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ARE REACTOR VESSEL LIDS CRACKING PHENOMENON ACCIDENTAL OR FORESEABLE

2. Contents . AbstractI.INTRODUCTIONII. MODELING AND CALCULATION II. A. Mathematical Modeling of Coolant Pressure OscillationsII.B. Results of EFOCP CalculationsIII MEASUREMENTS ON NPP WITH WWER-1000III. A. Means of measurementsIII. B. Measurement's results and analysisIV. NOVOVORONEZH NPP A

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ARE REACTOR VESSEL LIDS CRACKING PHENOMENON ACCIDENTAL OR FORESEABLE

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    1. ARE REACTOR VESSEL LIDS CRACKING PHENOMENON ACCIDENTAL OR FORESEABLE? Proskuryakov Konstantin N., Moscow Power Engineering Institute (Technical University), 14, Krasnokazarmennaya str. 111250, Moscow, Russia. Phone: +07 (095) 362-73-51, Fax: +07 (095) 362-73-51 e-mail: prosk@npp.mpei.ac.ru V.U. Hiretdinov OKB Gidropress, 21, Ordjonikidze str., 142103, Podolsk, Moscow region, Russia, phone: 007 (0967) 69-18-13, Fax: 007 (0967), 54 25-16, E-mail: grpress@grpress.podolsk.ru

    2. 2 Contents Abstract I.INTRODUCTION II. MODELING AND CALCULATION II. A. Mathematical Modeling of Coolant Pressure Oscillations II.B. Results of EFOCP Calculations III MEASUREMENTS ON NPP WITH WWER-1000 III. A. Means of measurements III. B. Measurements results and analysis IV. NOVOVORONEZH NPP AND DAVID- BESSE-1 NPP INCIDENTIII. V. PERSPECTIVES IY. CONCLUSIONS

    3. 3 Abstract some original academic and engineering R&D results carried out which that time advanced the current requirements of nuclear power. Acoustic method of early detection of steam cavity appearance in reactor core had been published before the accident at TMI-2. The researches lead to understanding new factors that at present time are characterized as concealed dynamic processes in NPP equipments. These processes play important role in the formation of vessel lid cracks and are not considered in design documentations and not indicated by ordinary operating diagnostic systems. . Dominant effects on the vibration level are imposed by the pressure oscillations of coolant under critical combination of thermalhydraulic and operating parameters. Results could be useful for achieving strategic goals: prevention of sudden failures and lifetime management.

    4. 4 I.INTRODUCTION In the present time priority problems of NPP improving is increase of : lifetime, an economic efficiency and safety. The perfection of diagnostic methods and devices, forecasting, expert judgment of equipment condition and reactor installation heat-transfer agent condition is one of the most important problems. One of the main factors determining the dynamic loads on the equipments is the movement of the fluid flow. . -

    5. 5 Resonant destruction of constructions takes place in cases when eigenfrequencies of oscillations of the coolant pressure (EFOCP) begin to be equal to the eigenfrequencies of structural oscillations. To prevent the appearance of the conditions for resonance interaction , it is necessary to provide the different frequencies for the self oscillations in the separated elements of the circulating system and also in the parts of the system formed by the comprising of these elements.

    6. 6

    7. 7 II. MODELING AND CALCULATION The principles of the mechanic systems mathematical modeling are known. The method of reference nodes in stiffness matrix, rod model or super cells (as in sub construction method) can be used as simplified models.

    8. 8 II. A. Mathematical Modeling of Coolant Pressure Oscillations Up to presence we worked out the range of necessary models and algorithms to evaluate eigenfrequences of acoustical, parametrical and system oscillation in NPP with WWER and PWR. The number of papers illustrates how these first time results were obtained: Analyzes show the topicality of researches taking in NPPD MPEI (TU), which that time advanced the current requirements of nuclear power. It should be noted, that we published recommendations about early detection of coolant boiling in WWER and PWR reactor core before the accident at "Three Mile Island" NPP. Specialists recognized this recommendation capability for prevention of operating personal actions, which consequently led to one of the most severe accident in international practice of nuclear power plants operating.

    9. 9 Interdisciplinary and composite character of R&D provided in particular possibility of: - forecasting dynamical processes and the existence of the system effects not covered by design documentation; - forecasting effects that depends not only on the parameters of the thermal-hydraulic processes, but also on the geometrical dimensions of the heat-removal circuit and the stage of the process of cooling down the core.

