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Sample records for main steam line

  1. Development of LBB Piping Evaluation Diagram for APR 1000 Main Steam Line Piping

    International Nuclear Information System (INIS)

    Yang, J. S.; Jeong, I. L.; Park, C. Y.; Bai, S. Y.

    2010-01-01

    This paper presents the piping evaluation diagram (PED) to assess the applicability of Leak-Before- Break(LBB) for APR 1000 main steam line piping. LBB-PED of APR 1000 main steam line piping is independent of its piping geometry and has a function of the loads applied in piping system. Also, in order to evaluate LBB applicability during construction process with only the comparative evaluation of material properties between actually used and expected, the expected changes of material properties are considered in the LBB-PED. The LBB-PED, therefore, can be used for quick LBB evaluation of APR 1000 main steam line piping of both design and construction

  2. Quad Cities Unit 2 Main Steam Line Acoustic Source Identification and Load Reduction

    International Nuclear Information System (INIS)

    DeBoo, Guy; Ramsden, Kevin; Gesior, Roman

    2006-01-01

    The Quad Cities Units 1 and 2 have a history of steam line vibration issues. The implementation of an Extended Power Up-rate resulted in significant increases in steam line vibration as well as acoustic loading of the steam dryers, which led to equipment failures and fatigue cracking of the dryers. This paper discusses the results of extensive data collection on the Quad Cities Unit 2 replacement dryer and the Main Steam Lines. This data was taken with the intent of identifying acoustic sources in the steam system. Review of the data confirmed that vortex shedding coupled column resonance in the relief and safety valve stub pipes were the principal sources of large magnitude acoustic loads in the main steam system. Modifications were developed in sub-scale testing to alter the acoustic properties of the valve standpipes and add acoustic damping to the system. The modifications developed and installed consisted of acoustic side branches that were attached to the Electromatic Relief Valve (ERV) and Main Steam Safety Valve (MSSV) attachment pipes. Subsequent post-modification testing was performed in plant to confirm the effectiveness of the modifications. The modifications have been demonstrated to reduce vibration loads at full Extended Power Up-rate (EPU) conditions to levels below those at Original Licensed Thermal Power (OLTP). (authors)

  3. Assessment of vibration anomalies of main steam lines at Palo Verde-3

    International Nuclear Information System (INIS)

    Amr, A.; Landstrom, C.; Maxwell, H.; Miller, J.S.; Lynch, J.J.

    1996-01-01

    Historically, flow induced vibration in piping systems that transport liquid has presented problems for plant designers. When evaluating a vibration problem, it is always important to determine the forcing frequencies from different phenomena and the natural frequencies of the system as an integral part of establishing the root cause of the problem. Since in most cases of large vibration and noise levels, the natural frequency of the system and the frequency of the flow induced vibration are very close, determining the natural frequency of the system is important. Palo Verde Unit-3 exhibited a vibration problem where identification of the root cause was difficult. A Palo Verde team was created which consisted of engineers from different on-site departments and support from consultants. The process used to determine the root cause for the vibration/noise problem on Main Steam Supply System (MSSS) steam line 2 at Palo Verde Unit 3 is discussed in this paper. Since the root cause was not readily apparent, a finite element model was constructed to determine the natural frequency of the piping system. The finite element model consisted of a portion of the main steam lines, including a sample line which traverses the main steam line

  4. NPP Krsko Containment Response Following Main Steam Line Break

    International Nuclear Information System (INIS)

    Spalj, S.; Grgic, D.; Cavlina, N.

    2000-01-01

    This paper presents the calculation of thermohydraulic environmental parameters (pressure and temperature) inside containment of Krsko NPP after postulated Main Steam Line Break (MSLB) accident. This analysis was done as a part of the ambient parameters specification in the frame of the NPP Krsko Equipment Qualification (EQ) project. The RELAP5/mod2 computer code was used for the determination of MSLB mass and energy release and computer code GOTHIC was used to calculate pressure and temperature profiles inside NPP Krsko containment. The analysis was performed for spectrum of break sizes to account for possible steam superheating during accidents with smaller break sizes. (author)

  5. Flow Instabilities and Main Steam Line Vibrations in a Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Henriksson, Mats; Westin, Johan; Granhall, Tord; Andersson, Lars; Bjerke, Lars-Erik

    2002-01-01

    Severe vibrational problems occurred in the main steam system of a PWR nuclear power plant, about 18 months after a steam generator replacement had been carried out. The magnitude of the vibrations reached levels at which the operators had to reduce power in order to stay within the operating limits imposed by the nuclear inspectorate. To solve the problem the following analyses methods were employed: - Testing the influence on vibration level from different modes of plant operation; - Analyses of plant measurement data; - Calculations of: hydraulic behaviour of the system, structural dynamic behaviour of the system, flow at the steam generator outlet. Scale model testing of the steam generator outlet region. Hydraulic flow disturbances in the main steam system were measured using pressure and strain gauges, which made it possible to track individual pressure pulses propagating through the main steam system. Analyses showed that the pressure pulses causing the vibration originated from the vicinity of the steam generator outlet. By using computer codes for network fluid flow analyses the pressure pulses found in the measurement traces could be generated in calculations. Careful studies of the flow at the steam generator outlet region, using model testing in a 1:3 scale model as well as transient 3D CFD calculations, gave clear indications that flow separation occurred at the steam generator outlet nozzle and at the first bend. Finally, by substituting the outlet nozzle for a different design with a multi-port nozzle, the steam line vibration problem has been solved. (authors)

  6. Simulation of a main steam line break with steam generator tube rupture using trace

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo, S.; Querol, A.; Verdu, G. [Departamento de Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Camino de Vera s/n, 46022, Valencia (Spain)

    2012-07-01

    A simulation of the OECD/NEA ROSA-2 Project Test 5 was made with the thermal-hydraulic code TRACE5. Test 5 performed in the Large Scale Test Facility (LSTF) reproduced a Main Steam Line Break (MSLB) with a Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR). The result of these simultaneous breaks is a depressurization in the secondary and primary system in loop B because both systems are connected through the SGTR. Good approximation was obtained between TRACE5 results and experimental data. TRACE5 reproduces qualitatively the phenomena that occur in this transient: primary pressure falls after the break, stagnation of the pressure after the opening of the relief valve of the intact steam generator, the pressure falls after the two openings of the PORV and the recovery of the liquid level in the pressurizer after each closure of the PORV. Furthermore, a sensitivity analysis has been performed to know the effect of varying the High Pressure Injection (HPI) flow rate in both loops on the system pressures evolution. (authors)

  7. AREVA main steam line break fully coupled methodology based on CATHARE-ARTEMIS - 15496

    International Nuclear Information System (INIS)

    Denis, L.; Jasserand, L.; Tomatis, D.; Segond, M.; Royere, C.; Sauvage, J.Y.

    2015-01-01

    The CATHARE code developed since 1979 by AREVA, CEA, EDF and IRSN is one of the major thermal-hydraulic system codes worldwide. In order to have at disposal realistic methodologies based on CATHARE for the whole transient and accident analysis in Chapter 15 of Safety Reports, a coupling with the code ARTEMIS was developed. ARTEMIS is the core code in AREVA's new reactor simulator system ARCADIA, using COBRA-FLX to model the thermal-hydraulics in the core. The Fully Coupled Methodology was adapted to the CATHARE-ARTEMIS coupling to perform Main Steam Line Break studies. This methodology, originally applied to the MANTA-SMART-FLICA coupling, is dedicated to Main Steam Line Break transients at zero power. The aim of this paper is to present the coupling between CATHARE and ARTEMIS and the application of the Fully Coupled Methodology in a different code environment. (authors)

  8. Design and performance of General Electric boiling water reactor main steam line isolation valves

    International Nuclear Information System (INIS)

    Rockwell, D.A.; van Zylstra, E.H.

    1976-08-01

    An extensive test program has been completed by the General Electric Company in cooperation with the Commonwealth Edison Company on the basic design type of large main steam line isolation valves used on General Electric Boiling Water Reactors. Based on a total of 40 tests under simulated accident conditions covering a wide range of mass flows, mixture qualities, and closing times, it was concluded that the commercially available valves of this basic type will close completely and reliably as required. Analytical methods to predict transient effects in the steam line and valve after postulated breaks were refined and confirmed by the test program

  9. Numerical simulations of pressure fluctuations at branch piping in BWR main steam line

    International Nuclear Information System (INIS)

    Morita, Ryo; Inada, Fumio; Yoshikawa, Kazuhiro; Takahashi, Shiro

    2009-01-01

    The power uprating of a nuclear power plant may increase/accelerate degradation phenomena such as flow-induced vibration and wall thinking. A steam dryer was damaged by a high cycle fatigue due to an acoustic-induced vibration at the branch piping of safety relief valves (SRVs) in main steam lines. In this study, we conducted the numerical simulations of steam/air flow around a simplified branch piping to clarify the basic characteristics of resonance. LES simulations were conducted in ordinary pressure/temperature air and steam under BWR plant conditions. In both cases, the excitation of the pressure fluctuations at the branch was observed under some inlet velocity conditions. These fluctuations and inlet conditions were normalized and the obtained results were compared. The normalized results showed that the range and maximum amplitude of pressure fluctuations were almost the same in low-pressure/temperature air and high-pressure/temperature steam. We found that ordinary pressure/temperature air experiments and simulations can possibly clarify the characteristics of the resonance in high-pressure/temperature steam. (author)

  10. Characteristics of fluctuating pressure generated in BWR main steam lines

    International Nuclear Information System (INIS)

    Takahashi, Shiro; Okuyama, Keita; Tamura, Akinori

    2009-01-01

    The BWR-3 steam dryer in the Quad Cities Unit 2 Nuclear Power Plant was damaged by high cycle fatigue due to acoustic-induced vibration. The dryer failure was as attributed to flow-induced acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSLs). The acoustic resonance was considered to be generated by interaction between the sound field and an unstable shear layer across the closed side branches with SRV stub pipes. We have started a research program on BWR dryers to develop their loading evaluation methods. Moreover, it has been necessary to evaluate the dryer integrity of BWR-5 plants which are the main type of BWR in Japan. In the present study, we used 1/10-scale BWR tests and analyses to investigate the flow-induced acoustic resonance and acoustic characteristics in MSLs. The test apparatus consisted of a steam dryer, a steam dome and 4 MSLs with 20 SRV stub pipes. A finite element method (FEM) was applied for the calculation of three-dimensional wave equations in acoustic analysis. We demonstrated that remarkable fluctuating pressures occurred in high and low frequency regions. High frequency fluctuating pressures was generated by the flow-induced acoustic resonance in the SRV stub pipes. Low frequency fluctuating pressure was generated in an MSL with the dead leg. The frequency of the latter almost coincided with the natural frequency of the MSL with the dead leg. The amplitude of the fluctuating pressures in the multiple stub pipes became more intense because of interaction between them compared with that in the single stub pipe. Acoustic analysis results showed that the multiple stub pipes caused several natural frequencies in the vicinity of the natural frequency of the single stub pipe and several modes of the standing wave in the MSLs. (author)

  11. Evaluation of acoustic resonance at branch section in main steam line. Part 1. Effects of steam wetness on acoustic resonance

    International Nuclear Information System (INIS)

    Uchiyama, Yuta; Morita, Ryo

    2011-01-01

    The power uprating of the nuclear power plant (NPP) is conducted in United States, EU countries and so on, and also is planned in Japan. However, the degradation phenomena such as flow-induced vibration and wall thinning may increase or expose in the power uprate condition. In U.S. NPP, the dryer had been damaged by high cycle fatigue due to acoustic-induced vibration under a 17% extended power uprating (EPU) condition. This is caused by acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSL). Increased velocity by uprating excites the pressure fluctuations and makes large amplitude resonance. To evaluate the acoustic resonance at the stub pipes of SRVs in actual BWR, it is necessary to clarify the acoustic characteristics in steam flow. Although there are several previous studies about acoustic resonance, most of them are not steam flow but air flow. Therefore in this study, to investigate the acoustic characteristics in steam flow, we conducted steam flow experiments in each dry and wet steam conditions, and also nearly saturated condition. We measured pressure fluctuation at the top of the single stub pipe and in main steam piping. As a result, acoustic resonance in dry steam flow could be evaluated as same as that in air flow. It is clarified that resonance amplitude of fluctuating pressure at the top of the stub pipe in wet steam was reduced to one-tenth compared with that in dry. (author)

  12. R.B. pressure and temperature transient following main steam line break

    International Nuclear Information System (INIS)

    Das, M.; Bhawal, R.N.; Prakash, P.

    1989-01-01

    The R.B. containment plays an important role in mitigating the consequences of any accident core. The analysis of Main Steam Line Break (MSLB), though not of relevance from activity release considerations, is essentially from structural integrity point of view. In this paper the outline of the likely scenario is drawn and the approach for thermal hydraulic simulation of the system for carrying out transient blowdown analysis is discussed. The results of the containment pressure and temperature transient analysis are also presented. (author). 4 refs., 7 figs

  13. RELAP5 analysis of PKL, main steam line break test

    Energy Technology Data Exchange (ETDEWEB)

    Jonnet, J.R.; Stempniewicz, M.M., E-mail: stempniewicz@nrg.eu; With, A. de; Wakker, P.H.

    2013-12-15

    Highlights: • RELAP5/MOD 3.2 code validation is performed by analyzing a main steam line break test in the PKL large scale test facility. • The RELAP5 model reproduces well the important events of the PKL test. • RELAP5 transient results show noticeable sensitivity to small differences in the initial conditions. • Accurate prediction of the coolant temperature is essential for the assessment of potential core re-criticality. - Abstract: PKL is a large scale test facility of the primary system owned by AREVA NP GmbH. It is used for extensive experimental investigations to study the integral behavior of Pressurized Water Reactor (PWR) plants under accident conditions. Since 2001, the test program is a part of an international cooperation project (SETH, followed by PKL1 and PKL2) set up by the OECD. The aim of the present work was to perform a short validation study of the thermo-hydraulics code RELAP5. A model of the PKL test facility has been developed, tested and applied to one of the experiments performed at the PKL. The chosen experiment was the test G3.1. In that experiment, a main steam line break occurs, causing a rapid depressurization of the affected steam generator. This leads to an increase of the heat transfer from the primary to the secondary side and thereby to a fast cool-down transient on the primary side. The main objective of this analysis was the qualification of the RELAP5 code results against heat transfer from the primary to the secondary side in both affected and intact loops, and temperatures in the primary system. The calculation results have been compared to the experimental results. It was concluded that the most important events during the test are reproduced relatively well by the model. The calculated coolant temperature in the core is higher than in the experiment. The minimum temperature is about 5% higher than measured. The secondary pressures in SG-1, 3, and 4 is in very good agreement with the experimental value, but in the

  14. Lungmen ABWR containment analyses during short-term main steam line break LOCA using GOTHIC

    International Nuclear Information System (INIS)

    Chen, Yen-Shu; Yuann, Yng-Ruey; Dai, Liang-Che

    2012-01-01

    Highlights: ► The Lungmen ABWR containment responses due to the main steam line break are analyzed. ► In the Lungmen FSAR, the peak drywell temperature is greater than the designed value. ► GOTHIC is used to calculate the containment responses in this study. ► With more realistic conditions, the drywell temperature can be reasonably suppressed. - Abstract: Lungmen Nuclear Power Plant in Taiwan is a GE-designed twin-unit Advanced Boiling Water Reactor (ABWR) plant with rated thermal power of 3926 MWt. Both units are currently under construction. In the Lungmen Final Safety Analysis Report (FSAR) section 6.2, the calculated peak drywell temperature during the short-term Main Steam Line Break (MSLB) event is 176.3 °C, which is greater than the designed temperature of 171.1 °C. It resulted in a controversial issue in the FSAR review process conducted by the Atomic Energy Council in Taiwan. The purpose of this study is to independently investigate the Lungmen ABWR containment pressure and temperature responses to the MSLB using the GOTHIC program. Blowdown conditions are either calculated by using a simplified reactor vessel volume in GOTHIC model, or provided by the RELAP5 transient analysis. The blowdown flow rate from the steam header side is calculated with a more reasonable pressure loss coefficient of the open main steam isolation valves, and the peak drywell temperature is then reduced. By using the RELAP5 blowdown data, the peak drywell temperature can be further reduced because of the initial liquid entrainment in the blowdown flow. The drywell space is either treated as a single volume, or separated into a upper drywell and a lower drywell to reflect the real configuration of the Lungmen containment. It is also found that a single drywell volume may not present the overheating of the upper drywell. With more realistic approaches and assumptions, the drywell temperature can be reasonably below the design value and the Lungmen containment integrity

  15. Structural analysis of steam generator internals following feed water main steam line break: DLF approach

    International Nuclear Information System (INIS)

    Bhasin, Vivek; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1993-01-01

    In order to evaluate the possible release of radioactivity in extreme events, some postulated accidents are analysed and studied during the design stage of Steam Generator (SG). Among the various accidents postulated, the most important are Feed Water Line Break (FWLB) and Main Steam Line Break (MSLB). This report concerns with dynamic structural analysis of SG internals following FWLB/MSLB. The pressure/drag-force time histories considered were corresponding to the conditions leading to the accident of maximum potential. The SG internals were analysed using two approaches of structural dynamics. In first approach simplified DLF method was adopted. This method yields an upper bound values of stresses and deflection. In the second approach time history analysis by Mode Superposition Technique was adopted. This approach gives more realistic results. The structure was qualified as per ASME B and PV Code SecIII NB. It was concluded that in all the components except perforated flow distribution plate, the stress values based on elastic analysis are within the limits specified by ASME Code. In case of perforated flow distribution plate during the MSLB transient the stress values based on elastic analysis are higher than the ASME Code limits. Therefore, its limit load analysis had to be done. Finally, the collapse pressure evaluated using limit load analysis was shown to be within the limits of ASME B and PV Code SecIII Nb. (author). 31 refs., 94 figs., 16 tabs

  16. Main steam line break accident simulation of APR1400 using the model of ATLAS facility

    Science.gov (United States)

    Ekariansyah, A. S.; Deswandri; Sunaryo, Geni R.

    2018-02-01

    A main steam line break simulation for APR1400 as an advanced design of PWR has been performed using the RELAP5 code. The simulation was conducted in a model of thermal-hydraulic test facility called as ATLAS, which represents a scaled down facility of the APR1400 design. The main steam line break event is described in a open-access safety report document, in which initial conditions and assumptionsfor the analysis were utilized in performing the simulation and analysis of the selected parameter. The objective of this work was to conduct a benchmark activities by comparing the simulation results of the CESEC-III code as a conservative approach code with the results of RELAP5 as a best-estimate code. Based on the simulation results, a general similarity in the behavior of selected parameters was observed between the two codes. However the degree of accuracy still needs further research an analysis by comparing with the other best-estimate code. Uncertainties arising from the ATLAS model should be minimized by taking into account much more specific data in developing the APR1400 model.

  17. HEXTRAN-SMABRE calculation of the 6th AER Benchmark, main steam line break in a WWER-440 NPP

    International Nuclear Information System (INIS)

    Haemaelaeinen, A.; Kyrki-Rajamaeki, R.

    2003-01-01

    The sixth AER benchmark is the second AER benchmark for couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a double end break of one main steam line in a WWER-440 plant. The core is at the end of its first cycle in full power conditions. In VTT HEXTRAN2.9 is used for the core kinetics and dynamics and SMABRE4.8 as a thermal hydraulic model for the primary and secondary loop. The plant model for SMABRE consists mainly of two input models, Loviisa model and a standard WWER-440/213 plant model. The primary loop includes six separate loops, the pressure vessel is divided into six parallel channels in SMABRE and the whole core calculation is performed in the core with HEXTRAN. The horizontal steam generators are modelled with heat transfer tubes in five levels and vertically with two parts, riser and downcomer. With this kind of detailed modelling of steam generators there occurs strong flashing after break opening. As a sequence of the main steam line break at nominal power level, the reactor trip is followed quite soon. The liquid temperature continues to decrease in one core inlet sector which may lead to recriticality and neuron power increase. The situation is very sensitive to small changes in the steam generator and break flow modelling and therefore several sensitivity calculations have been done. Also two stucked control rods have been assumed. Due to boric acid concentration in the high pressure safety injection subcriticality is finally guaranteed in the transient (Authors)

  18. Development of phenomena identification and ranking table for APR1400 main steam line break

    International Nuclear Information System (INIS)

    Song, J. H.; Chung, B. D.; Jeong, J. J.

    2003-01-01

    A Phenomena Identification and Ranking Table (PIRT) was developed for the Main Steam Line Break (MSLB) event of an APR-1400 (Advanced Power Reactor-1400). A team of experts from research institutes, industries, and regulatory bodies participated in the development. The selected event was a double-ended steam line break at full power with the reactor coolant pump running. The panel selected the fuel performance as the primary safety criterion for ranking. The plant design data, the results of APR-1400 safety analysis, and the results of additional best estimate analysis by MARS2.1 were utilized. Three phases of pre-trip, rapid cool-down, and safety injection phase were identified. Then, the ranking of a system, components, phenomenon/process based on the relative importance to the primary evaluation criterion were followed for each time phase. Finally, the knowledge-level for each important process in the component was ranked in terms of the existing knowledge. The highly ranked phenomena identified for APR-1400 MSLB are tube wall heat transfer at the steam generator shell, void distribution at the steam generator shell, liquid entrainment in the separators, mixture level in the separators, boron mixing in the upper down comer, boron transport and thermal mixing in the lower plenum, stored energy release in the upper head, and flow to and/from the upper head. The PIRT will be used as a guide in planning cost effective experimental programs and code development efforts, especially for the quantification of the process and/or phenomena, which have a high importance but low knowledge level

  19. Simulation and analysis of a main steam line transient with isolation valves closure and subsequent pipe break

    Energy Technology Data Exchange (ETDEWEB)

    Stevanovic, Vladimir; Studovic, Milovan [Faculty of Mechanical Engineering, University of Belgrade, Belgrade (Yugoslavia); Bratic, Aleksandar [Thermal Power Plant Nikola Tesla (Yugoslavia)

    1993-11-01

    Simulation and analysis of a real main steam line break transient at the coal fired 300 MW Drmno Thermal Power Plant have been performed by the computer code TEA-01. The methods and procedures used could be applied to a nuclear power plant. 9 refs., 6 figs.

  20. Code Assessment of SPACE 2.19 using LSTF 10% Main Steam-Line-Break Test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minhee; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The Safety and Performance Analysis Code for Nuclear Power Plants (SPACE) has been developed in recent years by the Korea Hydro and Nuclear Power Co. through collaborative works with other Korean nuclear industries and research institutes. As a result of the development, the 2.19 version of the code was released through the successive various verification and validation works. In this study, results produced by the SPACE 2.19 code were compared with the experimental data from JAERI's LSTF Test Run SBSL- 01 for a 10% main steam line break transient in a pressurized water reactor. The LSTF 10% main steam line break test were simulated using the SPACE 2.19 for code V and V work. The overall comparisons between the SPACE 2.19 code prediction and the LSTF Test Run SB-SL-01 experimental data are reasonably satisfactory. The comparisons were conducted in terms of the variations of mass flow rate, void fraction, pressure, collapsed liquid level, temperature, and system flow rate for the transient. In addition, the input model was modified for simulation accuracy of PZR pressure based on the calculated results. The correction of PORV setpoint affects to simulate the PORV open and close phenomena similarly with experiments. From the modification, the computed results show a reasonable agreement with experimental data in overall transient time.

  1. Development and application of an entrainment model for the PWR U-tube steam generators for main steam line break analysis

    International Nuclear Information System (INIS)

    Song, Dong-Soo; Park, Young-Chan

    2004-01-01

    The purpose of this paper is to present the analyses that were performed to develop and use an entrainment model for pressurized water reactor U-tube steam generators (SG) for main steam line break (MSLB) analyses. The entrainment model was developed using the RETRAN-3D computer program, and the model was benchmarked against experimental data of moisture carryover during a simulated MSLB accident. The application methodology was also developed to incorporate into the MSLB mass and energy release calculations for Kori Unit 1. This methodology utilizes LOFTRAN and RETRAN-3D codes in an iterative sequence of cases in which the LOFTRAN nuclear steam supply system model provides boundary conditions for the RETRAN-3D broken loop steam generator model, and the RETRAN-3D model provides the entrainment data that is input back into the LOFTRAN model. FORTRAN programs were developed to facilitate the sequencing of these iterative calculations. As a result of applying the entrainment model to Kori Unit 1, the temperature calculated inside Containment during MSLB accident using the CONTEMP-LT computer program decreased by about 25degC. Consequently this entrainment model provides a significant benefit by decreasing the temperature envelop for environment qualification as well as decreasing the peak Containment pressure. (author)

  2. Analysis of containment parameters during the main steam line break with the failure of the feedwater control valves

    International Nuclear Information System (INIS)

    Fabjan, L.; Petelin, S.; Mavko, B.; Gortnar, O.; Tiselj, I.

    1992-01-01

    U.S. Nuclear Regulatory Commission (NRC) information notice 91-69: 'Errors in Main Steam Line Break Analyses for Determining Containment Parameters' shows the possibility of an accident which could lead to beyond design containment pressure and temperature. Such accident would be caused by the continuation of feedwater flow following a main stream line break (MSLB) inside the containment. Krsko power plant already experienced problems with main feedwater control valves. For that reason, analysis of MSLB has been performed taking into account continuous feedwater addition scenario and different containment safety systems capabilities availability. Steam and water released into the containment during MSLB was calculated using RELAP5/MOD2 computer code. The containment response to MSLB was calculated using CONTEMPT-LT/028 computer code. The results indicated that the continuous feedwater flow following a MSLB could lead to beyond design containment pressure. The peak pressure and temperature depend on isolation time for main- and auxiliary-feedwater supply. In the case of low boron concentration injection, the core recriticality is characteristic for this type of accidents. It was concluded that the presented analysis of MSLB with continuous feedwater addition scenario is the worst case for containment design

  3. Signal analysis of acoustic and flow-induced vibrations of BWR main steam line

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Paredes, G., E-mail: gepe@xanum.uam.mx [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Prieto-Guerrero, A. [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Núñez-Carrera, A. [Comisión Nacional de Seguridad Nuclear y Salvaguardias, Doctor Barragán 779, Col. Narvarte, México, D.F. 03020 (Mexico); Vázquez-Rodríguez, A. [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Centeno-Pérez, J. [Instituto Politécnico Nacional, Escuela Superior de Física y Matemáticas Unidad Profesional “Adolfo López Mateos”, Av. IPN, s/n, México, D.F. 07738 (Mexico); Espinosa-Martínez, E.-G. [Departamento de Sistemas Energéticos, Universidad Nacional Autónoma de México, México, D.F. 04510 (Mexico); and others

    2016-05-15

    Highlights: • Acoustic and flow-induced vibrations of BWR are analyzed. • BWR performance after extended power uprate is considered. • Effect of acoustic side branches (ASB) is analyzed. • The ASB represents a reduction in the acoustic loads to the steam dryer. • Methodology developed for simultaneous analyzing the signals in the MSL. - Abstract: The aim of this work is the signal analysis of acoustic waves due to phenomenon known as singing in Safety Relief Valves (SRV) of the main steam lines (MSL) in a typical BWR5. The acoustic resonance in SRV standpipes and fluctuating pressure is propagated from SRV to the dryer through the MSL. The signals are analyzed with a novel method based on the Multivariate Empirical Mode Decomposition (M-EMD). The M-EMD algorithm has the potential to find common oscillatory modes (IMF) within multivariate data. Based on this fact, we implement the M-EMD technique to find the oscillatory mode in BWR considering the measurements obtained collected by the strain gauges located around the MSL. These IMF, analyzed simultaneously in time, allow obtaining an estimation of the effects of the multiple-SRV in the MSL. Two scenarios are analyzed: the first is the signal obtained before the installation of the acoustic dampers (ASB), and the second, the signal obtained after installation. The results show the effectiveness of the ASB to damp the strong resonances when the steam flow increases, which represents an important reduction in the acoustic loads to the steam dryer.

  4. Signal analysis of acoustic and flow-induced vibrations of BWR main steam line

    International Nuclear Information System (INIS)

    Espinosa-Paredes, G.; Prieto-Guerrero, A.; Núñez-Carrera, A.; Vázquez-Rodríguez, A.; Centeno-Pérez, J.; Espinosa-Martínez, E.-G.

    2016-01-01

    Highlights: • Acoustic and flow-induced vibrations of BWR are analyzed. • BWR performance after extended power uprate is considered. • Effect of acoustic side branches (ASB) is analyzed. • The ASB represents a reduction in the acoustic loads to the steam dryer. • Methodology developed for simultaneous analyzing the signals in the MSL. - Abstract: The aim of this work is the signal analysis of acoustic waves due to phenomenon known as singing in Safety Relief Valves (SRV) of the main steam lines (MSL) in a typical BWR5. The acoustic resonance in SRV standpipes and fluctuating pressure is propagated from SRV to the dryer through the MSL. The signals are analyzed with a novel method based on the Multivariate Empirical Mode Decomposition (M-EMD). The M-EMD algorithm has the potential to find common oscillatory modes (IMF) within multivariate data. Based on this fact, we implement the M-EMD technique to find the oscillatory mode in BWR considering the measurements obtained collected by the strain gauges located around the MSL. These IMF, analyzed simultaneously in time, allow obtaining an estimation of the effects of the multiple-SRV in the MSL. Two scenarios are analyzed: the first is the signal obtained before the installation of the acoustic dampers (ASB), and the second, the signal obtained after installation. The results show the effectiveness of the ASB to damp the strong resonances when the steam flow increases, which represents an important reduction in the acoustic loads to the steam dryer.

  5. Valve for closing a steam line

    International Nuclear Information System (INIS)

    Meyer, W.; Potrykus, G.

    1976-01-01

    Instead of several control elements, the quick-closing valve, especially in the main-steam line between steam generator and turbine of a power station has the valve cone itself as the only movable part, acting with its inner surface as a piston within a second cylinder space. The valve shaft is at the same time a piston rod with a stepped piston at the upper end. This piston is loaded in a cylinder at the upspace below the valve cover on one hand by a spring, on the other hand by its own medium. Two non-return valves, one of it in a bore of the valve cone, connect the first-mentioned cylinder space with the steam-loaded inlet resp. outlet side of the valve. For controlling the valve, a magnet valve is sufficient. By automatic control of the valve cone coupled with several pistons several control lines can be omitted. There are also no pressurized control lines outside the valve which could be damaged by exterior influences. (ERA) [de

  6. Steam line break analysis in CAREM-25 reactor

    International Nuclear Information System (INIS)

    Zanocco, Pablo; Gimenez, Marcelo O.; Vertullo, Alicia; Schlamp, Miguel A.; Garcia, Alicia E.

    2000-01-01

    The main objective of this report is to analyze the reactor response during a steam line break postulated accident with RELAP5, a plant code using a separated flow model. The steam line break caused a rapid blowdown of the secondary coolant increasing the heat removal in the steam generator. As a consequence and due to reactor features the core power is also increased. As maximum removed power in the secondary side is highly dependant on the total water volume evaporated during the accident a detailed model of feed water and outlet steam pipes is provided. Different cases are analyzed with and without feedwater system and considering the fail or success of the First Shutdown System. In all the sequences the DNBR and CPR remain above the minimum safety values established by design. Further calculations concerning depressurization ramps and steam generator feed water pumps response during depressurization are advised. (author)

  7. Steam Line Break Analysis in CAREM-25 Reactor

    International Nuclear Information System (INIS)

    Zanocco, Pablo; Gimenez, Marcelo; Vertullo, Alicia; Garcia, A; Schlamp, Miguel

    2000-01-01

    The main objective of this report is to analyze the reactor response during a steam line break postulated accident with RELAP5, a plant code using a separated flow model.The steam line break caused a rapid blowdown of the secondary coolant increasing the heat removal in the steam generator.As a consequence and due to reactor features the core power is also increased.As maximum removed power in the secondary side is highly dependant on the total water volume evaporated during the accident a detailed model of feed water and outlet steam pipes is provided.Different cases are analyzed with and without feedwater system and considering the fail or success of the First Shutdown System.In all the sequences the DNBR and CPR remain above the minimum safety values established by design.Further calculations concerning depressurization ramps and steam generator feed water pumps response during depressurization are advised

  8. Coupled simulation of steam line break accident

    International Nuclear Information System (INIS)

    Royer, E.; Raimond, E.; Caruge, D.

    2000-01-01

    The steam line break is a PWR type reactor design accident, which concerns coupled physical phenomena. To control these problems simulation are needed to define and validate the operating procedures. The benchmark OECD PWR MSLB (Main Steam Line Break) has been proposed by the OECD to validate the feasibility and the contribution of the multi-dimensional tools in the simulation of the core transients. First the benchmark OECD PWR MSLB is presented. Then the analysis of the three exercises (system with pinpoint kinetic, three-dimensional core and whole system with three-dimensional core) are discussed. (A.L.B.)

  9. Study of condensation heat transfer following a main steam line break inside containment

    Energy Technology Data Exchange (ETDEWEB)

    Cho, J.H.; Elia, F.A. Jr.; Lischer, D.J. [Stone & Webster Engineering Corporation, Boston, MA (United States)

    1995-09-01

    An alternative model for calculating condensation heat transfer following a main stream line break (MSLB) accident is proposed. The proposed model predictions and the current regulatory model predictions are compared to the results of the Carolinas Virginia Tube Reactor (CVTR) test. The very conservative results predicted by the current regulatory model result from: (1) low estimate of the condensation heat transfer coefficient by the Uchida correlation and (2) neglecting the convective contribution to the overall heat transfer. Neglecting the convection overestimates the mass of steam being condensed and does not permit the calculation of additional convective heat transfer resulting from superheated conditions. In this study, the Uchida correlation is used, but correction factors for the effects of convection an superheat are derived. The proposed model uses heat and mass transfer analogy methods to estimate to convective fraction of the total heat transfer and bases the steam removal rate on the condensation heat transfer portion only. The results predicted by the proposed model are shown to be conservative and more accurate than those predicted by the current regulatory model when compared with the results of the CVTR test. Results for typical pressurized water reactors indicate that the proposed model provides a basis for lowering the equipment qualification temperature envelope, particularly at later times following the accident.

  10. Analyses in support of installation of steam-dump-to-atmosphere valves at steam lines of the Dukovany NPP

    International Nuclear Information System (INIS)

    Kral, P.

