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"Overall Plant Design Descriptions, VVER, Water-Cooled, Water-Moderated, Energy Reactor," Nuclear Safety Institute, Russian Academy of Science, Moscow, April 1995.
V. I. Baranenko, A. I. Piontkovskii, S. P. Khmaryuk, N. N. Davidenko, A. G. Shalaev, N. D. Kukharev, E. O. Vorontsov, A. V. Ermolaev, V. A. Zakharov, and A. F. Kukharev, "Reliability of High-Pressure Recovery System in the Power-Generating Units of Nuclear Power Plants with VVER-1000 and -440 Nuclear Reactors," Translated from Atomnaya Energiya, Vol. 78, No. 2, pp. 133-138, February 1995.
A. Kryukov. P. Platonov, Ya. Shtrombakh, V. Nikolaev, E. Klausniter, C. Leitz, C.-Y. Rieg, "Investigation of Samples Taken from Kozloduy Unit 2 Reactor Pressure Vessel," Nuclear Engineering and Design, Vol. 160, No. 1-2, pp. 59-76, 1996.
Abstract - Within the framework of the 6 month WANO program, small samples were cut from the inside surface of the Kozloduy NPP unit 2 reactor pressure vessel to assess the actual condition of the pressure vessel material before and after annealing. The actual values of the weld metal characteristics required for estimating radiation-limited lifetime--the ductile-to-brittle transition temperature (DBTT) in the initial state (Tko) and the phosphorus and copper contents which affect the radiation stability of steel--were not determined during manufacturing. The Kozloduy unit 2 pressure vessel had no surveillance program. Radiation stability was evaluated using dependencies based on analysis results for surveillance samples taken from other VVER-440 reactors. For this reason, the actual pressure vessel characteristics and their changes in the course of reactor operation, as well as comparison of experimental with calculated data were the principle objective of the study. Instrumented impact test were carried out on sub-size specimens of base and weld metal. Correlation dependencies were used with standard tests to determine DBTTs for the base and weld metal (in accordance with Russian standards): base metal before annealing 40 °C, after annealing 16 °C: weld metal before annealing 212 °C, after annealing 70 °C. The estimate value of Tko, for the initial, unirradiated weld metal, was 50 °C. The experimental results were compared with a prediction of the extent of radiation-induced embrittlement of Kozloduy unit 2 pressure vessel materials. It was confirmed that the radiation-induced embrittlement of the base metal does not impose any limits on the radiation-limited lifetime of the pressure vessel. The predicted increase in the DBTT of the weld metal as a result of irradiation (about 165 °C) is practically equal to the experimental result (162 °C). However, the value of Tf obtained from tests before annealing (212 °C) is about 40 °C higher than the estimated value, i.e. the calculation does not produce a conservative estimate. This was explained by a low estimate of Tko (10 °C) which had been calculated using data from chemical analysis of the weld metal, performed by the manufacturer. The investigations on the samples, however, yielded an estimated value of Tko = 50 °C. The effectiveness of annealing in restoring the mechanical properties of irradiated VVER-440 reactor pressure vessels was confirmed. Recovery annealing lowered the DBTT of the weld metal by 85% or more of its radiation-induced shift.
T. G. Apostolov, Kr. D. Ilieva, T. M. Petrova, "Calculation Results from Radiation Embrittlement of VVER Pressure Vessel at Kozloduy NPP," Nuclear Engineering and Design, Vol. 160, No. 1-2, pp. 153-158, 1996.
Abstract - A calculational procedure for the evaluation of the transition temperature shift on the basis of neutron fluence has been applied for assessing the reactor pressure vessel (RPV) embrittlement and life time for a VVER- 440/230. The calculated results are lower than the passport values, because the real fuel regimes, the low-leakage schemes and loadings with dummy cassettes have been taken into consideration in neutron fluence calculations. The temperature of the outer wall of the RPV has been measured. No significant deviation between the measurement and the data given in the reactor passport has been observed. This shows the correct application of the calculational procedure.
