Tests in an MR reactor of experimental fuel elements of the VVÉR-1000 reactor

Conclusions In an MR reactor performance tests of 16 fuel assemblies, with elements having essentially the same structure as standard VVÉR-100 fuel elements, were carried out. Tests of five more fuel assemblies are continuing. Of the 16 assemblies, 13 were studied in a hot laboratory.The tests in the MR, carried out at high loads and with a large number of transition processes, as well as the postreactor studies, indicated that fuel elements of the specified design (with initial helium pressures of 1.96–2.45 MPa) have a high reliability. None of the elements of the fuel assemblies studied malfunctioned due to design defects or faults in their fabrication. During the tests the jackets were subject to a little oxidation and hydrogenation (zirconium-oxide film<3 m thick, hydrogen content less than 0.008% by mass), and their plasticity remained high (the relative elongation at the working temperature remained at the 20% level).Translated from Atomnaya Énergiya, Vol. 62, No. 5, pp. 312–317, May, 1987.     

Plutonium charging of the VVÉR-440 water-cooled water-moderated reactor

Conclusions Our study has been limited to a neutron-physical analysis of the active zone in the presence of a fresh charge. Allowing for the disposition of the fuel assemblies this reflects a reasonably practical situation. The difficulties associated with local peaks of energy evolution may be quite easily overcome by using different fuel enrichments inside the assembly.Our example of the charging of the active zone with plutonium is of a hypothetical character, since there is far more plutonium in this arrangement than is required in any practical situation. The calculations show that there are no problems as regards the distribution of heat evolution with respect to the radius of the active zone.It would be extremely desirable to pursue this investigation with due allowance for the factors involved in the burnup of the nuclear fuel. Certain experimental work will be required in this connection so as to provide confirmation of the validity of the computer calculations. The nonuniformity of the distribution of energy evolution will be smoothed as the nuclear fuel is impoverished during burnup.Another important aspect to be studied is that of determining the weight of the control rods. We feel that a study of reactivity problems will reveal some more rigorous limitations than those deduced from the study of energy distribution undertaken in the present investigation.State Technical Scientific-Research Center. Laboratory of Nuclear-Power Technology, Helsinki, Finland. Translated from Atomnaya Énergiya, Vol. 40, No. 4, pp. 283–286, April, 1976.     

Using zirconium spacer lattices in the VVÉR-440 reactor

RNTs "Kurchatov Institute." Experimental Design Office "Gidropress." Translated from Atomnaya Énergiya, Vol. 73, No. 5, pp. 410–411, November, 1992.     

Handling Irradiated RBMK-1000 and VVÉR-1000 Fuel during the Growth of Nuclear Power

On-site storage facilities, consisting of ponds with water, for irradiated RBMK-1000 fuel are now close to being filled. To continue operating nuclear power plants with RBMK reactors, it is necessary to select one possible method for handling irradiated fuel.A variant of long-term storage followed by reprocessing is examined and considerations are presented for future use of reprocessed irradiated RBMK and VVÉR fuel as fuel for an initial load for naturally-safe fast reactor. Important points in handling irradiated RBMK-1000 fuel include economic assessments and requirements for a strategy for development of nuclear power in Russia based on closure of the nuclear fuel cycle with radiation-equivalent burial of wastes and utilization of accumulated plutonium for fast reactors. 3 figures, 2 tables, 8 references.     

Analysis of criticality in shipment and storage of fuel at a nuclear power plant with a VVÉR reactor

The substantiation of nuclear safety during shipment and storage of fresh and spent fuel at nuclear power plants with VVéR reactors is examined in the light of the more stringent nuclear safety rules. Possible technical measures for satisfying the safety criterion are examined, for example, the concept of subcritical fresh fuel. An example of the estimation of the probability of the formation of a critical mass as result of fuel assemblies falling randomly out of a container is presented. Certain characteristic features of the calculation of the neutron-physical characteristics of fuel in a cooling pond are presented, for example, the nonconservative nature of a separate analysis in the infinite approximation. 4 figures, 5 references. OKB “Gidropress”. Translated from Atomnaya éneriya, Vol. 87, No. 1, pp. 11–16, July, 1999.     