    1998-01-01

    Four conservative analyses were carried out with a view to examining the cooldown capacity of the super-emergency feedwater pump (SEFWP) → steam generator (SG) → steam dump to atmosphere/main steam line (SDA/MSL) chain. This emergency cooldown capacity was investigated for a postulated accident associated with a main steam header break + main feedwater header break + closing of all main steam lines, and for an artificial accident with SCRAM + isolation of all MSLs + loss of feedwater. The RELAP5/MOD3.1 code and a detailed 3-loop input model of the Dukovany plant were employed. Conservative assumptions with respect to the initial reactor power, decay heat evolution, and other input parameters were applied. The results gave evidence that the capacity of both the 2SEFWP → 2SG → 2SDA/SG and 1SEFWP → 1SG → 1SDA/SG chains is sufficient for the decay heat to be removed from the reactor; however, a considerably long time allowing for a sufficient drop of the decay heat is necessary for a deep cooldown of the primary circuit. For the event encompassing main steam header break + main feedwater header break with isolation of all MSLs and with cooling by 2SEFWPs, a time-consuming calculation gave evidence of the feasibility of passing to the water-water regime and primary system cooldown to below 93 deg C in the hot legs

  11. Analysis of the OECD main steam line break benchmark using ANC-K/MIDAC code

    International Nuclear Information System (INIS)

    Aoki, Shigeaki; Tahara, Yoshihisa; Suemura, Takayuki; Ogawa, Junto

    2004-01-01

    A three-dimensional (3D) neutronics and thermal-and-hydraulics (T/H) coupling code ANC-K/MIDAC has been developed. It is the combination of the 3D nodal kinetic code ANC-K and the 3D drift flux thermal hydraulic code MIDAC. In order to verify the adequacy of this code, we have performed several international benchmark problems. In this paper, we show the calculation results of ''OECD Main Steam Line Break Benchmark (MSLB benchmark)'', which gives the typical local power peaking problem. And we calculated the return-to-power scenario of the Phase II problem. The comparison of the results shows the very good agreement of important core parameters between the ANC-K/MIDAC and other participant codes. (author)

  12. Response of the steam generator VVER 1000 to a steam line break

    International Nuclear Information System (INIS)

    Novotny, J.; Novotny, J. Jr.

    2003-01-01

    Dynamic effects of a steam line break in the weld of the steam pipe and the steam collector on the steam generator system are analyzed. Modelling of a steam line break may concern two cases. The steam line without a restraint and the steam line protected by a whip restraint with viscous elements applied at the postulated break cross-section. The second case is considered. Programme SYSTUS offers a special element the stiffness and viscous damping coefficients of which may be defined as dependent on the relative displacement and velocity of its nodes respectively. A circumferential crack is simulated by a sudden decrease of longitudinal and lateral stiffness coefficients of these special SYSTUS elements to zero. The computation has shown that one can simulate the pipe to behave like completely broken during a time interval of 0,0001 s or less. These elements are used to model the whip restraint with viscous elements and viscous dampers of the GERB type as well. In the case of a whip restraint model the stiffness coefficient-displacement relation and damping coefficient - velocity relation are chosen to fit the given characteristics of the restraint. The special SYSTUS elements are used to constitute Maxwell elements modelling the elasto-plastic and viscous properties of the GERB dampers applied to the steam generator. It has been ascertained that a steam line break at the postulated weld crack between the steam pipe and the steam generator collector cannot endanger the integrity of the system even in a case of the absence of a whip restraint effect. (author)

  13. Neutronic calculations for Angra-1 steam line break accident

    International Nuclear Information System (INIS)

    Ponzoni Filho, Pedro; Sato, Sadakatu

    2000-01-01

    The reduction of boron concentration in the Boron Injection Tank (BIT), to the room temperature solubility level, makes necessary a reanalysis of the steam line break accident of Angra 1 NPP. This paper describes the neutronic calculation related to this reanalysis. The main steps of the work were: review of reactivity parameters used in the accident simulation; search of xenon profiles that cause the most severe core power distribution; calculation of hot channel factors and other neutronic parameters necessary for DNBR determination. The final conclusion, related to the steam line break accident, states the BIT concentration may be reduced to 2000 ppm. (author)

  14. Evaluation of a main steam line break with induced, multiple tube ruptures: A comparison of NUREG 1477 (Draft) and transient methodologies Palo Verde Nuclear Generating Station

    Energy Technology Data Exchange (ETDEWEB)

    Parrish, K.R.

    1995-09-01

    This paper presents the approach taken to analyze the radiological consequences of a postulated main steam line break event, with one or more tube ruptures, for the Palo Verde Nuclear Generating Station. The analysis was required to support the restart of PVNGS Unit 2 following the steam generator tube rupture event on March 14, 1993 and to justify continued operation of Units 1 and 3. During the post-event evaluation, the NRC expressed concern that Unit 2 could have been operating with degraded tubes and that similar conditions could exist in Units 1 and 3. The NRC therefore directed that a safety assessment be performed to evaluate a worst case scenario in which a non-isolable main steam line break occurs inducing one or more tube failures in the faulted steam generator. This assessment was to use the generic approach described in NUREG 1477, Voltage-Based Interim Plugging Criteria for Steam Generator Tubes - Task Group Report. An analysis based on the NUREG approach was performed but produced unacceptable results for off-site and control room thyroid doses. The NUREG methodology, however, does not account for plant thermal-hydraulic transient effects, system performance, or operator actions which could be credited to mitigate dose consequences. To deal with these issues, a more detailed analysis methodology was developed using a modified version of the Combustion Engineering Plant Analysis Code, which examines the dose consequences for a main steam line break transient with induced tube failures for a spectrum equivalent to 1 to 4 double ended guillotine U-tube breaks. By incorporating transient plant system responses and operator actions, the analysis demonstrates that the off-site and control room does consequences for a MSLBGTR can be reduced to acceptable limits. This analysis, in combination with other corrective and recovery actions, provided sufficient justification for continued operation of PVNGS Units 1 and 3, and for the subsequent restart of Unit 2.

  15. Evaluation of a main steam line break with induced, multiple tube ruptures: A comparison of NUREG 1477 (Draft) and transient methodologies Palo Verde Nuclear Generating Station

    International Nuclear Information System (INIS)

    Parrish, K.R.

    1995-01-01

    This paper presents the approach taken to analyze the radiological consequences of a postulated main steam line break event, with one or more tube ruptures, for the Palo Verde Nuclear Generating Station. The analysis was required to support the restart of PVNGS Unit 2 following the steam generator tube rupture event on March 14, 1993 and to justify continued operation of Units 1 and 3. During the post-event evaluation, the NRC expressed concern that Unit 2 could have been operating with degraded tubes and that similar conditions could exist in Units 1 and 3. The NRC therefore directed that a safety assessment be performed to evaluate a worst case scenario in which a non-isolable main steam line break occurs inducing one or more tube failures in the faulted steam generator. This assessment was to use the generic approach described in NUREG 1477, Voltage-Based Interim Plugging Criteria for Steam Generator Tubes - Task Group Report. An analysis based on the NUREG approach was performed but produced unacceptable results for off-site and control room thyroid doses. The NUREG methodology, however, does not account for plant thermal-hydraulic transient effects, system performance, or operator actions which could be credited to mitigate dose consequences. To deal with these issues, a more detailed analysis methodology was developed using a modified version of the Combustion Engineering Plant Analysis Code, which examines the dose consequences for a main steam line break transient with induced tube failures for a spectrum equivalent to 1 to 4 double ended guillotine U-tube breaks. By incorporating transient plant system responses and operator actions, the analysis demonstrates that the off-site and control room does consequences for a MSLBGTR can be reduced to acceptable limits. This analysis, in combination with other corrective and recovery actions, provided sufficient justification for continued operation of PVNGS Units 1 and 3, and for the subsequent restart of Unit 2

  16. Simulation of the OECD Main-Steam-Line-Break Benchmark Exercise 3 Using the Coupled RELAP5/PANTHER Codes

    International Nuclear Information System (INIS)

    Schneidesch, Christophe R.; Zhang Jinzhao

    2004-01-01

    The RELAP5 best-estimate thermal-hydraulic system code has been coupled with the PANTHER three-dimensional neutron kinetics code via the TALINK dynamic data exchange control and processing tool. The coupled RELAP5/PANTHER code package has been qualified and will be used at Tractebel Engineering (TE) for analyzing asymmetric pressurized water reactor (PWR) accidents with strong core-system interactions. The Organization for Economic Cooperation and Development/U.S. Nuclear Regulatory Commission PWR main-steam-line-break benchmark problem was analyzed as part of the qualification efforts to demonstrate the capability of the coupled code package of simulating such transients. This paper reports the main results of TE's contribution to the benchmark Exercise 3

  17. CATHENA analysis of CANDU 6 steam generators for steam main break at a remote location

    Energy Technology Data Exchange (ETDEWEB)

    Liu, T. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2009-07-01

    CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a nonequilibrium, two-phase, two fluid network analysis code that has been in use for over two decades by various groups in Canada and around the world. It is the primary Thermalhydraulics network analysis tool used by Atomic Energy Canada Ltd. (AECL) in the design, safety and licensing analysis of power and research reactors as well as test facilities. In the thermalhydraulic model, the liquid and vapor phases may have different pressures, velocities, and temperatures. The objective of the present paper is to present the detailed modeling of a CANDU 6 Steam Generator (SG) using the transient, thermalhydraulics network code CATHENA. The model represents the secondary side, primary side and the main steam system including the main steam line up to the assumed break location. The present model is designed such that the transient pressure drops across Tube Support Plates (TSP) could be extracted. The resistances of degraded/fouled TSPs were modeled by using the increased/reduced flow area of the TSPs. CATHENA then calculates the flow resistance in two-phase flow based on the area contraction/expansion at the TSPs. Three sets of simulations were performed; one with the degraded steam generator data provided by the utility users, and the other two with waterlanced (cleaned with high pressure water jet) TSPs. One run assumed the flow area increased by 25 percent, the other run assumed the flow area increased by 50 percent. on the hot side of the SG. No significant changes in the break discharge rates were observed between the runs. However, the steam generator downcomer flow for the waterlanced case did not reverse during the blowdown as was calculated for the degraded case. As expected, the pressure drop across the TSPs was decreased in the waterlanced cases comparing with degraded cases. The CATHENA simulation provides estimates of the velocity, density, and quality in the tube bundle as well as

  18. The effect of steam separataor efficiency on transient following a steam line break

    International Nuclear Information System (INIS)

    Choi, J.H.; Ohn, M.Y.; Lee, N.H.; Hwang, S.T.; Lee, S.K.

    1996-01-01

    Detailed thermalhydraulic simulations for CANDU 6 steam line break inside containment are performed to predict the response of the primary and secondary circuits. The analysis is performed using the thermalhydraulic computer code, CATHENA, with a coupled primary and secondary circuit model. A two-loop representation of the primary and secondary circuits is modelled. The secondary circuit model includes the feedwater line from the deaerator storage tank, multi-node steam generators and the steam line up to the turbine. Two cases were carried out using different assumptions for the efficiency of the steam separators. Case 1 assumes the efficiency of the steam separators becomes zero when the water level in the steam drum increases to the elevation of primary cyclones, or the outlet flow from the steam generator becomes higher than 150 % of normal flow. Case 2 assumes the efficiency becomes zero only when the water level in the steam drum reaches the elevation of primary cyclones. The simulation results show that system responses are sensitive to the assumption for the efficiency of the steam separators and case 1 gives higher discharge energy. Fuel cooling is assured, since primary circuit is cooled down sufficiently by the steam generators for both cases. (author)

  19. Containment fan cooler heat transfer calculation during main steam line break for Maanshan PWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw; Kao, Lain-Su, E-mail: lskao@iner.gov.tw

    2013-10-15

    Highlights: • Evaluate component cooling water (CCW) thermal response during MSLB for Maanshan. • Using GOTHIC to calculate CCW temperature and determine time required to boil CCW. • Both convective and condensation heat transfer from the air side are considered. • Boiling will not occur since T{sub B} is sufficiently longer than CCW pump restart time. -- Abstract: A thermal analysis has been performed for the Containment Fan Cooler Unit (FCU) during Main Steam Line Break (MSLB) accident, concurrent with loss of offsite power, for Maanshan PWR plant. The analysis is performed in order to address the waterhammer and two-phase flow issues discussed in USNRC's Generic Letter 96-06 (GL 96-06). Maanshan plant is a twin-unit Westinghouse 3-loop PWR currently operated at rated core thermal power of 2822 MWt for each unit. The design basis for containment temperature is Main Steam Line Break (MSLB) accident at power of 2830.5 MWt, which results in peak vapor temperature of 387.6 °F. The design is such that when MSLB occurs concurrent with loss of offsite power (MSLB/LOOP), both the coolant pump on the secondary side and the fan on the air side of the FCU loose power and coast down. The pump has little inertia and coasts down in 2–3 s, while the FCU fan coasts down over much longer period. Before the pump is restored through emergency diesel generator, there is potential for boiling the coolant in the cooling coils by the high-temperature air/steam mixture entering the FCU. The time to boiling depends on the operating pressure of the coolant before the pump is restored. The prediction of the time to boiling is important because it determines whether there is potential for waterhammer or two-phase flow to occur before the pump is restored. If boiling occurs then there exists steam region in the pipe, which may cause the so called condensation induced waterhammer or column closure waterhammer. In either case, a great amount of effort has to be spent to

  20. Main Steam Line Break Analysis for the Fully Passive Safety System of SMART

    International Nuclear Information System (INIS)

    Kim, Seong Wook; Chun, Ji Han; Bae, Kyoo Hwan; Kim, Keung Koo

    2013-01-01

    The standard design approval of SMART (System-integrated Modular Advanced ReacTor) developed by KAERI and KEPCO consortium was issued on July 4, 2012. Although SMART has enhanced safety compared to the conventional reactor, there is a demand to meet the 'passive safety performance requirements' after the Fukushima accident. The passive safety performance requirements are the capabilities to maintain the plant at a safe shutdown condition for a minimum of 72 hours without AC power supply or operator action in case of design basis accident (DBA). To satisfy the requirements, KAERI is developing a safety enhanced SMART by adopting a passive safety injection system. The passive safety injection system developed for SMART is a gravity-driven injection system, which consists of four trains, each of which includes a pressure balance line, core makeup tank (CMT), safety injection tank (SIT) and injection line. The CMT plays an important role to inject borated water into the RCS to prevent or dissolve the return to power (re-criticality) condition during the event of increase in heat removal by the secondary system. The main steam line break accident (MSLB) is the most limiting accident for an increase in heat removal by the secondary system. In this study, the safety analysis results of MSLBs at hot full power condition and at hot zero power condition in view of re-criticality are given. The MSLB accident has been analyzed for the SMART adopting fully passive safety system in the aspect of re-criticality. The results show that the core remains subcritical condition throughout the transient due to the borated water injected by the CMT. As further works, many kinds of analyses and sensitivity studies should be performed for the design establishment and improvement of the fully passive system of SMART

  1. Evaluation of acoustic resonance at branch section in main steam line. Part 2. Proposal of method for predicting resonance frequency in steam flow

    International Nuclear Information System (INIS)

    Uchiyama, Yuta; Morita, Ryo

    2012-01-01

    Flow-induced acoustic resonances of piping system containing closed side-branches are sometimes encountered in power plants. Acoustic standing waves with large amplitude pressure fluctuation in closed side-branches are excited by the unstable shear layer which separates the mean flow in the main piping from the stagnant fluid in the branch. In U.S. NPP, the steam dryer had been damaged by high cycle fatigue due to acoustic-induced vibration under a power uprating condition. Our previous research developed the method for evaluating the acoustic resonance at the branch sections in actual power plants by using CFD. In the method, sound speed in wet steam is evaluated by its theory on the assumption of homogeneous flow, although it may be different from practical sound speed in wet steam. So, it is necessary to consider and introduce the most suitable model of practical sound speed in wet steam. In addition, we tried to develop simplified prediction method of the amplitude and frequency of pressure fluctuation in wet steam flow. Our previous experimental research clarified that resonance amplitude of fluctuating pressure at the top of the branch in wet steam. However, the resonance frequency in steam condition could not be estimated by using theoretical equation as the end correction in steam condition and sound speed in wet steam is not clarified as same reason as CFD. Therefore, in this study, we tried to evaluate the end correction in each dry and wet steam and sound speed of wet steam from experimental results. As a result, method for predicting resonance frequency by using theoretical equation in each wet and dry steam condition was proposed. (author)

  2. Imitative modeling automatic system Control of steam pressure in the main steam collector with the influence on the main Servomotor steam turbine

    Science.gov (United States)

    Andriushin, A. V.; Zverkov, V. P.; Kuzishchin, V. F.; Ryzhkov, O. S.; Sabanin, V. R.

    2017-11-01

    The research and setting results of steam pressure in the main steam collector “Do itself” automatic control system (ACS) with high-speed feedback on steam pressure in the turbine regulating stage are presented. The ACS setup is performed on the simulation model of the controlled object developed for this purpose with load-dependent static and dynamic characteristics and a non-linear control algorithm with pulse control of the turbine main servomotor. A method for tuning nonlinear ACS with a numerical algorithm for multiparametric optimization and a procedure for separate dynamic adjustment of control devices in a two-loop ACS are proposed and implemented. It is shown that the nonlinear ACS adjusted with the proposed method with the regulators constant parameters ensures reliable and high-quality operation without the occurrence of oscillations in the transient processes the operating range of the turbine loads.

  3. Detection of leaks in steam lines by distributed fibre-optic temperature sensing (DTS)

    Energy Technology Data Exchange (ETDEWEB)

    Craik, N G [Maritime Nuclear, Fredericton, N.B. (Canada)

    1997-12-31

    This paper describes an instrumentation system concept which should be capable of early detection of a leak-before-break in main steam lines. Distributed fibre-optic Temperature Sensing (DTS) systems have been used in commercial application for a few years now, but in other industries and applications. DTS uses very long fibre optical cable both as a temperature sensor and as a means of bringing the information back from the sensor to the terminal equipment. The entire length of the fibre is sensitive to temperature and each resolvable section of fibre is equivalent to a point sensor. This commercially available DTS system could be adapted to indicate leaks in steam lines. The fibre-optic cable could either be run either just underneath the aluminium sheathing covering the installation over a steam line, or between the two layers of insulation. This would detect an increase in the temperature of the insulation due to a steam leak. 1 ref., 4 figs.

  4. Detection of leaks in steam lines by distributed fibre-optic temperature sensing (DTS)

    International Nuclear Information System (INIS)

    Craik, N.G.

    1996-01-01

    This paper describes an instrumentation system concept which should be capable of early detection of a leak-before-break in main steam lines. Distributed fibre-optic Temperature Sensing (DTS) systems have been used in commercial application for a few years now, but in other industries and applications. DTS uses very long fibre optical cable both as a temperature sensor and as a means of bringing the information back from the sensor to the terminal equipment. The entire length of the fibre is sensitive to temperature and each resolvable section of fibre is equivalent to a point sensor. This commercially available DTS system could be adapted to indicate leaks in steam lines. The fibre-optic cable could either be run either just underneath the aluminium sheathing covering the installation over a steam line, or between the two layers of insulation. This would detect an increase in the temperature of the insulation due to a steam leak. 1 ref., 4 figs

  5. Engineering task plan for steam line ramp calculations

    International Nuclear Information System (INIS)

    DeSantis, G.N.; Freeman, R.D.

    1994-01-01

    The purpose of this document is to provide an approved work plan to perform calculations that verify the load limits of a proposed ramp over a steam line at the back side (East side) of SY Farm in support of work package 2W-94-00812/K. The objective of this supporting document is to provide Operations with a set of checked calculations that verify the ramp over the steam line at SY Farm will support a fully loaded concrete mixer truck without affecting the steam line. The calculations will be performed by an engineers from Facility Systems and independently checked and reviewed by another engineer. The calculations may then be added to the work package. If Operations decides to make any configuration changes to the steam line or surrounding area, Operations shall have these changes documented by an Engineering Change Notice (ECN). This ECN can be done by Facility Systems or any other engineering organization at the direction of Operations

  6. Main Steam Line Break Mass/Energy and Pressure/Temperature Analysis for the Environmental Qualification

    International Nuclear Information System (INIS)

    Park, Yong-Chan; Song, Dong-Soo; Jun, Hwang-Yong

    2006-01-01

    The Main steam line break(MSLB) occurring inside a reactor containment structure may result in significant releases of high energy fluid to the containment, possibly result in high containment pressure and temperature. The MSLB accident, along with the Loss Of Coolant Accident, is a design basis accident for determining the peak containment pressure and temperature. The analysis for a MSLB for inside containment should be performed to justify the structural integrity and equipment qualification in accordance with revision 1 of Reg. Guide 1.89. Rev1(1984), which is also required as part of obtaining the extended operating license for WestingHouse(WH) 3-Loops Nuclear Power Plant(NPP). Now, the WH NPP has been performed power uprating. Therefore, all initial conditions, setpoints and uncertainties were considered with MSLB analysis for environment qualification(EQ). The transient was analyzed to determine the worst set of mass and energy releases that impact the EQ aspects of safety related equipment inside containment. The most limiting single failure in this event was determined by a sensitivity study. The MSLB event was analyzed for a full set of power conditions and break sizes

  7. Effects of Secondary Circuit Modeling on Results of Pressurized Water Reactor Main Steam Line Break Benchmark Calculations with New Coupled Code TRAB-3D/SMABRE

    International Nuclear Information System (INIS)

    Daavittila, Antti; Haemaelaeinen, Anitta; Kyrki-Rajamaeki, Riitta

    2003-01-01

    All of the three exercises of the Organization for Economic Cooperation and Development/Nuclear Regulatory Commission pressurized water reactor main steam line break (PWR MSLB) benchmark were calculated at VTT, the Technical Research Centre of Finland. For the first exercise, the plant simulation with point-kinetic neutronics, the thermal-hydraulics code SMABRE was used. The second exercise was calculated with the three-dimensional reactor dynamics code TRAB-3D, and the third exercise with the combination TRAB-3D/SMABRE. VTT has over ten years' experience of coupling neutronic and thermal-hydraulic codes, but this benchmark was the first time these two codes, both developed at VTT, were coupled together. The coupled code system is fast and efficient; the total computation time of the 100-s transient in the third exercise was 16 min on a modern UNIX workstation. The results of all the exercises are similar to those of the other participants. In order to demonstrate the effect of secondary circuit modeling on the results, three different cases were calculated. In case 1 there is no phase separation in the steam lines and no flow reversal in the aspirator. In case 2 the flow reversal in the aspirator is allowed, but there is no phase separation in the steam lines. Finally, in case 3 the drift-flux model is used for the phase separation in the steam lines, but the aspirator flow reversal is not allowed. With these two modeling variations, it is possible to cover a remarkably broad range of results. The maximum power level reached after the reactor trip varies from 534 to 904 MW, the range of the time of the power maximum being close to 30 s. Compared to the total calculated transient time of 100 s, the effect of the secondary side modeling is extremely important

  8. Large scale multi-zone creep finite element modelling of a main steam line branch intersection

    International Nuclear Information System (INIS)

    Payten, Warwick

    2006-01-01

    A number of papers detail the non-linear creep finite element analysis of branch pieces. Predominately these models have incorporated only a single material zone representing the parent material. Multi-zone models incorporating weld material and heat affected zones have primarily been two-dimensional analyses, in part due to the large number of elements required to adequately represent all of the zones. This paper describes a non-linear creep analysis of a main steam line branch intersection using creep properties to represent the parent metal, weld metal, and heat affected zone (HAZ), the stress redistribution over 100,000 h is examined. The results show that the redistribution leads to a complex stress state, particularly at the heat affected zone. Although, there is damage on the external surface of the branch piece as expected, the results indicate that the damage would be more widespread through extensive sections of the heat affected zone. This would appear to indicate that the time between damage indications on the surface using techniques such as replication and full thickness damage may be more limited then previously expected

  9. Effects of the Pressurized Water Reactor Main Steam Line Break Location on the Blowdown Loading

    International Nuclear Information System (INIS)

    Jo, Jong Chull; Kang, Soon Ho; Chan, Won Joon

    2016-01-01

    The thermal hydraulic analysis has been performed generally using a simple lumped model or one dimensional numerical model. However, those models have limitations in predicting the transient variations of the steam velocity, pressure and hydrodynamic load at a local point and the most conservative condition. Furthermore, it cannot be confirmed if the blowdown loads predicted by either of the models are conservative to evaluate every part of the SG internal structures. In this study, the transient hydraulic response of the SG secondary side to the MSLB case where the pipe break is assumed to occur at the SG outlet nozzle end, another weld point on the MSL, was numerically simulated using a CFD code. The present CFD calculation results was compared with those in ref. to investigate the effect of break location (friction loss) on the blowdown load in the SG secondary side. The result shows that the friction loss along the steam line span between the SG nozzle end and the MSIV would cause reduction in steam velocity disturbance or dynamic pressure. It implies that the consequence of the MSLB at the SG nozzle end would be much severer that those of other MSLB cases where the break locations are far from the SG. Therefore, to assure a conservative safety evaluation of the SG structural integrity, the blowdown loading on the SG internal structures including tubes during a MSLB accident in terms of the transient steam velocity, dynamic pressure and decompression wave fluctuations should be assessed for the MSLB case where the break is assumed to occur at the SG nozzle end.

  10. Effects of the Pressurized Water Reactor Main Steam Line Break Location on the Blowdown Loading

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Jong Chull; Kang, Soon Ho; Chan, Won Joon [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The thermal hydraulic analysis has been performed generally using a simple lumped model or one dimensional numerical model. However, those models have limitations in predicting the transient variations of the steam velocity, pressure and hydrodynamic load at a local point and the most conservative condition. Furthermore, it cannot be confirmed if the blowdown loads predicted by either of the models are conservative to evaluate every part of the SG internal structures. In this study, the transient hydraulic response of the SG secondary side to the MSLB case where the pipe break is assumed to occur at the SG outlet nozzle end, another weld point on the MSL, was numerically simulated using a CFD code. The present CFD calculation results was compared with those in ref. to investigate the effect of break location (friction loss) on the blowdown load in the SG secondary side. The result shows that the friction loss along the steam line span between the SG nozzle end and the MSIV would cause reduction in steam velocity disturbance or dynamic pressure. It implies that the consequence of the MSLB at the SG nozzle end would be much severer that those of other MSLB cases where the break locations are far from the SG. Therefore, to assure a conservative safety evaluation of the SG structural integrity, the blowdown loading on the SG internal structures including tubes during a MSLB accident in terms of the transient steam velocity, dynamic pressure and decompression wave fluctuations should be assessed for the MSLB case where the break is assumed to occur at the SG nozzle end.

  11. Signal analysis of steam line acoustics

    International Nuclear Information System (INIS)

    Martin, C. Samuel

    2003-01-01

    upstream and downstream between two locations. The method shows that the direction of the larger magnitude wave -- whether propagating upstream or downstream -- is directly related to the slope of the unwrapped phase angle versus frequency correlation. Indeed, the slope of this line can be related to the acoustic velocity of the wave. The method is then applied to dynamic pressure recordings obtained in a nuclear steam system. Plots of cross-spectra phase versus frequency taken in straight runs of steam piping yield correlations that are nearly linear, and, moreover, the slope of the line is closely related to the acoustic velocity at the corresponding steam pressure. (author)

  12. Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

    International Nuclear Information System (INIS)

    Mendler, O.J.; Takeuchi, K.; Young, M.Y.

    1986-10-01

    The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results

  13. Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

    Energy Technology Data Exchange (ETDEWEB)

    Mendler, O J; Takeuchi, K; Young, M Y

    1986-10-01

    The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results.

  14. Condensation of the steam in the horizontal steam line during cold water flooding

    International Nuclear Information System (INIS)

    Strubelj, L.; Tiselj, I.

    2006-01-01

    Direct contact condensation and condensation induced water-hammer in a horizontal pipe was experimentally investigated at PMK-2 test facility of the Hungarian Atomic Energy Research Institute KFKI. The experiment is preformed in the horizontal section of the steam line of the PMK-2 integral test facility. As liquid water floods the horizontal part of the pipeline, the counter current horizontally stratified flow is being observed. During the flooding of the steam line, the vapour-liquid interface area increases and therefore the vapour condensation rate and the vapour velocity also increase. Similar phenomena can occur in the cold/hot leg of the primary loop of PWR nuclear power plant during loss of coolant accident, when emergency core cooling system is activated. Water level at one cross-section and four local void fraction and temperature at the top of steam line was measured and compared with simulation. Condensed steam increases the water temperature that is why the local temperature measurements are the most important information, from which condensation rate can be estimated, since mass of condensed steam was not measured. Free surface simulation of the experiment with thermal phase change model is presented. Surface renewal concept with small eddies is used for calculation of heat transfer coefficient. With surface renewal theory we did not get results similar to experiment, that is why heat transfer coefficient was increased by factor 20. In simulation with heat transfer coefficient calculated with surface renewal concept bubble entrapment is due to reflection of the wave from the end of the pipe. When heat transfer coefficient is increased, condensation rate and steam velocity are also increased, bubble entrapment is due to Kelvin-Helmholtz instability of the free surface, and the results become similar to the measurements. (author)

  15. Process of characterization of vibration in Cofrentes NPP SRVs - scale model of main steam line; Proceso de caracterizacion de vibraciones en SRVs de C.N. Cofrentes-Modelo a escala linea de vapor principal

    Energy Technology Data Exchange (ETDEWEB)

    Galbally, D.; Hernando, J.; Garcia, G.; Barral, M.

    2014-07-01

    The Cofrentes Nuclear power plant has experienced different events anomalous related to its relief and system (SRVs) main steam safety valves. After various studies is determined that the existence of dynamics of pressure oscillations in the interior of the main steam lines is the cause of many of the events that occurred in the SRVs. To monitor these vibrations, Iberdrola performed the installation of a measuring system of vibration in SRVs and actuators during the recharge 18 (September - October 2011) with a total of 40 accelerometers distributed in 6 of the 16 existing valves. (Author)

  16. Sensitivity Studies for Main Steam Line Break Exercises 2 and 3 with RELAP5/PANBOX

    International Nuclear Information System (INIS)

    Boeer, Rainer; Knoll, Alfred

    2003-01-01

    This paper presents and discusses results obtained with the nuclear plant safety analysis code system RELAP5/PANBOX (R/P/C) for the return-to-power scenario of exercises 2 and 3 of the Organization for Economic Cooperation and Development/Nuclear Energy Agency Main Steam Line Break (MSLB) Benchmark. Both the external and internal coupling options of R/P/C have been considered for exercise 3; i.e., the COBRA module of PANBOX was used to calculate the core thermal hydraulics in the external coupling option, whereas the core thermal hydraulics of RELAP5 was used in the internal coupling option. For the representation of thermal-hydraulic channels, a fine channel geometry based on the 177 fuel assemblies was selected for the external coupling option, and a coarse channel geometry based on 19 coarse channels has been investigated for the internal coupling option. The comparison of the results shows very good agreement of important core parameters between the considered coupling variants. Both exercises 2 and 3 have been investigated with respect to local safety parameters like fuel centerline temperatures and minimum departure from nucleate boiling ratios using the on-line hot subchannel analysis capability of R/P/C in the external coupling option. The results show that both quantities are far from the safety-related limits.The benchmark demonstrates, that R/P/C - as part of the integrated CASCADE-3D core analysis system of Framatome ANP GmbH - has proven to be a powerful tool for detailed analyses of an MSLB accident

  17. The modification of main steam safety valves in Qinshan phase Ⅱ expansion project

    International Nuclear Information System (INIS)

    Chen Haiqiao

    2012-01-01

    The main steam safety valves of NPP steam system are second- class nuclear safety component. It used to limit the pressure of SG secondary side and main steam system via emitting steam into the environment. At present, the main steam safety valves have mechanical valves and assisted power valves. According to the experience of power plants at home and abroad, including Qinshan Phase Ⅱ unit 1/2 experience feedback, Qinshan Phase Ⅱ expansion project made modification on valve type, setting value and valve body. This paper introduce the characteristics of different safety valve types, the modification of main steam safety valves and the modification analysis on safety issues.security and impact on the other systems in Qinshan Phase Ⅱ expansion project. (author)

  18. Three-dimensional neutron kinetics-thermal-hydraulics VVER 1000 main steam line break analysis by RELAP5-3D code

    International Nuclear Information System (INIS)

    Frisani, A.; Parisi, C.; D'Auria, F.

    2007-01-01

    After the development and the assessment of Three-Dimensional (3D) Neutron Kinetics (NK) - 1D Thermal-Hydraulics (TH) coupled codes analyses methods, deterministic nuclear safety technology is nowadays producing noticeable efforts for the validation of 3D NK - 3D TH coupled codes analyses methods too. Thus, the purpose of this work was to address the capability of the RELAP5-3D 3D NK-3D TH code to reproduce VVER 1000 Nuclear Power Plant (NPP) core dynamic in simulating the mixing effects that could happen in the vessel downcomer and lower plenum during some scenarios. The work was developed in three steps. The first step dealt with the 3D TH modeling of the Kozloduy-6 VVER 1000 reactor pressure vessel. Then this model was validated following a Steam Generator Isolation transient. The second step has been the development of a 3D NK nodalization for the reactor core region. Then the 3D NK model was directly coupled with the previously developed 3D TH model. The third step was the calculation of a Main Steam Line Break (MSLB) transient. The 3D NK global nuclear parameters were then compared with the 0-D results showing a good agreement; nevertheless only the 3D NK- 3D TH model allowed the calculation of each single assembly power trend for this strong NK-TH asymmetric transient. (author)

  19. Validation cases of CATHARE 2 for VVER-1000 main steam line break analysis

    International Nuclear Information System (INIS)

    Kolev, Nikolay P.; Petrov, Nikolay; Donov, Jordan; Sabotinov, Luben; Nikonov, Sergey

    2008-01-01

    Recent coupled code benchmarks identified coolant mixing in the reactor vessel as an unresolved issue in the analysis of complex plan transients with reactivity insertion. Thus, Phase 2 of the OECD VVER-1000 Coolant Transient Benchmark (V1000CT-2) was defined. The benchmark includes calculation of vessel mixing tests and main steam line break (MSLB) analysis. The reference plant is Kozloduy-6 in Bulgaria. The general objective is the assessment of system codes for VVER safety analysis and specifically for their use in the analysis of reactivity transients. A specific objective is the testing of different scale mixing models (mixing matrix, multi-1D, coarse-3D and CFD), and analysis of MSLB transients with improved vessel thermal hydraulic models. The benchmark is sponsored by CEA-France and OECD and is jointly prepared by CEA and INRNE, in collaboration with the Kozloduy NPP, IRSN and PSU. This paper summarizes CATHARE2 code assessment calculations using multi-1D vessel thermal hydraulics with cross flow. Test cases are the OECD V1000CT-1 pump start-up benchmark and the V1000CT-2 benchmarks. Emphasis is put on vessel mixing aspects. Separate effects in the lower plenum as well as component and integral system tests are considered. The comparison shows that a six-sector vessel mixing model informed by plant data or validated CFD calculations in the initial state was able to correctly reproduce the channel average temperatures at the core inlet as well as the vessel outlet temperatures. Testing at system level including code-to-experiment and CATHARE-ATHLET comparison shows that the considered CATHARE VVER-1000 system model is capable of MSLB simulation. (author)

  20. MHTGR steam generator on-line heat balance, instrumentation and function

    International Nuclear Information System (INIS)

    Klapka, R.E.; Howard, W.W.; Etzel, K.T.; Basol, M.; Karim, N.U.

