“Romashka” reactor-converter

On August 14, 1964, the “Romashka” reactor-converter—the world's first setup with direct (thermoelectric) conversion of heat from²³⁵U fission into electricity—was started up at the I. V. Kurchatov Institute of Atomic Energy and service-life nuclear tests were begun. In 1999, 35 years have passed since the “Romashka” reactor-converter was first started up and trials began. “Romashka” successfully operated for about 15000 hours and generated ∼6100 kW·h of electricity. The startup and successful tests of the “Romashka” reactor-converter recieved in its time high marks from the world scientific community and they demonstrated that the world's first operating nuclear reactor-converter, capable of generating electricity without the participation of any moving working bodies or mechanisms, was developed in the short time of about 3 years in the Soviet Union, and it was shown experimentally that the system was capable of prolonged operation. The technology created in the process of developing the “Romashka” reactor-converter and, specifically, the thermoelectric silicon-germanium transducers have found application in operating space nuclear (reactor and isotopic) power systems. 3 figures. Russian Science Center “Kurchatov Institute”. Translated from Atomnaya énergiya, Vol. 88, No. 3, pp. 176–183, March, 2000.     

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.     

High-temperature heat pipes for power modules of space nuclear power systems

The results of an experimental analysis of the technology for producing high-temperature heat pipes and choosing their structure so as to ensure a long service life and high performance characteristics required for the power modules of space nuclear power systems with an out-of-core energy conversion system are examined. It is estimated that the operation of molybdenum-lithium heat pipes, fabricated using the technology developed, for the power modules of such nuclear power systems will be stable for >10⁵ h. 2 figures, 1 table, 13 references. State Enterprise Krasnaya Zvezda. Translated from Atomnaya énergiya, Vol. 89, No. 1, pp. 82–86, July, 2000.     

Limiting thermionic energy conversion characteristics

The limiting current–voltage characteristics, electronic efficiency, and the efficiency taking account of the radiative heat losses from the emitter for a diode thermionic energy converter operating in the regime of a low-voltage arc discharge are calculated. It is shown that for low-temperature energy converters (emitter temperature 1300–1600 K, current density 2–4 A/cm²) the efficiency does not exceed 20% even at low collector temperature – half the emitter temperature. The requirements for the work function of the electrodes that must be satisfied to obtain the limiting characteristics are determined. The relations and computational results obtained can be used to develop advanced thermionic power-generation facilities.     

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.     

Preparation and removal for reprocessing of spent VVR-2 and or nuclear fuel from the Russian Science Center Kurchatov Institute

The details of the preparation and removal of spent nuclear fuel from the Institute’s VVR-2 and OR research reactors for chemical reprocessing are presented. The spent fuel is represented by fuel assemblies which have different shapes and contain EK-10 fuel elements with similar construction and UO₂–Mg 10% enrichment kernels or S-36 fuel elements with U–Al alloy kernels with 36% enrichment. The storage conditions for the spent fuel are described. The details of the procedures developed to identify fuel assemblies by type of fuel elements are presented. The choice of the TUK-19 shipment container for loading and transporting spent fuel for reprocessing is validated. The details of the loading of spent fuel assemblies into TUK-19 are described; these operations are performed by workers under a protective layer of water in a handling room specially designed for such purposes. Translated from Atomnaya énergiya, Vol. 106, No. 4, pp. 201–209, April, 2009.     

Calculation of ρR-parameter and Energy Gain for Aneutronic Fusion in Degenerate P–11 B Plasma

Aneutronic fusion reactions are more safe and clean than the other reactions. One of the important candidate for these reactions is P–¹¹ B. This reaction in characteristic conditions creates degenerate plasma. In a Fermi-degenerate plasma, the electronic stopping of a slow ion is smaller than given by the classical formula, because some transitions between the electron states are forbidden. The bremsstrahlung losses are then smaller, so that the nuclear burning of an aneutronic fuel is more efficient. Practical obstacles in this regime that must be overcome before net energy can be realized include the compression of the fuel to an ultra dense state and the creation of a hot spot. In this paper, ρR parameter (Lawson’s criterion) and energy gain for P–¹¹ B are given.     