Taking Account of the Spectral History of Fuel Burnup during the Preparation of the Neutron-Physical Constants for VVÉR-1000 Fuel Assemblies

The results of an analysis of the influence of the fuel burnup conditions on the two-group neutron physical constants of VVÉR-1000 fuel assemblies are described. The spectral index proposed by Spanish physicists and taking account of the characteristics of the fuel burnup regime is analyzed theoretically and experimentally. A modified version of the spectral index is developed and analyzed. Calculations are performed using the GETERA code. The modified spectral index is used in the HARD-NUT program system to analyze the fuel loads of the No. 2 unit of the Kalinin nuclear power plant. The results of changing the computational duration of a run after adopting a spectral index are presented. 4 figures, 1 table, 3 references.     

Neutron–Noise Vibrational Diagnostics of a VVÉR-440 Core Shaft

A method is presented for estimating the trajectory of the motion of a VVÉR-440 reactor core shaft on the basis of ionization-chamber noise. It is demonstrated that the statistics of the singular points of the trajectory makes it possible to monitor the wear of the core shaft–reactor vessel keyed expansion joints.     

Activity of radionuclides in the coolant of the secondary loop of a nuclear power plant with VVÉR-440 reactors

Conclusions Analysis of the values presented for the concentrations of certain radionuclides measured in the water of the primary and secondary loops shows the presence of a constant leakage of water from the primary into the secondary loop equal to 19±12 g/h on the average for a pressure drop of 7·10⁶ N/m² between the loops.The probability of this leakage increasing temporarily by a factor of 100 to 1 liter/h is 1.2·10^–1-1.6· 10^–2 (reactor eff. days)^–1 or 1.2·10^–8–1.6·10^–9 (kWh)^–1 of generated electrical energy.The data obtained can be used to analyze the radiological consequences of such leaks for various heattransfer schemes in nuclear power plants for supplying heat.Translated from Atomnaya Énergiya, Vol. 46, No. 1, pp. 31–35, January, 1979.     

Steam generators of the power-generating units of nuclear power plants with VVÉR-1000 reactors

Special Design Office "Gidropress." Translated from Atomnaya énergiya, Vol. 77, No. 2, pp. 100–107, August, 1994.     

Industrial processing of the spent fuel of npp operating with water-moderated VVÉR-440 reactors

Mayak Production Complex. Sverdlovsk Scientific-Research Institute of Chemical Machinery. Research-cum-production Complex VNIPIET. Ministry of Atomic Energy, Russian Federation. A. A. Bochvar All-Union Scientific-Research Institute of Inorganic Materials. Translated from Atomnaya Énergiya, Vol. 72, No. 5, pp. 432–436, May, 1992.     

Measurement of the Coolant Flow Rate through a VVÉR-440 Fuel Channel on the Basis of Fluctuations of Direct-Charge Sensor Signals

The coolant velocity is one factor determining the vibrational state of fuel assemblies. If this factor exceeds the design value, vibrational anomalies can appear and shorten the service life of the fuel assembly.The noise method for measuring the coolant velocity on the basis of the fluctuation component of the signals from direct-charge sensors promises high accuracy.     