J. Kysela, M. Zmitko, V. A. Yurmanov, V. F. Tiapkov, "Primary Coolant Chemistry in VVER Units," Nuclear Engineering and Design, Vol. 160, No. 1-2, pp. 185-192, 1996.
Abstract - Operation experience with VVER coolant chemistry has been reviewed. The paper describes the results of measurements of radiation fields in primary system components and occupational doses that are compared with radiation control philosophy based on boron-potassium-ammonia coordinated water chemistry, modified and hydrazine water chemistry. The difference in water chemistry guidelines between VVERs and different current operational practices at VVER utilities are outlines. Special emphasis is given to the ammonia-hydrazine water chemistry on some VVER plants, high temperature filtration and silver behavior.
Krassimira Lieva, Tihomir Apostolov, Ivan Penev, Sergey Belousov, Evgeny Taskaev, Stoyan Antonov, Temenuga Petrova, Petjo Tzokov, Zahari Bojadjiev, "Verification of the VVER-440 Pressure Vessel Neutron Fluence Calculation by Activity Measurement," Nuclear Engineering and Design, Vol. 160, No. 1-2, pp.257-260, 1996.
Abstract - To qualify the calculation methodology, measurements of neutron flux responses of a VVER-440/230 reactor pressure vessel have been carried out. The activity of shavings sampled out from the inner pressure vessel wall of Unit 1 of Kozloduy NPP after the 14th cycle has been measured. Calculation of the expected activity at the shaving positions has been carried out, taking into account the local power distribution. Comparison of calculated and measured activity values has indicated that the computed value for the fast neutron fluence is underestimated by up to 20%.
P. Varpasuo, "The Seismic Reliability of VVER-1000 NPP Prestressed Containment Building," Nuclear Engineering and Design, Vol. 160, No. 3, pp.387-398, 1996.
Abstract - The failure probability assessment of the containment building is an essential feature of the Level 2 PSA studies of nuclear power plants. The primary purpose of this paper is to demonstrate the methodology of evaluating containment seismic induced probability of failure without containment pressurization. the Loviisa, Finland site is one of the most seismically stable in the world and the numerically evaluated seismic induced failure probabilities are not representative for other sites. In addition, the containment concept described in this paper is not the typical Russian design which uses helical tendons in the cylindrical part of the structure and has a ring girder at the spring line of the structure. So the conclusions reached are applicable only to the containment configuration described in the paper. The geometry of the containment was determined by its preliminary design. The seismic hazard of the plant site was assessed during Level 1 PSA of the Loviisa plant. The initial information for seismic fragility analysis of the containment is the seismic response of the structure. The structural model for response analysis was the stick model. The stress analysis of the containment was carried out using the shell element model. The fragility evaluation of the containment was performed with the PROBAN-program. The structure was modeled as a parallel system consisting of the most heavily stressed elements. The resulting fragility curve gives the conditional probability of failure as a function of peak ground acceleration. The seismic hazard and the fragility were convolved to obtain the annual nonexceedance probability distribution for the collapse frequency of the structure.
K. D. Ilieva, S. Y. Antonov, and S. I. Belousov, "Calculation Modeling of Detector Activity in the VVER Pressurized-Water Reactor Vessel Surveillance," Nuclear Science and Engineering, Vol. 122, No. 1, pp. 131-135, 1996.
M. Antonopoulosdomis, K. Mourtzanos, and G. Por, "Identification of the Excitation Source of the Pressure-Vessel Vibration in a Soviet Built WWER PWR with Signal Transmission Path-Analysis," Annals of Nuclear Energy, Vol.23, No. 12, pp. 989-995, 1996.
J. Zd'arek, L. Pecinka, and F. Kaspar, "The LBB Project for the New Generation of WWER Plants and Lessons Learned from the WWER 440 Model 230 LBB Project," Nuclear Engineering and Design, Vol. 159, No. 1, pp. 55-62, 1995.
