, opoka, silicic acid, liquid glass, and diatomite is examined. The additions were used to obtain matrices that satisfy the requirements of safe storage of cemented radwastes. The partition coefficients of ¹³⁷Cs in partially evaporated bottoms are determined for all additives studied. The most effective additive for solidification of partially evaporated bottoms of VVER is diatomite. Matrices with diatomite have strength 50–81 kg/cm², the rate of leaching of ¹³⁷Cs ~ 10^–3–10^–4 g/(cm²·day) and the fill with respect to salts reaches 20.9 wt.%. On the solidification of partially evaporated RBMK bottoms the most effective hardening additives are aerosil and microsilica and the most effective sorbing additives are bentonite, opoka, and diatomite. The matrices so obtained have strength 59–93 kg/cm², ¹³⁷Cs leach rate ~ 10^–3–10^–4 g/(cm²·day) and contain to 25.1 wt.% salts.     

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.     

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.     

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.     

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.     

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.     

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.     

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.     

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.     

Compaction of Radioactive Thermal-Insulation Materials and Construction Debris by Melting in a Cold Crucible

A technology is proposed for reprocessing radioactive thermal-insulation materials and construction debris, produced at nuclear power plants and radiochemical facilities, by melting the wastes in a induction furnace with a cold crucible. The results show that the final volume of the thermal-insulation wastes can be decreased by a factor of 40 and that of the construction debris by a factor of 2.5. The high hydrolytic stability of the final materials produced as a result of melting (the ¹³⁵Cs and ⁹⁰Sr leach rate is less than 1·10⁻⁷ g/(cm²·day) allows these wastes subsequently to be stored and buried. Recommendations are developed for an apparatus-technological scheme and for electrotechnical equipment for a commercial facility. __________ Translated from Atomnaya Energiya, Vol. 99, No. 3, pp. 167–171, September, 2005.     

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.     

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.     

Analysis of the characteristics of <Superscript>238</Superscript>U, <Superscript>232</Superscript>Th, <Superscript>237</Superscript>Np,and <Superscript>231</Superscript>Pa radiators for fission ionization chambers for measuring the neutron flux density in the ITER diverter

Computational results, obtained by analyzing possible schemes of nuclear transformations of each of four threshold fission radiators ²³⁸U, ²³²Th, ²³⁷Np, and ²³¹Pa, for fission ionization chambers are presented. The influence of the nuclear reactions (n, ƒ), (n, γ), and (n, 2n) on the characteristics of fission ionization chambers is taken into account in the nuclear transformation schemes for all four radiators. The results are presented in the form of a dependence of the sensitivity of the fission ionization chambers on the neutron fluence in the range 10²¹–10²⁴ cm⁻². The effect of 0.2 and 1 g/cm² thick boron screens is examined. Translated from Atomnaya énergiya, Vol. 106, No. 1, pp. 42–47, January, 2009.     

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.     

In search of nuclear fusion in electrolytic cells and in metal/gas systems

It has been reported recently in the literature that unexpected thermal and nuclear effects (production of excess heat, neutrons, γ-rays, and tritium) can occur during the electrolysis of heavy water at palladium or titanium electrodes, or during temperature and pressure cycling of the titanium/deuterium gas system. We have attempted to reproduce some of these experiments. A variety of electrochemical cells having palladium cathodes in the form of wires, tubes, sheets, and rods have been used to electrolyze heavy water containing 0.1 mol.dm⁻³ LiOH, 0.1 mol.dn⁻³ LiOD or 0.5 mol.dm⁻³ D₃PO₄. Current densities of up to 200 mA.cm⁻² were applied. The mass of the palladium cathodes covered the range from 1–40 grams and the surface area varied from 8–140 cm². Neutron detection systems with low constant backgrounds were used to search for neutron emission during electrolysis. These included³He- and¹⁰BF₃-based detectors. After running some of the cells for more than 30 days, no neutron emission above background could be detected. This puts upper limits of 0.5 s⁻¹ and 2×10⁻²³ fus. D-D.s⁻¹ on the neutron emission and the fusion rate, respectively. A sensitive and accurate heat-flow calorimeter was built and used to monitor the energy balance of some of the cells during electrolysis. No unexpected heat effects were observed. This puts an upper limit of 0.13 W.cm⁻³ on the specific excess power. No enrichment of the electrolyte in tritium was evident after electrolysis. Experiments were also performed with the titanium/ deuterium gas system. These consisted of exposing titanium metal to a deuterium gas pressure of 40 atmospheres, lowering the temperature to −196°C, releasing the pressure and gradually warming the titanium to room temperature. No neutron emission above background was observed during these experiments, which puts upper limits of 0.5 s⁻¹ and 4×10⁻²⁵ fus.D-D.s⁻¹ on the neutron emission and fusion rate, respectively. Submitted toJournal of Fusion Energy as part of the Proceedings of the Workshop on Cold Fusion Phenomena held in Santa Fe in May 1989.     

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.     

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.     

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.