Taking account of the effect of the curvature of VVÉR-1000 fuel assemblies on the power of the fuel elements

A comparative analysis is made of the deterministic and statistical methods of taking into account the effect of the curvature of VVéR-1000 fuel assemblies on the power of fuel elements. The fuel-element distribution of the energy release in the core for any random distribution of the gaps between the fuel assemblies is simulated, using the MEX code, on the basis of precise calculations (MCU code) and design calculations (BIPR-7 and PERMAK codes). The Monte Carlo method (Zazor code) was used to model the nominal density distribution of gaps in the core for different degrees of curvature of the fuel assemblies. It is shown that the power gain, obtained for the fuel elements by the probabilistic-statistical method, due to the curvature of the fuel assemblies is smaller and makes it possible to substantiate core safety with large perturbations, in contrast to the deterministic “maximum gaps near the most-energy stressed fuel element” method. 5 figures, 1 table, 3 references. Special Design Office “Gidropress.” Translated from Atomnaya énergiya, Vol. 87, No. 3, pp. 210–213, September, 1999.     

Evaluation of the containment vessel failure time from operator’s notifications during a severe accident

This paper aims at clarifying the potential and the limit of the method to surmise the timing of the containment vessel (CV) failure utilizing the Emergency Action Levels (EALs) issued as a nuclear operator’s notification in an early phase of the severe accident (SA). We analyzed the timings of the EALs issued in all kinds of the SA sequences of several PWR plant models by using the SA analysis code, MAAP. We found high correlations between the timing of SE41 (EAL issued at CV pressure of 0.5 design pressure) and the timing of the CV failure in the typical scenarios, e.g. over-pressure failures. We could therefore establish an evaluating method to estimate the time for a CV failure. This method has the potential to support the decision-making in nuclear emergency preparedness.     

Improved method for measuring the γ-ray absorbed dose rate in models of iron shielding of a thermonuclear reactor

An improved method of measuring the absorbed γ-ray dose rate usingCaSO ₄ andSrSO ₄ type thermoluminescent detectors in models of iron shielding of a thermonuclear reactor is described. The reactionT(d, n)⁴ He served as a neutron source. The method obtained makes it possible to determine the absorbed γ-ray dose rate in shielding without using computed information and relying only on experimental data on the rates of nuclear reactions in threshold detectors. 7 figures, 1 table, 9 references. Moscow Engineering-Physics Institute. Translated from Atomnaya énergiya, Vol. 86, No. 3, pp. 219–225, March, 1999.     

Use of Refractory Coatings in Systems for Melt Containment in a Serious Accident at a Nuclear Power Plant with a VVÉR Reactor

A computational analysis is performed of the use of zirconium dioxide ceramic in melt containment systems with relatively thin heat-conducting structures. The results of an investigation of the physicochemical processes for the interaction of melt with zirconium dioxide ceramic are presented. It is shown that the use of zirconium dioxide ceramic together with a passive water source gives the required melt cooling parameters for 3 to 24 h depending on the cooling regime. 4 figures, 9 references.     

A new approach to analysis of water chemistry indicators in the first loop of reactor installations in nuclear power plants with VVÉR reactors

A new approach is developed for analyzing a complex of data reflecting the state of the coolant in the first loop of reactor installations of nuclear power plants with VVéR reactors. The method is based on the treating the measurements of the concentration of the components of the coolant during a run as nonuniform time series and desribing them using a flicker-noise model or continuous wavelet analysis. It is shown that this method can be used to obtain objective information about the relations between the physicochemical processes leading to the transformation of various components of the coolant (hydrogen, ammonia, hydrazine, metal ions—products of corrosion) in the technological regime without using any speculative models. 8 figures, 1 table, 3 references. Special Design Office “Gidropress”. Translated from Atomnaya énergiya, Vol. 87, No. 3, pp. 194–199, September, 1999.     

Probabilistic Method for Estimating the Effect of the Gaps between Fuel Assemblies on the Energy Release in a VVéR-1000 Core

A method is presented for analyzing the influence of perturbations and uncertainties in the core on the energy release of fuel elements. The method uses a system of programs – design, precision, and specially developed programs – and it uses as initial data the results of complicated thermomechanical calculations of the curvature of fuel assemblies. A criterion is formulated for the usefulness of fuel-assembly profiling. The method as a computational tool makes it possible to develop improved profiling schemes which give higher heat-engineering margins of safety.