Abstract - As part of the safety enhancement of WWER-type nuclear power plants (NPPs) with model 213 and 320 reactors operated or constructed in the Czech and Slovak Republics, the leak-before-break (LBB) methodology is applied to main circulating pipe and pressurizer surge lines. The typical feature of all these NPPs is the horizontal type steam generator. The resultant effect is that the lengths of all safety significant piping are longer in comparison with standard western PWRs. For this reason the complex dynamical model is used for seismic response assessment performed as part of LBB methodology. In this paper the results of decoupling studies and the new set of experiments concerning both leak rates and accomplished diagnostics measurements are described.
J. Sievers, X. Liu, P. Rajamaki, H. Talja, and H. Raiko, "Comparative Analyses Concerning Integrity of a VVER-440 Reactor Pressure Vessel," Nuclear Engineering and Design, Vol. 159, No. 1, pp. 63-68, 1995.
Abstract - Comparative analyses have been performed with 3-D finite element (FE) models for a six loop cladded reactor pressure vessel with an assumed circumferential surface crack. In the analyses, a thermomechanical transient due to loss of coolant with high pressure injection has been considered. The transient is characterized by an axisymmetric cooling phase which is followed by an asymmetric, strip-like cooling period. For the fracture assessment, stress density factors calculated from local J-integral values have been compared. It has been shown that with different FE programs and meshes, the crack loading can be calculated within a 10% scatter. Furthermore, simplified fracture methods show a much larger scatter, mainly due to approximation of the non- linear stress distribution by membrane and bending stress.
J. Bohmert, F. Bergner, M. Grosse, H.-W. Viehrig, "Influence of the Depth Position on the Neutron Embrittlement of the VVER Reactor Pressure Vessel Steel 15CrMoV(A) Consequences for the Assessment of Reactor Safety," Nuclear Engineering and Design, Vol. 159, No. 2-3, pp. 131-141, 1995.
Abstract - The dependence of the mechanical properties on the depth position in the unirradiated state and after irradiation to neutron fluences of approximately 5x1018 and 70x1018 cm-2 (E > 0.5 MeV) is tested on a forging made out of VVER 440 reactor pressure vessel (RPV) steel 15CrMoV. The near-surface position shows a higher strength and power transition temperature than the positions greater than 1/4 wall thickness. Irradiation does not change these differences in a significant manner. The testing of specimens from the 1/4 depth position within the surveillance program, as normally laid down in the legal rules relating to nuclear power plants, results in a conservative safety assessment against brittle failure up to the EOL fluence. On taking into account fluence attenuation, the flux effect, the toughness gradually increases from the inside to the outside of the wall after longer RPV operating times.
Heikki Keinanen, Heli Talja, Rauno Rintamaa, Kari Torronen, Ralf Ahlstrand, Pekka Nurkkala, George Karzov, Boris Timofeev, Alexander Blumin, "Crack Initiation and Arrest in a Pressure Vessel Made of VVER-440 Reactor Pressure Vessel Steel," Nuclear Engineering and Design, Vol. 158, No. 2-3, pp. 217-226, 1995.
Abstract - A joint pressure vessel integrity research program involving three partners is being carried out during 1990-1995. The partners are the Central Research Institute of Structural Materials "Prometey" from Russia, IVO International Ltd (IVO) from Finland, and the Technical Research Center of Finland (VTT). The main objective of the research program is to increase the reliability of the VVER-440 reactor pressure vessel safety analysis. This is achieved by providing material property data for the VVER-440 pressure vessel steel, and by producing experimental understanding of the crack behavior in pressurized thermal shock loading for the validation of different fracture assessment methods. The program is divided into four parts: pressure vessel tests, material characterization, computational fracture analyses, and evaluation of the analysis methods. The testing program comprises tests on two model pressure vessels with artificial axial outer surface flaws. The first model vessel had circumferential weld seam at the mid-length of the vessel. A special embrittling heat treatment is applied to the vessels before tests to simulate the fracture toughness at the end-of-life condition of a real reactor pressure vessel. The sixth test on the first model led to crack initiation followed by arrest. After the testing phase, material characterization was performed. Comparison of calculated and experimental data generally led to a good correlation, although the work is being continued to resolve the discrepancies between the measured initiation followed by arrest. After the testing phase, material characterization was performed. Comparison of calculated and experimental data generally led to a good correlation, although the work is being continued to resolve the discrepancies between the measured initiation and arrest properties of the material.