    10. 10 The complicity of mentioned approaches explain the reason why up to presence the problem of resonance interaction between coolant and mechanical structures of NPP and effective way of their prevention is not solved but must be included in the schedule of urgent R&D. Specially developed computer codes are used for ? substantiation serviceability of these systems. That fact, that modern code and, accordingly, the models incorporated in them are far from perfect, it is marked by ? number of authors. The greatest discrepancy of values flow rate and pressure predicted by codes to the data of experimental measurements is marked in area pulsating modes, i.e. in those modes when thermal hydraulic circuit shows itself as nonlinear oscillating system.

    11. 11 the reasons of loss of stability are caused by approach of ? study of phase transformations in the coolant at reactor core and approach of integrated parameters of system to maximum permissible values. Our Q- factor of steam generating ducts parametrical similarity criterion i.e.- Criterion of steam generating ducts parametrical similarity (CSGDPS) shows the strong influence of hydraulic resistances on two phases flow stability:

    12. 12 It is necessary emphasis that the hydraulic resistances of PSIS experimental units much more than in real systems. Corresponding reconstructions of PSIS experimental units based on these criterion indications were realized in Russia and will be realized by USA .

    13. 13 II.B. Results of EFOCP Calculations The composition of the acoustic elements for the primary circuit of NPP with WWER-1000 have been shown in Fig.1 and have been marked with numbers from 1 to 12.

    14. 14 Fig.1: Acoustic scheme for two-loop of reactor WWER-1000

    15. 15 Table The results of the calculations of EFOCP for the different reactor operation conditions for WWER-1000.

    16. 16 III MEASUREMENTS ON NPP WITH WWER-1000 The experiments done in operating reactors, including also the measurements done in first power unit of Volgodonskaya NPP (Vo NPP) with reactor WWER-1000 show that, the vibration properties of the reactor internals could be considered normal ones in the vast range of the changing operating parameters of reactor units.

    17. 17 Systems of preoperational vibrodynamic monitoring in a set of SPM during commission of NPP with VVER is presented on the Fig. 2. On this figures number 01 04 corresponds to different diagnostic systems as follows: 01: System of monitoring the reactor internals vibration characteristics (63 channels). 02: System of monitoring the reactor internals noise characteristics (12 channels). 03: System of monitoring the vibration characteristics of reactor fuel assembly simulators (78 channels). 04: System of monitoring the pressure pulsations and the vibration of the reactor plant equipment components (84 channels). 04: System of monitoring the pressure pulsations and the vibration of the reactor plant equipment components (84 channels).

    18. 18 They are intended for determination of : parameters of hydrodynamic disturbances (pressure pulsations in the in-pile part of the circuit, as well as in MCP loops); characteristics of dynamic response of the equipment monitored (vibration accelerations, velocities and displacements); indices of vibration state (dynamic deformations and stresses); interrelations of reactor vibration state and dynamic behavior of the internals components and equipment .

    19. 19 Fig.2. Systems of preoperational vibrodynamic monitoring in a set of SPM during commission of NPP with VVER.

    20. 20 III. A. Means of measurements For measuring of dynamic stresses, coolant pressure pulsations and equipment vibrations of MCL different types of transducers were used: tensometers, pressure pulsation and vibrations gauges. Primary measuring instruments is shown on Fig.3

    21. 21 Primary measuring instruments

    22. 22

    23. 23 Multi-channel software-hardware complexes, developed and manufactured according to the presented descriptions, were applied for preoperational dynamic measurements at Kalinin NPP (Unit 3), Khmelnitsky NPP (Unit 2), Rovno NPP (Unit 4) and Tianwan NPP (Units 1&2). The experience accumulated in the development will allow in future to use the data of software-hardware complex for solution of a number of problems in justification of vibration strength of RP equipment at NPP with VVER under operational loads and in the analysis of accident situations.

    24. 24 III. B. Measurements results and analysis Fig.5 The spectrum of vibro acceleration. Transducer 02BI-1 (installed at top reactor unit). For working MCP 1,2,3,4, P=15.9 MPa, T=270C.

    25. 25 The outstanding characteristic for this spectrum is the dominant peak in APSD of vibro accelerations equal to three times of rotation frequency of the pump, equal to 49.8 Hz. Its value is in some orders of magnitudes higher than the others, including the rotational frequency of the MCP equal 16.6 Hz. This feature of the spectrum of the vibrations of reactor vessel unit is appeared only in the shown regime, i.e. in the absence of the fission energy As it is clear from Fig.6 this pick is less in the operational condition of the reactor at low power. It is clear that the dominant effect for the vibration is the pulsations of coolant pressure. These verify the results of measurements of pressure pulsations given in Fig. 7 and Fig.8.

