Investigation of the Mixing of Coolant Flows in a VVÉR Vessel

The mixing of flows in a flow-through channel of a VVÉR-1000 reactor in the case where a flow with high or low boron concentration and a flow whose temperature is different from the temperature in the reactor are fed into the same circuit is studied. To this end investigations were performed on a bench at the Special Design Office Gidropress with a model VVÉR-1000 reactor on a 1:5 scale. Regimes with the main circulation pump switched on and with natural circulation restored for an accident with a small leak are studied.Intercircuit mixing in the VVÉR-1000 reactor setup is also studied in the case where the temperature of the coolant changes in one of the circuits (regimes with a break in a steam pipe and closure of the steam channel of the turbine) or with asymmetric feeding of boron into the first loop when one of the tanks in the system for rapid introduction of boron fails. An analytical solution of the problem and the results of measurements performed on units with VVÉR-1000 reactors are presented.     

Measurement of VVÉR Control Rod Efficiency

The reasons for the large discrepancies between the computed and experimentally measured values of the efficiency of control rods observed during startup experiments on a VVÉR reactor are discussed. The spatial distribution of the neutron flux has the greatest effect. A method for eliminating the influence of spatial effects on the measurements of control rod efficiency is proposed. The possibilities of the proposed method are illustrated for examples of the analysis of experiments performed on power-generating units with VVÉR-440 and -1000 reactors.     

Radiation Embrittlement of VVÉR-1000 Vessel Materials

Radiation embrittlement of VVÉR-1000 vessel materials has been studied much less than for VVÉR-440 reactors. In the present paper the results of an investigation of the first batches of control samples of VVÉR-1000 vessel materials are discussed. The chemical composition of the materials is characterized by a low content of harmful impurities (copper and phosphorus) and a high nickel content (up to 1.9% in some weld seams). The actual rate of radiation embrittlement of the material studied is comparable to the embrittlement calculated using the Russian standards. The dependence of radiation embrittlement of VVÉR-1000 vessel materials on the metallurgical variables and the damaging dose is studied. The investigation showed that nickel greatly intensifies the radiation embrittlement. New relations were developed for determining the actual rate of radiation embrittlement of VVÉR-1000 reactor vessel materials and assessment of its conservativeness.     

Testing of experimental fuel elements for VVÉR-1000 reactors in MR to high fuel burnup

Conclusions The results of testing in a multipurpose reactor and post-irradiation examinations indicate satisfactory performance of the fuel element for the VVÉR-1000, which is designed for a 3-year run. In addition to the computational data, the experimental data were used to substantiate the performance of fuel elements when fuel burnup is increased and atomic power plants are switched from the VVÉR-1000 to a 3-year cycle (with an average burnup of 40 MW-day/kg).I. V. Kurchatov Institute of Atomic Energy. Translated from Atomnaya Énergiya, Vol. 72, No. 2, pp. 116–120. February, 1992.     

Computational and Experimental Investigations of Accident Regimes in Standard Problems on the ISB-VVÉR Thermophysical Stand

Five standard problems of investigating accidents with loss of coolant in the first loop are studied for the first time on the basis of experiments which were performed in 1993–2001 on the ISB-VVÉR two-circuit integrated thermophysical stand, simulating the first loop of a reactor system with a VVÉR-1000 reactor. The objective of the investigations was to verify the computational thermohydraulic codes developed in this country and abroad – TECh', KORSAR, ATHLET, CATHARE, and RELAP. The results of the verification calculations on the whole agreed well with the experimental data. Most processes and phenomena which can occur in VVÉR-1000 accidents with a small and average-size coolant leak were reproduced in the experiments. Analysis of the results showed that these computational codes can be used to simulate the processes occurring during an accident.     

VVÉR-Based Power-Generating Units with Transcritical Coolant Parameters

The possibility of increasing the thermodynamic efficiency of power-generating units based on single-loop VVÉR with transcritical coolant–working body parameters is studied. The heat-utilizing turbine part of the power-generating unit is examined. Experience in developing and building turbines with transcritical parameters of the working body for fossil fuel power engineering facilitates the development of such turbines for nuclear power production.     

Fit of Simple Analytic Functions to the Time Dependences of Nuclide Concentrations

Simple analytical functions are used to fit the results of theoretical numerical calculations of the time dependences of the specific mass of ²³⁵U, fission fragments, and the products of activition of fuel nuclei for VVÉR-440 and -1000 and RBMK-1000 reactors with an error of up to several percent. The possibilities of this formalism are examined for the example of the damaged RBMK-1000 reactor at Chernobyl.     