    1991-09-01

    Instrumentation is used to measure the Modular High Temperature Gas-Cooled Reactor (MHTGR) steam generator dissimilar metal weld temperature during start-up testing. Additional instrumentation is used to determine an on-line heat balance which is maintained during the 40 year module life. In the process of calibrating the on-line heat balance, the helium flow is adjusted to yield the optimum boiling level in the steam generator relative to the dissimilar metal weld. After calibration is complete the weld temperature measurement is non longer required. The reduced boiling level range results in less restrictive steam generator design constraints

  1. Evaluation of material integrity on electricity generator water steam cycles component (Main Steam Pipe)

    International Nuclear Information System (INIS)

    Sudardjo; Histori; Triyadi, Ari

    1998-01-01

    The evaluation of material integrity on electricity generator component has been done. That component was main steam pipe of Unit II Suralaya Coal Fired Power Plant. evaluation was done by replication technique. The damage was found are two porosity's, from two point samples of six points sample population. Based on cavity evaluation in steels, which proposed by Neubauer and Wedel that porosity's still at class A damage. For class A damage, its means no remedial action would be required until next major scheduled maintenance outage. That porosity's was grouped on isolated cavities and not need ti repair that main steam pipe component less than three year after replication test

  2. Technical evaluation: 300 Area steam line valve accident

    International Nuclear Information System (INIS)

    1993-08-01

    On June 7, 1993, a journeyman power operator (JPO) was severely burned and later died as a result of the failure of a 6-in. valve that occurred when he attempted to open main steam supply (MSS) valve MSS-25 in the U-3 valve pit. The pit is located northwest of Building 331 in the 300 Area of the Hanford Site. Figure 1-1 shows a layout of the 300 Area steam piping system including the U-3 steam valve pit. Figure 1-2 shows a cutaway view of the approximately 10- by 13- by 16-ft-high valve pit with its various steam valves and connecting piping. Valve MSS-25, an 8-in. valve, is located at the bottom of the pit. The failed 6-in. valve was located at the top of the pit where it branched from the upper portion of the 8-in. line at the 8- by 8- by 6-in. tee and was then ''blanked off'' with a blind flange. The purpose of this technical evaluation was to determine the cause of the accident that led to the failure of the 6-in. valve. The probable cause for the 6-in. valve failure was determined by visual, nondestructive, and destructive examination of the failed valve and by metallurgical analysis of the fractured region of the valve. The cause of the accident was ultimately identified by correlating the observed failure mode to the most probable physical phenomenon. Thermal-hydraulic analyses, component stress analyses, and tests were performed to verify that the probable physical phenomenon could be reasonably expected to produce the failure in the valve that was observed

  3. Numerical simulation of a 374 tons/h water-tube steam boiler following a feedwater line break

    International Nuclear Information System (INIS)

    Deghal Cheridi, Amina Lyria; Chaker, Abla; Loubar, Ahcène

    2016-01-01

    Highlights: • We simulate the behavior of a steam boiler during feed-water line break accident. • To perform accident analysis of the steam boiler, Relap5/Mod3.2 system code is used. • A Relap5 model of the boiler is developed and qualified at the steady state level. • A good agreement between Relap5 results and available experimental data. • The Relap5 model predicts well the main transient features of the boiler. - Abstract: To ensure the operational safety of an industrial water-tube steam boiler it is very important to assess various accident scenarios in real plant working conditions. One of the most challenging scenarios is the loss of feedwater to the steam boiler. In this paper, a simulation of the behavior of an industrial water-tube radiant steam boiler during feedwater line break accident is discussed. The simulation is carried out using the RELAP5 system code. The steam boiler is installed in an Algerian natural gas liquefaction complex. The simulation shows the capabilities of RELAP5 system code in predicting the behavior of the steam boiler at both steady state and transient working conditions. From another side, the behavior of the steam boiler following the accident shows how the control system can successfully mitigate the effects and consequences of such accident and how the evaporator tubes can undergo a severe damage due to an uncontrolled increase of the wall temperature in case of failure of this system.

  4. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    International Nuclear Information System (INIS)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B.

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked ampersand influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs

  5. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B. [and others

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked & influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs.

  6. Simulation of main steam and feedwater system of full scope simulator for Qinshan 300 MW Nuclear Power Unit

    International Nuclear Information System (INIS)

    Zhao Xiaoyu

    1996-01-01

    The simulation of main steam and feedwater system is the most important and maximal part in secondary circuit model, including all of main steam and feedwater's thermal-hydraulic properties, except heat-exchange of secondary side of steam generator. It simulates main steam header, steam power in each stage of turbine, moisture separator-reheater, deaerator, condenser, high pressure and low pressure heater, auxiliary feedwater and main steam bypass in full scope

  7. Comparison of the updated solutions of the 6th dynamic AER Benchmark - main steam line break in a NPP with WWER-440

    International Nuclear Information System (INIS)

    Kliem, S.

    2003-01-01

    The 6 th dynamic AER Benchmark is used for the systematic validation of coupled 3D neutron kinetic/thermal hydraulic system codes. It was defined at The 10 th AER-Symposium. In this benchmark, a hypothetical double ended break of one main steam line at full power in a WWER-440 plant is investigated. The main thermal hydraulic features are the consideration of incomplete coolant mixing in the lower and upper plenum of the reactor pressure vessel and an asymmetric operation of the feed water system. For the tuning of the different nuclear cross section data used by the participants, an isothermal re-criticality temperature was defined. The paper gives an overview on the behaviour of the main thermal hydraulic and neutron kinetic parameters in the provided solutions. The differences in the updated solution in comparison to the previous ones are described. Improvements in the modelling of the transient led to a better agreement of a part of the results while for another part the deviations rose up. The sensitivity of the core power behaviour on the secondary side modelling is discussed in detail (Authors)

  8. Operational reliability of high pressure steam lines of pearlitic steels after 150-200 thousand h service

    International Nuclear Information System (INIS)

    Veksler, E.Ya.; Chajkovskij, V.M.; Osasyuk, V.V.

    1980-01-01

    Usage of both calculational and physical methods is recommended to estimate a service operating life of long-term working steam line materials. Application of these methods is demonstrated when studying steam line bends made of 12MKh and 12Kh1MF pearlitic steels. Good coincidence of results for the determination of residual durability of steam lines is obtained using these two methods [ru

  9. Analysis of the main steam line break accident with loss of offsite power using the fully coupled RELAP5/PANTHER/COBRA code package

    International Nuclear Information System (INIS)

    Ruben Van Parys; Sandrine Bosso; Christophe Schneidesch; Jinzhao Zhang

    2005-01-01

    Full text of publication follows: A coupled thermal hydraulics-neutronics code package (RELAP5/PANTHER/COBRA) has been qualified for accident analysis at Tractebel Engineering. In the TE coupled code package, the best estimate thermal-hydraulic system code, RELAP5/MOD2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via a dynamic data exchange control and processing tool, TALINK. An interface between PANTHER code and the sub-channel thermal-hydraulic analysis code COBRA-IIIC is developed in order to perform online calculation of Departure from Nucleate Boiling Ratio (DNBR). The TE coupled code package has been applied to develop a MSLB accident analysis methodology using the TE deterministic bounding approach. The methodology has been applied for MSLB accident analysis in support of licensing of the power up-rate and steam generator replacement of the Doel 2 plant. The results of coupled thermal-hydraulic and neutronic analysis of SLB show that there exists an important margin in the traditional FSAR MSLB accident analysis. As a specific licensing requirement, the main steam line break accident with loss of offsite power has to be analyzed. In the standard methodology with the coupled RELAP5/PANTHER code, and some corrective methods has to be taken in order to overcome the limitations due to the close-channel T/H model in PANTHER at low flow conditions. The results show that the steam line break accident with loss of offsite power is far less limiting. In order to verify the effect of the cross-flow at low flow conditions, the fully dynamic coupling of RELAP5/PANTHER/COBRA code package is used for reanalysis of this case, in which the PANTHER close-channel T/H model is replaced by the COBRA sub-channel T/H model with crossflow option. It has been demonstrated that, although the consideration of cross-flow in this challenging situation may lead to higher core return to power and slightly lower DNBR than in the standard methodology

  10. Analysis of the main steam line break accident with loss of offsite power using the fully coupled RELAP5/PANTHER/COBRA code package

    Energy Technology Data Exchange (ETDEWEB)

    Ruben Van Parys; Sandrine Bosso; Christophe Schneidesch; Jinzhao Zhang [Nuclear Department, Suez-Tractebel Engineering, avenue Ariane 5, B-1200 Brussels (Belgium)

    2005-07-01

    Full text of publication follows: A coupled thermal hydraulics-neutronics code package (RELAP5/PANTHER/COBRA) has been qualified for accident analysis at Tractebel Engineering. In the TE coupled code package, the best estimate thermal-hydraulic system code, RELAP5/MOD2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via a dynamic data exchange control and processing tool, TALINK. An interface between PANTHER code and the sub-channel thermal-hydraulic analysis code COBRA-IIIC is developed in order to perform online calculation of Departure from Nucleate Boiling Ratio (DNBR). The TE coupled code package has been applied to develop a MSLB accident analysis methodology using the TE deterministic bounding approach. The methodology has been applied for MSLB accident analysis in support of licensing of the power up-rate and steam generator replacement of the Doel 2 plant. The results of coupled thermal-hydraulic and neutronic analysis of SLB show that there exists an important margin in the traditional FSAR MSLB accident analysis. As a specific licensing requirement, the main steam line break accident with loss of offsite power has to be analyzed. In the standard methodology with the coupled RELAP5/PANTHER code, and some corrective methods has to be taken in order to overcome the limitations due to the close-channel T/H model in PANTHER at low flow conditions. The results show that the steam line break accident with loss of offsite power is far less limiting. In order to verify the effect of the cross-flow at low flow conditions, the fully dynamic coupling of RELAP5/PANTHER/COBRA code package is used for reanalysis of this case, in which the PANTHER close-channel T/H model is replaced by the COBRA sub-channel T/H model with crossflow option. It has been demonstrated that, although the consideration of cross-flow in this challenging situation may lead to higher core return to power and slightly lower DNBR than in the standard methodology

  11. Isolating valve, especially in main-steam pipes of power plants

    International Nuclear Information System (INIS)

    Karpenko, A.N.

    1977-01-01

    The valve for PWRs and BWRs, with diameters up to 1.25 m, for temperatures from -180 0 C to about 600 0 C and pressures up to over 50 bar, is designed for high reliability and long useful life. Two circular valve discs are moved as isolating elements in their plane across the steam direction and brought before the valve seat within a valve chamber. Shortly before reaching this final position, each disc is rotated by a small amount about its axis. Only after reaching the final position a double-wedge, further pushed forward between both discs, produces the necessary contact pressure. By revolving and frictionless closing caking together at high stresses and temperature variation is prevented and permanent tightness assured. The valve body is moved in a cylinder, cast on the valve housing, by means of a stepped piston. Its larger diameter is guided in a second cylinder flanged on above. In the cover of the second cylinder a pilot valve is mounted being controlled over 2 parallel solenoid valves by means of compressed air. In normal operation process steam from the valve chamber serves to move the stepped piston with the valve chamber. On closing of a bore, connecting both cylinder spaces, by the pilot valve the main valve is opened. If the pilot valve is opened the steam through the connecting bore is acting on both piston stages and closing the main valve. On loss of steam (pipe break) or for testing purposes one or the other cylinder space over solenoid valves is acted upon by auxiliary energy or evacuated, the main valve thus being controlled. (HP) [de

  12. HTGR power plant hot reheat steam pressure control system

    International Nuclear Information System (INIS)

    Braytenbah, A.S.; Jaegtnes, K.O.

    1975-01-01

    A control system for a high temperature gas cooled reactor (HTGR) power plant is disclosed wherein such plant includes a plurality of steam generators. Dual turbine-generators are connected to the common steam headers, a high pressure element of each turbine receiving steam from the main steam header, and an intermediate-low pressure element of each turbine receiving steam from the hot reheat header. Associated with each high pressure element is a bypass line connected between the main steam header and a cold reheat header, which is commonly connected to the high pressure element exhausts. A control system governs the flow of steam through the first and second bypass lines to provide for a desired minimum steam flow through the steam generator reheater sections at times when the total steam flow through the turbines is less than such minimum, and to regulate the hot reheat header steam pressure to improve control of the auxiliary steam turbines and thereby improve control of the reactor coolant gas flow, particularly following a turbine trip. (U.S.)

  13. Why extraction lines and heaters in the turbine-condenser steam space should be lagged

    Energy Technology Data Exchange (ETDEWEB)

    Burns, J.M.; Haynes, C.J.

    1998-07-01

    Deregulated utilities face conditions today that necessitate their nuclear and fossil steam plants have the best possible heat rates. The low pressure turbine exhaust and condenser areas are known to be particularly sensitive to betterment. One relatively modest but cost effective heat rate improvement and one whose function and design is often misunderstood is the insulation of the extraction lines and heaters that are located within the turbine-condenser steam space. This paper discusses the dynamic environment of that turbine exhaust region and quantifies the application and benefit of stainless steel lagging to the extraction lines and heater shells within. The paper first focuses on the high energy, non-uniform steam flows of the turbine exhaust and how that impacts the heat losses, mechanical design and support of any components located inside that space. It then examines and quantifies the varieties of heat transfer from the heaters and extraction lines to the passing lower temperature, moist, high velocity turbine exhaust steam as it travels to the condenser. A new relationship is developed that defines the predominantly evaporative heat transfer mechanism on the exterior surfaces in contact with the exhaust steam. For a typical 630 MW fossil plant with three heater of different temperature levels in the steam space as exemplified by the US Generation fossil fired Brayton Point 3, the paper determined the additional condenser heat load and extra extraction steam. The paper lastly concluded that in this case, lagging the larger diameter lines of the lowest pressure heater and the heater itself is likely not cost-effective.

  14. Steam turbines for nuclear power plants

    International Nuclear Information System (INIS)

    Kosyak, Yu.F.

    1978-01-01

    Considered are the peculiarities of the design and operation of steam turbines, condensers and supplementary equipment of steam turbines for nuclear power plants; described are the processes of steam flow in humid-steam turbines, calculation and selection principles of main parameters of heat lines. Designs of the turbines installed at the Charkov turbine plant are described in detail as well as of those developed by leading foreign turbobuilding firms

  15. Leak before break evaluation for main steam piping system made of SA106 Gr.C

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Kyoung Mo; Jee, Kye Kwang; Pyo, Chang Ryul; Ra, In Sik [Korea Power Engineering Company, Seoul (Korea, Republic of)

    1997-04-01

    The basis of the leak before break (LBB) concept is to demonstrate that piping will leak significantly before a double ended guillotine break (DEGB) occurs. This is demonstrated by quantifying and evaluating the leak process and prescribing safe shutdown of the plant on the basis of the monitored leak rate. The application of LBB for power plant design has reduced plant cost while improving plant integrity. Several evaluations employing LBB analysis on system piping based on DEGB design have been completed. However, the application of LBB on main steam (MS) piping, which is LBB applicable piping, has not been performed due to several uncertainties associated with occurrence of steam hammer and dynamic strain aging (DSA). The objective of this paper is to demonstrate the applicability of the LBB design concept to main steam lines manufactured with SA106 Gr.C carbon steel. Based on the material properties, including fracture toughness and tensile properties obtained from the comprehensive material tests for base and weld metals, a parametric study was performed as described in this paper. The PICEP code was used to determine leak size crack (LSC) and the FLET code was used to perform the stability assessment of MS piping. The effects of material properties obtained from tests were evaluated to determine the LBB applicability for the MS piping. It can be shown from this parametric study that the MS piping has a high possibility of design using LBB analysis.

  16. Simulation and analysis of main steam control system based on heat transfer calculation

    Science.gov (United States)

    Huang, Zhenqun; Li, Ruyan; Feng, Zhongbao; Wang, Songhan; Li, Wenbo; Cheng, Jiwei; Jin, Yingai

    2018-05-01

    In this paper, after thermal power plant 300MW boiler was studied, mat lab was used to write calculation program about heat transfer process between the main steam and boiler flue gas and amount of water was calculated to ensure the main steam temperature keeping in target temperature. Then heat transfer calculation program was introduced into Simulink simulation platform based on control system multiple models switching and heat transfer calculation. The results show that multiple models switching control system based on heat transfer calculation not only overcome the large inertia of main stream temperature, a large hysteresis characteristic of main stream temperature, but also adapted to the boiler load changing.

  17. Evaluation of the Main Steam Line Break Accident for the APR+ Standard Design using MARS-KS

    Energy Technology Data Exchange (ETDEWEB)

    Park, M. H.; Kim, Y. S.; Hwang, Min Jeong; Sim, S. K. [Environment Energy Technology, Daejeon (Korea, Republic of); Bang, Young Seok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-05-15

    As a part of licensing evaluation of the APR+ (Advanced Power Reactor +) standard design, Korea Institute of Nuclear Safety(KINS) performed safety evaluation of the APR+ Standard Safety Analysis Report(SSAR). The results of the safety evaluation of the APR+ Main Steam Line Break(MSLB) accident is presented for the most limiting post-trip return-to-power case with the single failure assumption of the Loss Of Offsite Power(LOOP). MARS-KS regulatory safety analysis code has been used to evaluate safety as well as the system behavior during MSLB accident. The MARS-KS analysis results are compared with the results of the MSLB safety analysis presented in the SSAR of the APR+. Independent safety evaluation has been performed using MARS-KS regulatory safety analysis code for the APR+ MSLB accident inside containment for the limiting case of the full power post-trip return-to-power. The results of MARS-KS analysis were compared with the results of the MSLB safety analysis presented in the APR+ SSAR. Due to higher cooldown of the MARS-KS analysis, the MARS-KS analysis results in a higher positive reactivity insertion into the core by the negative moderator and fuel temperature reactivity coefficients than the APR+ SSAR analysis. Both results show no return-to-power during the limiting case of the MSLB inside containment. However, APR+ SSAR moderator temperature reactivity insertion should be evaluated against the MARS-KS moderator density reactivity insertion for is conservatism. This study also clearly shows asymmetric thermal hydraulic behavior during the MSLB accident at intact and affected sides of the downcomer and the core. These asymmetric phenomena should be further investigated for the effects on the system design.

  18. RELAP/MOD1.5 analysis of steam line break transients for a 3-loop and a 4-loop Westinghouse nuclear steam supply system

    International Nuclear Information System (INIS)

    Peeler, G.B.; McDonald, T.A.; Kennedy, M.F.

    1984-01-01

    RELAP/MOD1.5 (Cycle 31 and 34) calculations were made to assess the assumptions used by Westinghouse (W) to analyze mainsteam line break transients. Models of a W 3-loop and 4-loop nuclear steam supply system were used. Sensitivity studies were performed to determine the effect of the availability of offsite power, break size and initial core power. Comparison with W results indicated that if the assumptions used by W are replicated within the RELAP5 framework, then the W methodology for prediction of the Nuclear Steam Supply System (NSSS) response is conservative for steam line break transients

  19. Main trends of upgrading the 1000 MW steam turbine

    International Nuclear Information System (INIS)

    Drahy, J.

    1990-01-01

    Parameters are compared for the 1000 MW steam turbine manufactured by the Skoda Works, Czechoslovakia, and turbines in the same power range by other manufacturers, viz. ABB, Siemens/KWU, GEC and LMZ. The Skoda turbine compares well with the other turbines with respect to all design parameters, and moreover, enables the most extensive heat extraction for district heating purposes. The main trends in upgrading this turbine are outlined; in particular, they include an additional increase in the heat extraction, which is made possible by a new design of the low-pressure section or by using a ''satellite'' turbine. The studies performed also indicate that the output of the full-speed saturated steam turbine can be increased to 1300 MW. An experimental turbine representing one flow of the high-pressure part of the 1000 MW turbine is being built on the 1:1 scale. It will serve to verify the methods of calculation of the wet steam flow and to experimentally test the high-pressure part over a wide span of the parameters. (Z.M.). 1 tab., 3 figs., 7 refs

  20. Evaluation on operation of liquid relief valves for steam line break accidents by RELAP5/CANDU+ code

    International Nuclear Information System (INIS)

    Yang, C. Y.; Bang, Y. S.; Kim, H. J.

    2001-01-01

    A development of RELAP5/CANDU+ code for regulatory audits of accident analysis of CANDU nuclear power plants is on progress. This paper is undertaken in a procedure of a verification and validation for RELAP5/CANDU+ code by analyzing main steam line break accidents of WS 2/3/4. Following the ECC injection in sequence of the steam line breaks, the mismatch in heat transfer between the primary and the secondary systems makes pressure of the primary system instantly peaked to the open setpoint of liquid relief valves. The event sequence follows the result of WS 2/3/4 FSAR, but there is a difference in pressure transient after ECC injection. Sensitivity analysis for main factors dependent on the peak pressure such as control logics of liquid relief valves. ECC flow path and feedwater flow is performed. Because the pressure increase is continued for a long time and its peaking is high, open and close of the liquid relief valves are repeated several times, which is obviously different from those of WS 2/3/4 FSAR. As a result, it is evaluated that conservative modeling for the above variables is required in the analysis

  1. Coupled RELAP5/PANTHER/COBRA steam line break accident analysis in support of licensing DOEL 2 power uprate and steam generator replacement

    International Nuclear Information System (INIS)

    Zhang, J.; Bosso, S.; Henno, X.; Ouliddren, K.; Schneidesch, C.R.; Hove, W. van

    2004-01-01

    The nuclear reactor accident analyses using best estimate codes provide a better understanding and more accurate modeling of the key physical phenomena, which allows a more realistic evaluation of the conservatism and margins in the final safety analysis report (FSAR) accident analysis. The use of the best estimate codes and methods is necessary to meet the increasing technical, licensing and regulatory requirements for major plant changes (e.g. steam generator replacement), power uprate, core design optimization (cycle extension), as well as Periodic Safety Review. Since 1992, Tractebel Engineering (TE) has developed and applied a deterministic bounding approach to FASR accident analysis using the best estimate system thermal hydraulic code RELAP5/MOD2.5 and the subchannel thermal hydraulic code COBRA-3C. This approach has been accepted by the Belgian Safety Authorities, and turned out to be cost effective for most of the non-LOCA transient analyses. Since this approach adapts a decoupled modeling of the core responses, the analysis results often involved too large un-quantified conservatisms, due to either simplistic approximations for asymmetric accidents with strong 3D core neutronics - plant thermal hydraulics interactions, or additional penalties introduced from 'incoherent' initial/boundary conditions for separate plant and core analyses. Therefore, an external dynamic coupling between the RELAP5/MOD2.5 code and the 3-D neutronic code PANTHER was implemented since 1997 via the transient analysis code linkage program TALINK. Furthermore, a static linkage between the PANTHER code and the COBRA-3C code was developed for on-line calculation of (Departure from Nucleate Boiling Ratio (DNBR). TE intends to use the coupled code package for licensing non-symmetric FSAR accident analysis. The TE coupled code package has been applied to develop a main steam line break (MSLB) accident analysis methodology [using the TE deterministic bounding approach. The methodology

  2. Rupture of steam lines between blocks D and G

    International Nuclear Information System (INIS)

    1999-01-01

    Analysis of steam lines rupture between blocks D and G of Ignalina NPP was performed. Model for evaluation of thermo hydrodynamic parameters was developed. Structural analysis of the shaft building was done as well. State of the art codes such as RELAP5, ALGOR, NEPTUNE were used in these calculations

  3. OECD/NEZ Main Steam Line Break Benchmark Problem Exercise I Simulation Using the SPACE Code with the Point Kinetics Model

    International Nuclear Information System (INIS)

    Kim, Yohan; Kim, Seyun; Ha, Sangjun

    2014-01-01

    The Safety and Performance Analysis Code for Nuclear Power Plants (SPACE) has been developed in recent years by the Korea Nuclear Hydro and Nuclear Power Co. (KHNP) through collaborative works with other Korean nuclear industries. The SPACE is a best-estimated two-phase three-field thermal-hydraulic analysis code to analyze the safety and performance of pressurized water reactors (PWRs). The SPACE code has sufficient features to replace outdated vendor supplied codes and to be used for the safety analysis of operating PWRs and the design of advanced reactors. As a result of the second phase of the development, the 2.14 version of the code was released through the successive various V and V works. The topical reports on the code and related safety analysis methodologies have been prepared for license works. In this study, the OECD/NEA Main Steam Line Break (MSLB) Benchmark Problem Exercise I was simulated as a V and V work. The results were compared with those of the participants in the benchmark project. The OECD/NEA MSLB Benchmark Problem Exercise I was simulated using the SPACE code. The results were compared with those of the participants in the benchmark project. Through the simulation, it was concluded that the SPACE code can effectively simulate PWR MSLB accidents

  4. Substantiation of vibration strength of nuclear reactor and steam generator internals. Main problems

    International Nuclear Information System (INIS)

    Fyodorov, V.G.; Sinyavasky, V.F.

    1977-01-01

    The report details the scope and priority of studies necessary for substantiation of vibration strength of steam generator tube bundles and reactor fuel assemblies, and design modifications helping to reduce flow-induced vibration of the internals specified. Steam generator tube bundles are studied on the basis of a standard establishing vibration requirements at various stages of design, manufacture and operation of a steam generator at a nuclear power station. The main vibration characteristics of tubes obtained through model and full-scale tests are compared with calculation results. Results are provided concerning test-stand vibration tests of fuel elements and fuel assemblies. (author)

  5. Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

    Energy Technology Data Exchange (ETDEWEB)

    Brown, C.S., E-mail: csbrown3@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Raleigh, NC 27695-7909 (United States); Zhang, H., E-mail: Hongbin.Zhang@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3870 (United States); Kucukboyaci, V., E-mail: kucukbvn@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Sung, Y., E-mail: sungy@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2016-12-01

    Highlights: • Best estimate plus uncertainty (BEPU) analyses of PWR core responses under main steam line break (MSLB) accident. • CASL’s coupled neutron transport/subchannel code VERA-CS. • Wilks’ nonparametric statistical method. • MDNBR 95/95 tolerance limit. - Abstract: VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was applied to simulate core behavior of a typical Westinghouse-designed 4-loop pressurized water reactor (PWR) with 17 × 17 fuel assemblies in response to two main steam line break (MSLB) accident scenarios initiated at hot zero power (HZP) at the end of the first fuel cycle with the most reactive rod cluster control assembly stuck out of the core. The reactor core boundary conditions at the most DNB limiting time step were determined by a system analysis code. The core inlet flow and temperature distributions were obtained from computational fluid dynamics (CFD) simulations. The two MSLB scenarios consisted of the high and low flow situations, where reactor coolant pumps either continue to operate with offsite power or do not continue to operate since offsite power is unavailable. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this demonstration of BEPU application, 59 full core simulations were performed for each accident scenario to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. A parametric goodness-of-fit approach was also applied to the results to obtain the MDNBR value at the 95/95 tolerance limit. Initial sensitivity analysis was performed with the 59 cases per accident scenario by use of Pearson correlation coefficients. The results show that this typical PWR core

  6. Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

    International Nuclear Information System (INIS)

    Brown, C.S.; Zhang, H.; Kucukboyaci, V.; Sung, Y.

    2016-01-01

    Highlights: • Best estimate plus uncertainty (BEPU) analyses of PWR core responses under main steam line break (MSLB) accident. • CASL’s coupled neutron transport/subchannel code VERA-CS. • Wilks’ nonparametric statistical method. • MDNBR 95/95 tolerance limit. - Abstract: VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was applied to simulate core behavior of a typical Westinghouse-designed 4-loop pressurized water reactor (PWR) with 17 × 17 fuel assemblies in response to two main steam line break (MSLB) accident scenarios initiated at hot zero power (HZP) at the end of the first fuel cycle with the most reactive rod cluster control assembly stuck out of the core. The reactor core boundary conditions at the most DNB limiting time step were determined by a system analysis code. The core inlet flow and temperature distributions were obtained from computational fluid dynamics (CFD) simulations. The two MSLB scenarios consisted of the high and low flow situations, where reactor coolant pumps either continue to operate with offsite power or do not continue to operate since offsite power is unavailable. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this demonstration of BEPU application, 59 full core simulations were performed for each accident scenario to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. A parametric goodness-of-fit approach was also applied to the results to obtain the MDNBR value at the 95/95 tolerance limit. Initial sensitivity analysis was performed with the 59 cases per accident scenario by use of Pearson correlation coefficients. The results show that this typical PWR core

  7. Maintaining steam/condensate lines

    International Nuclear Information System (INIS)

    Russum, S.A.

    1992-01-01

    Steam and condensate systems must be maintained with the same diligence as the boiler itself. Unfortunately, they often are not. The water treatment program, critical to keeping the boiler at peak efficiency and optimizing operating life, should not stop with the boiler. The program must encompass the steam and condensate system as well. A properly maintained condensate system maximizes condensate recovery, which is a cost-free energy source. The fuel needed to turn the boiler feedwater into steam has already been provided. Returning the condensate allows a significant portion of that fuel cost to be recouped. Condensate has a high heat content. Condensate is a readily available, economical feedwater source. Properly treated, it is very pure. Condensate improves feedwater quality and reduces makeup water demand and pretreatment costs. Higher quality feedwater means more reliable boiler operation

  8. Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kral, P. [Nuclear Research Inst. Rez (Switzerland)

    1995-12-31

    The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal `MSH Rupture` leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS. 7 refs.

  9. Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kral, P [Nuclear Research Inst. Rez (Switzerland)

    1996-12-31

    The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal `MSH Rupture` leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS. 7 refs.

  10. Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

    International Nuclear Information System (INIS)

    Kral, P.

    1995-01-01

    The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal 'MSH Rupture' leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS

  11. Shearon Harris steam generator channel head drain line leakage

    International Nuclear Information System (INIS)

    Bauer, P.A.

    1992-01-01

    All three Shearon Harris steam generators were equipped with Inconel 600 drain penetrations inserted into clearance holes on the bottom center line of the plenums, roll expanded into the plenum shell, and seal welded to the stainless steel cladding. Eddy current inspections showed axial cracks in the drain lines of B and C generators, but not on the leaking A generator. The drain lines of the three generators were repaired by cutting off the pipe under the plenum, applying Inconel 600 cladding to the underside of the plenum by a temper bead process, spot facing the overlay cladding and welding a new Inconel 600 pipe coupling to the clad surface. 3 figs

  12. Method to detect steam generator tube leakage

    International Nuclear Information System (INIS)

    Watabe, Kiyomi

    1994-01-01

    It is important for plant operation to detect minor leakages from the steam generator tube at an early stage, thus, leakage detection has been performed using a condenser air ejector gas monitor and a steam generator blow down monitor, etc. In this study highly-sensitive main steam line monitors have been developed in order to identify leakages in the steam generator more quickly and accurately. The performance of the monitors was verified and the demonstration test at the actual plant was conducted for their intended application to the plants. (author)

  13. Thermal expansion measurement of turbine and main steam piping by using strain gages in power plants

    International Nuclear Information System (INIS)

    Na, Sang Soo; Chung, Jae Won; Bong, Suk Kun; Jun, Dong Ki; Kim, Yun Suk

    2000-01-01

    One of the domestic co-generation plants have undergone excessive vibration problems of turbine attributed to external force for years. The root cause of turbine vibration may be shaft alignment problem which sometimes is changed by thermal expansion and external force, even if turbine technicians perfectly performed it. To evaluate the alignment condition from plant start-up to full load, a strain measurement of turbine and main steam piping subjected to thermal loading is monitored by using strain gages. The strain gages are bonded on both bearing housing adjusting bolts and pipe stoppers which installed in the x-direction of left-side main steam piping near the turbine inlet in order to monitor closely the effect of turbine under thermal deformation of turbine casing and main steam piping during plant full load. Also in situ load of constant support hangers in main steam piping system is measured by strain gages and its results are used to rebalance the hanger rod load. Consequently, the experimental stress analysis by using strain gages turns out to be very useful tool to diagnose the trouble and failures of not only to stationary components but to rotating machinery in power plants

  14. Tar formation in a steam-O2 blown CFB gasifier and a steam blown PBFB gasifier (BabyHPR) : Comparison between different on-line measurement techniques and the off-line SPA sampling and analysis method

    NARCIS (Netherlands)

    Meng, X.; Mitsakis, P.; Mayerhofen, M.; De Jong, W.; Gaderer, M.; Verkooijen, A.H.M.; Spliethoff, H.

    2012-01-01

    Two on-line tar measurement campaigns were carried out using an atmospheric pressure 100 “”kWth steam-O2 blown circulating fluidized bed (CFB) gasifier at the Delft University of Technology (TUD) and a 30–40kWth steam blown pressurized bubbling fluidized bed (PBFB) gasifier BabyHPR (Heatpipe

  15. Welded joints engineering design of the primary circuit, surge line and main steam piping of the Angra 2 reactor

    International Nuclear Information System (INIS)

    Volta, Angelo Roberto; Couto, Jose Gonzalo Villaverde

    1995-01-01

    The erection of nuclear systems of a Nuclear Power Station is under international requests, that results in a detailed elaboration of documents for the performance of welds. NUCLEN as an engineering design company, responsible for the erection of Angra 2, developed a suitable software program for the elaboration of welding procedure qualifications, tests and examination sequence plans and heat treatment plans applied to primary circuit, surgeline and main steam piping. The paper shows the employed methodology for the elaboration of these documents, as well as the requested engineering design of welding technology and testability in order to assure the stipulated quality level, according to requirements of the specifications, codes and norms. (author). 6 refs

  16. Steam line rupture experiments with the PPOOLEX test facility

    International Nuclear Information System (INIS)

    Laine, J.; Puustinen, M.