J. Krell, B. Mangel, and H. Rebohm, "Analysis of Rod-Drop Measurements at a VVER-440 Reactor," Nuclear Engineering and Design, Vol. 159, No. 2-3, pp. 265-271, 1995.
Abstract - During startup of the fifth unit of Nord nuclear power plant (NPP) several different rod-drop measurements were performed. The reactivity values obtained experimentally by a reactimeter on the basis of inverse point-kinetics differed, in some cases considerably, from values obtained by stationary k- calculations. In order to analyze these differences, calculations with a three- dimensional kinetic code were performed, which clearly showed that the local flux changes caused by dropped rods are responsible for the differences between calculated and measured rod worths. Furthermore, it is possible to use these calculations for the evaluation of correction coefficients. These coefficients, applied to the measured reactivity values, lead to corrected reactivity values, which are in fairly good agreement with calculated values. In our opinion this combination of calculation and measurement seems to be a suitable approach to the determination of rod worth under conditions typical for a NPP with VVER-440-type reactors.
H. Karwat, "The Evaluation of the Bubble Condenser Containment of VVER-440/213 Plants," Nuclear Engineering and Design, Vol. 157, No. 3. pp. 361-374, 1995.
Abstract - A number of pressurized water reactors of the VVER-440 type have a containment building which is connected to a bubble condenser unit reducing the design pressure of the entire containment system. The bubble condenser acts as a pressure suppression system by condensation of released steam. Some characteristic features of this system are presented. Referring to experience with the assessment of Western pressure suppression system containments of boiling water reactors (BWRs) the experimental and analytical background of the bubble condenser design has been studied and assessed. These test rigs serve to prove the thermal efficiency of this arrangement. Characteristic properties of pressures suppression systems of Western BWRs are compared with the properties of bubble condensers and some existing test rigs. An international working group has identified a number of open questions for which neither experimental evidence nor sound analytical predictions exist. Of main interest are the response of bubble condenser structures to oscillatory condensation processes, the simultaneous interaction of a large number of gap- cap units and additionally some unresolved scale-up problems. Based on these findings a supplementary experimental research program has been proposed to answer these open questions. The suitability of existing test rigs and of a new experimental model have been assessed. The results are summarized. In view of the importance of the containment effectiveness for public safety the proposed research work deserves a high priority within a program to assist countries utilizing VVER-440 reactors in confirming an adequate level of reactor safety.
Zsolt Techy, Gabon Lajtha, Robert Taubner, "Accident Loads for a VVER-440/213 Containment," Nuclear Engineering and Design, Vol. 157, No.3, pp. 375-385, 1995.
Abstract - Specific features of the VVER-440/213 containment are the subdivided rectangular building and the localization system including the bubbler trays and air traps. Accident loads are calculated for a large break loss of coolant accident (LBLOCA). The maximum pressure and temperature loads are calculated with different codes during the blowdown phase of the LBLOCA. The uncertainty margins of the maximum pressure are given in this case. Sensitivity studies are performed for different leakage rates and hydraulic data of the containment. The effects of the active and passive spray systems on the depressurization are presented in this paper. The maximum pressure loads are also examined in case of degraded conditions of the localization system.
Jan Murani, "Accident Localization System with Jet Condensers for VVER 440-V 230 NPP at Jaslovske Bohunise," Nuclear Engineering and Design, Vol. 157, No.3, pp. 387-393, 1995.