    26. 26 Fig.6. The spectrum of vibro acceleration. Transducer B11-2 (installed at lid of reactor unit). For working MCP 1,2,3,4, P=16 MPa, Tout. =281C., Tinl. =278C.

    27. 27 Fig.7. The spectrum of pressure pulsations. Transducer BP-8 (reactor internals, reactor barrel, exit from loop No 4). For working MCP 1,2,3,4, P=15.9 MPa, T=270C.

    28. 28 Fig.8. The spectrum of pressure pulsations. Transducer BP-11 (Main circulating pipeline, loop1, exit from reactor. For working MCP 1,2,3,4, P=16 MPa, Tout. =318C, Tinl. =287C

    29. 29 By using the values of EFOCP from the Table, it may reveal the reason for the increase of vibrations level in reactor lid unit at zero power condition. The values of EFOCP of reactor core at zero power level are equal to 49.3 Hz, i.e. its value is practically equal to the frequency oscillations of coolant that is caused by the rotation of MCP and hence strengthens them. The values of EFOCP RC at nominal reactor operating condition is equal to 41.5 Hz, i.e. its outside the range of mutual resonance of coolant pressure oscillations caused by the MCP rotations and the EFCPO RC. By reduction the power level of reactor, the values of EFCPO RC would approach to the range of mutual resonance effects.

    30. 30 IV. NOVOVORONEZH NPP AND DAVID- BESSE-1 NPP INCIDENT

    31. 31

    32. 32 V. PERSPECTIVES

    33. 33 The researches carried out, allow determining procedure for control of residual resource of reactor lid units and the timely warning to the personnel about achievement of a maximum permissible level of dynamic weariness of metal of reactor lid unit. This procedure is based on: ?) application of non-destructive testing for obtaining information on dynamic processes in the coolant and lid unit ; b) - substantiation of the types of detectors, their quantities and their placements; c) substantiation of periodicity of measurements of the specified signals

    34. 34 d) determination of ranges of reactor power level change and thermal-hydraulic parameters of the coolant in transient modes in which it is necessary to carry out continuous monitoring of signals, e) the statistical analysis of signal information in system of not-destructive testing, f) determination of the residual resource of metal of reactor lid unit in the zone of prospective destruction g) documentation of the results of measurements, the analysis and an estimation of a residual resource.

    35. 35 Fig. 9 Reactor vessel lid

    36. 36 Fig. 10 Damaged areas in the lid of the reactor vessel

    37. 37 IY. CONCLUSIONS Practice proves the existence of the concealed physical phenomena and processes in thermal hydraulic circuits of NPP, which can not be foreseen in design and normative documents and not predicted by the thermal hydraulic computer codes. the Non-Destructive Examination (NDE) system for the reactor vessel lid material does not exist. Obviously, it is necessary for all countries that own operating nuclear reactors of vessel types. These tasks could be accomplished in the framework of International Science and Technology Center (ISTC) project. The effectiveness of these works would be provided by the development of the original models of coolant oscillations for single and two-phase models and reliable computational models, and in some cases with predetermined measured data from NPPs.

    38. 38 Besides, the knowledge of know-how for development of methods and tools for diagnosis of cyclic load conditions providing resonance interaction with reactor vessel lid structures and also monitoring of the cracking appearance and its growth could be achieved The carried researches lead to understanding new factors that at present time are characterized as concealed dynamic processes in NPP equipments. These concealed dynamic processes in reactor vessel lid play important role in the formation of cracks and are not considered in design documentations. Based on these researches, new tools should be made preventing sudden occurrence of accidental conditions like the one in Three Mile Island-2 accident.

    39. 39 Summary Comparison of calculated and measured EFOCP values permits to understand the conditions that provide their increasing. Endorsement of hypothesis, that one of the main reasons of reactor vessel lid cracking is high cyclic load is worked out. . Valid demonstration predictability of thermalhydraulic parameters combinations that could provide resonant between mechanical and acoustical oscillations is presented. Control actions to prevent resonance interaction between coolant and equipment are recommended. Results are intended for vibroacoustic NPP equipment certification

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