    2008-07-01

    The results of the steam line rupture experiment series in 2007 with the scaled down PPOOLEX test facility designed and constructed at Lappeenranta University of Technology are reported. The test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. Air was blown into the dry well compartment and from there through a DN200 blowdown pipe to the condensation pool. Altogether five experiments, each consisting of several blows (tests), were carried out. The main purpose of the experiment series was to study the initial phase of a postulated steam line break accident inside a BWR containment. Specifically, thermal stratification in the dry well compartment and ejection of water plug from the blowdown pipe were of interest. In addition, the effect of counterpressure on bubble dynamics was studied. A temperature difference of approximately 15 deg. C between the upper and lower part of the dry well was measured. In the wet well gas space, a temperature difference of more than 30 deg. C was registered. These were measured during the compression period of the tests. Towards the end of the tests the temperature differences tended to disappear. To get a more detailed picture of temperature distribution in the wet well, especially close to the water level, a dense net of measurements is required in future experiments. In longer experiments, heat conduction to structures and heat losses to surroundings should also be taken into account. Ejection of water plugs from the blowdown pipe did not cause notable loads to the structures due to the suppressing effect of the dry well compartment. The maximum measured pressure pulse at the pool bottom was only 10 kPa and the maximum strain amplitude at the pool bottom rounding was negligible both in axial and circumferential direction. As the counterpressure of the system increased, but the flow rate remained the same, the maximum size of the air bubbles at the blowdown pipe outlet got smaller and

  17. Steam line rupture experiments with the PPOOLEX test facility

    Energy Technology Data Exchange (ETDEWEB)

    Laine, J.; Puustinen, M. (Lappeenranta Univ. of Technology, Nuclear Safety Research Unit (Finland))

    2008-07-15

    The results of the steam line rupture experiment series in 2007 with the scaled down PPOOLEX test facility designed and constructed at Lappeenranta University of Technology are reported. The test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. Air was blown into the dry well compartment and from there through a DN200 blowdown pipe to the condensation pool. Altogether five experiments, each consisting of several blows (tests), were carried out. The main purpose of the experiment series was to study the initial phase of a postulated steam line break accident inside a BWR containment. Specifically, thermal stratification in the dry well compartment and ejection of water plug from the blowdown pipe were of interest. In addition, the effect of counterpressure on bubble dynamics was studied. A temperature difference of approximately 15 deg. C between the upper and lower part of the dry well was measured. In the wet well gas space, a temperature difference of more than 30 deg. C was registered. These were measured during the compression period of the tests. Towards the end of the tests the temperature differences tended to disappear. To get a more detailed picture of temperature distribution in the wet well, especially close to the water level, a dense net of measurements is required in future experiments. In longer experiments, heat conduction to structures and heat losses to surroundings should also be taken into account. Ejection of water plugs from the blowdown pipe did not cause notable loads to the structures due to the suppressing effect of the dry well compartment. The maximum measured pressure pulse at the pool bottom was only 10 kPa and the maximum strain amplitude at the pool bottom rounding was negligible both in axial and circumferential direction. As the counterpressure of the system increased, but the flow rate remained the same, the maximum size of the air bubbles at the blowdown pipe outlet got smaller and

  18. HELB Analysis for ESBWR Reactor Building and Main Steam Tunnel

    Energy Technology Data Exchange (ETDEWEB)

    Noguera Oliva, O.

    2011-07-01

    The Reactor Building compartments and tbe Main Steam Tunnel are modeled using GOTHIC 7.2a. These models are based on Control Volumes (Rooms/Compartments/Regions), Flow Paths (junctions such as vent path or any opening) and Boundary Conditions (Mass and energy releases and outside conditions). Due to the different break locations, four models are built to analyze the short-term pressurization response. Are shown the cases analyzed, the results obtained and the models used for this purpose.

  19. The Concept of Steam Pressure Control by Changing the Feedwater Flow during Heatup Operation for an Integral Reactor with a Once-Through Steam Generator

    International Nuclear Information System (INIS)

    Seo, Jae Kwang; Choi, Ki Yong; Kang, Han Ok; Kim, Young In; Yoon, Ju Hyeon; Zee, Sung Qunn

    2005-01-01

    The design features of a once-through steam generator (OTSG) for an integral reactor are significantly different from the commercial U-tube type steam generator from several aspects such as the general arrangement, size, operation conditions, and so on. Therefore a sufficient understanding of the thermal-hydraulic characteristics of the OTSG is essential for the design of the nuclear steam supply system (NSSS) and the power conversion system (PCS). It is also necessary to develop operation procedures complying to the unique design features of the OTSG of interest. The OTSG is sized to produce a sufficiently superheated steam during a normal power operation and therefore the secondary system can be simple relative to that of the other types of steam generators. For the plant adopting the OTSG, the steam pressure in the secondary circuit (tube side of the OTSG) is controlled to be constant during a normal power operation. Constant steam pressure is realized by regulating the control valve on the main steam line dedicated for this purpose. However during a heatup operation, at which the fluid state at the exit of the OTSG is a single phase hot water or two phases, it is not proper to use the control valve on the main steam line due to a control problem at low and multi-phase flow conditions and possibly an erosion problem. For these reasons, another dedicated line called a startup cooling line is used during a heatup condition. There may be several operational conditions for the secondary fluid required to pass through during heatup operation, depending on the design of the PCS. In general, there are two conditions: One is a condition for a vacuum operation for the condenser and another is an entry condition for a steam pressure control operation for an auxiliary power system. In this study, the concept of using a simple startup cooling line with a fixed flow resistance and changing the feedwater flow for the pressure control of the PCS during a heatup period are

  20. A study on the evaluation of vibration effect and the development of vibration reduction method for Wolsung unit 1 main steam piping

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyun; Kim, Yeon Whan [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center; Kim, Tae Ryong; Park, Jin Ho [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of)

    1996-08-01

    The main steam piping of nuclear power plant which runs between steam generator and high pressure turbine has been experienced to have a severe effect on the safe operation of the plant due to the vibration induced by the steam flowing inside the piping. The imposed cyclic loads by the vibration could result in the degradation of the related structures such as connection parts between main instruments, valves, pipe supports and building. The objective of the study is to reduce the vibration level of Wolsung nuclear power plant unit 1 main steam pipeline by analyzing vibration characteristics of the piping, identifying sources of the vibration and developing a vibration reduction method .The location of the maximum vibration is piping between the main steam header and steam chest .The stress level was found to be within the allowable limit .The main vibration frequency was found to be 4{approx}6 Hz which is the same as the natural frequency from model test .A vibration reduction method using pipe supports of energy absorbing type(WEAR)is selected .The measured vibration level after WEAR installation was reduced about 36{approx}77% in displacement unit (author). 36 refs., 188 figs.

  1. A coupled 3-D kinetics/system thermal-hydraulic analysis of main steam line break accident for Optimized Power Reactor 1000

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Yung Kwon; Choi, Chul Jin; Kim, Eun Kee; Lee, Sang Yong [Korea Power Engineering Company, Inc, 150 Deokjin-dong, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

    2006-07-01

    This paper presents the results of the coupled 3-D neutronics/thermal-hydraulic analysis of hypothetical main steam line break (MSLB) accident for Optimized Power Reactor 1000. One of the major concerns of this accident is a return-to-power occurrence accompanied with extremely large radial peaking near the stuck Control Element Assembly (CEA). The conventional point kinetics application does not properly account for this kind of asymmetric and local core behavior. Therefore, the current licensing method of point kinetics application introduces some uncertainties and conservatisms in the physics parameters generation, e.g., the static net scram rod worth, moderator cooldown reactivity, Doppler reactivity, and a 3-D peaking factor. The recently developed UNICORN-TM code system is applied for the 3-D coupled calculation, where neutronics code MASTER is coupled with the best-estimate system transient code RETRAN. The 3-D coupled results were assessed in comparison with those by point kinetics application using stand-alone RETRAN application. To quantify the 3-D reactivity benefits over point kinetics, both calculations assumed the accidents to be initiated from the same core state, e.g., end of cycle burnup, fuel and CEA configuration with the same initial moderator and Doppler temperature coefficient, and with initial system thermal-hydraulic condition. The core physics parameters required for point kinetics application were produced using MASTER with the method and procedure consistent with the current licensing application. The occurrence of return-to-power was simulated by intentionally reducing the net CEA worth in order to assess the spatial power distribution and local T-H effect on the dynamic reactivity feedback. The results have demonstrated that the 3-D analysis removes some of the conservatisms inherent in point kinetics analysis mainly caused by the inability to properly account for local reactivity feedback effects during return-to-power transient

  2. A coupled 3-D kinetics/system thermal-hydraulic analysis of main steam line break accident for Optimized Power Reactor 1000

    International Nuclear Information System (INIS)

    Jin, Yung Kwon; Choi, Chul Jin; Kim, Eun Kee; Lee, Sang Yong

    2006-01-01

    This paper presents the results of the coupled 3-D neutronics/thermal-hydraulic analysis of hypothetical main steam line break (MSLB) accident for Optimized Power Reactor 1000. One of the major concerns of this accident is a return-to-power occurrence accompanied with extremely large radial peaking near the stuck Control Element Assembly (CEA). The conventional point kinetics application does not properly account for this kind of asymmetric and local core behavior. Therefore, the current licensing method of point kinetics application introduces some uncertainties and conservatisms in the physics parameters generation, e.g., the static net scram rod worth, moderator cooldown reactivity, Doppler reactivity, and a 3-D peaking factor. The recently developed UNICORN-TM code system is applied for the 3-D coupled calculation, where neutronics code MASTER is coupled with the best-estimate system transient code RETRAN. The 3-D coupled results were assessed in comparison with those by point kinetics application using stand-alone RETRAN application. To quantify the 3-D reactivity benefits over point kinetics, both calculations assumed the accidents to be initiated from the same core state, e.g., end of cycle burnup, fuel and CEA configuration with the same initial moderator and Doppler temperature coefficient, and with initial system thermal-hydraulic condition. The core physics parameters required for point kinetics application were produced using MASTER with the method and procedure consistent with the current licensing application. The occurrence of return-to-power was simulated by intentionally reducing the net CEA worth in order to assess the spatial power distribution and local T-H effect on the dynamic reactivity feedback. The results have demonstrated that the 3-D analysis removes some of the conservatisms inherent in point kinetics analysis mainly caused by the inability to properly account for local reactivity feedback effects during return-to-power transient

  3. Main steam system piping response under safety/relief valve opening events

    International Nuclear Information System (INIS)

    Swain, E.O.; Esswein, G.A.; Hwang, H.L.; Nieh, C.T.

    1980-01-01

    The stresses in the main steam branch pipe of a Boiling Water Reactor due to safety/relief valve blowdown has been measured from an in situ piping system test. The test results were compared with analytical results. The predicted stresses using the current state of art analytical methods used for BWR SRV discharge transient piping response loads were found to be conservative when compared to the measured stress values. 3 refs

  4. TRAC analysis of steam-generator overfill transients for TMI-1

    International Nuclear Information System (INIS)

    Bassett, B.

    1983-01-01

    A reactor safety issue concerning the overfilling of once-through steam generators leading to combined primary/secondary blowdown has been raised recently. A series of six calculations, performed with the LWR best-estimate code, TRAC-PD2, on a Babcock and Wilcox Plant (TMI-1), was performed to investigate this safety issue. The base calculation assumed runaway main feedwater to one steam generator causing it to overfill and to break the main steam line. Four additional calculations build onto the base case with combinations of a pump-seal failure, a steam-generator tube rupture, and the pilot-operated relief valve not reseating. A sixth calculation involved only the rupture of a single steam-generator tube. The results of these analyses indicate that for the transients investigated, the emergency cooling system provided an adequate make-up coolant flow to mitigate the accidents

  5. Durability of bends in high-temperature steam lines under the conditions of long-term operation

    Science.gov (United States)

    Katanakha, N. A.; Semenov, A. S.; Getsov, L. B.

    2015-04-01

    The article presents the results of stress-strain state computations and durability of bent and steeply curved branches of high-temperature steam lines carried out on the basis of the finite element method using the modified Soderberg formula for describing unsteady creep processes with taking the accumulation of damage into account. The computations were carried out for bends made of steel grades that are most widely used for manufacturing steam lines (12Kh1MF, 15Kh1M1F, and 10Kh9MFB) and operating at different levels of inner pressure and temperature. The solutions obtained using the developed creep model are compared with those obtained using the models widely used in practice.

  6. Energy and exergy analysis of the turbo-generators and steam turbine for the main feed water pump drive on LNG carrier

    International Nuclear Information System (INIS)

    Mrzljak, Vedran; Poljak, Igor; Mrakovčić, Tomislav

    2017-01-01

    Highlights: • Two low-power steam turbines in the LNG carrier propulsion plant were investigated. • Energy and exergy efficiencies of both steam turbines vary between 46% and 62%. • The ambient temperature has a low impact on exergy efficiency of analyzed turbines. • The maximum efficiencies area of both turbines was investigated. • A method for increasing the turbo-generator efficiencies by 1–3% is presented. - Abstract: Nowadays, marine propulsion systems are mainly based on internal combustion diesel engines. Despite this fact, a number of LNG carriers have steam propulsion plants. In such plants, steam turbines are used not only for ship propulsion, but also for electrical power generation and main feed water pump drive. Marine turbo-generators and steam turbine for the main feed water pump drive were investigated on the analyzed LNG carrier with steam propulsion plant. The measurements of various operating parameters were performed and obtained data were used for energy and exergy analysis. All the measurements and calculations were performed during the ship acceleration. The analysis shows that the energy and exergy efficiencies of both analyzed low-power turbines vary between 46% and 62% what is significantly lower in comparison with the high-power steam turbines. The ambient temperature has a low impact on exergy efficiency of analyzed turbines (change in ambient temperature for 10 °C causes less than 1% change in exergy efficiency). The highest exergy efficiencies were achieved at the lowest observed ambient temperature. Also, the highest efficiencies were achieved at 71.5% of maximum developed turbo-generator power while the highest efficiencies of steam turbine for the main feed water pump drive were achieved at maximum turbine developed power. Replacing the existing steam turbine for the main feed water pump drive with an electric motor would increase the turbo-generator energy and exergy efficiencies for at least 1–3% in all analyzed

  7. The main features of control and operation of steam turbines at nuclear power plants

    International Nuclear Information System (INIS)

    Czinkoczky, B.

    1981-01-01

    The output and speed control of steam turbines at nuclear power plants as well as the combination of both controls are reviewed and evaluated. At the same time the tasks of unit control at nuclear power plants, the control of steady main steam pressure and medium pressure of primary circuit, further the connection of reactor and turbine controls and the self-controlling properties of pressurized water reactor are dealt with. Hydraulic and electro-hydraulic speed control, the connection of cach-up dampers and speed control and the application of electro-hydraulic signal converters are discussed. The accomplishment of protection is also described. (author)

  8. Moisture separator for steam generator level measurement system

    International Nuclear Information System (INIS)

    Cantineau, B.J.

    1987-01-01

    A steam generator level measurement system having a reference leg which is kept full of water by a condensation pot, has a liquid/steam separator in the connecting line between the condensation pot and the steam phase in the steam generator to remove excess liquid from the steam externally of the steam generator. This ensures that the connecting line does not become blocked. The separator pot has an expansion chamber which slows down the velocity of the steam/liquid mixture to aid in separation, and a baffle, to avoid liquid flow into the line connected to the condensate pot. Liquid separated is returned to the steam generator below the water level through a drain line. (author)

  9. Diagnostics of metal state in steam lines

    International Nuclear Information System (INIS)

    Gofman, Yu.M.; Kazantseva, N.S.; Losev, L.Ya.; Nevolina, G.S.

    1986-01-01

    A series of micropore detection methods is suggested: light microscopy, electron microscopy, hydrostatic weighing; and comparative investigations of pore-formation processes in 12Kh1MF steel steam lines, which have operated for about 100 thousand hours at t=550 deg C and 47-55 MPa stresses are conducted using these methods. It is shown, that electron microscpy method can be applied at the early stages damaging, when embrionic micropores of 0.1 μm in size appear. Optical metallography allows one to detect pores of about 1 μm in size. Damage in density using the hydrostatic weighing method is estimated in the following way: at creep stages 1-2-0.1; at stage 3-0.4-0.6; at predestruction stage the degree of damage equals to 0.7-0.8

  10. Development of the On-line Acoustic Leak Detection Tool for the SFR Steam Generator Protection

    International Nuclear Information System (INIS)

    Kim, Tae-Joon; Jeong, Ji-Young; Kim, Jong-Man; Kim, Byung-Ho; Kim, Seong-O

    2007-01-01

    The successful detection of a water/steam into a sodium leak in the SFR SG (steam generator) at an early phase of a leak origin depends on the fast response and sensitivity of a leak detection system. This intention of an acoustic leak detection system is stipulated by a key impossibility of a fast detecting of an intermediate leak by the present nominal systems such as the hydrogen meter. Subject of this study is to introduce the detection performance of an on-line acoustic leak detection tool discriminated by a back-propagation neural network with a preprocessing of the 1/m Octave band analysis, and to introduce the status of an on-line development being developed with the acoustic leak detection tool(S/W) in KAERI. For a performance test, it was used with the acoustic signals for a sodium-water reaction from the injected steam into water experiments in KAERI, the acoustic signals injected from the water into the sodium obtained in IPPE, and the background noise of the PFR superheater

  11. Dynamic load in suppression pool during BWR main steam safety relief valve actuation

    International Nuclear Information System (INIS)

    Tsukada, Hiroshi; Yamaguchi, Hirokatsu; Morita, Terumichi

    1979-01-01

    BWRs are so designed that the exhaust steam from main steam safety relief valves is led to pressure suppression pools, and the steam is condensed in pool water, but at this time, dynamic load seems to arise in the pool water. In Tokai No. 2 Power Station, a Mark-2 containment vessel was adopted to improve the reliability as much as possible and to obtain the design with margin. In this report, the result of actual machine test in Tokai No. 2 Power Station and the method of reducing the load are described. When a relief valve works, the discharge of water in exhaust pipes into a suppression pool, the exhaust of air in exhaust pipes and repeated expansion and contraction of bubbles in pool water, and the exhaust of steam and condensation occur. As for the construction of the suppression pool in Tokai No. 2 Power Station, cross-shaped quencher and the structure with jet deflector were installed. The test plan and the test result with an actual machine are reported. The soundness of the Mark-2 containment vessel and the structures in the pool was proved. The differential pressure acting on the structures was negligibly small. The measured pulsating pressure was in the range from 0.84 to -0.39 kg/cm 2 . (Kako, I.)

  12. A Study on the Main Steam Safety Valve Opening Mechanism by Flashing on NPPs

    International Nuclear Information System (INIS)

    Kim, Bae Joo

    2009-01-01

    A safety injection event happened by opening of the Main Steam Safety Valve at Kori unit 1 on April 16, 2005. The safety valves were opened at the lower system pressure than the valve opening set point due to rapid system pressure drop by opening of the Power Operated Relief Valve installed at the upstream of the Main Steam System. But the opening mechanism of safety valve at the lower set point pressure was not explained exactly. So, it needs to be understood about the safety valve opening mechanism to prevent a recurrence of this kind of event at a similar system of Nuclear Power Plant. This study is aimed to suggest the hydrodynamic mechanism for the safety valve opening at the lower set point pressure and the possibility of the recurrence at similar system conditions through document reviewing for the related previous studies and Kori unit 1 event

  13. A standing pressure wave hypothesis of oscillating forces generated during a steam line break

    International Nuclear Information System (INIS)

    Tinoco, H.

    2001-01-01

    A rapid glance at the figure depicting the net forces acting on the reactor vessel and internals, as obtained through a CFD simulation of a BWR steam line break, reveals an amazing oscillating regularity of these forces which is in glaring contrast to the chaotic behaviour of the steam pressure field in the steam annulus. Assuming that the decompression process excites and maintains standing pressure waves in the annular cylindrical region constituted by the steam annulus, it is possible to reconstruct the net forces acting on the reactor vessel and internals through the contribution of almost only the first dispersive mode. If a Neumann boundary condition is assumed at the section connecting the steam annulus to the steam dome, the frequency predicted is approximately % 5.9 higher than that of the CFD simulations. However, this connecting section allows wave transmission, and a more appropriate boundary condition should be one of the Robin type. Therefore, this section is modelled as an absorbing wall, and the corresponding normal impedance is calculated using the CFD simulations. Week non-linear effects can also be observed in the calculated forces through the presence of the first subharmonic. By the methodology described above, an estimate of the forces acting on the reactor vessel and internals of unit 3 of Forsmark Nuclear Power Plant has been obtained. (author)

  14. Type GQS-1 high pressure steam manifold water level monitoring system

    International Nuclear Information System (INIS)

    Li Nianzu; Li Beicheng; Jia Shengming

    1993-10-01

    The GQS-1 high pressure steam manifold water level monitoring system is an advanced nuclear gauge that is suitable for on-line detecting and monitor in high pressure steam manifold water level. The physical variable of water level is transformed into electrical pulses by the nuclear sensor. A computer is equipped for data acquisition, analysis and processing and the results are displayed on a 14 inch color monitor. In addition, a 4 ∼ 20 mA output current is used for the recording and regulation of water level. The main application of this gauge is for on-line measurement of high pressure steam manifold water level in fossil-fired power plant and other industries

  15. Main results of assessing integrity of RNPP-3 steam generator heat exchange tubes in accident management

    International Nuclear Information System (INIS)

    Shugajlo, Al-j P.; Mustafin, M.A.; Shugajlo, Al-r P.; Ryzhov, D.I.; Zhabin, O.I.

    2017-01-01

    Tubes integrity evaluation under accident conditions considering drain of SG and current technical state of steam exchange tubes is an important question regarding SG long-term operation and improvement of accident management strategy.The main investigation results prepared for heat exchange surface of RNPP-3 steam generator are presented in this research aimed at assessing integrity of heat exchange tubes under accident conditions, which lead to full or partial drain of heat exchange surface, in particular during station blackout.

  16. Level-Swell Prediction With RETRAN-3D And Its Application To A BWR Steam-Line-Break Analysis

    International Nuclear Information System (INIS)

    Aounallah, Y.; Hofer, K.

    2003-01-01

    Level-swell experiments have often been simulated using system codes, such as TRAC and RELAP, but only cursory assessments have been performed with the operational-transient code RETRAN-3D, the main system code used within the STARS project. The present study, initiated in the framework of a BWR Steam-Line-Break (SLB) accident scenario, addresses this lacuna by performing RETRAN simulations of the General Electric Level-Swell experiments, and by investigating their implications on power plant accident analyses. Parameters to which the predicted level swell is sensitive have been identified, and recommendations on code options are made. The SLB analysis objective was to determine the amount of steam and liquid discharged through the break under specified boundary conditions, and to gauge the results against reference values. The impact of the nodalization of the upper part of the reactor pressure vessel was investigated and found to play an important role, whereas the level swell induced from flashing was found not to be the predominant factor for these simulations. (author)

  17. Maintenance and repair of LMFBR steam generators

    International Nuclear Information System (INIS)

    Verriere, P.; Alanche, J.; Minguet, J.L.

    1984-06-01

    After some general remarks on the French fast neutron system, this paper presents the state of the program for the construction of fast reactor in France. Then, the general design of Super Phenix 1 steam generator components is outlined and, the in-service monitoring systems and protective devices with which they are equiped are briefly described. The methods used, in the event of leakage, for leak location, steam generator inspection, steam generator repair and putting the affected loop back into service, are discussed. There are two main lines of research, relating respectively to the means of water leak detection in sodium and the inspection arrangements that will be used either periodically, or following a sodium-water reaction. Finally, after a brief description of the steam generator, this paper describes the four incidents (leaks) that occurred on the Phenix steam generator in the course of 1982 and 1983, and the subsequent repair operations

  18. The applicability of CFD to simulate and study the mixing process and the thermo-hydraulic consequences of a main steam line break in PWR model

    Directory of Open Access Journals (Sweden)

    Farkas Istvan

    2017-01-01

    Full Text Available This paper focuses on the validation and applicability of CFD to simulate and analyze the thermo-hydraulic consequences of a main steam line break. Extensive validation data come from experiments performed using the Rossendorf coolant mixing model facility. For the calculation, the range of 9 to 12 million hexahe¬dral cells was constructed to capture all details in the interrogation domain in the system. The analysis was performed by running a time-dependent calculation, Detailed analyses were made at different cross-sections in the system to evaluate not only the value of the maximum and minimum temperature, but also the loca¬tion and the time at which it occurs during the transient which is considered to be indicator for the quality of mixing in the system. CFD and experimental results were qualitatively compared; mixing in the cold legs with emergency core cooling systems was overestimated. This could be explained by the sensitivity to the bound¬ary conditions. In the downcomer, the experiments displayed higher mixing: by our assumption this related to the dense measurement grid (they were not modelled. The temperature distribution in the core inlet plane agreed with the measurement results. Minor deviations were seen in the quantitative comparisons: the maximum temperature difference was 2ºC.

  19. Development of main steam safety valve set pressure evaluating system

    International Nuclear Information System (INIS)

    Oketani, Koichiro; Manabe, Yoshihisa.

    1991-01-01

    A main steam safety valve set pressure test is conducted for all valves during every refueling outage in Japan's PWRs. Almost all operations of the test are manually conducted by a skilled worker. In order to obtain further reliability and reduce the test time, an automatic test system using a personnel computer has been developed in accordance with system concept. Quality assurance was investigated to fix system specifications. The prototype of the system was manufactured to confirm the system reliability. The results revealed that this system had high accuracy measurement and no adverse influence on the safety valve. This system was concluded to be applicable for actual use. (author)

  20. Steam generator tube failures

    International Nuclear Information System (INIS)

    MacDonald, P.E.; Shah, V.N.; Ward, L.W.; Ellison, P.G.

    1996-04-01

    A review and summary of the available information on steam generator tubing failures and the impact of these failures on plant safety is presented. The following topics are covered: pressurized water reactor (PWR), Canadian deuterium uranium (CANDU) reactor, and Russian water moderated, water cooled energy reactor (VVER) steam generator degradation, PWR steam generator tube ruptures, the thermal-hydraulic response of a PWR plant with a faulted steam generator, the risk significance of steam generator tube rupture accidents, tubing inspection requirements and fitness-for-service criteria in various countries, and defect detection reliability and sizing accuracy. A significant number of steam generator tubes are defective and are removed from service or repaired each year. This wide spread damage has been caused by many diverse degradation mechanisms, some of which are difficult to detect and predict. In addition, spontaneous tube ruptures have occurred at the rate of about one every 2 years over the last 20 years, and incipient tube ruptures (tube failures usually identified with leak detection monitors just before rupture) have been occurring at the rate of about one per year. These ruptures have caused complex plant transients which have not always been easy for the reactor operators to control. Our analysis shows that if more than 15 tubes rupture during a main steam line break, the system response could lead to core melting. Although spontaneous and induced steam generator tube ruptures are small contributors to the total core damage frequency calculated in probabilistic risk assessments, they are risk significant because the radionuclides are likely to bypass the reactor containment building. The frequency of steam generator tube ruptures can be significantly reduced through appropriate and timely inspections and repairs or removal from service

  1. Experiment Operating Specification for the Semiscale MOD-2C feedwater and steam line break experiment series. Appendix S-FS-6 and 7

    International Nuclear Information System (INIS)

    Boucher, T.J.; Owca, W.A.

    1985-05-01

    This document is the Semiscale MOD-2C feedwater and steam line break experiment series Experiment Operating Specification Appendix for tests S-FS-6 and S-FS-7. Test S-FS-6 is the third test in the series and simulates a 100% break in a steam generator bottom feedwater line downstream of the check valve accompanied by compounding factors (such as check valve failure, loss-of-offsite power at SIS and SIS delayed until low steam generator pressure signal). The test is terminated after plant stabilization and recovery procedures including unaffected loop steam and feed, pressurizer heater operation, pressurizer auxiliary spray operation, and normal charging/letdown operation. Test S-FS-7 is the fourth test in the series and simulates a 14.3% break in a steam generator bottom feedwater line downstream of the check valve, accompanied by compounding factors. The test is terminated after plant stabilization procedures including unaffected loop steam and feed, pressurizer heater operation, and normal charging/letdown operation. The test was followed by an affected loop secondary refill after isolating the break. The Appendix contains information on the major fluid systems, initial experiment conditions, experiment boundary conditions, and sequence of experiment events. Also included is a discussion of the scaling criteria and philosophy used to develop the experiment initial and boundary conditions and system configuration

  2. Hydrodynamic and acoustic analysis in 3-D of a section of main steam line to EPU conditions; Analisis hidrodinamico y acustico en 3D de una seccion de linea de vapor principal a condiciones de EPU

    Energy Technology Data Exchange (ETDEWEB)

    Centeno P, J.; Castillo J, V.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico); Nunez C, A.; Polo L, M. A., E-mail: baldepeor21@hotmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    The objective of this word is to study the hydrodynamic and acoustic phenomenon in the main steam lines (MSLs). For this study was considered the specific case of a pipe section of the MSL, where is located the standpipe of the pressure and/or safety relief valve (SRV). In the SRV cavities originates a phenomenon known as whistling that generates a hydrodynamic disturbance of acoustic pressure waves with different tones depending of the reactor operation conditions. In the SRV cavities the propagation velocity of the wave can originate mechanical damage in the structure of the steam dryer and on free parts. The importance of studying this phenomenon resides in the safety of the integrity of the reactor pressure vessel. To dissipate the energy of the pressure wave, acoustic side branches (ASBs) are used on the standpipe of the SRVs. The ASBs are arrangements of compacted lattices similar to a porous medium, where the energy of the whistling phenomenon is dissipate and therefore the acoustic pressure load that impacts in particular to the steam dryers, and in general to the interns of the vessel, diminishes. For the analysis of the whistling phenomenon two three-dimensional (3-D) models were built, one hydrodynamic in stationary state and other acoustic for the harmonic times in transitory regimen, in which were applied techniques of Computational Fluid Dynamics. The study includes the reactor operation analysis under conditions of extended power up rate (EPU) with ASB and without ASB. The obtained results of the gauges simulated in the MSL without ASB and with ASB, show that tones with values of acoustic pressure are presented in frequency ranges between 160 and 200 Hz around 12 MPa and of 7 MPa, respectively. This attenuation of tones implies the decrease of the acoustic loads in the steam dryer and in the interns of the vessel that are designed to support pressures not more to 7.5 MPa approximately. With the above-mentioned is possible to protect the steam dryer

  3. Effects of aging and service wear on main steam isolation valves and valve operators

    International Nuclear Information System (INIS)

    Clark, R.L.

    1996-03-01

    In recent years main steam isolation valve (MSIV operating problems have resulted in significant operational transients (e.g., spurious reactor trips, steam generator dry out, excessive valve seat leakage), increased cost, and decreased plant availability. A key ingredient to an engineering-oriented reliability improvement effort is a thorough understanding of relevant historical experience. A detailed review of historical failure data available through the Institute of Nuclear Power Operation's Nuclear Plant Reliability Data System has been conducted for several types of MSIVs and valve operators for both boiling-water reactors and pressurized-water reactors. The focus of this review is on MSIV failures modes, actuator failure modes, consequences of failure on plant operations, method of failure detection, and major stressors affecting both valves and valve operators

  4. Measure Guideline. Steam System Balancing and Tuning for Multifamily Residential Buildings

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jayne [Partnership for Advanced Residential Retrofit (PARR), Chicago, IL (United States); Ludwig, Peter [Partnership for Advanced Residential Retrofit (PARR), Chicago, IL (United States); Brand, Larry [Partnership for Advanced Residential Retrofit (PARR), Chicago, IL (United States)

    2013-04-01

    This guideline provides building owners, professionals involved in multifamily audits, and contractors insights for improving the balance and tuning of steam systems. It provides readers an overview of one-pipe steam heating systems, guidelines for evaluating steam systems, typical costs and savings, and guidelines for ensuring quality installations. It also directs readers to additional resources for details not included here. Measures for balancing a distribution system that are covered include replacing main line vents and upgrading radiator vents. Also included is a discussion on upgrading boiler controls and the importance of tuning the settings on new or existing boiler controls. The guideline focuses on one-pipe steam systems, though many of the assessment methods can be generalized to two-pipe steam systems.

  5. STEAM ENHANCED REMEDIATION RESEARCH FOR DNAPL IN FRACTURED ROCK, LORING AIR FORCE BASE, LIMESTONE, MAINE

    Science.gov (United States)

    This report details a research project on Steam Enhanced Remediation (SER) for the recovery of volatile organic compounds from fractured limestone that was carried out at the Quarry at the former Loring Air Force Base in Limestone, Maine. This project was carried out by USEPA, Ma...

  6. Examination of failed studs from No. 2 steam generator at the Maine Yankee Nuclear Power Station

    International Nuclear Information System (INIS)

    Czajkowski, C.