Abstract - Compared to the current world standard, the operational safety of the V1 (VVER 230) nuclear power plant (NPP) is unsatisfactory and it does not correspond to present requirements as to nuclear safety. Further NPP operation after 1995 is conditional on nuclear safety enhancement to a level comparable with that in West European countries. This aim should be achieved by a principal reconstruction involving in addition to others also backfitting the V1 NPP with technical facilities aimed at coping with a design basis accident (DBA). To cope with such an accident the Power Equipment Research Institute (VUEZ) designed an accident localization system with jet condensers. This system consists of (a) an air trap (one for each unit, mutually interconnected) with an expansion bell enclosed within, placed on a plate with 200 pipes of jet condensers passing through, and (b) a connecting duct between the hermetic zone and the air trap. The vertical jet condenser is an essential element of the system designed for steam condensation. Apart from condensation it serves as a water seal separating units 1 and 2. Demonstration tests of the jet condenser (model 1:1) condensing function were carried out at the testing unit of the All-Union Research Institute for NPP Operation (VNIIAES), Moscow in Kashir, 11-22 September 1992. These experiments concerning the dynamics and overpressure in the free space above the pool were close to the conditions in the air trap during DBA. The jet condenser height was proved to be sufficient to ensure the sealing function. Design and experimental work has been implemented in close cooperation with Russian experts Mr. V.N. Bulynin from the VNIIAES, Moscow, and Mr. M.V. Kuznecov from the Scientific and Engineering Center for Nuclear and Radiological Safety, Moscow.
Sergey I. Belousov, Krassimira D. Ilieva, and Stoyan Y. Antonov, "Three-Dimensional Neutron Flux Calculations for the VVER Pressure Vessel," Nuclear Technology, Vol. 111, No. 2, pp 270-274, August 1995.
Abstract - The neutron flux values at the sites important for the pressure vessels of the VVER-1000 and VVER-440 reactors have been calculated by the three-dimensional TORT code and the synthesis method approximation. The synthesis method is widely used now for neutron fluence routine calculations in metal embrittlement surveillance. The three-dimensional neutron flux evaluation by the synthesis method is based on the two-dimensional and one- dimensional solutions of the transport equation. The comparison of the results obtained by both methods confirms the good consistency within 3% for integral neutron flux with energy > 0.5 Mev, used for metal damage estimation, according to Russian reactor standards. Further investigations on the calculation validity will be based on comparisons with measurements of the threshold detector activities, monitored in the air shell behind the reactor pressure vessels of the Kozloduy nuclear power plant.
V. V. Bulavin, V. I. Pavelko, and D. F. Gutsev, "Investigation of the Characteristics of Vibrational Diagnostics of VVER-100 Reactors Under Operating Conditions," Translated from Atomnaya Energiya, Vol. 79, No. 5, pp. 343-349, November 1995.
E. A. Ivanov, I. V. Pyrkov, and L. P. Kham'yanov, "Monitoring of the Tightness of the Seal of Steam Generators in Nuclear Power Plants with VVER-1000 Reactors after Reconstruction of the Water Feed and Purge Systems," Translated from Atomnaya Energiya, Vol. 79, No. 5, pp. 350-353, November 1995.
A. V. Gordeev and B. G. Ershov, "Calculation of the pH of Water Coolant in the First Loop of a VVER Reactor," Translated from Atomnaya Energiya, Vol. 79, No. 5, pp. 360-366, November 1995.
A. G. Godizov and A. D. Efanov, "Physical-Mathematical Model of Volume Condensation and Transport of Aerosols Inside the Containment Shell of a VVER Reactor," Translated from Atomnaya Energiya, Vol.79, No. 5, pp. 376-381, November 1995
A. M. Kirichenko and M. V. Sigal, "Analysis of the Technical Level and the Operating Conditions of a Nuclear Power Plant with a VVER Reactor," Translated from Atomnaya Energiya, Vol. 78, No. 3, pp. 160-166, March 1995.