Efficiency of producing additional power in units of nuclear power stations containing water-cooled-water-moderated reactors

Conclusions There is a basic possibility to raise the maximum power of a unit containing the VVÉR-1000 reactor in the course of the fuel charge burn-up and with lowering the coefficient of the energy-release nonuniformity in the reactor core. It is more advantageous economically to obtain additional power in carrying the peak load. With the duration of such operating conditions for 1000–2000 h/yr savings can amounts to 1–4 million rubles per year per 100 MW of additional power when compared to peaking GTP.It is necessary to analyze in more detail the safety as well as technical and economic indexes of the VVÉR-1000 and of the entire power unit both under normal operating conditions and in emergency situation for the proposed elevated power mode.It is necessary to develop a procedure of experimental substantiation of a possible forcing level for the operating units of the NPS containing a VVÉR. The interested agencies should be involved in this development. Technical and economic advisability of making up a set of power-unit equipment suitable to carry short load peaks and prolonged elevation of the electric loads in the EPS should be determined on this basis and changes and additions should be introduced when developing new designs.Translated from Atomnaya Énergiya, Vol. 61, No. 6, pp. 397–401, December, 1986.     

Analysis of the Possibility of Using Fuel Based on Reclaimed Uranium in VVÉR Reactors

A current problem is to show that reclaimed uranium can be, in principle, brought into the VVÉR fuel cycle. The possibility of using fuel based on reprocessed uranium in VVÉR is analyzed. The requirements for the initial isotopic composition of test batches of fuel pellets with 4% effective enrichment are determined, the compensation coefficient is calculated, the requirements for monitoring the isotopic composition are determined, and the technlogy for fabricating fuel pellets from relaimed fuel is determined. It is shown that the basic neutron-physical characteristics satisfy the restrictions adopted in the VVÉR-440 and -1000 designs. The effect of radiation on the public and the environment as a result of switching to fuel fabricated from reclaimed uranium is the same as for the standard oxide fuel.     

Using the Offset–Power Diagram to Monitor the Local Linear Power Density of Fuel Elements in a VVÉR-1000 Core

A method for real-time monitoring of jumps in the local linear power density of fuel elements in a VVÉR-1000 core is described. Monitoring jumps in the local power is one component of the improved algorithms for reactor control which ease the operating conditions for the fuel assemblies.     

Maintaining an Equilibrium Offset as an Effective Method for Suppressing Xenon Oscillations in VVÉR-1000

A method is described for suppressing xenon oscillations of the vertical distribution of energy release in VVÉR-1000 by maintaining an instantaneous offset in accordance with the instantaneous vertical distribution of xenon nuclei along the reactor core. Computational examples of the suppression of oscillations are presented. 2 figures, 3 references.     

Transmutation of Long-Lived Actinides in Power Reactors

Heavy- and light-water power reactors can be used for partial transmutation of nuclear wastes, thereby making it possible to limit the accumulation of long-lived radionuclides in storage sites for spent nuclear fuel to relatively low levels. During the operation of PHWR-880 in the self-service regime, the equilibrium radiotoxicity in long-term storage and the time when it is reached are 4–5 times less than in the analogous regime for a VVÉR-1000 reactor.     

Long-Lived-Radionuclide Transmutation Scenarios

The results of multigroup calculations of continuous irradiation of Np, Am, and Cm in VVÉR-1000, PHWR-880, Superphoenix-1200, BREST-1000, and ÉLYaU-800 reactors are used to compare transmutation efficiency. The sources of continuous replenishment for the transmuters were Np, Am, and Cm extracted after a 3-yr holding period from the VVÉR and Superphoenix spent fuel. It is shown that the most effective transmuter is a subcritical liquid-fuel ÉLYaU system with an average thermal-neutron flux in the blanket 2·10¹⁵ sec^–1·cm^–2. For solid-fuel reactors, the continuous-irradiation model makes it possible to describe approximately the multiple transmutation regime. In the foreseeable future, one-time transmutation of Np, Am, and Cm in a solid-fuel reactor followed by storage in a long-term storage facility is feasible. The results of different computational variants for such regimes show that for transmutation in 10 yr in PHWR the radiotoxicity of Np, Am, and Cm accumulated in long-term storage reaches an equilibrium in no longer than 100 yr.     

16N γ-Ray Diagnostics of a Nuclear Reactor in a Nuclear Power Plant

The AIRM system, which uses the ¹⁶N activity in the VVÉR-1000 reactor at the Kalinin nuclear power plant to measure the thermal power of the reactor and the coolant flow rate, and similar systems used in nuclear power plants with PWR reactors are described.The influence of a change in the neutron spectrum as a result of the loss of ²³⁵U and the accumulation of ²³⁹Pu and ²⁴¹Pu in the VVÉR-1000 core over a run on the yield of the reaction ¹⁶O(n,p)¹⁶N and, correspondingly, the influence of the changes in the indications of the thermal power measured according to ¹⁶N are taken into account.     