    1983-02-01

    Three studs removed from service on the primary manway cover from steam generator No. 2 of the Maine Yankee station were sent to Brookhaven National Laboratory (BNL) for examination. The examination consisted of visual/dye penetrant examination, optical metallography and Scanning Electron Microscopy/Energy Dispersive Spectroscopy (SEM/EDS) evaluation. One bolt was through cracked and its fracture face was generally transgranular in nature with numerous secondary intergranular cracks. The report concludes that the environmenally assisted cracking of the stud was due to the interaction of the various lubricants used with steam leaks associated with this manway cover

  7. Analysis of steam line break of SMART using RETRAN-3D/INT

    International Nuclear Information System (INIS)

    Kim, Tae-Wan; Kim, Jong-Won; Park, Goon-Cherl

    2003-01-01

    RETRAN-3D has been modified to be suitable to safety analysis for integral type marine reactor with modular helical-coiled steam generator cassettes. The modified RETRAN-3D, RETRAN-3D/INT, has helical coil heat conductor model and heat transfer coefficient models for tube and shell sides of helical-coiled steam generator. In addition, moving models are added to simulate the effect of ship motions such as inclination, heaving, rolling and so on. RETRAN-3D/INT has been verified with natural circulation experiment conducted in Seoul National University and the analysis results for the first Japanese nuclear ship, MUTSU. In this study, the safety analysis for SMART, which has been developed by Korea Atomic Energy Research Institute, is performed to examine the applicability of RETRAN-3D/INT to the safety analysis of SMART. The steam line break is selected as reference case. The break type is assumed to the guillotine break. The loss of offsite power is considered as a coincident event and the failure of single train of passive residual heat removal system is assumed as single failure. From the results, it is found that RETRAN-3D/INT can appropriately simulate the transient of SMART and the improvement of non-condensable gas model is required. (author)

  8. Three Mile Island Unit 1 Main Steam Line Break Three-Dimensional Neutronics/Thermal-Hydraulics Analysis: Application of Different Coupled Codes

    International Nuclear Information System (INIS)

    D'Auria, Francesco; Moreno, Jose Luis Gago; Galassi, Giorgio Maria; Grgic, Davor; Spadoni, Antonino

    2003-01-01

    A comprehensive analysis of the double ended main steam line break (MSLB) accident assumed to occur in the Babcock and Wilcox Three Mile Island Unit 1 (TMI-1) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy, in cooperation with the University of Zagreb, Croatia. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development-Committee on the Safety of Nuclear Installations-Nuclear Science Committee pressurized water reactor MSLB benchmark.Thermal-hydraulic system codes (various versions of Relap5), three-dimensional (3-D) neutronics codes (Parcs, Quabbox, and Nestle), and one subchannel code (Cobra) have been adopted for the analysis. Results from the following codes (or code versions) are assumed as reference:1. Relap5/mod3.2.2, beta version, coupled with the 3-D neutron kinetics Parcs code parallel virtual machine (PVM) coupling2. Relap5/mod3.2.2, gamma version, coupled with the 3-D neutron kinetics Quabbox code (direct coupling)3. Relap5/3D code coupled with the 3-D neutron kinetics Nestle code.The influence of PVM and of direct coupling is also discussed.Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by Pennsylvania State University in cooperation with GPU Nuclear Corporation (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission. The comparison among the results obtained by adopting the same thermal-hydraulic nodalization and the coupled code version is discussed in this paper.The capability of the control rods to recover the accident has been demonstrated in all the cases as well as the capability of all the codes to predict the time evolution of the assigned transient. However, one stuck control rod caused some 'recriticality' or 'return to power' whose magnitude is largely affected by boundary and initial conditions

  9. TRAC-PF1/MOD 1 independent assessment: Semiscale MOD-2A feedwater-line break (S-SF-3) and steam-line break (S-SF-5) tests

    International Nuclear Information System (INIS)

    Dobranich, D.

    1985-11-01

    The TRAC-PF1/MOD1 independent assessment project at Sandia is part of an overall effort funded by the NRC to determine the ability of various systems codes to predict the detailed thermal/hydraulic response of LWRs during accident and off-normal conditions. As part of this effort, calculations for Semiscale Mod-2A test S-SF-3, a feedwater-line break test, and S-SF-5, a steam-line break test, were performed with TRAC-PF1/MOD1. Most aspects of both the S-SF-3 and S-SF-5 steady-state calculations were found to be in good agreement with data. However, the need for a better steam separator model was identified from the S-SF-3 calculation. Overall, the qualitative behavior of both transients was calculated reasonably well; however, there were some discrepancies in the prediction of the quantitative behavior. The results for the S-SF-3 transient calculation indicate that the primary-to-secondary heat transfer degradation began too early. This was possibly due to overprediction of entrainment in the steam generator boiler, leading to an incorrect calculation of the secondary-side fluid distribution during the steady state. However, there was insufficient data to verify this. Results for the S-SF-5 transient calculation indicate that the primary-side fluid temperature response to a steam-line break was in reasonable agreement with data but the pressure response did not coincide with the data. Uncertainties in the data are sufficient to prevent us from determining the exact cause of this discrepancy, but there is indirect evidence that the calculated rate of phase change in the pressurizer was incorrect. 16 refs., 73 figs., 13 tabs

  10. Development of a wheat-Aegilops searsii substitution line with positively affecting Chinese steamed bread quality.

    Science.gov (United States)

    Du, Xuye; Ma, Xin; Min, Jingzhi; Zhang, Xiaocun; Jia, Zhenzhen

    2018-03-01

    A wheat- Aegilops searsii substitution line GL1402, in which chromosome 1B was substituted with 1S s from Ae. searsii , was developed and detected using SDS-PAGE and GISH. The SDS-PAGE analysis showed that the HMW-GS encoded by the Glu-B1 loci of Chinese Spring was replaced by the HMW-GS encoded by the Glu-1S s loci of Ae. searsii . Glutenin macropolymer (GMP) investigation showed that GL1402 had a much higher GMP content than Chinese Spring did. A dough quality comparison of GL1402 and Chinese Spring indicated that GL1402 showed a significantly higher protein content and middle peak time (MPT), and a smaller right peak slope (RPS). Quality tests of Chinese steamed bread (CSB) showed that the GL1402 also produced good steamed bread quality. These results suggested that the substitution line is a valuable breeding material for improving the wheat processing quality.

  11. Numerical and computational aspects of the coupled three-dimensional core/ plant simulations: organization for economic cooperation and development/ U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-II. 2. TRAB-3D/SMABRE Calculation of the OECD/ NRC PWR MSLB Benchmark

    International Nuclear Information System (INIS)

    Daavittila, A.; Haemaelaeinen, A.; Kyrki-Rajamaki, R.

    2001-01-01

    All three exercises of the OECD/NRC Pressurized Water Reactor (PWR) Main-Steam-Line-Break (MSLB) Benchmark were calculated at VTT Energy. The SMABRE thermal-hydraulics code was used for the first exercise, the plant simulation with point-kinetics neutronics. The second exercise was calculated with the TRAB-3D three-dimensional reactor dynamics code. The third exercise was calculated with the combination TRAB-3D/SMABRE. Both codes have been developed at VTT Energy. The results of all the exercises agree reasonably well with those of the other participants; thus, instead of reporting the results, this paper concentrates on describing the computational aspects of the calculation with the foregoing codes and on some observations of the sensitivity of the results. In the TRAB-3D neutron kinetics, the two-group diffusion equations are solved in homogenized fuel assembly geometry with an efficient two-level nodal method. The point of the two-level iteration scheme is that only one unknown variable per node, the average neutron flux, is calculated during the inner iteration. The nodal flux shapes and cross sections are recalculated only once in the outer iteration loop. The TRAB-3D core model includes also parallel one-dimensional channel hydraulics with detailed fuel models. Advanced implicit time discretization methods are used in all submodels. SMABRE is a fast-running five-equation model completed by a drift-flux model, with a time discretization based on a non-iterative semi-implicit algorithm. For the third exercise of the benchmark, the TMI-1 models of TRAB-3D and SMABRE were coupled. This was the first time these codes were coupled together. However, similar coupling of the HEXTRAN and SMABRE codes has been shown to be stable and efficient, when used in safety analyses of Finnish and foreign VVER-type reactors. The coupling used between the two codes is called a parallel coupling. SMABRE solves the thermal hydraulics both in the cooling circuit and in the core

  12. Long-term rupture strength as a criterion of operational durability of steam line metal

    International Nuclear Information System (INIS)

    Gofman, Yu.M.

    2000-01-01

    The method for substantiation of the steam line service life prolongation, depending on the achieved level of the metal vulnerability to damage, is proposed. The methodology for evaluating the metal state is developed on the basis of the durability bond with the level of the vulnerability to damage through micropores and the ferrite dislocation structure state. The main changes in the metal at the 1-3 stages of its creep are presented. The micropores are absent at the 1 stage. the micropores of about 0.1 μm in diameter are identified at the beginning of the 2 stage. The ferrite grains on the transition from the 2 to the 3 creep stage are mainly fragmentary. There takes place further micropores growth on the grain boundaries up to 1 - 3 μm. Significant number of recrystallized volumes in the ferrite is observed at the 3 creep stage. The number of micropores of 1 - 3μm in size sharply increases, and, as a rule, chains of micropores are observed. The pores of 5 μm in size are formed at the pre-destruction stage, the fusion whereof leads to microcracks formation [ru

  13. Inversion mechanisms for OH main lines astrophysical masers

    International Nuclear Information System (INIS)

    Elitzur, M.

    1977-01-01

    Excitation processes that can lead to inversion of the main lines of the OH ground state are discussed. Due to the frequency dependence of the emission coefficient of dust, far-IR emitted by warm enough dust can excite the upper halves of the Λ-doublets of rotational levels more strongly than the lower halves. The cascade back to the ground state will then invert the main lines and it is shown that this mechanism can explain rather well the main lines emission from OH-IR stars. The main lines masers associated with compact HII regions are discussed extensively. It is argued that the most plausible explanation for them is a model based on the mechanism suggested some time ago by Johnston where the inversion is due to collisional excitation by streams of uni-directional electrons. (author)

  14. Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

    International Nuclear Information System (INIS)

    Jeong, Jae Jun; Chung, Bub Dong

    2005-09-01

    For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis

  15. Preliminary Analysis of a Steam Line Break Accident with the MARS-KS code for the SMART Design with Passive Safety Systems

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Doohyuk; Ko, Yungjoo; Suh, Jaeseung [Hannam Univ., Daejeon (Korea, Republic of); Bae, Hwang; Ryu, Sunguk; Yi, Sungjae; Park, Hyunsik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    SMART has been developed by KAERI, and SMART-Standard Design Approval (SDA) was recently granted in 2012. A SMART design with Passive Safety System (PSS) features (called SMART-PSS) is being developed and added to the standard design of SMART by KAERI to improve its safety system. Active safety systems such as safety injection pumps will be replaced by a passive safety system, which is actuated only by the gravity force caused by the height difference. All tanks for the passive safety systems are higher than the injection nozzle, which is located around the reactor coolant pumps (RCPs). In this study, a preliminary analysis of the main steam line break accident (MSLB) was performed using the MARS-KS code to understand the general behavior of the SMART-PSS design and to prepare its validation test with the SMART-ITL (FESTA) facility. An anticipated accident for the main steam line break (MSLB) was performed using the MARS-KS code to understand the thermal-hydraulic behaviors of the SMART-PSS design. The preliminary analysis provides good insight into the passive safety system design features of the SMART-PSS and the thermal-hydraulic characteristics of the SMART design. The analysis results of the MSLB showed that the core water collapsed level inside the core support barrel was maintained high over the active core top level during the transient period. Therefore, the SMART-PSS design has satisfied the requirements to maintain the plant at a safe shutdown condition during 72 hours without AC power or operator action after an anticipated accident.

  16. Steam condenser

    International Nuclear Information System (INIS)

    Masuda, Fujio

    1980-01-01

    Purpose: To enable safe steam condensation by providing steam condensation blades at the end of a pipe. Constitution: When high temperature high pressure steam flows into a vent pipe having an opening under water in a pool or an exhaust pipe or the like for a main steam eacape safety valve, non-condensable gas filled beforehand in the steam exhaust pipe is compressed, and discharged into the water in the pool. The non-condensable gas thus discharged from the steam exhaust pipe is introduced into the interior of the hollow steam condensing blades, is then suitably expanded, and thereafter exhausted from a number of exhaust holes into the water in the pool. In this manner, the non-condensable gas thus discharged is not directly introduced into the water in the pool, but is suitable expanded in the space of the steam condensing blades to suppress extreme over-compression and over-expansion of the gas so as to prevent unstable pressure vibration. (Yoshihara, H.)

  17. Modal analysis of main steam line piping under high energy line break condition

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae-Jin; Kim, Seung Hyun; Je, Sang-Yun; Chang, Yoon-Suk [Kyung Hee University, Yongin (Korea, Republic of)

    2016-10-15

    If HELB (High Energy Line Break) occurs in NPPs (Nuclear Power Plants), not only environmental effect like release of radioactive material but also secondary structural defects should be considered. Jet impingement phenomenon caused by sudden pipe rupture may lead to severe damage on neighboring safe-related components and other structure. Lots of studies have been conducted to assess dynamic behaviors of the SG and MSL piping while pipe whip restraints and jet impingement shields are taken into account during design stage. Arroyo et al. performed modal analyses of a simple square component to examine the jet impingement phenomenon. Also, structural characteristics were predicted to assure structural integrity against the HELB. In this study, we examined dynamic characteristics of SG and MSL piping in a typical 1000MWe NPP. Simulation was performed by using two commercial computational softwares. In particular, modal analyses were conducted to determine mode shapes and natural frequencies of the structure and maximum displacements. The data obtain from each software were compared and observation was discussed in relation to the jet impingement phenomenon. In this research, modal analyses on the SG and MSL piping were carried out to get natural frequencies, vibration mode shapes and maximum displacements. Thereby, the following key finding was observed. (1) Maximum displacement was calculated at the top of SG outlet nozzle with y-directional bending at the third mode. (2) The differences between two models were respectively 7% in natural frequencies and less than 1% in maximum displacements.

  18. Increase of Steam Moisture in the BWR-Facility KKP 1

    International Nuclear Information System (INIS)

    Noack, Volker

    2002-01-01

    Main steam moisture in a BWR facility is determined by steam quality at core outlet and efficiency of steam separators and steam dryers. Transport of water with steam is accompanied by transport of radionuclides out of RPV resulting in enhanced radiation level in the main steam system. A remarkable increase of main steam moisture started at KKP 1 in 1997. In the following years increase of steam outlet moisture started at lower and lower core mass flow rates. Dose rate in main steam system increased simultaneously. Core mass flow rate and thus thermal power had to be reduced during stretch out operation to keep the main steam moisture below the specified boundary of 0.2 %. This boundary also guarantees, that radiological exposure remains far below approved values. The increase of main steam moisture corresponds with the application of low leakage core loading. Low leakage core loading results in enhanced steam generation in the center and in reduced steam generation in the outer zones of the core. It can be shown, that the uneven steam generation in the core became stronger over the years. Therefore, steam quality at inlet of the outer steam separators was getting lower. This resulted in higher carry over of water in this steam separators and steam dryers, thus explaining the increasing main steam moisture. KKP 1 started in 2000 with spectral shift operation. As one should expect, this resulted in reduced steam moisture. It remains the question of steam moisture in case of stretch out operation. Countermeasures are briefly discussed. (authors)

  19. A fault detection and diagnosis in a PWR steam generator

    International Nuclear Information System (INIS)

    Park, Seung Yub

    1991-01-01

    The purpose of this study is to develop a fault detection and diagnosis scheme that can monitor process fault and instrument fault of a steam generator. The suggested scheme consists of a Kalman filter and two bias estimators. Method of detecting process and instrument fault in a steam generator uses the mean test on the residual sequence of Kalman filter, designed for the unfailed system, to make a fault decision. Once a fault is detected, two bias estimators are driven to estimate the fault and to discriminate process fault and instrument fault. In case of process fault, the fault diagnosis of outlet temperature, feed-water heater and main steam control valve is considered. In instrument fault, the fault diagnosis of steam generator's three instruments is considered. Computer simulation tests show that on-line prompt fault detection and diagnosis can be performed very successfully.(Author)

  20. Practical use of valve seating machine with remote control system for main steam isolation valve at N.P.S

    International Nuclear Information System (INIS)

    Ito, Sadao; Noda, Hiroshi; Sadamura, Morito; Utsunomiya, Yasushi.

    1975-01-01

    The main steam isolation valves in BWR power stations are installed at the boundary of reactor containment vessels, and 2 valves in each main steam system total 8 valves in a plant. They are pneumatically operated Y type globe valves for preventing the release of radioactive substances in the atmosphere in case of the breaking of main steam pipes and also preventing the loss of coolant in case of the breaking of recirculating equipments. Therefore careful leak test, inspection, and seat-fitting are carried out to the valves at each regular maintenance. The manual maintenance work is difficult because of narrow space and the reduction of exposure, and the seat-fitting work requires the skill of high degree, therefore Okano Valve Manufacturing Co. and Tokyo Electric Power Co. jointly started the research and development of an automatic valve seating machine, and successfully put it to practical use in Fukushima No.1 Nuclear Power Station in Nov. 1974. First, the problems in the manual seat-fitting work were investigated, and the means to mechanically solve them were materialized with a prototype machine. After its mock-up test, an actual machine was designed and manufactured. The test result showed remarkable reduction of exposure and labor-saving, and the leak evaluation was sufficiently below the allowable value. (Kako, I.)

  1. Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jae Jun; Chung, Bub Dong

    2005-09-15

    For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis.

  2. 46 CFR 61.15-5 - Steam piping.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Steam piping. 61.15-5 Section 61.15-5 Shipping COAST... Periodic Tests of Piping Systems § 61.15-5 Steam piping. (a) Main steam piping shall be subjected to a... removed and the piping thoroughly examined. (b) All steam piping subject to pressure from the main boiler...

  3. Vibrations as a contributor to the cracking of PGV-1000 steam generator cold collector

    International Nuclear Information System (INIS)

    Verezemskij, V.G.

    1997-01-01

    The results of multiple investigations of cold collector ruptures at NPPs with WWER-1000 reactor as a complex and multi-parameter process are generalized. It is shown that the temperature of cold collector operation (280-290 deg C) at which environment corrosion effects are maximum has an important role for revealing the factors causing the damage. When the reactor plant operates under stationary and transient conditions the primary coolant circuit equipment, main circulation pipelines and main steam lines become involved into vibrations with different intensities as a result of pressure pulsations at reverse and multiple frequencies of the main circulation pumps connected with formation of standing pressure waves in the primary circuit and steam lines. The analysis made gives an opportunity to conclude that dynamic loads (vibrations) play the leading part in reaching the limits of cold collector metal cyclic strength and its cracking. It means that the measures for cold collector service life increasing should be directed on vibration amplitude lowering and cyclic stress decreasing

  4. Steam-generator-tube-rupture transients for pressurized-water reactors

    International Nuclear Information System (INIS)

    Dobranich, D.; Henninger, R.J.; DeMuth, N.S.

    1982-01-01

    Steam generator tube ruptures with and without concurrent main-steam-line break are investigated for pressurized water reactors supplied by the major US vendors. The goal of these analyses is to provide thermodynamic and flow conditions for the determination of iodine transport to the environment and to provide an evaluation of the adequacy of the plant safety systems and operating procedures for controlling these transients. The automatic safety systems of the plant were found to be adequate for the mitigation of these transients. Emergency injection system flows equilibrated with the leakage flows and prevented core uncovery. Sufficient time was afforded by the plant safety systems for the operators to identify the problem and to take appropriate measures

  5. Cycle improvement for nuclear steam power plant

    International Nuclear Information System (INIS)

    Silvestri, G.J. Jr.

    1976-01-01

    A pressure-increasig ejector element is disposed in an extraction line intermediate to a high pressure turbine element and a feedwater heater. The ejector utilizes high pressure fluid from a reheater drain as the motive fluid to increase the pressure at which the extraction steam is introduced into the feedwater heater. The increase in pressure of the extraction steam entering the feedwater heater due to the steam passage through the ejector increases the heat exchange capability of the extraction steam thus increasing the overall steam power plant efficiency

  6. Fission product retention during faults involving steam generator tube rupture

    International Nuclear Information System (INIS)

    Rodliffe, R.S.

    1983-08-01

    In some PWR fault conditions, such as stuck open safety relief valve in the secondary circuit or main steam line break, the release of fission products to the atmosphere may be increased by the leakage of primary coolant into the secondary circuit following steam generator tube rupture. The release may be reduced by retention either within the primary circuit or within the affected steam generator unit (SGU). The mechanisms leading to retention are reviewed and quantified where possible. The parameters on which any analysis will be most critically dependent are identified. Fission product iodine and caesium may be retained in the secondary side of a SGU either by partition to retained water or by droplet deposition on surfaces and subsequent evaporation to dryness. Two extreme simplifications are considered: SGU 'dry', i.e. the secondary side is steam filled, and SGU 'wet', i.e. the tube bundle is covered with water. Consideration is given to: the distribution of fission products between gaseous and aerosol forms; mechanisms for droplet formation, deposition and resuspension; fission product retention during droplet or film evaporation primary coolant mixing and droplet scrubbing in a wet SGU; and the performance of moisture separators and steam driers. (author)

  7. Three-Dimensional Modeling of a Steam-Line Break in a Boiling Water Reactor

    International Nuclear Information System (INIS)

    Tinoco, Hernan

    2002-01-01

    Because of weld problems, the core grids of Units 1 and 2 at the Forsmark nuclear power plant have been replaced by grids of a new design, consisting of a single machined piece without welds. The qualifying structural analysis has been carried out considering dynamic loads, which implies that even loss-of-coolant accidents have to be included. Therefore, a detailed time description of the loads acting on the different internal parts of the reactor is needed. To achieve sufficient space and time resolution, a computational fluid dynamics (CFD) analysis was considered to be a viable alternative.A CFD analysis of a steam-line break in the boiling water reactor of Unit 2 is the subject of this work. The study is based on the assumption that the timescale of the transient analysis is smaller than the relaxation time of the water-steam system. Therefore, a simulation of only the upper, steam part of the reactor with no two-phase effects (flashing) is feasible.The results obtained display a rather complex behavior of the decompression process, forcing the analysis of the pressure field to be accomplished through animation. In contrast, the computed instantaneous forces over different internal parts oscillate regularly and are approximately twice the forces estimated in the past by simpler methods, with frequencies of 30 to 40 Hz; top amplitudes of ∼1.64 MN; and relatively low damping, ∼25% after 0.5 s.According to the present results, this type of modeling is physically meaningful for simulation timescales smaller than the water-steam relaxation time, i.e., ∼0.5 s at reactor conditions. At larger times, a two-phase model is necessary to describe the decompression process since two-phase effects are dominant. The results have not yet been validated with experiments, but validation computations will be run in the future for comparison with results of the Marviken tests

  8. Regulatory analysis for the resolution of Generic Issue 125.II.7 ''Reevaluate Provision to Automatically Isolate Feedwater from Steam Generator During a Line Break''

    International Nuclear Information System (INIS)

    Basdekas, D.L.

    1988-09-01

    Generic Issue 125.II.7 addresses the concern related to the automatic isolation of auxiliary feedwater (AFW) to a steam generator with a broken steam or feedwater line. This regulatory analysis provides a quantitative assessment of the costs and benefits associated with the removal of the AFW automatic isolation and concludes that no new regulatory requirements are warranted. 21 refs., 7 tabs

  9. Design of on-line steam generator leak monitoring system based on Cherenkov counting technique

    International Nuclear Information System (INIS)

    Dileep, B.N.; D'Cruz, S.J.; Biju, P.; Jashi, K.B.; Prabhakaran, V.; Venkataramana, K.; Managanvi, S.S.

    2006-01-01

    The methodology developed by Nuclear Power Corporation of India Ltd. for identification of leaky Steam Generator (SG) by monitoring 134 I activity in the blow down water is a very high sensitive method. However, this technique can not be put into use as an on-line system. A new method of on-line detection of SG leak and identify the offending SG based on Cherenkov counting technique is explained in this paper. It identifies the leak by detecting Cherenkov radiation produced by the hard beta emitting radio nuclides escaped into feed water during leak in an operating reactor. A simulated system shows that a leak rate of 2 kg/h can be detected by the proposed system, while coolant 134 I activity is 3.7 MBq/l (100μCi/l). (author)

  10. ANALYSIS AND SIMULATION OF MAIN MAGNET TRANSMISSION LINE EFFECT

    Energy Technology Data Exchange (ETDEWEB)

    ZHANG,W.; MARNERIS, I.; SANDBERG, J.

    2007-06-25

    A main magnet chain forms a pair of transmission lines. Pulse-reflection-caused voltage and current differentiation throughout the magnet chain can have adverse effect on main magnet field quality. This effect is associated with magnet system configuration, coupling efficiency, and parasitic parameters. A better understanding of this phenomenon will help us in new design and existing system upgrade. In this paper, we exam the transmission line effect due to different input functions as well as configuration, coupling, and other parameters.

  11. Application of PHADEC method for the decontamination of radioactive steam piping components

    International Nuclear Information System (INIS)

    Lo Frano, R.; Pilo, F.; Aquaro, D.

    2013-01-01

    The dismantling of nuclear plants is a complex activity that originates often a large quantity of radioactive contaminated residue. In this paper the attention was focused on the PHADEC (Phosphoric Acid Decontamination) plant adopted for the clearance of Caorso NPP (in Italy) metallic systems and components contaminated by Co 60 (produced by the neutron capture in the iron materials), like the main steam lines, moisture separator of the turbine buildings, etc.. The PHADEC plant consists in a chemical off line treatment: the crud, deposited along the steam piping during life plant as an example, is removed by means of acid attacks in ponds coupled to a high pressure water washing. Due to the fact that the removed contaminated layers, essentially, iron oxides of various chemical composition, depend on components geometry, type of contamination and time of treatment in the PHADEC plant, it becomes of meaningful importance to suggest a procedure capable to improve the control of the PHADEC process parameters. This study aimed thus at the prediction and optimization of the mentioned treatment time in order to improve the efficiency of the plant itself and to achieve, in turn, the minimization of produced wastes. To the purpose an experimental campaign was carried out by analysing several samples, i.e. taken along the main steam piping line. Smear tests as well as metallographic analyses were carried out in order to determine respectively the radioactivity distribution and the crud composition on the inner surface of the components. Moreover the radioactivity in the crud thickness was measured. These values allowed finally to correlate the residence time in the acid attack ponds to the level of the achieved decontamination. (authors)

  12. Replacement of outboard main steam isolation valves in a boiling water reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    Schlereth, J.R.; Pennington, D.

    1996-12-01

    Most Boiling Water Reactor plants utilize wye pattern globe valves for main steam isolation valves for both inboard and outboard isolation. These valves have required a high degree of maintenance attention in order to pass the plant local leakage rate testing (LLRT) requirements at each outage. Northern States Power made a decision in 1993 to replace the outboard valves at it`s Monticello plant with double disc gate valves. The replacement of the outboard valves was completed during the fall outage in 1994. During the spring outage in April of 1996 the first LLRT testing was performed with excellent results. This presentation will address the decision process, time requirements and planning necessary to accomplish the task as well as the performance results and cost effectiveness of replacing these components.

  13. Replacement of outboard main steam isolation valves in a boiling water reactor plant

    International Nuclear Information System (INIS)

    Schlereth, J.R.; Pennington, D.

    1996-01-01

    Most Boiling Water Reactor plants utilize wye pattern globe valves for main steam isolation valves for both inboard and outboard isolation. These valves have required a high degree of maintenance attention in order to pass the plant local leakage rate testing (LLRT) requirements at each outage. Northern States Power made a decision in 1993 to replace the outboard valves at it's Monticello plant with double disc gate valves. The replacement of the outboard valves was completed during the fall outage in 1994. During the spring outage in April of 1996 the first LLRT testing was performed with excellent results. This presentation will address the decision process, time requirements and planning necessary to accomplish the task as well as the performance results and cost effectiveness of replacing these components

  14. Turbine main engines

    CERN Document Server

    Main, John B; Herbert, C W; Bennett, A J S

    1965-01-01

    Turbine Main Engines deals with the principle of operation of turbine main engines. Topics covered include practical considerations that affect turbine design and efficiency; steam turbine rotors, blades, nozzles, and diaphragms; lubricating oil systems; and gas turbines for use with nuclear reactors. Gas turbines for naval boost propulsion, merchant ship propulsion, and naval main propulsion are also considered. This book is divided into three parts and begins with an overview of the basic mode of operation of the steam turbine engine and how it converts the pressure energy of the ingoing ste

  15. Pressure waves transient occurred in the steam generators feedwater lines of the Atucha-1 Nuclear Power Plant

    International Nuclear Information System (INIS)

    Balino, J.L.; Carrica, P.M.; Larreteguy, A.E.

    1993-01-01

    The pressure transient occurred at Atucha I Nuclear Power Plant in March 1990 is simulated. The transient was due to the fast closure of a flow control valve at the steam generators feedwater lines. The system was modelled, including the actuation of the relief valves. The minimum closure time for no actuation of the relief valves and the evolution of the velocity and piezo metric head for different cases were calculated. (author)

  16. Application of PHADEC method for the decontamination of radioactive steam piping components of Caorso plant

    International Nuclear Information System (INIS)

    Lo Frano, R.; Aquaro, D.; Fontani, E.; Pilo, F.

    2014-01-01

    Highlights: • Application of PHADEC chemical off-line methodology. • Decontamination of radioactive steam piping components of Caorso turbine building. • Experimental characterization of metallic components, e.g., by SEM analysis. • Measure of the efficiency of treatment by means of the reduction of activity and vs. the treatment time. • Minimization of secondary waste produced during decontamination activity of Caorso BWR plant. - Abstract: The dismantling of nuclear plants is a complex activity that originates often a large quantity of radioactive contaminated residue. In this paper the attention was focused on the PHADEC (PHosphoric Acid DEContamination) plant adopted for the clearance of Caorso NPP (in Italy) metallic systems and components contaminated by Co60 (produced by the neutron capture in the iron materials), like the main steam lines, moisture separator of the turbine buildings, etc. The PHADEC plant consists in a chemical off line treatment: the crud, deposited along the steam piping during life plant as an example, is removed by means of acid attacks in ponds coupled to a high pressure water washing. Due to the fact that the removed contaminated layers, essentially, iron oxides of various chemical composition, depend on components geometry, type of contamination and time of treatment in the PHADEC plant, it becomes of meaningful importance to suggest a procedure capable to improve the control of the PHADEC process parameters. This study aimed thus at the prediction and optimization of the mentioned treatment time in order to improve the efficiency of the plant itself and to achieve, in turn, the minimization of produced wastes. To the purpose an experimental campaign was carried out by analysing several samples, i.e., taken along the main steam piping line. Smear tests as well as metallographic analyses were carried out in order to determine respectively the radioactivity distribution and the crud composition on the inner surface of the

  17. Analysis of Communication between Main Control Room Operators in Decision-making Process in Steam Generator Tube Rupture Accident

    International Nuclear Information System (INIS)

    Petkov, M.; Petkov, G.

    2006-01-01

    The paper presents an investigation results for Main Control Room operators' reliability by Performance Evaluation of Teamwork method, based on FSS-1000 training archives in KNPP in case of Steam Generator Tube Rupture accident. The advantages of operators' teamwork are shown: a) group decision-making vs. individual one: b) positive influence of crew initiated communication consisting of orders and reports that are required by instruction. (authors)

  18. Steam purity in PWRs

    International Nuclear Information System (INIS)

    Hopkinson, J.; Passell, T.

    1982-01-01

    Reports that 2 EPRI studies of PWRs prove that impure steam triggers decay of turbine metals. Reveals that EPRI is attempting to improve steam monitoring and analysis, which are key steps on the way to deciding the most cost-effective degree of steam purity, and to upgrade demineralizing systems, which can then reliably maintain that degree of purity. Points out that 90% of all cracks in turbine disks have occurred at the dry-to-wet transition zone, dubbed the Wilson line. Explains that because even very clean water contains traces of chemical impurities with concentrations in the parts-per-billion range, Crystal River-3's secondary loop was designed with even more purification capability; a deaerator to remove oxygen and prevent oxidation of system metals, and full-flow resin beds to demineralize 100% of the secondary-loop water from the condenser. Concludes that focusing attention on steam and water chemistry can ward off cracking and sludge problems caused by corrosion

  19. Oahu Sewer Main Lines, Oahu County HI, 2016, Honolulu GIS

    Data.gov (United States)

    U.S. Environmental Protection Agency — Linear features representing sewer main lines as maintained by Honolulu ENV Department of Environmental Services. Includes an inventory of sewer mains used for...

  20. Condensation induced water hammer in steam supply system

    International Nuclear Information System (INIS)

    Andrews, P.B.; Antaki, G.A.; Rawls, G.B.; Gutierrez, B.J.

    1995-01-01

    The accidental mixing of steam and water usually leads to condensation induced water hammer. This phenomenon is not uncommon in the power and process industries, and is of particular concern due to the high energies which accompany steam transients. The paper discusses the conditions which lead to a recent condensation induced water hammer in a 150 psig steam supply header. The ensuing structural damage, inspection and repairs are described. Finally, a list of design, maintenance and operational cautions are presented to help minimize the potential for condensation induced water hammer in steam lines

  1. Condensation induced water hammer in steam supply system

    International Nuclear Information System (INIS)

    Andrews, P.B.; Antaki, G.A.; Rawls, G.B.; Gutierrez, B.J.

    1995-01-01

    The accidental mixing of steam and water usually leads to condensation induced water hammer. THis phenomenon is not uncommon in the power and process industries, and is of particular concern due to the high energies which accompany steam transients. The paper discusses the conditions which lead to a recent condensation induced water hammer in a 150 psig steam supply header. The insuing structural damage, inspection and repairs are described. Finally, a list of design cautions are presented to help minimize the potential for condensation induced water hammer in steam lines

  2. Comparative Analyses on OPR1000 Steam Generator Tube Rupture Event Emergency Operational Guideline

    International Nuclear Information System (INIS)

    Lee, Sang Won; Bae, Yeon Kyoung; Kim, Hyeong Teak

    2006-01-01

    The Steam Generator Tube Rupture (SGTR) event is one of the important scenarios in respect to the radiation release to the environment. When the SGTR occurs, containment integrity is not effective because of the direct bypass of containment via the ruptured steam generator to the MSSV and MSADV. To prevent this path, the Emergency Operational Guideline of OPR1000 indicates the use of Turbine Bypass Valves (TBVs) as an effective means to depressurize the main steam line and prevent the lifting of MSSV. However, the TBVs are not operable when the offsite power is not available (LOOP). In this situation, the RCS cool-down is achieved by opening the both intact and ruptured SG MSADV. But this action causes the large amount of radiation release to the environment. To minimize the radiation release to the environment, KSNP EOG adopts the improved strategy when the SGTR concurrently with LOOP is occurred. However, these procedures show some duplicated procedure and branch line that might confusing the operator for optimal recovery action. So, in this paper, the comparative analysis on SGTR and SGTR with LOOP is performed and optimized procedure is proposed

  3. Dual turbine power plant and method of operating such plant, especially one having an HTGR steam supply

    International Nuclear Information System (INIS)

    Braytenbah, A.S.; Jaegtnes, K.O.

    1977-01-01

    A power plant including dual steam turbine-generators connected to pass superheat and reheat steam from a steam generator which derives heat from the coolant gas of a high temperature gas-cooled nuclear reactor is described. Associated with each turbine is a bypass line to conduct superheat steam in parallel with a high pressure turbine portion, and a bypass line to conduct superheat steam in parallel with a lower pressure turbine portion. Auxiliary steam turbines pass a portion of the steam flow to the reheater of the steam generator and drive gas blowers which circulate the coolant gas through the reactor and the steam source. Apparatus and method are disclosed for loading or unloading a turbine-generator while the other produces a steady power output. During such loading or unloading, the steam flows through the turbine portions are coordinated with the steam flows through the bypass lines for protection of the steam generator, and the pressure of reheated steam is regulated for improved performance of the gas blowers. 33 claims, 5 figures

  4. LWRA analysis of inadvertent closing of the main steam isolation valve in NPP Krsko

    International Nuclear Information System (INIS)

    Feretic, D.; Cavlina, N.; Grgic, D.; Spalj, S.