J. R. Secker, R. W. Miller, L. T. Mayhue, and R. N. Milanova, "Advanced Designs and Operating Strategies to Enhance the Safety, Operability, and Efficiency of VVER-1000 Reactors," Nuclear Science and Engineering, Vol. 121, No.1, pp. 142-152, 1995
P. E. Filimonov and Yu. A. Krainov, " Supressing Axial Oscillations of the Energy Distribution in a VVER-1000 Reactor without Half-Length Control Rods," Translated from Atomnaya Energiya, Vol. 78, No. 5, pp. 338-339, May 1994.
E. A. Ivanov, I. V. Pyrkov, and L. P. Kham'yanov, "Model of Accumulation of Radionuclides in Boiler Water of Steam Generators in Nuclear Plants with VVER-440 and -1000 Reactors," Translated from Atomnaya Energiya, Vol. 77, No. 1, pp. 58-63, July 1994.
E. A. Ivanov, I. V. Pyrkov, and L. P. Kham'yanov, "Methods for Diagnostics of Coolant Leaks in the First Loop into the Boiler Water of Steam Generators in Nuclear Power Plants with VVER-440 and -1000 Reactors," Translated from Atomnaya Energiya, Vol. 77, No. 1, pp. 51-59, July 1994.
V. F. Titov, "Steam Generators of the Power-Generating Units of Nuclear Power Plants with VVER-1000 Reactors," Translated from Atomnaya Energiya, Vol. 77, No. 2, pp. 100-107, August 1994.
E. A. Ivanov, I. V. Pyrkov, and L. P. Kham'yanov, "SNESK - Program for Calculating First- Loop Coolant Leaks into the Boiler Water of the Steam Generators of a Nuclear Power plant with a VVER Reactor," Translated from Atomnaya Energiya, Vol. 77, No. 6, pp. 457-458, December 1994.
V. V. Stekol'nikov and V. G. Fedorov, "From VVER-210 to VVER-2000: Experience in the Development and Improvement of Reactor Designs for Nuclear Power Plants," Translated from Atomnaya Energiya, Vol. 76, No. 4, pp. 310-318, April 1994.
J. Tuunanen, H. Tuomisto, and P. Raussi, "Experimental and Analytical Atudies of Boric Acid Concentrations in a VVER-440 Reactor During the Long-Term Cooling Period of Loss- of-Coolant Accidents," Nuclear Engineering and Design, Vol. 148, No.2-3, pp. 217-231, 1994.
Abstract - Concentrating and mixing of boric acid (H3BO3) during the long- term cooling of loss-of-coolant accidents (LOCAS) in the Loviisa VVER-440 reactors has been studied with the REWET-II and VEERA facilities. To get more detailed information boric acid mass transfer, a specific facility was built to simulate boron mixing in the lower plenum of the reactor. The experiments with the VEERA facility showed that in the VVER-440 reactor fuel bundles the mixing is complete due to boiling and U-tube oscillations and, hence, concentration distribution of boric acid in the bundles is uniform. The U-tube oscillations proved to be an important mechanism in Transferring concentrated boric acid from the core to the lower plenum. The experiments demonstrated that crystallization of boric acid in the core simulator is possible, if a stable long-term cooling situation with water boiling in the core continues long enough. In the experiments the crystallization of boric acid in the core simulator led to a flow blockage of the fuel bundle and overheating of the rod simulators when the flow through the core ceased. Experiments results were used to develop a computational model for calculations of boric acid concentrations in the reactor during LOCAS. The development work was supported with a series of RELAP5/MOD3 small-break LOCA analyses. The results of the RELAP5/MOD3 calculations were used to determine the boundary conditions under which concentrating of boric acid might occur. Reactor analysis showed that the crystallization of boric acid in the reactor is not possible during the long-term cooling period of LOCAs this is mainly due to the fact that the ice-condenser in the Loviisa plant contains sodium tetraborate Na2B4O710H2O (borax), which enters the reactor when emergency core cooling water is taken from the sump. Borax increases greatly the solubility of boric acid in water and, hence, decreases the risk of crystallization.