Detection of trace quantities of10B in biosamples using track detectors

A new variant of a solid-state detector adapted for estimating the¹⁰B content in biological tissue was developed. A new polymeric composition based on oligoester acrylates was used as the detector. The number of tracks is a linear function of the boron content in the concentration range 0.5–200 μg¹⁰B/g inbiological sample samples with mass 100–500 μg. The detector can be recommended for deterning¹⁰B in biological tissues, mineralogy, and agrochemistry, 3 figures, 2 references. Translated from Atomnaya énergiya, Vol. 88, No. 3, pp. 194–197, March, 2000.     

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.     

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.     

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.     

New-generation space thermionic nuclear power systems with out-of-core electricity generating systems

The basic problems of the development of long service life thermionic nuclear power systems for space applications for supplying power to transport power modules, intended for placing spacecraft in a geostationary orbit using electrorocket motors and for providing prolonged operation of onboard systems for up to 10–15 years, are examined. Concepts for power systems with multi- and single-element thermionic low-temperature energy-conversion systems, which are placed outside the reactor core and in which heat is delivered to the emitters using molybdenum-lithium heat pipes at 1600–1670 K, are discussed as an example. Computational results are presented for the electro- and thermophysical characteristics of electricity generating systems and power systems with different variants of the design-technological solutions, including the possibility of increasing the service life by reserving a part of the thermionic electrical power. The mass/size characteristics of the power systems with basic variants of the systems are obtained and recommendations based on research and development work are made, 8 figures, 8 references. Translated from Atomnaya énergiya, Vol. 89, No. 1, pp. 22–34, July, 2000.     

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.     

Measurement of the cross sections of the reactions 14N(n, α)11B and 14N(n, t)12C at neutron energies 5.45–7.2 MeV

The cross sections of the reactions ¹⁴N(n, α)¹¹B and ¹⁴N(n, t)¹²C have been measured for neutron energies 5.46–7.2 MeV. The neutrons are generated in the reaction D(d, n) on a solid titanium target. The work employs digital spectrometry. The charged-particle detector is a pulsed ionization chamber with a Frisch grid, filled with a kyrpton-nitrogen mixture. The cross sections are measured for four groups of α particles α₀, α₁, α₂, and α₃ from the reaction ¹⁴N(n, α)¹¹B and for tritium from the reaction ¹⁴N(n, t)¹²C. The energy resolution of the spectrometer was 60 keV. The errors in determining the cross sections for the reactions (n, α) and (n, t) are 10–15%. The measurement results are compared with the ENDF/V VI evaluation. Good agreement is obtained in the neutron energy range 5.45–6.5 MeV. At higher energies, the discrepancy reaches 30%. __________ Translated from Atomnaya énergiya, Vol. 101, No. 4, pp. 307–311, October, 2006.     

Neutron-capture therapy with ultra-cold neutrons

The status of neutron-capture therapy of malignant tumors and its problems – damage to healthy tissue as a result of neutron transport to the irradiation location and presence in the therapeutic beam of a background consisting of γ rays and fast neutrons – are presented. To solve these problems, the authors have proposed using ultracold neutrons with energy less than 10^–7 eV, whose unique capability is to undergo total reflection from the surface of a condensed substance at any angle of incidence. Numerous works have demonstrated that such neutrons can be transported along neutron guides. The cross section for inelastic scattering of neutrons by hydrogen-containing substances – water, ethyl alcohol, and biological tissue – has been measured in an IR-8 beam of ultracold and very cold neutrons. At temperature 200–300 K, the experimental data are in very good agreement with calculations, but as temperature decreases further a discrepancy appears, which could be due to the inaccuracy of the model spectra of the oscillations hydrogen-containing substances used in the calculations. The use of ultracold neutrons opens up new possibilities of neutron-capture therapy for treating malignant tumors localized in body cavities or organs.     