    1996-01-01

    The paper describes the use of system code RELAP5/mod2 and analyzer code LWRA in analysis of inadvertent closing of the main steam isolation valve that happened in NPP Krsko on September, 25 1995. Three cases were calculated in order to address different aspects of the modelled transient. This preliminary calculation showed that, even though the real plant behaviour was not completely reproduced, such kind of analysis can help to better understand plant behaviour and to identify important phenomena in the plant during transient. The results calculated by RELAP5 and LWRA were similar and both codes indicated lack of better understanding of the plant systems status. The LWRA was more than 5 times faster than real time. (author)

  5. The 8-GeV transfer line injection into main ring

    International Nuclear Information System (INIS)

    Yang, M.J.

    1995-06-01

    Included in this report are a brief review of the design lattice of the 8-GeV beam transfer line and the Main Ring, the recent measurements on the 8-GeV line lattice function as well as that of the Main Ring at 8-GeV. The injection matching is a very important part of the MR operation. Mismatches such as energy, timing, or position are easily corrected because they cause oscillations which are visible on the Turn-By-Turn (TBT) TV monitor display. Mis-matches due to beta and dispersion functions are detected only by using the Flying Wire or by doing measurements during beam study. A new method which makes use of the available data from TBT hardware was used to obtain the beam phase space ellipse. Data taken from Main Ring at injection gives the beta function needed for transfer matching from 8-GeV line. The result of this measurement is also presented here

  6. STYLE, Steam Cycle Heat Balance for Turbine Blade Design in Marine Operation

    International Nuclear Information System (INIS)

    Love, J.B.; Dines, W.R.

    1970-01-01

    1 - Nature of physical problem solved: The programme carries out iterative steam cycle heat balance calculations for a wide variety of steam cycles including single reheat, live steam reheat and multistage moisture separation. Facilities are also available for including the steam-consuming auxiliaries associated with a marine installation. Though no attempt is made to carry out a detailed turbine blading design the programme is capable of automatically varying the blading efficiency from stage to stage according to local steam volume flow rate, dryness fraction and shaft speed. 2 - Method of solution: 3 - Restrictions on the complexity of the problem: Steam pressures to lie within range 0.2 to 5,000 lb/square inch abs steam temperatures to lie within range 50 to 1600 degrees F. Not more than 40 points per turbine expansion line; Not more than 10 expansion lines; Not more than 15 feed heaters. UNIVAC 1108 version received from FIAT Energia Nucleare, Torino, Italy

  7. Thermoelastic steam turbine rotor control based on neural network

    Science.gov (United States)

    Rzadkowski, Romuald; Dominiczak, Krzysztof; Radulski, Wojciech; Szczepanik, R.

    2015-12-01

    Considered here are Nonlinear Auto-Regressive neural networks with eXogenous inputs (NARX) as a mathematical model of a steam turbine rotor for controlling steam turbine stress on-line. In order to obtain neural networks that locate critical stress and temperature points in the steam turbine during transient states, an FE rotor model was built. This model was used to train the neural networks on the basis of steam turbine transient operating data. The training included nonlinearity related to steam turbine expansion, heat exchange and rotor material properties during transients. Simultaneous neural networks are algorithms which can be implemented on PLC controllers. This allows for the application neural networks to control steam turbine stress in industrial power plants.

  8. Operating experience of main steam isolation valves at Fessenheim and Bugey

    International Nuclear Information System (INIS)

    Dredemis, G.; Fourest, B.; Giroux, C.

    1985-07-01

    The paper presents the experience of Hopkinson MSIVs over about 40 reactor-years (1977 to 1984) of operation at Fessenheim and Bugey units (900 MWe PWR). The various problems encountered including ageing effects on auxiliary equipments and increases in closure time are discussed. The corrective actions undertaken by the utility and the safety assessment of these events performed by the french safety authorities are also described. This study is the synthesis of an in-depth analysis of Main Steam Isolation Valves (MSIV) and their auxiliary circuits equipping the Bugey and Fessenheim 900 MWe PWR nuclear power plants. These valves are different from those installed in the other French 900 MWe PWR reactors. The evaluation of the operation of these valves was made on the basis of incidents which occured during operation of the units or during the periodic tests, as well as anomalies discovered during maintenance operations. This analysis proved that the anomalies related to the design of the valves, as well as to their manufacture and installation, had been correctly dealt with. Furthermore, it should have also revealed potential anomalies due to ageing of the equipment

  9. Dual turbine power plant and a reheat steam bypass flow control system for use therein

    International Nuclear Information System (INIS)

    Braytenbah, A.S.; Jaegtnes, K.O.

    1977-01-01

    An electric power plant having dual turbine-generators connected to a steam source that includes a high temperature gas cooled nuclear reactor is described. Each turbine comprises a high pressure portion operated by superheat steam and an intermediate-low pressure portion operated by reheat steam; a bypass line is connected across each turbine portion to permit a desired minimum flow of steam from the source at times when the combined flow of steam through the turbine is less than the minimum. Coolant gas is propelled through the reactor by a circulator which is driven by an auxiliary turbine which uses steam exhausted from the high pressure portions and their bypass lines. The pressure of the reheat steam is controlled by a single proportional-plus-integral controller which governs the steam flow through the bypass lines associated with the intermediate-low pressure portions. At times when the controller is not in use its output signal is limited to a value that permits an unbiased response when pressure control is resumed, as in event of a turbine trip. 25 claims, 2 figures

  10. Steam Reforming of Bio-oil Model Compounds

    DEFF Research Database (Denmark)

    Trane, Rasmus; Jensen, Anker Degn; Dahl, Søren

    The steam reforming of bio-oil is a sustainable and renewable route to synthesis gas and hydrogen, where one of the main hurdles is carbon formation on the catalyst.......The steam reforming of bio-oil is a sustainable and renewable route to synthesis gas and hydrogen, where one of the main hurdles is carbon formation on the catalyst....

  11. Condition assessment of main structural members of steam schooner WAPAMA

    Science.gov (United States)

    Xiping Wang; James Wacker; Robert Ross; Brian Brashaw

    2008-01-01

    The historic American ship WAPAMA is the last surviving example of the wooden steam-powered schooners designed for the 19th- and 20th-century Pacific Coast lumber trade and coastal service. Since its launching in 1915, the WAPAMA has had a long and productive life in plying cargo and passengers along the stormy West Coast from Mexico to Alaska. As the sole survivor of...

  12. Steam cleaning device

    International Nuclear Information System (INIS)

    Karaki, Mikio; Muraoka, Shoichi.

    1985-01-01

    Purpose: To clean complicated and long objects to be cleaned having a structure like that of nuclear reactor fuel assembly. Constitution: Steams are blown from the bottom of a fuel assembly and soon condensated initially at the bottom of a vertical water tank due to water filled therein. Then, since water in the tank is warmed nearly to the saturation temperature, purified water is supplied from a injection device below to the injection device above the water tank on every device. In this way, since purified water is sprayed successively from below to above and steams are condensated in each of the places, the entire fuel assembly elongated in the vertical direction can be cleaned completely. Water in the reservoir goes upward like the steam flow and is drained together with the eliminated contaminations through an overflow pipe. After the cleaning has been completed, a main steam valve is closed and the drain valve is opened to drain water. (Kawakami, Y.)

  13. Model studies of the vertical steam generator thermal-hydraulic characteristics

    International Nuclear Information System (INIS)

    Desyatun, V.F.; Moskvichev, V.F.; Ulasov, V.M.; Morozov, V.G.; Burkov, V.K.; Grebennikov, V.N.

    1984-01-01

    Results of investigations conducted to clarify the calculation technique and to test the workability of the main elements and units of the PGV-250 vertical steam generator of saturated steam are considered. The steam generating capacity of the plant is 1486 t/h, thermal power is 792 MW. Steam generation follows a multiple circulation scheme. The heat surface comprises 330-shields. The investigations are carried out with a model which reproduces all the main elements of the steam generator xcluding the economizer section. The flow rates of feed water, generated steam and coolant of the first circuit as well as temperature, pressure and humidity of the generated steam past the separator are determined. The average heat transfer factors of the heat surface are calculated on the base of the data obtained and a conclusion is drawn on the correctness of the thermohydraulic calculation technique used in development of the PGV-250 steam generator design. Temperature pulsations and heat surface steaming are not observed. The steam humidity at the outlet and steam capture into sinking tubes are within permissible values

  14. Steam generator life management

    International Nuclear Information System (INIS)

    Tapping, R.L.; Nickerson, J.; Spekkens, P.; Maruska, C.

    1998-01-01

    Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in Pressurized Water Reactors (PWRs). CANDU steam generators are not immune to steam generator degradation, and the variety of CANDU steam generator designs and tube materials has led to some unexpected challenges. However, aggressive remedial actions, and careful proactive maintenance activities, have led to a decrease in steam generator-related station unavailability of Canadian CANDUs. AECL and the CANDU utilities have defined programs that will enable existing or new steam generators to operate effectively for 40 years. Research and development work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermal hydraulics, fouling, inspection and cleaning, as well as provision for specially tool development for specific problem solving. A major driving force is development of CANDU-specific fitness-for-service guidelines, including appropriate inspection and monitoring technology to measure steam generator condition. Longer-range work focuses on development of intelligent on-line monitoring for the feedwater system and steam generator. New designs have reduced risk of corrosion and fouling, are more easily inspected and cleaned, and are less susceptible to mechanical damage. The Canadian CANDU utilities have developed programs for remedial actions to combat degradation of performance (Gentilly-2, Point Lepreau, Bruce A/B, Pickering A/B), and have developed strategic plans to ensure that good future operation is ensured. This report shows how recent advances in cleaning technology are integrated into a life management strategy, discusses downcomer flow measurement as a means of monitoring steam generator condition, and describes recent advances in hideout return as a life management tool. The research and development program, as well as operating experience, has identified

  15. Main characteristics and design features of steam generators for VG-400 plant

    International Nuclear Information System (INIS)

    Golovko, V.F.; Grebennik, V.N.; Gol'tsev, A.O.; Ivanov, S.M.; Sergeev, A.I.; Pospelov, V.N.

    1988-01-01

    The description of a steam generator for the VG-400 plant performed in two variants depending on a heat-exchange surface arrangement (one-bundle coil and module-cassette construction) is given. (author)

  16. CANDU steam generator life management

    International Nuclear Information System (INIS)

    Tapping, R.L.; Nickerson, J.; Spekkens, P.; Maruska, C.

    1998-01-01

    Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in Pressurized Water Reactors (PWRs). CANDU steam generators are not immune to steam generator degradation, and the variety of CANDU steam generator designs and tube materials has led to some unexpected challenges. However, aggressive remedial actions, and careful proactive maintenance activities, have led to a decrease in steam generator-related station unavailability of Canadian CANDUs. AECL and the CANDU utilities have defined programs that will enable existing or new steam generators to operate effectively for 40 years. Research and development work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermalhydraulics, fouling, inspection and cleaning, as well as provision for specially tool development for specific problem solving. A major driving force is development of CANDU-specific fitness-for-service guidelines, including appropriate inspection and monitoring technology to measure steam generator condition. Longer-range work focuses on development of intelligent on-line monitoring for the feedwater system and steam generator. New designs have reduced risk of corrosion and fouling, are more easily inspected and cleaned, and are less susceptible to mechanical damage. The Canadian CANDU utilities have developed programs for remedial actions to combat degradation of performance (Gentilly-2, Point Lepreau, Bruce A/B, Pickering A/B), and have developed strategic plans to ensure that good future operation is ensured. The research and development program, as well as operating experience, has identified where improvements in operating practices and/or designs can be made in order to ensure steam generator design life at an acceptable capacity factory. (author)

  17. Steam Turbine Control Valve Stiction Effect on Power System Stability

    International Nuclear Information System (INIS)

    Halimi, B.

    2010-01-01

    One of the most important problems in power system dynamic stability is low frequency oscillations. This kind of oscillation has significant effects on the stability and security of the power system. In some previous papers, a fact was introduced that a steam pressure continuous fluctuation in turbine steam inlet pipeline may lead to a kind of low frequency oscillation of power systems. Generally, in a power generation plant, steam turbine system composes of some main components, i.e. a boiler or steam generator, stop valves, control valves and turbines that are connected by piping. In the conventional system, the turbine system is composed with a lot of stop and control valves. The steam is provided by a boiler or steam generator. In an abnormal case, the stop valve shuts of the steal flow to the turbine. The steam flow to the turbine is regulated by controlling the control valves. The control valves are provided to regulate the flow of steam to the turbine for starting, increasing or decreasing the power, and also maintaining speed control with the turbine governor system. Unfortunately, the control valve has inherent static friction (stiction) nonlinearity characteristics. Industrial surveys indicated that about 20-30% of all control loops oscillate due to valve problem caused by this nonlinear characteristic. In this paper, steam turbine control valve stiction effect on power system oscillation is presented. To analyze the stiction characteristic effect, firstly a model of control valve and its stiction characteristic are derived by using Newton's laws. A complete tandem steam prime mover, including a speed governing system, a four-stage steam turbine, and a shaft with up to for masses is adopted to analyze the performance of the steam turbine. The governor system consists of some important parts, i.e. a proportional controller, speed relay, control valve with its stiction characteristic, and stem lift position of control valve controller. The steam turbine has

  18. International examples of steam generator replacement

    International Nuclear Information System (INIS)

    Wiechmann, K.

    1993-01-01

    Since 1979-1980 a total of twelve nuclear power plants world-wide have had their steam generators replaced. The replacement of the Combustion steam generators in the Millstone-2 plant in the United States was completed very recently. Steam generator replacement activities are going on at present in four plants. In North Anna, the steam generators have been under replacement since January 1990. In Japan, preparations have been started for Genkai-1. Since January 1992, the two projects in Beznau-1, Switzerland, and Doel-3, Belgium, have bee planned and executed in parallel. Why steam generator replacement? There are a number of defect mechanisms which give rise to the need for early steam generator replacement. One of the main reasons is the use of Inconel-600 as material for the heating tubes. Steam generator heating tubes made of Inconel-600 have been known to exhibit their first defects due to stress corrosion cracking after less than one year of operation. (orig.) [de

  19. Pilot RCM application to the Diablo Canyon main stream system

    International Nuclear Information System (INIS)

    Groff, C.R.; Beckham, P.E.; Bych, K.H.

    1988-01-01

    In 1986 Pacific Gas ampersand Electric Company (PG ampersand E) became extremely interested in reliability-centered maintenance (RCM) after the initial review of two successful Electric Power Research Institute sponsored projects. RCM was visualized as a methodology to common sensitize the burgeoning preventive maintenance (PM) program at the Diablo Canyon plant. RCM could further the uses of predictive and condition-monitoring techniques, as well as eliminate maintenance on components whose failures were noncritical. An extensive review of maintenance and operation experience data, in conjunction with plant staff recommendations and a prioritization according to maintenance expenditures and operational/safety significance, produced the selected system: the turbine main steam supply system (main steam). The pilot project segmented the main steam system into eight subsystems to aid in analysis: (a) main steam isolation valves, (b) auxiliary feedwater pump turbine, (c) overpressure protection (steam dump), (d) main feedwater pump turbines, (e) main steam, (f) main turbine, (g) steam blowdown, and (h) moisture separator reheaters. System analysis activities, including the preparation of functional failure analyses, failure modes and effects analyses, and logic model analyses, were conducted in parallel with corrective and preventive maintenance data-gathering activities to maximize project team personnel participation during the project. Results and lessons learned are summarized

  20. Many-body calculation of the coincidence L3 photoelectron spectroscopy main line of Ni metal

    International Nuclear Information System (INIS)

    Ohno, Masahide

    2008-01-01

    The partial singles L 3 photoelectron spectroscopy (PES) main line of Ni metal correlated with Auger electrons emitted by the localized L 3 -VV Auger decay is calculated by a many-body theory. The partial singles L 3 PES main line of Ni metal almost coincides in both line shape and peak kinetic energy (KE) with the singles one. The former main line peak shows a KE shift of only 0.01 eV toward the lower KE and a very small asymmetric line shape change compared to the singles one. The asymmetric line shape change and the peak KE shift of the partial singles L 3 main line are very small. However, they are due to the variation with photoelectron KE in the branching ratio of the partial Auger decay width in the partial singles L 3 PES main line by the photoelectron KE dependent imaginary part of the shakeup self-energy. The L 3 PES main line of Ni metal measured in coincidence with the L 3 -VV ( 1 G) Auger electron spectroscopy (AES) main line peak is the partial singles one modulated by a spectral function R a of a fixed energy Auger electron analyzer so that it should show only a symmetric line narrowing by R a compared to the singles one. The L 3 PES main line peak of Ni metal measured in coincidence with the delocalized band-like L 3 -VV AES peak or not completely split-off (or not completely localized) L 3 -VV ( 3 F) AES peak, will show an asymmetric line narrowing and a KE shift compared to the singles one. Thus, the L 3 PES main line of Ni metal in coincidence with various parts of the L 3 -VV AES spectrum depends on which part of the L 3 -VV AES spectrum a fixed energy Auger electron analyzer is set. The experimental verification is in need

  1. Advanced on-line monitoring of power plant water/steam quality

    Energy Technology Data Exchange (ETDEWEB)

    Perboni, G.; Rocchini, G.; Sigon, F. [Ente Nazionale per l`Energia Elettrica, Milan (Italy)

    1995-03-01

    To improve the behaviour and resistance of materials in the water-steam cycle critical components (steam generator, condensate heaters, turbine) it is necessary to adopt proper actions for promoting formation and integrity of surface protective oxide layers and preventing general and localised corrosion and transport processes of corrosion products throughout the cycle. In this report two important topics are reported: steam side corrosion in the low pressure turbines induced by the `first condensate` in the final stages of the turbine, and the stability of the oxides layers as a function of the condensate chemistry, with particular attention to the transport of corrosion products to the boiler. Furthermore an innovative technique for monitoring some physico-chemical parameters at the actual fluid temperature (150-300C) using new electrochemical sensors improved by ENEL/CRAM is studied: pH, conductivity, corrosion rate, corrosion and redox potentials.ENEL/CRAM validated on lab-scale testing loops these sensors and carried out the following programs: calibration procedures, reliability of the response, long-term stability and assessment of a reduced maintenance. Applications of the electrochemical methods to fossil fired units have demonstrated their validity for monitoring the cycle chemistry and the resistance to corrosion of structural materials in real time.

  2. Electricity from geothermal steam

    Energy Technology Data Exchange (ETDEWEB)

    Wheatcroft, E L.E.

    1959-01-01

    The development of the power station at Wairakei geothermal field is described. Wairakei is located at the center of New Zealand's volcanic belt, which lies within a major graben which is still undergoing some degree of downfaulting. A considerable number of wells, some exceeding 610 m, have been drilled. Steam and hot water are produced from both deep and shallow wells, which produce at gauge pressures of 1.5 MPa and 0.6 MPa, respectively. The turbines are fed by low, intermediate, and high pressure mains. The intermediate pressure turbine bank was installed as a replacement for a heavy water production facility which had originally been planned for the development. Stage 1 includes a 69 MW plant, and stage 2 will bring the capacity to 150 MW. A third stage, which would bring the output up to 250 MW had been proposed. The second stage involves the installation of more high pressure steam turbines, while the third stage would be powered primarily by hot water flashing. Generation is at 11 kV fed to a two-section 500 MVA board. Each section of the board feeds through a 40 MVA transformer to a pair of 220 V transmission lines which splice into the North Island grid. Other transformers feed 400 V auxiliaries and provide local supply.

  3. What is geothermal steam worth?

    International Nuclear Information System (INIS)

    Thorhallsson, S.; Ragnarsson, A.

    1992-01-01

    Geothermal steam is obtained from high-temperature boreholes, either directly from the reservoir or by flashing. The value of geothermal steam is similar to that of steam produced in boilers and lies in its ability to do work in heat engines such as turbines and to supply heat for a wide range of uses. In isolated cases the steam can be used as a source of chemicals, for example the production of carbon dioxide. Once the saturated steam has been separated from the water, it can be transported without further treatment to the end user. There are several constraints on its use set by the temperature of the reservoir and the chemical composition of the reservoir fluid. These constraints are described (temperature of steam, scaling in water phase, gas content of steam, well output) as are the methods that have been adopted to utilize this source of energy successfully. Steam can only be transported over relatively short distances (a few km) and thus has to be used close to the source. Examples are given of the pressure drop and sizing of steam mains for pipelines. The path of the steam from the reservoir to the end user is traced and typical cost figures given for each part of the system. The production cost of geothermal steam is estimated and its sensitivity to site-specific conditions discussed. Optimum energy recovery and efficiency is important as is optimizing costs. The paper will treat the steam supply system as a whole, from the reservoir to the end user, and give examples of how the site-specific conditions and system design have an influence on what geothermal steam is worth from the technical and economic points of view

  4. Steam generator leak detection using acoustic method

    International Nuclear Information System (INIS)

    Goluchko, V.V.; Sokolov, B.M.; Bulanov, A.N.

    1982-05-01

    The main requirements to meet by a device for leak detection in sodium - water steam generators are determined. The potentialities of instrumentation designed based on the developed requirements have been tested using a model of a 550 kw steam generator [fr

  5. Model for transient simulation in a PWR steam circuit

    International Nuclear Information System (INIS)

    Mello, L.A. de.

    1982-11-01

    A computer code (SURF) was developed and used to simulate pressure losses along the tubes of the main steam circuit of a PWR nuclear power plant, and the steam flow through relief and safety valves when pressure reactors its thresholds values. A thermodynamic model of turbines (high and low pressure), and its associated components are simulated too. The SURF computer code was coupled to the GEVAP computer code, complementing the simulation of a PWR nuclear power plant main steam circuit. (Author) [pt

  6. Eddy current testing of steam generator tubes

    International Nuclear Information System (INIS)

    Neumaier, P.

    1981-01-01

    A rotating probe is described for improving the inspection of tubes and end plate in steam generators. The method allows a representation of the whole defect, consequently the observer is able to determine directly the type of defect, signal processing in-line or off-line is possible [fr

  7. Three Mile Island Nuclear Station steam generator chemical cleaning

    International Nuclear Information System (INIS)

    Hansen, C.A.

    1992-01-01

    The Three Mile Island-1 steam generators were chemically cleaned in 1991 by the B and W Nuclear Service Co. (BWNS). This secondary side cleaning was accomplished through application of the EPRI/SGOG (Electric Power Research Institute - Steam Generator Owners Group) chemical cleaning iron removal process, followed by sludge lancing. BWNS also performed on-line corrosion monitoring. Corrosion of key steam generator materials was low, and well within established limits. Liquid waste, subsequently processed by BWNS was less than expected. 7 tabs

  8. Steam System Opportunity Assessment for the Pulp and Paper, Chemical Manufacturing, and Petroleum Refining Industries: Main Report

    Energy Technology Data Exchange (ETDEWEB)

    2002-10-01

    This report assesses steam generation and use in the pulp and paper, chemical, and petroleum refining industries, and estimates the potential for energy savings from implementation of steam system performance and efficiency improvements.

  9. Steam system opportunity assessment for the pulp and paper, chemical manufacturing, and petroleum refining industries: Main report

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2002-10-01

    This report assesses steam generation and use in the pulp and paper, chemical, and petroleum refining industries, and estimates the potential for energy savings from implementation of steam system performance and efficiency improvements.

  10. Liquid metal steam generator

    International Nuclear Information System (INIS)

    Wolowodiuk, W.

    1975-01-01

    A liquid metal heated steam generator is described which in the event of a tube failure quickly exhausts out of the steam generator the products of the reaction between the water and the liquid metal. The steam is generated in a plurality of bayonet tubes which are heated by liquid metal flowing over them between an inner cylinder and an outer cylinder. The inner cylinder extends above the level of liquid metal but below the main tube sheet. A central pipe extends down into the inner cylinder with a centrifugal separator between it and the inner cylinder at its lower end and an involute deflector plate above the separator so that the products of a reaction between the liquid metal and the water will be deflected downwardly by the deflector plate and through the separator so that the liquid metal will flow outwardly and away from the central pipe through which the steam and gaseous reaction products are exhausted. (U.S.)

  11. 400-MWe consolidated nuclear steam system (CNSS): 1200-MWt/conceptual design

    International Nuclear Information System (INIS)

    1977-06-01

    A 1200-MWt consolidated nuclear steam system (CNSS) conceptual design is described. The concept, derived from nuclear merchant ship propulsion steam systems but distinctly different from those systems in detail, incorporates the steam generators within the reactor pressure vessel. This configuration eliminates primary coolant circulating piping external to the reactor pressure vessel since the primary coolant circulating pumps are mounted in the pressure vessel head. So arranged, the maximum piping break that must be assumed is that of the pressurizer surge line, which is substantially smaller than a primary coolant circulating line. A fracture of the pressurizer surge line would result in substantially lower mass and energy release rates of the primary coolant during the assumed loss-of-coolant accident. This in turn makes practical a pressure-suppression containment rather than the ''dry'' containment commonly used for pressurized water reactors

  12. BWR 2 % main recirculation line break LOCA tests RUNs 915 and 920 without HPCS in ROSA-III program

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Nakamura, Hideo; Anoda, Yoshinari; Kumamaru, Hiroshige; Yonomoto, Taisuke; Koizumi, Yasuo; Tasaka, Kanji

    1987-03-01

    This report presents the experimental results of BWR LOCA integral tests, RUNs 915 and 920, which are performed in the ROSA-III program simulating 2 % main recirculation line break LOCA tests with and without pressure control system operation. The ROSA-III test facility simulates a BWR system with volume scale of 1/424 and has four half-length electrically heated fuel bundles, two active recirculation loops, four types of ECCS's, and steam and feedwater systems. The report presents (1) the experimental results of 2 % small break LOCA phanomena in the ROSA-III system and (2) the effects of the pressure control system on the LOCA phenomena. The pressure control system contributed to (A) prevent bulk flashing in the early blowdown phase, (B) early closure of MSIV by L2 level trip, (C) early actuation of ADS by L1 level trip. However, the core thermal responses of the two tests were similar because of the similar mass inventory in PV after the ADS actuation in both tests. (author)

  13. Design specifications to ensure flow-induced vibration and fretting-wear performance in CANDU steam generators and heat exchangers

    International Nuclear Information System (INIS)

    Janzen, V.P.; Han, Y.; Pettigrew, M.J.

    2009-01-01

    Preventing flow-induced vibration and fretting-wear problems in steam generators and heat exchangers requires design specifications that bring together specific guidelines, analysis methods, requirements and appropriate performance criteria. This paper outlines the steps required to generate and support such design specifications for CANDU nuclear steam generators and heat exchangers, and relates them to typical steam-generator design features and computer modeling capabilities. It also describes current issues that are driving changes to flow-induced vibration and fretting-wear specifications that can be applied to the design process for component refurbishment, replacement or new designs. These issues include recent experimental or field evidence for new excitation mechanisms, e.g., the possibility of in-plane fluidelastic instability of U-tubes, the demand for longer reactor and component lifetimes, the need for better predictions of dynamic properties and vibration response, e.g., two-phase random-turbulence excitation, and requirements to consider system 'excursions' or abnormal scenarios, e.g., a main steam line break in the case of steam generators. The paper describes steps being taken to resolve these issues. (author)

  14. Steam feeding redundancy for turbine-drives of feed pumps at WWER-1000 NPP

    International Nuclear Information System (INIS)

    Nesterov, Yu.V.; Shmukler, B.I.

    1987-01-01

    The system of steam supply for feed pump driving turbines (T) at the South Ukrainian Unit 1 according to the centralized redundancy principle is described. T is feeded through the collector of water auxiliary sytem (CWAS) to which steam from the third steam extraction line of turbine is supplied under thenormal regime. Under the reduction of turbine load, live steam from the steam generator is supplied to CWAS through the pressure regulator, possesing 10 s speed of responce. In this case the level reduction in the steam generator makes up 170 mm

  15. 49 CFR 192.369 - Service lines: Connections to cast iron or ductile iron mains.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 3 2010-10-01 2010-10-01 false Service lines: Connections to cast iron or ductile iron mains. 192.369 Section 192.369 Transportation Other Regulations Relating to Transportation... ductile iron mains. (a) Each service line connected to a cast iron or ductile iron main must be connected...

  16. Technological investigations and efficiency analysis of a steam heat exchange condenser: Conceptual design of a hybrid steam condenser

    OpenAIRE

    Kapooria, R K; Kumar, S; Kasana, K S

    2008-01-01

    Most of the electricity being produced throughout the world today is from steam power plants. At the same time, many other competent means of generating electricity have been developed viz. electricity from natural gas, MHD generators, biogas, solar cells, etc. But steam power plants will continue to be competent because of the use of water as the main working fluid which is abundantly available and is also reusable. The condenser remains among one of the key components of a steam power plant...

  17. IAEA activities on steam generator life management

    International Nuclear Information System (INIS)

    Gueorguiev, B.; Lyssakov, V.; Trampus, P.

    2002-01-01

    The International Atomic Energy Agency (IAEA) carries out a set of activities in the field of Nuclear Power Plant (NPP) life management. Main activities within this area are implemented through the Technical Working Group on Life Management of NPPs, and mostly concentrated on studies of understanding mechanisms of degradation and their monitoring, optimisation of maintenance management, economic aspects, proven practices of and approaches to plant life management including decommissioning. The paper covers two ongoing activities related to steam generator life management: the International Database on NPP Steam Generators and the Co-ordinated Research Project on Verification of WWER Steam Generator Tube Integrity (WWER is the Russian designed PWR). The lifetime assessment of main components relies on an ability to assess their condition and predict future degradation trends, which to a large extent is dependent on the availability of relevant data. Effective management of ageing and degradation processes requires a large amount of data. Several years ago the IAEA started to work on the International Database on NPP Life Management. This is a multi-module database consisting of modules such as reactor pressure vessels materials, piping, steam generators, and concrete structures. At present the work on pressure vessel materials, on piping as well as on steam generator is completed. The paper will present the concept and structure of the steam generator module of the database. In countries operating WWER NPPs, there are big differences in the eddy current inspection strategy and practice as well as in the approach to steam generator heat exchanger tube structural integrity assessment. Responding to the need for a co-ordinated research to compare eddy current testing results with destructive testing using pulled out tubes from WWER steam generators, the IAEA launched this project. The main objectives of the project are to summarise the operating experiences of WWER

  18. Unsteady coupling effects of wet steam in steam turbines flows

    International Nuclear Information System (INIS)

    Blondel, Frederic

    2014-01-01

    In addition to conventional turbomachinery problems, both the behavior and performances of steam turbines are highly dependent on the vapour thermodynamic state and the presence of a liquid phase. EDF, the main French electricity producer, is interested in further developing its' modelling capabilities and expertise in this area to allow for operational studies and long-term planning. This PhD thesis explores the modelling of wetness formation and growth in a steam turbine and an analysis of the coupling between the liquid phase and the main flow unsteadiness. To this end, the work in this thesis took the following approach. Wetness was accounted for using a homogeneous model coupled with transport equations to take into account the effects of non-equilibrium phenomena, such as the growth of the liquid phase and nucleation. The real gas attributes of the problem demanded adapted numerical methods. Before their implementation in the 3D elsA solver, the accuracy of the chosen models was tested using a developed one-dimensional nozzle code. In this manner, various condensation models were considered, including both poly-dispersed and monodispersed behaviours of the steam. Finally, unsteady coupling effects were observed from several perspectives (1D, 1D - 3D, 3D), demonstrating the ability of the method of moments to sustain unsteady phenomena which were not apparent in a simple monodispersed model. (author)

  19. Modeling and Simulation of U-tube Steam Generator

    Science.gov (United States)

    Zhang, Mingming; Fu, Zhongguang; Li, Jinyao; Wang, Mingfei

    2018-03-01

    The U-tube natural circulation steam generator was mainly researched with modeling and simulation in this article. The research is based on simuworks system simulation software platform. By analyzing the structural characteristics and the operating principle of U-tube steam generator, there are 14 control volumes in the model, including primary side, secondary side, down channel and steam plenum, etc. The model depends completely on conservation laws, and it is applied to make some simulation tests. The results show that the model is capable of simulating properly the dynamic response of U-tube steam generator.

  20. Open channel steam generator feedwater system

    International Nuclear Information System (INIS)

    Kim, R.F.; Min-Hsiung Hu.

    1985-01-01

    A steam generator which utilizes a primary fluid to vaporize a secondary fluid is provided with an open flow channel and elevated discharge nozzle for the introduction of secondary fluid. The discharge nozzle is positioned above a portion of the inlet line such that the secondary fluid passes through a vertical section of inlet line prior to its discharge into the open channel. (author)

  1. On line instrument systems for monitoring steam turbogenerators

    Science.gov (United States)

    Clapis, A.; Giorgetti, G.; Lapini, G. L.; Benanti, A.; Frigeri, C.; Gadda, E.; Mantino, E.

    A computerized real time data acquisition and data processing for the diagnosis of malfunctioning of steam turbogenerator systems is described. Pressure, vibration and temperature measurements are continuously collected from standard or special sensors including startup or stop events. The architecture of the monitoring system is detailed. Examples of the graphics output are presented. It is shown that such a system allows accurate diagnosis and the possibility of creating a data bank to describe the dynamic characteristics of the machine park.

  2. Analysis methodology for the post-trip return to power steam line break event

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chul Shin; Kim, Chul Woo; You, Hyung Keun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-06-01

    An analysis for Steam Line Break (SLB) events which result in a Return-to-Power (RTP) condition after reactor trip was performed for a postulated Yonggwang Nuclear Power Plant Unit 3 cycle 8. Analysis methodology for post-trip RTP SLB is quite different from that of non-RTP SLB and is more difficult. Therefore, it is necessary to develop a methodology to analyze the response of the NSSS parameters to the post-trip RTP SLB events and the fuel performance after the total reactivity exceeds the criticality. In this analysis, the cases with and without offsite power were simulated crediting 3-D reactivity feedback effect due to a local heatup in the vicinity of stuck CEA and compared with the cases without 3-D reactivity feedback with respect to post-trip fuel performance. Departure-to Nucleate Boiling Ratio (DNBR) and Linear Heat Generation Rate (LHGR). 36 tabs., 32 figs., 11 refs. (Author) .new.