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.     

Improvement of the design of the series-produced fuel element of the VVÉR-440 reactor to increase burnup

Conclusion The fuel elements for the VVÉR-440, which were developed for a 4-yr run, operate satisfactorily to an average burnup of 59 MW-day/kg of uranium in the most-stressed fuel element with average burnup of the unloaded fuel of 40 MW-day/kg. Through the introduction of end beveling of the pellets, the mobility of the fuel column is increased. Furthermore, the presence of bevels makes possible a reduction of the number of chips and elimination of process crumbs, and the high mobility of the fuel column makes it possible, during outfitting of the fuel element, to eliminate the axial gaps between pellets, which are unacceptable for safety reasons. The increase in the initial helium gage pressure has made possible substantial improvement of the thermomechanical characteristics of the fuel elements and avoidance of high fuel temperatures and large gas release. When the conversion is made to a 4-yr fuel run, the number of fuel assemblies refueled each year decreases from 117 to 90 per power-generating unit, natural uranium consumption is reduced by 11–12% [2], zirconium consumption is lowered, and the effective capacity of burned-up fuel stores is increased.I. V. Kurchatov Institute of Atomic Energy. Mashinostroitel'nyi Zavod Production Association. All-Union Scientific Research Institute of Aviation Materials. Translated from Atomnya Énergiya, Vol. 72, No. 2, pp. 121–124, February, 1992.     

Computational and Experimental Investigation of the Dynamical Characteristics of a Model of a VVÉR-1000 Reactor Pit

A finite-element method is proposed for determining numerically the characteristic frequencies and forms of the oscillations of a shell structure in a liquid. An experimental method for determining the charcteristic frequencies and forms of the oscillations of the structural elements by holographic interferometry is examined. The results obtained for the characteristic frequencies and forms of the oscillations of a model of a VVÉR-1000 reactor pit by the finite-element and holographic interferometry methods are presented. The calculations agree well with the experimental data. This shows that the computational model adopted for the oscillations of bodies in liquids is adequate and that the algorithms developed for simulating the vibrations can be verified by holographic interferometry.     

Testing of improved algorithms controlling energy release in VVÉR-1000 in maneuvering regimes at the Tianwan nuclear power plant (China)

The salient features of improved algorithms adopted at the Tianwan nuclear power plant for controlling the energy released in a VVéR-1000 core are examined. The optimal configuration of the controlling groups is chosen in the first two power-generating units, the reactor power can be varied automatically under the control of an automatic power regulator, the boron regulation system makes it possible to determine the first-loop makeup automatically, and a modern version of the Imitator Reactora program has been installed. The results of testing the algorithms in the No. 1 unit in regimes with single and cyclic (daily) power maneuvers are presented. The tests of single power maneuvers were combined with dynamic tests of equipment. The operation of the power-generating unit in a daily load schedule was tested separately. Five daily load-change cycles were conducted during these tests. __________ Translated from Atomnaya énergiya, Vol. 103, No. 5, pp. 277–282.     

New Generation of Microfuel-Based Fuel Elements for VVÉR Reactors

The results of development work on a new generation of fuel elements based on microfuel for VVÉR reactors using the basic data from post-reactor investigations and bench tests in experiments simulating LOCA for existing fuel elements with ceramic fuel are presented. It is shown that cermet fuel elements will make it possible to realize most fully the advantages of such fuel, specifically, to develop a sealed first loop and to simplify and reduce the cost of safety, automatic control, radiation protection, coolant puification, and other systems. For example, cermet fuel elements in VVÉR-1500 reactors can improve safety under various operating conditions, maneuverability, vibrational strength, fuel assembly lifetime, and geometric stability of fuel elements.     

Fast-neutron fluence determination with diamond sensors

A new method of determining the fluence with an integrating diamond sensor is developed. The method is based on the results of many experiments on combined irradiation in VVÉR-1000 of diamond sensors and control samples for which activity data on the fast-neutron fluence were obtained from the ⁵⁴Mn by a computational-experimental method. The expansion of the crystal lattice of irradiated diamond, which is measured with a comparatively high accuracy, is used as the measured property. The new method is free of many of the drawbacks of the activation method, and its relative accuracy is higher than the accuracy of the computational-experimental method, especially for a long holding time in the reactor.__________Translated from Atomnaya Énergiya, Vol. 98, No. 2, pp. 118–123, February, 2005.