K. Hashimoto, M. Hirose, and T. Shibata, "Interpretation of Positive Scram Reactivity in the RBMK-1000 Reactor," Annals of Nuclear Energy, Vol. 21, No. 4, pp.211-217, 1994.
Abstract - The Positive scram reactivity in the Chernobyl-4 reactor is interpreted by breaking it up into three components by the use of the exact perturbation method. The three components are absorption, diffusion, and slowing down. The negative reactivity contribution of the absorber elements inserted is significantly reduced by the distortion of the axial flux shape and the removal of water from the bottom of the core by graphite follower contributes to the positive reactivity predominantly through absorption and diffusion effects. The total scram reactivity thus becomes positive. Further, Theoretical expression of the higher-order reactivity component induced by the flux distortion is developed on the basis of explicit higher-order perturbation formulation, The expression shows that the positive magnitude of the higher- order term is inversely proportional to the lambda-mode eigenvalue separation, The significantly small separation between the fundamental and the first axial- harmonic eigenvalue, which indicates spatial decoupling of the core is confirmed in the lambda-mode eigenvalue calculation, It is suggested that to ensure total scram reactivity is negative, the axial burnup distribution should be flattened and the follower dimensions should be modified.
SA. W. Lomperski and J. Kouhia, "Natural circulation Experiments with a VVER Reactor Geometry," Nuclear Engineering and Design, Vol. 147, No. 3, pp. 409-424, 1994.
Abstract - Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after Soviet-designed VVER pressurized water reactors. Using stepwise inventory reduction and small-break experiments, primary loop flow behaviour was studied over a range of coolant inventories. The tests revealed a trend toward decreasing primary side mass flow rate with inventory. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs and coolant flow into the hot legs changed from single to two-phase flow. The cause of this flow interruption was attributed to the hot legs loop seals, which are a unique feature of the VVER geometry. Finally, an experiment was conducted to demonstrate how loop seal refilling behaviour at low coolant inventories depends upon the steam flow rate through an individual hot leg. It was shown that loop seal refilling results when low steam velocities permit countercurrent flow in the upflow side of the seal.
V. I. Baranenko, V. S. Kirov, V. P. Kravchenko, V. A. Korovkin, and N. A. Fridman, "Temperature Oscillation in a High-Pressure Regeneration System of a Nuclear Power Plant with a VVER-1000 Reactor," Translated from Atomnaya Energiya, Vol. 76, No. 2, pp. 93-98, February 1994.
A. S. Timonin and S. A. Tsimbalov, "Quality of the Installment of Intrareactor Thermocouples in VVER Channels," Translated from Atomnaya Energiya, Vol. 76, No. 3, pp. 227-229, March 1994.
V. A. Kiselev, E. Yu. Rivkin, Yu. I. Smirnov, L. M. Sokov, and A. V. Sudakov, "Application of Leak-Before-Break Concept to Integrity and Safety of PWR Primary Piping with WWER- 1000," Nuclear Engineering and Design, Vol.151, No.2-3, pp. 409-424, 1994.
Abstract - The general methodology and criteria used for the leak-before-break (LBB) concept for the PWR primary circuit with WWER-1000 are described. The application of the LBB concept requires the following additional projections. First, the candidate system should be highly qualified piping, performed in accordance with the applicable regulations and guidelines, carefully screened to verify that they are not subjected to any disqualifying failure mechanism, Therefore the piping service experience and design bases are screened. Next, the analysis of leak-detection capabilities is performed. The postulated crack must be of sufficient size that the resulting leakage will be detected by the leak-detection system installed. Finally, fractured-mechanics analysis is performed to demonstrate that the crack will not cause pipe rupture even if seismic loads are applied before the flawed section is discovered and repaired. The calculated and experimental results obtained to date are very encouraging and show the applicability of the LBB concept for PWR primary circuit with WWER-1000.
L. M. Luzanova, V. N. Miglo, and P. D. Slavyagin, "Normalizing the Maximum Permissible Seal Failure of the Fuel Cladding of VVER and the Activity of the Fission Products in the Coolant," Translated from Atomnaya Energiya, Vol. 74, No. 6, pp. 491-497, June 1993.