  3. Steam 80 steam generator instrumentation

    International Nuclear Information System (INIS)

    Carson, W.H.; Harris, H.H.

    1980-01-01

    This paper describes two special instrumentation packages in an integral economizer (preheater) steam generator of one of the first System 80 plants scheduled to go into commercial operation. The purpose of the instrumentation is to obtain accurate operating information from regions of the secondary side of the steam generator inaccessible to normal plant instrumentation. In addition to verification of the System 80 steam generator design predictions, the data obtained will assist in verification of steam generator thermal/hydraulic computer codes developed for generic use in the industry

  4. Condition monitoring of steam generator by estimating the overall heat transfer coefficient

    International Nuclear Information System (INIS)

    Furusawa, Hiroaki; Gofuku, Akio

    2013-01-01

    This study develops a technique for monitoring in on-line the state of the steam generator of the fast-breeder reactor (FBR) “Monju”. Because the FBR uses liquid sodium as coolant, it is necessary to handle liquid sodium with caution due to its chemical characteristics. The steam generator generates steam by the heat of secondary sodium coolant. The sodium-water reaction may happen if a pinhole or crack occurs at the thin metal tube wall that separates the secondary sodium coolant and water/steam. Therefore, it is very important to detect an anomaly of the wall of heat transfer tubes at an early stage. This study aims at developing an on-line condition monitoring technique of the steam generator by estimating overall heat transfer coefficient from process signals. This paper describes simplified mathematical models of superheater and evaporator to estimate the overall heat transfer coefficient and a technique to diagnose the state of the steam generator. The applicability of the technique is confirmed by several estimations using simulated process signals with artificial noises. The results of the estimations show that the developed technique can detect the occurrence of an anomaly. (author)

  5. Evaluation of the PTS potential in a WWER-1000 following a steam line break

    International Nuclear Information System (INIS)

    Beghini, M.; D'Auria, F.; Galassi, G.M.; Vitale, E.

    1997-01-01

    A qualified nodalization for WWER-1000 is available at DCMN (Dipartimento di Costruzioni Meccaniche e Nucleari) of University of Pisa that is suitable for running with the thermohydraulic system code Relap5/mod3.2. The nodalization consists of about 1400 hydraulic nodes and more than 5000 mesh points for conduction heat transfer. The four loops of the NPP are separately modelled. Detailed information about the plant hardware has been gotten from contacts with Eastern Organizations in Bulgaria, Russia and Ukraine. The qualification of the nodalization has been achieved at a steady state level utilizing a procedure available at DCMN and at a transient level on the basis of operational (planned) transients performed in the Bulgarian Kozloduy-5 NPP and of the unplanned transient occurred at the Ukrainian Zaporosche NPP (April 1995). Data measured in steam generators have also been utilized. The nodalization has been widely applied to the analysis of accident scenarios in WWER-1000, including Large Break LOCA, Small Break LOCA, ATWS, Loss of Feedwater and Station Blackout. The present activity aims at evaluating the potential for PTS (Pressurized Thermal Shock) following a steam line break accident. The thermalhydraulic results were employed as input for a parametric Fracture Mechanics analysis based on conservative hypothesis of the shape and localization of a pre-existing defect. Stress analysis evidenced the effect of partial cooling of the vessel and gave some general indications of the risk for unstable crack propagation under the simulated PTS conditions. (author). 30 refs, 17 figs, 4 tabs

  6. Evaluation of the PTS potential in a WWER-1000 following a steam line break

    Energy Technology Data Exchange (ETDEWEB)

    Beghini, M; D` Auria, F; Galassi, G M; Vitale, E [Universita degli Studi di Pisa, Dipt. di Costruzioni Meccaniche e Nucleari, Pisa (Italy)

    1997-09-01

    A qualified nodalization for WWER-1000 is available at DCMN (Dipartimento di Costruzioni Meccaniche e Nucleari) of University of Pisa that is suitable for running with the thermohydraulic system code Relap5/mod3.2. The nodalization consists of about 1400 hydraulic nodes and more than 5000 mesh points for conduction heat transfer. The four loops of the NPP are separately modelled. Detailed information about the plant hardware has been gotten from contacts with Eastern Organizations in Bulgaria, Russia and Ukraine. The qualification of the nodalization has been achieved at a steady state level utilizing a procedure available at DCMN and at a transient level on the basis of operational (planned) transients performed in the Bulgarian Kozloduy-5 NPP and of the unplanned transient occurred at the Ukrainian Zaporosche NPP (April 1995). Data measured in steam generators have also been utilized. The nodalization has been widely applied to the analysis of accident scenarios in WWER-1000, including Large Break LOCA, Small Break LOCA, ATWS, Loss of Feedwater and Station Blackout. The present activity aims at evaluating the potential for PTS (Pressurized Thermal Shock) following a steam line break accident. The thermalhydraulic results were employed as input for a parametric Fracture Mechanics analysis based on conservative hypothesis of the shape and localization of a pre-existing defect. Stress analysis evidenced the effect of partial cooling of the vessel and gave some general indications of the risk for unstable crack propagation under the simulated PTS conditions. (author). 30 refs, 17 figs, 4 tabs.

  7. Determination of stresses caused by fluctuation of acoustic load in the steam dryers of a BWR

    International Nuclear Information System (INIS)

    Centeno P, J.; Quezada G, S.; Prieto G, A.; Vazquez R, A.; Espinosa P, G.; Nunez C, A.

    2014-10-01

    The extended power up-rate (EPU) in a nuclear power plant cause various problems in BWR components also in the steam system. This due to increased steam flow generated in the reactor and is conveyed to the turbine by the four main steam lines (MSL). One of the most serious problems is the generation of acoustic pressure loads in the metal structure of the steam dryer which eventually leads to fatigue failure and even the appearance of cracks, and in turn it causes loose parts that are entrained by the steam and transported in the MSL. This problem is due to the fluctuation of load acoustics caused by the union of the safety or relief valves (SRV) with the MSL, spreading through these to reach the reactor pressure vessel (RPV) where the effect of resonance of the acoustic wave is amplified and impacts directly in the supporting structure of the steam dryer, skirt and the panels where the mixture liquid-steam is dried, by centrifugation effect and runoff of liquid water. Efforts in the steam dryer operating conditions of EPU for two cases will be analyzed in this work, the first is before the installation of Acoustic Side Branch (ASB), and in the second case we consider the installation of said ASB in the standpipes of SRV. The analysis was performed with numerical experiments on a platform for computational fluid dynamics with virtual geometries previously designed based on the actual components of the reactor and steam system. The model to study is delimited by the top of the RPV, the steam dryer and a section of each of the four MSL with ten standpipes of SRV. With the obtained data and considering the mechanical-structural properties of the steam dryer material, we can evaluate the mechanical resistance to impacts by acoustic pressure load and its possible deformation or cracking. (Author)

  8. Literature investigation of air/steam ingress through small cracks in concrete wall under pressure differences

    International Nuclear Information System (INIS)

    Jiang, J.T.

    2008-01-01

    Traditionally within CANDU safety analysis, a loss coefficient of ∼2.8 is used to characterize turbulent flow leakage through narrow, sharp-edged cracks into, and out of Steam Protected Rooms (SPRs). In the event of main steam line break (MSLB), the pressure differences observed between SPRs and the surrounding area of the powerhouse range from 0.01kPa to 0.1 kPa. The relatively low pressure differences, coupled with narrow crack sizes, for instance, below 1 mm, may result in laminar flow leakage pathways as opposed to the turbulent variety assumed in analysis. The main purpose of this paper is thus (a) to calculate the loss coefficient for laminar flow through small cracks; and (b) to assess the effect of steam ingress to SPRs when the flow through some or all of the room leakage area is assumed to be laminar. Based on the literature review, the loss coefficient for laminar flow, through 1 mm crack size at 0.1 kPa pressure difference, ranges from 10 to about 65. This value represents an increase in loss coefficient of 3 ∼ 22 times the loss coefficient used for SPR safety analysis. The actual volumetric leakage rate is therefore 3 ∼ 8 times smaller than the amount previously applied. This paper demonstrates how the traditional loss coefficient used in safety analysis is extremely conservative in the analysis of the SPRs steam ingress phenomenon. (author)

  9. Literature investigation of air/steam ingress through small cracks in concrete wall under pressure differences

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, J.T. [McMaster Univ., Engineering Physics Dept., Hamilton, Ontario (Canada)], E-mail: jiangj3@mcmaster.ca

    2008-07-01

    Traditionally within CANDU safety analysis, a loss coefficient of {approx}2.8 is used to characterize turbulent flow leakage through narrow, sharp-edged cracks into, and out of Steam Protected Rooms (SPRs). In the event of main steam line break (MSLB), the pressure differences observed between SPRs and the surrounding area of the powerhouse range from 0.01kPa to 0.1 kPa. The relatively low pressure differences, coupled with narrow crack sizes, for instance, below 1 mm, may result in laminar flow leakage pathways as opposed to the turbulent variety assumed in analysis. The main purpose of this paper is thus (a) to calculate the loss coefficient for laminar flow through small cracks; and (b) to assess the effect of steam ingress to SPRs when the flow through some or all of the room leakage area is assumed to be laminar. Based on the literature review, the loss coefficient for laminar flow, through 1 mm crack size at 0.1 kPa pressure difference, ranges from 10 to about 65. This value represents an increase in loss coefficient of 3 {approx} 22 times the loss coefficient used for SPR safety analysis. The actual volumetric leakage rate is therefore 3 {approx} 8 times smaller than the amount previously applied. This paper demonstrates how the traditional loss coefficient used in safety analysis is extremely conservative in the analysis of the SPRs steam ingress phenomenon. (author)

  10. Surface-enhanced light olefin yields during steam cracking

    NARCIS (Netherlands)

    Golombok, M.; Kornegoor, M.; Brink, van den P.; Dierickx, J.; Grotenbreg, R.

    2000-01-01

    Various papers have shown enhanced olefin yields during steam cracking when a catalytic surface is introduced. Our studies reveal that increased light olefin yields during catalytic steam cracking are mainly due to a surface volume effect and not to a traditional catalytic effect. Augmentation of

  11. Testing improved steam separators in third energy block of Chernobyl AES

    Energy Technology Data Exchange (ETDEWEB)

    Novosel' skiy, O Yu; Karasev, V B; Sakovich, E V; Lyutov, M A; An' kov, V I

    1984-12-01

    Improved steam separating drums are described. These have a four-row arrangement of the pipe unions of the steam-water supply lines, with increased diameter. Two collectors 400 mm in diameter from adjacent separators drums go to a single steam line 600 mm in diameter, which goes to the turbine. Tests of the system were conducted at a pressure of 7 MPa and at thermal power of 65, 83, 93, and 100% of rated value to determine the dependence of the moisture content at the output from the separator drum on the mass level at constant thermal power of the block. The separator drums, 2600 mm in diameter, were found to have a reserve for maximum permissible moisture content, and the moisture content of the steam in the central pipe unions did not exceed 0.02% with a level in the drums 200 mm above the rated level. Thus maintenance of the level above the submerged perforated plate by 100 mm above the planned level permits an increase of the water reserve in the multiple forced circulation circuit by 28 m/sup 3/ and does not hinder an increase of the steam productivity of the block.

  12. Anticipated transient without scram analysis of the simplified boiling water reactor following main steam isolation valve closure with boron injection

    International Nuclear Information System (INIS)

    Khan, H.J.; Cheng, H.S.; Rohatgi, U.S.

    1996-01-01

    The simplified boiling water reactor (SBWR) operating in natural circulation is designed with many passive safety features. An anticipated transient without scram (ATWS) initiated by inadvertent closure of the main steam isolation valve (MSIV) in an SBWR has been analyzed using the RAMONA-4B code of Brookhaven National Laboratory. This analysis demonstrates the predicted performance of the SBWR during an MSIV closure ATWS, followed by shutdown of the reactor through injection of boron into the reactor core from the standby liquid control system

  13. The Creys Malville FBR Super Phenix steam generators

    International Nuclear Information System (INIS)

    Baque, P.; Zuber, T.; Saur, J.M.; Cambillard, E.

    1980-08-01

    After briefly recalling the French experience on sodium steam generators, the authors describe the design concepts of the Superphenix units and give their main characteristics. A short summary of the realized R and D program precedes the description of the four 750-MWt steam generators, the fabrication of which is in progress by Creusot-Loire at Chalon sur Saone (France). The studies started for the next French fast breeder reactors and their steam generators are mentioned

  14. Optimum thermal design of steam pipelines and its impact on environment pollution

    International Nuclear Information System (INIS)

    Abdallah, A.M.; Karameldin, A.

    1999-01-01

    The majority portion of electric power generated production all over the world - produced by conventional and nuclear fuels produced by steam. Moreover, steam is used extensively in electronic, food, seawater desalination, and many other industries. In the last fifty years, little improvements have been made on the thermal efficiency of steam boilers. The major developments have been carried out in the direction of maintaining this efficiency on low-grade fuel and reducing labor and maintenance charges. Because the annual cost of fuel (nuclear and non-nuclear) is often greater than the combined cost of other expenses in steam power plants, greater amount of money can be saved. Designing steam pipelines in such a way that minimizing the total annual cost of pipes can do this. This can be done by optimal design of the total annual cost of the pipe lines, which includes the cost of insulation material, the cost of burned fuel plus the cost of maintenance. To deal with such situation, a case study of a superheated main steam pipeline at Shobrah Elkhema power plant is investigated. A general simplified working formula for calculating the heat transfer coefficient round a tube has been correlated and verified to facilitate the development for of the heat transfer mathematical model together with the steam pipeline total cost algorithm. The total cost algorithm has been optimized and solved by a digital computer program derived specially for this study. Accordingly, the obtained results are presented in a graphical form and analyzed. The results revealed that the optimal steam pipeline insulation must be chosen carefully. The insulation thickness of 0.225 up to 0.235 m, covers the operating time of 10-20 years, and fuel price of 0.125 up to more than 0.2$/kg. The calculated optimized insulation thickness minimizes the emission of sulfur dioxide, nitrogen oxide and carbon dioxide from 375, 75 and 26,778 kg/m/y to less than 7, 1.7 and 590 kg/m.y respectively

  15. An analysis of main processes at small water-into-sodium leaks in the BN-350 and BN-600 NPP steam generators

    International Nuclear Information System (INIS)

    Poplavsky, V.M.

    1990-01-01

    The paper presents the main characteristics of emergency processes at small water-into-sodium leaks that took place during the BN-350 and BN-600 NPP steam generators operation. Leak characteristics are presented, the relationship between such parameters as leak rate and duration, its location in a tube bundle, mass of water ingress into sodium, and the character and size of a failure in the interaction zone is analyzed. (author). 5 refs, 3 figs, 2 tabs

  16. Digital implementation, simulation and tests in MATLAB of the models of Steam line, the turbines, the pressure regulator of a BWR type nucleo electric power plant

    International Nuclear Information System (INIS)

    Lopez R, A.

    2004-01-01

    In this phase of the project they were carried out exhaustive tests to the models of the steam lines, turbines and pressure regulator of a BWR type nucleo electric central for to verify that their tendencies and behaviors are it more real possible. For it, it was necessary to also analyze the transfer functions of the different components along the steam line until the power generator. Such models define alone the dominant poles of the system, what is not limitation to reproduce a wide range of anticipated transitoriness of a power station operation. In the same manner, it was integrated and proved the integrated model form with the models of feeding water of the SUN-RAH, simulating the nuclear reactor starting from predetermined entrances of the prospective values of the vessel. Also it was coupled with the graphic interface developed with the libraries DirectX implementing a specific monitoring panel for this system. (Author)

  17. An Isothermal Steam Expander for an Industrial Steam Supplying System

    Directory of Open Access Journals (Sweden)

    Chen-Kuang Lin

    2015-01-01

    Full Text Available Steam is an essential medium used in the industrial process. To ensure steam quality, small and middle scale boilers are often adopted. However, because a higher steam pressure (compared to the necessary steam pressure is generated, the boiler’s steam pressure will be reduced via a pressure regulator before the steam is directed through the process. Unfortunately, pressure is somewhat wasted during the reducing process. Therefore, in order to promote energy efficiency, a pressure regulator is replaced by a steam expander. With this steam expander, the pressure will be transformed into mechanical energy and extracted during the expansion process. A new type of isothermal steam expander for an industrial steam supplying system will be presented in the paper. The isothermal steam expander will improve the energy efficiency of a traditional steam expander by replacing the isentropic process with an isothermal expansion process. With this, steam condensation will decrease, energy will increase, and steam quality will be improved. Moreover, the mathematical model of the isothermal steam expander will be established by using the Schmidt theory, the same principle used to analyze Stirling engines. Consequently, by verifying the correctness of the theoretical model for the isothermal steam expander using experimental data, a prototype of 100 c.c. isothermal steam expander is constructed.

  18. Comparative analysis of methods for workload assessment of the main control room operators of NPP

    International Nuclear Information System (INIS)

    Georgiev, V.; Petkov, G.

    2008-01-01

    The paper presents benchmarking workload results obtained by a method for operator workload assessment – NASA Task Load Index, and a method for human error probability assessment - Performance Evaluation of Teamwork. Based on the archives of FSS-1000 training in the accident “Main Steam Line Tube Rupture at the WWER-1000 Containment” the capacities of the two methods for direct and indirect workload assessment are evaluated

  19. Wavelet network controller for nuclear steam generators

    International Nuclear Information System (INIS)

    Habibiyan, H; Sayadian, A; Ghafoori-Fard, H

    2005-01-01

    Poor control of steam generator water level is the main cause of unexpected shutdowns in nuclear power plants. Particularly at low powers, it is a difficult task due to shrink and swell phenomena and flow measurement errors. In addition, the steam generator is a highly complex, nonlinear and time-varying system and its parameters vary with operating conditions. Therefore, it seems that design of a suitable controller is a necessary step to enhance plant availability factor. The purpose of this paper is to design, analyze and evaluate a water level controller for U-tube steam generators using wavelet neural networks. Computer simulations show that the proposed controller improves transient response of steam generator water level and demonstrate its superiority to existing controllers

  20. Improving Steam System Performance: A Sourcebook for Industry

    Energy Technology Data Exchange (ETDEWEB)

    2002-06-01

    The sourcebook is a reference for industrial steam system users, outlining opportunities to improve steam system performance. This Sourcebook is designed to provide steam system users with a reference that describes the basic steam system components, outlines opportunities for energy and performance improvements, and discusses the benefits of a systems approach in identifying and implementing these improvement opportunities. The Sourcebook is divided into the following three main sections: Section 1: Steam System Basics--For users unfamiliar with the basics of steam systems, or for users seeking a refresher, a brief discussion of the terms, relationships, and important system design considerations is provided. Users already familiar with industrial steam system operation may want to skip this section. This section describes steam systems using four basic parts: generation, distribution, end use, and recovery. Section 2: Performance Improvement Opportunities--This section discusses important factors that should be considered when industrial facilities seek to improve steam system performance and to lower operating costs. This section also provides an overview of the finance considerations related to steam system improvements. Additionally, this section discusses several resources and tools developed by the U. S. Department of Energy's (DOE) BestPractices Steam Program to identify and assess steam system improvement opportunities. Section 3: Programs, Contacts, and Resources--This section provides a directory of associations and other organizations involved in the steam system marketplace. This section also provides a description of the BestPractices Steam Program, a directory of contacts, and a listing of available resources and tools, such as publications, software, training courses, and videos.

  1. Power generation from a 7700C heat source by means of a main steam cycle, a topping closed gas cycle and a ammonia bottoming cycle

    International Nuclear Information System (INIS)

    Tilliette, Z.P.

    1981-03-01

    For power generation, steam cycles make an efficient use of medium temperature heat sources. They can be adapted to dry cooling, higher power ratings and output increase in winter by addition of an ammonia bottoming cycle. Active development is carried out in this field by 'Electricite de France'. As far as heat sources at higher temperatures are concerned, particularly related to coal-fired or nuclear power plants, a more efficient way of converting energy is at first to expand a hot working fluid through a gas turbine. It is shown in this paper that a satisfactory result, for heat sources of about 770 0 C, is obtained with a topping closed gas cycle of moderate power rating, rejecting its waste heat into the main steam cycle. Attention has to be paid to this gas cycle waste heat recovery and to the coupling of the gas and steam cycles. This concept drastically reduces the importance of new technology components. The use and the significance of an ammonia bottoming cycle in this case are investigated

  2. Numerical simulation in steam injection process by a mechanistic approach

    Energy Technology Data Exchange (ETDEWEB)

    De Souza, J.C.Jr.; Campos, W.; Lopes, D.; Moura, L.S.S. [Petrobras, Rio de Janeiro (Brazil)

    2008-10-15

    Steam injection is a common thermal recovery method used in very viscous oil reservoirs. The method involves the injection of heat to reduce viscosity and mobilize oil. A steam generation and injection system consists primarily of a steam source, distribution lines, injection wells and a discarding tank. In order to optimize injection and improve the oil recovery factor, one must determine the parameters of steam flow such as pressure, temperature and steam quality. This study focused on developing a unified mathematical model by means of a mechanistic approach for two-phase steam flow in pipelines and wells. The hydrodynamic and heat transfer mechanistic model was implemented in a computer simulator to model the parameters of steam injection while trying to avoid the use of empirical correlations. A marching algorithm was used to determine the distribution of pressure and temperature along the pipelines and wellbores. The mathematical model for steam flow in injection systems, developed by a mechanistic approach (VapMec) performed well when the simulated values of pressures and temperatures were compared with the values measured during field tests. The newly developed VapMec model was incorporated in the LinVap-3 simulator that constitutes an engineering supporting tool for steam injection wells operated by Petrobras. 23 refs., 7 tabs., 6 figs.

  3. Crack formation in ferritic screws of main steam isolation valves in German boiling water reactors

    International Nuclear Information System (INIS)

    Steinmill, H.

    1992-01-01

    In connection with crack formations at screws of main steam isolation valves in boiling water reactors, detected for the first time in 1988 in the Federal Republic of Germany, metallographic and fractographic investigations and coating analyses of screw surfaces and crack flanks were performed in order to find out the causes. These and other investigations of damaged screws were accompanied in the years 1989 and 1990 by autoclave tests made in several laboratories. With a view to the mechanical stress of the screws, tightening tests and stress analyses were performed by means of FEM. Repeated autoclave tests were concluded recently by the Stuttgart MPA. Although these tests are not reported here, it can be stated that the results obtained fit in with the overall framework of the results summed up in this report. With regard to the kind of sample stress and the results obtained, two cases have to be distinguished in the autoclave tests discussed in this report. (orig.) [de

  4. Estimating probable flaw distributions in PWR steam generator tubes

    International Nuclear Information System (INIS)

    Gorman, J.A.; Turner, A.P.L.

    1997-01-01

    This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regarding uncertainties and assumptions in the data and analyses

  5. Fail-safety of the EBR-II steam generator system

    International Nuclear Information System (INIS)

    Chopra, P.S.; Stone, C.C.; Hutter, E.; Barney, W.K.; Staker, R.G.

    1976-01-01

    Fail-safe analyses of the EBR-II steam-generator system show that a postulated non-instantaneous leak of water or steam into sodium, through a duplex tube or a tubesheet, at credible leak rates will not structurally damage the evaporators and superheaters. However, contamination of the system and possible shell wastage by sodium-water reaction products may render the system inoperable for a period exceeding six months. This period would be shortened to three months if the system were modified by adding a remotely operated water dump system, a steam vent system, a secondary sodium superheater relief line, and a tubesheet leak-detection system

  6. Improving Steam System Performance: A Sourcebook for Industry, Second Edition

    Energy Technology Data Exchange (ETDEWEB)

    None

    2012-02-23

    This sourcebook is designed to provide steam system users with a reference that describes the basic steam system components, outlines opportunities for energy and performance improvements, and discusses the benefits of a systems approach in identifying and implementing these improvement opportunities. The sourcebook is divided into three main sections: steam system basics, performance improvement opportunities, and where to find help.

  7. Steam generator development in France for the Super Phenix project

    International Nuclear Information System (INIS)

    Robin, M.G.

    1975-01-01

    'Steam Generator Development for Super Phenix Project'. The development program of steam generators studied by Fives-Cail Babcock and Stein Industrie Companies, jointly with CEA end EDF, for the Super Phenix 1200 MWe Fast Breeder Power Plant, is presented. The main characteristics of both sodium heated steam generators are emphasized and experimental studies related to their key features are reported. (author)

  8. Modelling and simulation of the steam line, the high and low pressure turbines and the pressure regulator for the SUN-RAH nucleo electric university simulator

    International Nuclear Information System (INIS)

    Lopez R, A.

    2003-01-01

    In the following article the development of a simulator that allows to represent the dynamics of the following systems: steam line, nozzle, vapor separator, reheater, high pressure turbine, low pressure turbine, power generator and the pressure regulator of a nucleo electric power station. We start from the supposition that this plant will be modeled from a nuclear reactor type BWR (Boiling Water Reactor), using models of reduced order that represent the more important dynamic variables of the physical processes that happen along the steam line until the one generator. To be able to carry out the simulation in real time the Mat lab mathematical modeling software is used, as well as the specific simulation tool Simulink. It is necessary to point out that the platform on which the one is executed the simulator is the Windows operating system, to allow the intuitive use that only this operating system offers. The above-mentioned obeys to that the objective of the simulator it is to help the user to understand some of the dynamic phenomena that are present in the systems of a nuclear plant, and to provide a tool of analysis and measurement of variables to predict the desirable behavior of the same ones. The model of a pressure controller for the steam lines, the high pressure turbine and the low pressure turbine is also presented that it will be the one in charge of regulating the demand of the system according to the characteristics and critic restrictions of safety and control, assigned according to those wanted parameters of performance of this system inside the nucleo electric plant. This simulator is totally well defined and it is part of the University student nucleo electric simulator with Boiling Water Reactor (SUN-RAH), an integral project and of greater capacity. (Author)

  9. Stress analysis of LOFT steam generator blowdown cross-over line

    International Nuclear Information System (INIS)

    Singh, J.N.

    1978-01-01

    The purpose of this report is to demonstrate compliance of the LOFT Steam Generator Blowdown Cross-Over Piping with the ASME Boiler and Pressure Vessel Code, Section III, Subsection NC. Deadweight, thermal expansion, seismic, LOCE, and LOCA loads have been considered. With the addition of two snubbers, as shown in this report, the system conforms to all requirements

  10. Steam distribution and energy delivery optimization using wireless sensors

    Science.gov (United States)

    Olama, Mohammed M.; Allgood, Glenn O.; Kuruganti, Teja P.; Sukumar, Sreenivas R.; Djouadi, Seddik M.; Lake, Joe E.

    2011-05-01

    The Extreme Measurement Communications Center at Oak Ridge National Laboratory (ORNL) explores the deployment of a wireless sensor system with a real-time measurement-based energy efficiency optimization framework in the ORNL campus. With particular focus on the 12-mile long steam distribution network in our campus, we propose an integrated system-level approach to optimize the energy delivery within the steam distribution system. We address the goal of achieving significant energy-saving in steam lines by monitoring and acting on leaking steam valves/traps. Our approach leverages an integrated wireless sensor and real-time monitoring capabilities. We make assessments on the real-time status of the distribution system by mounting acoustic sensors on the steam pipes/traps/valves and observe the state measurements of these sensors. Our assessments are based on analysis of the wireless sensor measurements. We describe Fourier-spectrum based algorithms that interpret acoustic vibration sensor data to characterize flows and classify the steam system status. We are able to present the sensor readings, steam flow, steam trap status and the assessed alerts as an interactive overlay within a web-based Google Earth geographic platform that enables decision makers to take remedial action. We believe our demonstration serves as an instantiation of a platform that extends implementation to include newer modalities to manage water flow, sewage and energy consumption.

  11. 1000 MW steam turbine for Temelin nuclear power station

    International Nuclear Information System (INIS)

    Drahy, J.

    1992-01-01

    Before the end 1991 the delivery was completed of the main parts (3 low-pressure sections and 1 high-pressure section, all of double-flow design) of the first full-speed (3000 r.p.m.) 1000 MW steam turbine for saturated admission steam for the Temelin nuclear power plant. Description of the turbine design and of new technologies and tools used in the manufacture are given. Basic technical parameters of the steam turbine are as follows: maximum output of steam generators 6060 th -1 ; maximum steam flow into turbine 5494.7 th -1 ; output of turbo-set 1024 MW; steam conditions before the turbine inlet: pressure 5.8 MPa, temperature 273.3 degC, steam wetness 0.5%; nominal temperature of cooling water 21 degC; temperature of feed water 220.8 degC; maximum consumption of heat from turbine for heating at 3-stage heating of heating water 60/150 degC. (Z.S.) 7 figs., 2 refs

  12. A steam separator-superheater apparatus

    International Nuclear Information System (INIS)

    Androw, Jean; Bessouat, Roger; Peyrelongue, J.-P.

    1973-01-01

    Description is given of a separator-superheater apparatus comprising an outer enclosure containing a separating-unit and a steam superheating unit according to the main patent. The present addition relates to an improvement in that apparatus, characterized in that the separating unit and the superheating unit, mounted in two distinct portions of the outer enclosure, are divided into the same number of sub-units of each unit being identical and operating in parallel, and in that to each separator sub-unit is associated a superheater sub-unit, said sub-units being mounted in series and located in one in the other of the enclosure two portions, respectively. This can be applied to the treatment of the exhaust steam of a turbine high pressure body, prior to re-injecting said steam into the low pressure body [fr

  13. Steam turbine cycle

    International Nuclear Information System (INIS)

    Okuzumi, Naoaki.

    1994-01-01

    In a steam turbine cycle, steams exhausted from the turbine are extracted, and they are connected to a steam sucking pipe of a steam injector, and a discharge pipe of the steam injector is connected to an inlet of a water turbine. High pressure discharge water is obtained from low pressure steams by utilizing a pressurizing performance of the steam injector and the water turbine is rotated by the high pressure water to generate electric power. This recover and reutilize discharged heat of the steam turbine effectively, thereby enabling to improve heat efficiency of the steam turbine cycle. (T.M.)

  14. Steam generation at Rihand STPP Stage 1

    Energy Technology Data Exchange (ETDEWEB)

    1987-12-01

    The steam generation plant at Rihand in India has two 500 MW boilers. The boilers are of the balanced draught, single cell, radiant furnace type, and are controlled automatically. Cochran Thermax shell type auxillary steam boilers are used for preheating air to the main boilers and for heating fuel oil during storage and pumping. Electrostatic precipitators and ash handling plants are provided to keep dust and ash within limits. 2 figs.

  15. Radioactive gaseous waste processing device

    International Nuclear Information System (INIS)

    Kawabe, Ken-ichi; Uchiyama, Yoshio; Konno, Masanobu; Suzuki, Kunihiko; Kimura, Tadahiro.

    1991-01-01

    A main steam bypass line is disposed to a main steam line of an air extractor for directly sending diluting steams to an exhaust gas line disposed upstream of a preheater not by way of the extractor. Then, a hydrogen flowmeter is disposed to a hydrogen injection line of a hydrogen supply device for measuring the amount of hydrogen to be injected. Further, a control means is disposed to the main steam bypass line for controlling the injection of the diluting steams based on a signal from the hydrogen flowmeter. With such a constitution, the amount of the hydrogen gas supplied from the hydrogen supply device is detected by the hydrogen flowmeter. The control means disposed to the main steam bypass line or the control means disposed directly to the main steam line injects the diluting steams to the exhaust gas line based on the signal from the hydrogen flowmeter. This can reduce the hydrogen concentration in the exhaust gas upstream of the pre-heater to less than an explosive limit, to enable a countermeasure for preventing hydrogen explosion upon hydrogen injection. (T.M.)

  16. Condensation pool experiments with steam using DN200 blowdown pipe

    International Nuclear Information System (INIS)

    Laine, J.; Puustinen, M.

    2005-08-01

    This report summarizes the results of the condensation pool experiments with steam using a DN200 blowdown pipe. Altogether five experiment series, each consisting of several steam blows, were carried out in December 2004 with a scaled-down test facility designed and constructed at Lappeenranta University of Technology. The main purpose of the experiments was to increase the understanding of different phenomena in the condensation pool during steam discharge. (au)

  17. Thermal performances of molten salt steam generator

    International Nuclear Information System (INIS)

    Yuan, Yibo; He, Canming; Lu, Jianfeng; Ding, Jing

    2016-01-01

    Highlights: • Thermal performances of molten salt steam generator were experimentally studied. • Overall heat transfer coefficient reached maximum with optimal molten salt flow rate. • Energy efficiency first rose and then decreased with salt flow rate and temperature. • Optimal molten salt flow rate and temperature existed for good thermal performance. • High inlet water temperature benefited steam generating rate and energy efficiency. - Abstract: Molten salt steam generator is the key technology for thermal energy conversion from high temperature molten salt to steam, and it is used in solar thermal power station and molten salt reactor. A shell and tube type molten salt steam generator was set up, and its thermal performance and heat transfer mechanism were studied. As a coupling heat transfer process, molten salt steam generation is mainly affected by molten salt convective heat transfer and boiling heat transfer, while its energy efficiency is also affected by the heat loss. As molten salt temperature increased, the energy efficiency first rose with the increase of heat flow absorbed by water/steam, and then slightly decreased for large heat loss as the absorbed heat flow still rising. At very high molten salt temperature, the absorbed heat flow decreased as boiling heat transfer coefficient dropping, and then the energy efficiency quickly dropped. As the inlet water temperature increased, the boiling region in the steam generator remarkably expanded, and then the steam generation rate and energy efficiency both rose with the overall heat transfer coefficient increasing. As the molten salt flow rate increased, the wall temperature rose and the boiling heat transfer coefficient first increased and then decreased according to the boiling curve, so the overall heat transfer coefficient first increased and then decreased, and then the steam generation rate and energy efficiency of steam generator both had maxima.