Determination of the optimal parameters of a cascade of gas centrifuges

Conclusions The necessary relations which make it possible to implement the numerical-analytical method for determining the optimal parameters of a cascade of centrifuges were obtained. The fundamental distinction from conventional calculations of ideal cascades lies in the fact that high coefficients of utilization of the separation power with strictly fixed external parameters of the cascade are guaranteed. The method can be extended to other formulations of the problem in which the total number of centrifuges appears as a separate component of the target function. Moreover, it can be used for optimizing a real cascade with losses. Since the losses are small, it is sufficient to make several calculations. The first calculation is performed with the formulas presented for estimating N_i and C_i and the corresponding losses q_i of matter and of isotopes q_ui (i=l,...n). The next calculations are performed iteratively according to the prescribed values of q_i and q_ui. In these variants, the expressions for the transit fluxes τ_i′,τ_ui′ are corrected each time. Similar variations of calculations are also possible in other situations, for example, in the case when the number of centrifuges in the steps is small. The possibility of making such changes is due to the flatness of the target function — the target function increases very little when the parameters deviate from strictly best values. This work was supported by a grant from the Moscow Engineering Physics Institute (Goskonvuz) for 1996–1997. Ural State Technical University. Translated from Atomnaya énergiya, Vol. 84, No. 3, pp. 246–253. March, 1998.     

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.     

Determination of the multiplication of leakage neutrons and the fissile material mass in deeply subcritical systems

The previously developed analysis of measurements of neutron multiplication in deeply subcritical states of multiplying media by the α-Rossi method (method of double neutron-neutron coincidences) is supplemented by a new analysis of triple coincidences. To obtain the neutron multiplication and the effective neutron multiplication coefficient, it is no longer necessary to know the intensity of the source of spontaneous fissions – it is determined from experiments. This makes it possible to determine simultaneously the mass and enrichment of a fissile substance placed in a container. Measurements are performed and analyzed using uranium and plutonium blocks and fast-neutron detectors as well as AWCC (active well coincidence counters) neutron detectors. Previous measurements performed using the coincidence counters are analyzed. It is shown that without calibration of the ²⁴⁰Pu and ²³⁹Pu mass by measurement of counting rate of single neutrons as well as double and triple coincidences in ordinary experiments the error (1σ) in determining the mass fraction of ²⁴⁰Pu in plutonium is 0.3–0.8% and the error for all plutonium is 3–12%.     

Aerosol emissions from the destroyed power-generating unit of the Chernobyl nuclear power plant in 1986 and 2003–2005

The results of measurements of the volume activity and dispersity of aerosol carriers of β-emitting radionuclides during the acute phase of the accident and 20 years later are presented. It is determined that in August–October 1986, when samples were taken 10–50 m from the surface of the destroyed reactor, the concentration was about 1 kBq/m³, which is 100–1000 times higher than the value recorded in July–August of the same year from an aircraft flying at altitudes 200–1000 m. Thus, already in mid-summer 1986, because of the decrease in temperature, the aerosol emissions did not reach the survey altitude of the aircraft. Therefore, the sampling performed from the aircraft did not permit a quantitative assessment of the emissions of radioactive materials during this period of time. In 2003–2005, the total β activity was 10–100 times less than in fall 1986, because of the radioactive decay of ⁹⁵Zr, ⁹⁵Nb, ^103,106Ru, ¹³⁴Cs, ^141,144Ce, and other radionuclides. Since the ¹³⁷Cs concentration decreased negligibly, it seems that the roof constructed in 1986 above the Shelter was of little use. __________ Translated from Atomnaya énergiya, Vol. 100, No. 4, pp. 276–282, April, 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.