Yu. V. Shvyryaev, V. B. Morozov, A. F. Barsukov, A. A. Derevyankin, and G. V. Tokmachev, "Probabilistic Safety Assessment for Nuclear Stations Containing VVER Reactors," Translated from Atomnaya Energiya, Vol. 74, No. 6, pp. 459-466, June 1993.
A. B. Borzakov, V. N. Proselkov, Yu. K. Bibilashvili, Yu. V. Pimenov, A. G. Ioltukhovskii, V. K. Chistyalova, B. A. Zaletnykh, and A. A. Enin, "Analysis of the Operational Reliability of VVER-1000 Fuel Elements and Bundles in a Three-Year Fuel Cycle," Translated from Atomnaya Energiya, Vol. 74, No. 6, pp. 482-486, June 1993.
V. D. Sidorenko and A. S. Shcheglov, "Neutron-Physics Characteristics of the VVER Core Which Affect the Operability of the Fuel Elements," Translated from Atomnaya Energiya, Vol. 74, No. 6, pp. 533-535, June 1993.
S. S. Anikanov, V. G. Dunaev, and V. I. Mitin, "Controlling VVER-1000 Power Distribution During Power Changes," Translated from Atomnaya Energiya, Vol. 75, No. 1, pp. 3-8, July 1993.
A. M. Fedosov, "Influence of Burnable Absorbers on the Dehydration Effect of a RBMK Reactor," Translated from Atomnaya Energiya, Vol. 75, No. 1, pp. 67-69, July 1993.
V. I. Baranenko, V. S. Kirov, V. P. Kravchenko, V. A. Korovkin, and N. A. Fridman, "Damage to the Output Manifolds of Steam Generators at Nuclear Power Plants with VVER- 1000 Reactors," Translated from Atomnaya Energiya, Vol. 75, No. 5, pp. 391-394, November 1993.
V. A. Kazakov and V. A. Levadnyi, "Noise Method for Measuring Reactivities and Monitoring the Appearance of a Steam-Gas Bubble in VVER-1000," Translated from Atomnaya Energiya, Vol. 75, No. 5, pp. 341-345, November 1993.
D. Marschke, R. Hennig, and G. Daum, "The Material Concept for VVER Plants in Comparison with Specifications in German Nuclear Safety Standards," Kerntechnik, Vol. 58, No.6, pp. 330-327, 1993.
Abstract - During the course of safety analyses, the materials concept of VVER plants was evaluated and compared with that of Konvoi plants based on the KTA nuclear safety standards. The general materials and toughness concept is outlined using the example of reactor coolant systems main components. The types of steel used are not directly comparable to the materials used in Konvoi plants The specified examination coverage of individual parts is generally not as detailed as that stipulated in KTA, and corresponds more closely with the specification of other nuclear engineering codes and standards such as the ACME code. In contrast to Western codes and standards, the toughness of ferritic materials used in VVER plants is determined within the framework of the RTndt concept solely on the basis of Charpy impact tests and not by means of Pellini drop weight testing. A correlation between these two techniques still remains to be developed. The chemical analysis limits in the core region for Cu, P, and Ni for newer plants correspond to the specifications laid down in the KTA standards. As the values are considerably higher than these limits in older VVER pants, particularly in the weld metal, recovery annealing is necessary in these plants.
P. E. Filimonova, "Study of Renewed Criticality of Fuel in Severe Accidents of VVER Reactors," Translated from Atomnaya Energiya, Vol. 75, No. 1, pp. 71-73, November 1992.
V. P. Denisov, V. V. Stekol'nikov and V. A. Voznesenskii, "Modernizing the VVER: a Soviet perspective," Nuclear Engineering International, July 1991.
"Overall Plant Design Descriptions, VVER, Water-Cooled, Water-Moderated, Energy Reactor," DOE/NE-0084 Revision 1, 1987.
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