  18. Corrosion aspects in steam generators of nuclear power plants

    International Nuclear Information System (INIS)

    Visoni, E.; Santos Pinto, M. dos

    1988-01-01

    Steam generators of pressurized water reactors (PWR), transfer heat from a primary coolant system to a secondary coolant system. Primary coolant water is heated in the core and passes through the steam generator that transfer heat to the secondary coolant water. However, the steam generator is dead for ionic impurities, corrosion products and fabrication/maintenence residues. These impurities concentrate between crevice and cracks. Many types of degradation mechanisms affect the tubes. The tubes are dented, craked, ovalized, wasted, etc. This paper describes the main corrosion problems in steam generators and includes the corrective actions to considered to reduce or eliminate these corrosion problems. (author) [pt

  19. Steam reforming of commercial ultra-low sulphur diesel

    Energy Technology Data Exchange (ETDEWEB)

    Boon, J.; Van Dijk, E.; De Munck, S.; Van den Brink, R. [Energy research Centre of The Netherlands, ECN Hydrogen and Clean Fossil Fuels, P.O. Box 1, NL1755ZG Petten (Netherlands)

    2011-03-11

    Two main routes for small-scale diesel steam reforming exist: low-temperature pre-reforming followed by well-established methane steam reforming on the one hand and direct steam reforming on the other hand. Tests with commercial catalysts and commercially obtained diesel fuels are presented for both processes. The fuels contained up to 6.5 ppmw sulphur and up to 4.5 vol.% of biomass-derived fatty acid methyl ester (FAME). Pre-reforming sulphur-free diesel at around 475C has been tested with a commercial nickel catalyst for 118 h without observing catalyst deactivation, at steam-to-carbon ratios as low as 2.6. Direct steam reforming at temperatures up to 800C has been tested with a commercial precious metal catalyst for a total of 1190 h with two catalyst batches at steam-to-carbon ratios as low as 2.5. Deactivation was neither observed with lower steam-to-carbon ratios nor for increasing sulphur concentration. The importance of good fuel evaporation and mixing for correct testing of catalysts is illustrated. Diesel containing biodiesel components resulted in poor spray quality, hence poor mixing and evaporation upstream, eventually causing decreasing catalyst performance. The feasibility of direct high temperature steam reforming of commercial low-sulphur diesel has been demonstrated.

  20. Steam reforming of commercial ultra-low sulphur diesel

    Science.gov (United States)

    Boon, Jurriaan; van Dijk, Eric; de Munck, Sander; van den Brink, Ruud

    Two main routes for small-scale diesel steam reforming exist: low-temperature pre-reforming followed by well-established methane steam reforming on the one hand and direct steam reforming on the other hand. Tests with commercial catalysts and commercially obtained diesel fuels are presented for both processes. The fuels contained up to 6.5 ppmw sulphur and up to 4.5 vol.% of biomass-derived fatty acid methyl ester (FAME). Pre-reforming sulphur-free diesel at around 475 °C has been tested with a commercial nickel catalyst for 118 h without observing catalyst deactivation, at steam-to-carbon ratios as low as 2.6. Direct steam reforming at temperatures up to 800 °C has been tested with a commercial precious metal catalyst for a total of 1190 h with two catalyst batches at steam-to-carbon ratios as low as 2.5. Deactivation was neither observed with lower steam-to-carbon ratios nor for increasing sulphur concentration. The importance of good fuel evaporation and mixing for correct testing of catalysts is illustrated. Diesel containing biodiesel components resulted in poor spray quality, hence poor mixing and evaporation upstream, eventually causing decreasing catalyst performance. The feasibility of direct high temperature steam reforming of commercial low-sulphur diesel has been demonstrated.

  1. Enhancement of efficacy of process water monitors in detecting heavy water leak in steam generator blow down lines

    International Nuclear Information System (INIS)

    Mitra, S.R.; Kohale, S.D.; Parida, B.K.; Gathe, G.D.; Pati, C.K.; Mudgal, B.K.; Niraj; Pawar, S.K.

    2006-01-01

    The Steam Generator (SG) serves as an interface between primary and secondary cycle in Pressurized Heavy Water Reactor (PHWR). Failure of steam generator tubes result in leaking of active heavy water in the secondary closed loop. In Tarapur Atomic Power Station-3 and 4 (TAPS- 3 and 4), Scintillator detectors are provided to detect on line heavy water leakages in SG and moderator heat exchangers by monitoring Nitrogen-16 ( 16 N) and Oxygen-19 ( 19 O) activities. Efficacy of detection of these activities at designed detector position on SG blow down line in presence of background radiation field is analysed theoretically. The count rate of 19 O and 16 N estimated at the detector position inside Reactor Building (RB) shows that detectors only respond to very high leak rates due to presence of high ambient radiation level even though sensitivity is appreciably good. For detector position in RB in the accessible areas and out side the RE containment, the travel time for the blow down feed water becomes moderately and very long respectively resulting in poor sensitivity. However the results show that wherever background levels is low, the efficacy of leak detection becomes considerably better than the results obtained when detector is placed inside RB. The study was validated during the reactor operation by recording the detector count rates due to prevalent ambient radiation level near to the detectors. Subsequently the detectors were relocated in an area inside RB where relocation was feasible, travel time of the blow down feed water was moderate and the area had an relatively low ambient radiation level. This paper discusses the methodology adopted during the study and results obtained during theoretical estimation and practical validation. (author)

  2. Recent technology for nuclear steam turbine-generator units

    International Nuclear Information System (INIS)

    Moriya, Shin-ichi; Kuwashima, Hidesumi; Ueno, Takeshi; Ooi, Masao

    1988-01-01

    As the next nuclear power plants subsequent to the present 1,100 MWe plants, the technical development of ABWRs was completed, and the plan for constructing the actual plants is advanced. As for the steam turbine and generator facilities of 1,350 MWe output applied to these plants, the TC6F-52 type steam turbines using 52 in long blades, moisture separation heaters, butterfly type intermediate valves, feed heater drain pumping-up system and other new technologies for increasing the capacity and improving the thermal efficiency were adopted. In this paper, the outline of the main technologies of those and the state of examination when those are applied to the actual plants are described. As to the technical fields of the steam turbine system for ABWRs, the improvement of the total technologies of the plants was promoted, aiming at the good economical efficiency, reliability and thermal efficiency of the whole facilities, not only the main turbines. The basic specification of the steam turbine facilities for 50 Hz ABWR plants and the main new technologies applied to the turbines are shown. The development of 52 in long last stage blades, the development of the analysis program for the coupled vibration of the large rotor system, the development of moisture separation heaters, the turbine control system, condensate and feed water system, and the generators are described. (Kako, I.)

  3. An assessment of underground and aboveground steam system failures in the SRS waste tank farms

    International Nuclear Information System (INIS)

    Hsu, T.C.; Shurrab, M.S.; Wiersma, B.J.

    1997-01-01

    Underground steam system failures in waste tank farms at the Savannah River Site (SRS) increased significantly in the 3--4 year period prior to 1995. The primary safety issues created by the failures were the formation of sub-surface voids in soil and the loss of steam jet transfer and waste evaporation capability, and the loss of heating and ventilation to the tanks. The average annual cost for excavation and repair of the underground steam system was estimated to be several million dollars. These factors prompted engineering personnel to re-consider long-term solutions to the problem. The primary cause of these failures was the inadequate thermal insulation utilized for steam lines associated with older tanks. The failure mechanisms were either pitting or localized general corrosion on the exterior of the pipe beneath the thermal insulation. The most realistic and practical solution is to replace the underground lines by installing aboveground steam systems, although this option will incur significant initial capital costs. Steam system components, installed aboveground in other areas of the tank farms have experienced few failures, while in continuous use. As a result, piecewise installation of temporary aboveground steam systems have been implemented in F-area whenever opportunities, i.e., failures, present themselves

  4. Steam reforming of light oxygenates

    DEFF Research Database (Denmark)

    Trane-Restrup, Rasmus; Resasco, Daniel E; Jensen, Anker Degn

    2013-01-01

    Steam reforming (SR) of ethanol, acetic acid, acetone, acetol, 1-propanol, and propanal has been investigated over Ni/MgAl2O4 at temperatures between 400 and 700 degrees C and at a steam-to-carbon-ratio (S/C) of 6. The yield of H-2 and conversion increased with temperature, while the yield of by-...... of CH4. Significant deactivation of the catalyst was observed for all of the compounds and was mainly due to carbon formation. The carbon formation was highest for alcohols due to a high formation of olefins, which are potent coke precursors....

  5. Corrosion cracking of rotor steels of steam turbines

    International Nuclear Information System (INIS)

    Melekhov, R.K.; Litvintseva, E.N.

    1994-01-01

    Results of investigation of stress corrosion cracking of steam turbine materials in nuclear, fossil and geothermal power plants have been analysed. The role of factors that cause damage to rotor discs, mono block and welding rotors of steam turbines has been shown. These are yield stress and steel composition, stress intensity coefficient and crack growth rate, composition and temperature of the condensed steam and water, electrochemical conditions. The conclusion has been made about the state of stress corrosion cracking of the rotors materials, and main investigation trends which are necessary to solve this problem have been listed

  6. Ash particle erosion on steam boiler convective section

    Energy Technology Data Exchange (ETDEWEB)

    Meuronen, V

    1998-12-31

    In this study, equations for the calculation of erosion wear caused by ash particles on convective heat exchanger tubes of steam boilers are presented. A new, three-dimensional test arrangement was used in the testing of the erosion wear of convective heat exchanger tubes of steam boilers. When using the sleeve-method, three different tube materials and three tube constructions could be tested. New results were obtained from the analyses. The main mechanisms of erosion wear phenomena and erosion wear as a function of collision conditions and material properties have been studied. Properties of fossil fuels have also been presented. When burning solid fuels, such as pulverized coal and peat in steam boilers, most of the ash is entrained by the flue gas in the furnace. In bubbling and circulating fluidized bed boilers, particle concentration in the flue gas is high because of bed material entrained in the flue gas. Hard particles, such as sharp edged quartz crystals, cause erosion wear when colliding on convective heat exchanger tubes and on the rear wall of the steam boiler. The most important ways to reduce erosion wear in steam boilers is to keep the velocity of the flue gas moderate and prevent channelling of the ash flow in a certain part of the cross section of the flue gas channel, especially near the back wall. One can do this by constructing the boiler with the following components. Screen plates can be used to make the velocity and ash flow distributions more even at the cross-section of the channel. Shield plates and plate type constructions in superheaters can also be used. Erosion testing was conducted with three types of tube constructions: a one tube row, an in- line tube bank with six tube rows, and a staggered tube bark with six tube rows. Three flow velocities and two particle concentrations were used in the tests, which were carried out at room temperature. Three particle materials were used: quartz, coal ash and peat ash particles. Mass loss

  7. A Trajectory Correction based on Multi-Step Lining-up for the CLIC Main Linac

    CERN Document Server

    D'Amico, T E

    1999-01-01

    In the CLIC main linac it is very important to minimise the trajectory excursion and consequently the emittance dilution in order to obtain the required luminosity. Several algorithms have been proposed and lately the ballistic method has proved to be very effective. The trajectory method described in this Note retains the main advantages of the latter while adding some interesting features. It is based on the separation of the unknown variables like the quadrupole misalignments, the offset and slope of the injection straight line and the misalignments of the beam position monitors (BPM). This is achieved by referring the trajectory relatively to the injection line and not to the average pre-alignment line and by using two trajectories each corresponding to slightly different quadrupole strengths. A reference straight line is then derived onto which the beam is bent by a kick obtained by moving the first quadrupole. The other quadrupoles are then aligned on that line. The quality of the correction depends mai...

  8. Hydrogen-oxygen steam generator applications for increasing the efficiency, maneuverability and reliability of power production

    Science.gov (United States)

    Schastlivtsev, A. I.; Borzenko, V. I.

    2017-11-01

    The comparative feasibility study of the energy storage technologies showed good applicability of hydrogen-oxygen steam generators (HOSG) based energy storage systems with large-scale hydrogen production. The developed scheme solutions for the use of HOSGs for thermal power (TPP) and nuclear power plants (NPP), and the feasibility analysis that have been carried out have shown that their use makes it possible to increase the maneuverability of steam turbines and provide backup power supply in the event of failure of the main steam generating equipment. The main design solutions for the integration of hydrogen-oxygen steam generators into the main power equipment of TPPs and NPPs, as well as their optimal operation modes, are considered.

  9. Steam generators in indirect-cycle water-cooled reactors

    International Nuclear Information System (INIS)

    Fajeau, M.

    1976-01-01

    In the indirect cycle water-cooled nuclear reactors, the steam generators are placed between the primary circuit and the turbine. They act both as an energy transmitter and as a leaktigh barrier against fission or corrosion products. Their study is thus very important from a performance and reliability point of view. Two main types are presented here: the U-tube and the once-through steam generators [fr

  10. An Improved Steam Injection Model with the Consideration of Steam Override

    OpenAIRE

    He , Congge; Mu , Longxin; Fan , Zifei; Xu , Anzhu; Zeng , Baoquan; Ji , Zhongyuan; Han , Haishui

    2017-01-01

    International audience; The great difference in density between steam and liquid during wet steam injection always results in steam override, that is, steam gathers on the top of the pay zone. In this article, the equation for steam override coefficient was firstly established based on van Lookeren’s steam override theory and then radius of steam zone and hot fluid zone were derived according to a more realistic temperature distribution and an energy balance in the pay zone. On this basis, th...

  11. Steam turbine installations

    International Nuclear Information System (INIS)

    Bainbridge, A.

    1976-01-01

    The object of the arrangement described is to enable raising steam for driving steam turbines in a way suited to operating with liquid metals, such as Na, as heat transfer medium. A preheated water feed, in heat transfer relationship with the liquid metals, is passed through evaporator and superheater stages, and the superheated steam is supplied to the highest pressure stage of the steam turbine arrangement. Steam extracted intermediate the evaporator and superheater stages is employed to provide reheat for the lower pressure stage of the steam turbine. Only a major portion of the preheated water feed may be evaporated and this portion separated and supplied to the superheater stage. The feature of 'steam to steam' reheat avoids a second liquid metal heat transfer and hence represents a simplification. It also reduces the hazard associated with possible steam-liquid metal contact. (U.K.)

  12. Técnica de monitorado continuo (on – line para la evaluación del estado técnico de los turbogrupos de 64 y 100 MW. // Technique of continuous monitored (on - line for the evaluation of the technical state in steam turbine units of 64 and 100 MW.

    Directory of Open Access Journals (Sweden)

    F. de la Torre. Silva

    2001-04-01

    Full Text Available En este trabajo se presenta el resultado del estudio de factibilidad realizado a los turbogrupos de 64 y 100 MW de dosCentrales Termoeléctricas, sobre el empleo de técnicas de monitorado continuo “on line” para la evaluación del estadotécnico de estos turbogrupos.Palabras claves: Turbinas de vapor,vibraciones, monitorado continuo “on line”, diagnóstico.______________________________________________________________________Abstract:In this work an study of feasibility is presented. This study is carried out in steam turbine units of 64 and 100 MW, and show the use ofcontinuous monitored technique (on line for the evaluation of the technical state of these turbine units.Key Words: Steam turbines, vibrations, continuous monitoring on line, turbines supervision, Diagnosis,technical state evaluation.

  13. Design and operating experiences with 50MW steam generator

    International Nuclear Information System (INIS)

    Kawara, M.; Yamaki, H.; Kanamori, A.; Tanaka, K.; Takahashi, T.

    1975-01-01

    The main purpose of the 50 MW steam generator is to have experiences of manufacturing and operation with large scale steam generator including necessary research and development works which can be reflected on the design and fabrication of 'Monju' (Japan 300 MWe prototype LMFBR). The detailed design of the 50 MW steam, generator was begun on March, 1972 and succeeded in the demonstration of 72 hours continuous operation with full power on June, 1974. It has been successfully operated since then, the performances of which have been evaluated through various kinds of tests. In this paper, the following items are mainly discussed system design, thermal and hydraulic design, structure and fabrication and some experiences on testing operation including cleaning and sodium flushing of equipment, sodium level control system, the behavior of hydrogen detection system and general outlook of the performance. (author)

  14. Design and operating experiences with 50MW steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Kawara, M; Yamaki, H; Kanamori, A; Tanaka, K; Takahashi, T

    1975-07-01

    The main purpose of the 50 MW steam generator is to have experiences of manufacturing and operation with large scale steam generator including necessary research and development works which can be reflected on the design and fabrication of 'Monju' (Japan 300 MWe prototype LMFBR). The detailed design of the 50 MW steam, generator was begun on March, 1972 and succeeded in the demonstration of 72 hours continuous operation with full power on June, 1974. It has been successfully operated since then, the performances of which have been evaluated through various kinds of tests. In this paper, the following items are mainly discussed system design, thermal and hydraulic design, structure and fabrication and some experiences on testing operation including cleaning and sodium flushing of equipment, sodium level control system, the behavior of hydrogen detection system and general outlook of the performance. (author)

  15. A double parameters measurement of steam-water two-phase flow with single orifice

    International Nuclear Information System (INIS)

    Zhong Shuoping; Tong Yunxian; Yu Meiying

    1992-08-01

    A double parameters measurement of steam-water two-phase flow with single orifice is described. An on-line measurement device based on micro-computer has been developed. The measured r.m.s error of steam quality is less than 6.5% and the measured relative r.m.s. error of mass flow rate is less than 9%

  16. Large steam turbines for nuclear power stations. Output growth prospects

    International Nuclear Information System (INIS)

    Riollet, G.; Widmer, M.; Tessier, J.

    1975-01-01

    The rapid growth of the output of nuclear reactors, even if temporary settlement occurs, leads the manufacturer to evaluate, at a given time, technological limitations encountered. The problems dealing with the main components of turbines: steam path, rotors and stators steam valves, controle devices, shafts and bearings, are reviewed [fr

  17. AUTOMATIC CONTROL SYSTEM OF THE DRUM BOILER SUPERHEATED STEAM TEMPERATURE.

    Directory of Open Access Journals (Sweden)

    Juravliov A.A.

    2006-04-01

    Full Text Available The control system of the temperature of the superheated steam of the drum boiler is examined. Main features of the system are the PI-controller in the external control loop and introduction of the functional component of the error signal of the external control loop with the negative feedback of the error signal between the prescribed value of steam flowrate and the signal of the steam flowrate in the exit of the boiler in the internal control loop.

  18. Effects of blast wave to main steam piping under high energy line break condition by TNT model

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Hyun; Lee, Eung Seok; Chang, Yoon Suk [Kyung Hee University, Yongin (Korea, Republic of)

    2016-05-15

    The aim of this study is to examine effect of the blast wave according to pipe break position through FE (Finite Element) analyses. If HELB (High Energy Line Break) accident occurs in nuclear power plants, not only environmental effect such as release of radioactive material but also secondary structural defects should be considered. Sudden pipe rupture causes ejection of high temperature and pressure fluid, which acts as a blast wave around the break location. The blast wave caused by the HELB has a possibility to induce structural defects around the components such as safe-related injection pipes and other structures.

  19. Failure analysis of retired steam generator tubings

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hong Pyo; Kim, J. S.; Hwang, S. S. and others

    2005-04-15

    Degradation of steam generator leads to forced outage and extension of outage, which causes increase in repair cost, cost of purchasing replacement power and radiation exposure of workers. Steam generator tube rupture incident occurred in Uljin 4 in 2002, which made public sensitive to nuclear power plant. To keep nuclear energy as a main energy source, integrity of steam generator should be demonstrated. Quantitative relationship between ECT(eddy current test) signal and crack size is needed in assesment of integrity of steam generator in pressurized water reactor. However, it is not fully established for application in industry. Retired steam generator of Kori 1 has many kinds of crack such as circumferential and axial primary water stress corrosion crack and outer diameter stress corrosion crack(ODSCC). So, it can be used in qualifying and improving ECT technology and in condition monitoring assesment for crack detected in ISI(in service inspection). In addition, examination of pulled tube of Kori 1 retired steam generator will give information about effectiveness of non welded sleeving technology which was employed to repair defect tubes and remedial action which was applied to mitigate ODSCC. In this project, hardware such as semi hot lab. for pulled tube examination and modification transportation cask for pulled tube and software such as procedure of transportation of radioactive steam generator tube and non-destructive and destructive examination of pulled tube were established. Non-destructive and destructive examination of pulled tubes from Kori 1 retired steam generator were performed in semi hot lab. Remedial actions applied to Kori 1 retired steam generator, PWSCC trend and bulk water chemistry and crevice chemistry in Kori 1 were evaluated. Electrochemical decontamination technology for pulled tube was developed to reduce radiation exposure and enhance effectiveness of pulled tube examination. Multiparameter algorithm developed at ANL, USA was

  20. Device for achieving pressure balance in the steam generator of a power plant in case of a main-steam pipe or a feedwater pipe break

    International Nuclear Information System (INIS)

    Wietelmann, F.

    1978-01-01

    In order to increase the safety in the steam generator of a power plant in case of a pipe break, the possibility of a pressure balance between the feedwater inlet and the initial steam outlet chambers is allowed for. According to the invention, the partition wall separating these two chambers will exhibit several overflow openings, each of which will be provided with a closure and half of which may be opened to one side only, care having been taken that in case of an accident on occurrence of a certain differential pressure they will always be opened to the low-pressure side. As closures caps, which may be swing out of the way, or rupture diaphragms are mentioned. (UWI) 891 HP [de

  1. Nuclear power plant

    International Nuclear Information System (INIS)

    Nishio, Masahide

    1986-01-01

    Purpose: To provide a constitution capable of previously and reliably preventing radioactivity from releasing into the atmosphere upon occurrence of main steam pipe rupture accidents in a main steam tunnel chamber. Constitution: The outer circumference at the penetration portion of a nuclear reactor container is tightly closed and the main steam tunnel chamber has a tightly closed vessel structure, which is cooled by a local cooler during normal operation. The main steam tunnel chamber is in communication with a pressure control chamber by way of a release line and a releaf valve is disposed at the midway of the release line. Upon occurrence of rupture accident to the main steam pipes in the main steam tunnel chamber, while steams are issued from the ruptured portion, they are discharged through the release line to the suppression chamber and condensated. As a result, excess pressure in the main steam tunnel can be prevented and when the rupture accident is detected, the main steam isolation valve is closed rapidly to interrupt the steam feeding, whereby the steam released from the ruptured pipeways is stopped to avoid the radioactivity release to the atmosphere. (Kamimura, M.)

  2. Scale model test results for an inverted U-tube steam generator with comparisons to heat transfer correlations

    International Nuclear Information System (INIS)

    Boucher, T.J.

    1987-01-01

    To provide data for assessment and development of thermal-hydraulic computer codes, bottom main feedwater-line-break transient simulations were performed in a scale model (Semiscale Mod-2C) of a pressurized water reactor (PWR) with conditions typical of a PWR (15.0 MPa primary pressure, 600 K steam generator inlet plenum fluid temperatures, 6.2 MPa secondary pressure). The state-of-the-art measurements in the scale model (Type III) steam generator allow for the determination of U-tube steam generator allow for the determination of U-tube steam generator secondary component interactions, tube bundle local radial heat transfer, and tube bundle and riser vapor void fractions for steady state and transient operations. To enhance the understanding of the observed phenomena, the component interactions, local heat fluxes, local secondary convective heat transfer coefficients and local vapor void fractions are discussed for steady state, full-power and transient operations. Comparisons between the measurement-derived secondary convective heat transfer coefficients and those predicted by a number of correlations, including the Chen correlation currently used in thermal-hydraulic computer codes, show that none of the correlations adequately predict the data and points out the need for the formulation of a new correlation based on this experimental data. The unique information presented herein should be of the interest to anyone involved in modeling inverted U-tube steam generator thermal-hydraulics for forced convection boiling/vaporization heat transfer. 5 refs., 13 figs., 1 tab

  3. Manpower development for safe operation of nuclear power plant. China. Main steam bypass system operation and maintenance. Task: 6.1.6. Technical report

    International Nuclear Information System (INIS)

    Stubley, P.H.

    1994-01-01

    This mission concentrated on the Steam Bypass system of Qinshan Nuclear Power Plant. The system had experienced spurious opening of the bypass valves, disrupting the steam pressure control and the steam generator level control system. A series of commissioning type tests were defined which should allow the operators to revise the setpoints used in the control of the bypass system, and thus prevent spurious opening while maintaining the desired steam pressure control during power maneuvering. Training also included giving experience from other operating plants on aspects of steam and feedwater systems and components, especially as this experience affected maintenance or gave rise to problems. Steam generated maintenance experience is especially applicable, and a future mission is planned for an expert in this field. In addition other aspects of the Chinese nuclear program was assessed to guide future missions. This included assessment of operating procedures from an availability point of view

  4. Cogeneration steam turbines from Siemens: New solutions

    Science.gov (United States)

    Kasilov, V. F.; Kholodkov, S. V.

    2017-03-01

    The Enhanced Platform system intended for the design and manufacture of Siemens AG turbines is presented. It combines organizational and production measures allowing the production of various types of steam-turbine units with a power of up to 250 MWel from standard components. The Enhanced Platform designs feature higher efficiency, improved reliability, better flexibility, longer overhaul intervals, and lower production costs. The design features of SST-700 and SST-900 steam turbines are outlined. The SST-700 turbine is used in backpressure steam-turbine units (STU) or as a high-pressure cylinder in a two-cylinder condensing turbine with steam reheat. The design of an SST-700 single-cylinder turbine with a casing without horizontal split featuring better flexibility of the turbine unit is presented. An SST-900 turbine can be used as a combined IP and LP cylinder (IPLPC) in steam-turbine or combined-cycle power units with steam reheat. The arrangements of a turbine unit based on a combination of SST-700 and SST-900 turbines or SST-500 and SST-800 turbines are presented. Examples of this combination include, respectively, PGU-410 combinedcycle units (CCU) with a condensing turbine and PGU-420 CCUs with a cogeneration turbine. The main equipment items of a PGU-410 CCU comprise an SGT5-4000F gas-turbine unit (GTU) and STU consisting of SST-700 and SST-900RH steam turbines. The steam-turbine section of a PGU-420 cogeneration power unit has a single-shaft turbine unit with two SST-800 turbines and one SST-500 turbine giving a power output of N el. STU = 150 MW under condensing conditions.

  5. Thermo hydrodynamical analyses of steam generator of nuclear power plant

    International Nuclear Information System (INIS)

    Petelin, S.; Gregoric, M.

    1984-01-01

    SMUP computer code for stationary model of a U-tube steam generator of a PWR nuclear power plant was developed. feed water flow can enter through main and auxiliary path. The computer code is based on the one dimensional mathematical model. Among the results that give an insight into physical processes along the tubes of steam generator are distribution of temperatures, water qualities, heat transfer rates. Parametric analysis permits conclusion on advantage of each design solution regarding heat transfer effects and safety of steam generator. (author)

  6. Primary collector wall local temperature fluctuations in the area of water-steam phase boundary

    Energy Technology Data Exchange (ETDEWEB)

    Matal, O.; Klinga, J.; Simo, T. [Energovyzkum Ltd., Brno (Switzerland)

    1995-12-31

    A limited number of temperature sensors could be installed at the primary collector surface in the area of water - steam phase boundary. The surface temperatures as well WWER 440 steam generator process data were measured and stored for a long time and off-line evaluated. Selected results are presented in the paper. (orig.). 2 refs.

  7. Primary collector wall local temperature fluctuations in the area of water-steam phase boundary

    Energy Technology Data Exchange (ETDEWEB)

    Matal, O; Klinga, J; Simo, T [Energovyzkum Ltd., Brno (Switzerland)

    1996-12-31

    A limited number of temperature sensors could be installed at the primary collector surface in the area of water - steam phase boundary. The surface temperatures as well WWER 440 steam generator process data were measured and stored for a long time and off-line evaluated. Selected results are presented in the paper. (orig.). 2 refs.

  8. Steam Digest 2002

    Energy Technology Data Exchange (ETDEWEB)

    2003-11-01

    Steam Digest 2002 is a collection of articles published in the last year on steam system efficiency. DOE directly or indirectly facilitated the publication of the articles through it's BestPractices Steam effort. Steam Digest 2002 provides a variety of operational, design, marketing, and program and program assessment observations. Plant managers, engineers, and other plant operations personnel can refer to the information to improve industrial steam system management, efficiency, and performance.

  9. Thermo-mechanical lifetime assessment of components for 700 °C steam turbine applications

    International Nuclear Information System (INIS)

    Ehrhardt, F.

    2014-01-01

    ranges and hold periods. The effectiveness of the deformation model was verified in a Thermo-Mechanical Fatigue (TMF) experiment. The cycle definition was determined from a lifetime-limiting location in a 700°C high-pressure steam turbine rotor. The CF lifetime assessment methodology for Alloy 617 was verified with the help of post-test microstructural investigations. CF experiments with representative strain ranges and sufficiently long hold durations are to be performed to further investigate the apparent CF damage mechanism. The post-test metallographic investigation on the TMF test has revealed extensive creep damage development. The second main objective of this work was the identification of the maximum application temperatures of dissimilar metal welded (DMW) joints between the Ni-base Alloy 617 and a 1% CrMoV low alloy bainitic rotor steel as well as a higher-alloyed 10% Cr martensitic rotor steel. A testing procedure for the static creep and cyclic testing of the DMW joints was developed. The fatigue and CF endurance characteristics have been determined as well as the creep rupture (CR) strength curves, whereas CR strength extrapolation was only possible up to maximum rupture times of 30 kh. Post-test metallographic investigations revealed Type IV creep damage in the fine grained heat affected zone of the steels. Fusion line cracking associated with a Type I precipitate condition at the interface between Alloy 617 weld metal and the steel heat-affected zone was observed in the investigated Alloy 617 -1% CrMoV and Alloy 617 - 10% Cr DMW joints. Whereas the Type I carbide layer was apparent in Alloy 617 - 1% CrMoV DMW joint specimens even after short duration high temperature exposure, the situation was not clearly visible within the Alloy 617 - 10% Cr DMW joint. One major reason for the development of these Type I precipitates are the cross fusion line chemical composition gradients, mainly for the constituents C and Cr, which are obviously larger for the low

  10. Thermo-mechanical lifetime assessment of components for 700 °C steam turbine applications

    Energy Technology Data Exchange (ETDEWEB)

    Ehrhardt, F.

    2014-07-01

    ranges and hold periods. The effectiveness of the deformation model was verified in a Thermo-Mechanical Fatigue (TMF) experiment. The cycle definition was determined from a lifetime-limiting location in a 700°C high-pressure steam turbine rotor. The CF lifetime assessment methodology for Alloy 617 was verified with the help of post-test microstructural investigations. CF experiments with representative strain ranges and sufficiently long hold durations are to be performed to further investigate the apparent CF damage mechanism. The post-test metallographic investigation on the TMF test has revealed extensive creep damage development. The second main objective of this work was the identification of the maximum application temperatures of dissimilar metal welded (DMW) joints between the Ni-base Alloy 617 and a 1% CrMoV low alloy bainitic rotor steel as well as a higher-alloyed 10% Cr martensitic rotor steel. A testing procedure for the static creep and cyclic testing of the DMW joints was developed. The fatigue and CF endurance characteristics have been determined as well as the creep rupture (CR) strength curves, whereas CR strength extrapolation was only possible up to maximum rupture times of 30 kh. Post-test metallographic investigations revealed Type IV creep damage in the fine grained heat affected zone of the steels. Fusion line cracking associated with a Type I precipitate condition at the interface between Alloy 617 weld metal and the steel heat-affected zone was observed in the investigated Alloy 617 -1% CrMoV and Alloy 617 - 10% Cr DMW joints. Whereas the Type I carbide layer was apparent in Alloy 617 - 1% CrMoV DMW joint specimens even after short duration high temperature exposure, the situation was not clearly visible within the Alloy 617 - 10% Cr DMW joint. One major reason for the development of these Type I precipitates are the cross fusion line chemical composition gradients, mainly for the constituents C and Cr, which are obviously larger for the low

  11. Steam-Generator Integrity Program/Steam-Generator Group Project

    International Nuclear Information System (INIS)

    1982-10-01

    The Steam Generator Integrity Program (SGIP) is a comprehensive effort addressing issues of nondestructive test (NDT) reliability, inservice inspection (ISI) requirements, and tube plugging criteria for PWR steam generators. In addition, the program has interactive research tasks relating primary side decontamination, secondary side cleaning, and proposed repair techniques to nondestructive inspectability and primary system integrity. The program has acquired a service degraded PWR steam generator for research purposes. This past year a research facility, the Steam Generator Examination Facility (SGEF), specifically designed for nondestructive and destructive examination tasks of the SGIP was completed. The Surry generator previously transported to the Hanford Reservation was then inserted into the SGEF. Nondestructive characterization of the generator from both primary and secondary sides has been initiated. Decontamination of the channelhead cold leg side was conducted. Radioactive field maps were established in the steam generator, at the generator surface and in the SGEF

  12. Technical development and its application on steam generator replacement

    International Nuclear Information System (INIS)

    Morita, Sadahiko; Hanzawa, Katsumi; Sato, Hajime; Kannoto, Yasuo.

    1995-01-01

    Twenty-two PWR nuclear power plants are now under commercial operation in Japan. Eight of these plants are scheduled to have their steam generators replaced by up-graded units as a social responsibility for improved reliability, economy and easier maintenance. To carry out steam generator replacement, main coolant pipe cutting and restoration techniques, remote controlled welding machines and other remote controlled equipment, templating techniques with which the new steam generator primary nozzles will fit the existing primary pipes correctly were developed. An adequate training program was carried out to establish these techniques and they were then applied in replacement work on site. The steam generators of the three plants were replaced completely in 1994. These newly developed techniques are to be applied in upcoming plants and replaced plants will be much reliable. (author)

  13. Experience in adjusting of the level regulation system of steam generators of the Rovno NPP

    International Nuclear Information System (INIS)

    Patselyuk, S.N.; Sokolov, A.G.; Kazakov, V.I.; Dorosh, Yu.A.

    1984-01-01

    A system of feed water level control in steam generators at the Rovno NPP with WWER-440 reactors which comprises start-up as well as main regulators is described. The start-up regulator (single-pulsed with a signal by the level) keeps the level in the steam generator at loadings up to 30% of the nominal reactor power Nsub(nom.) The main regulator is connected in the three-pulsed circuit and it receives signals by steam and water flow rate and by the level in the steam generator. The main regulator has been started only at loadings above 40% Nsub(nom.). After reconstruction it was used in the 15-100% Nsub(nom.) range. Characteristics of the level control system in the steam generator at perturbations intoduced by the main circulating pump (MCP) and turbine disconnection as well as change in feed water flow rate have been studied. The studies have revealed that the system ensures necessary quality of control in stationary modes. The system operates stably at perturbations of feed water flow rate up to 50% Nsub(nom.). Perturbations by MCP connections and disconnections is most difficult for control system

  14. Steam drums

    International Nuclear Information System (INIS)

    Crowder, R.

    1978-01-01

    Steam drums are described that are suitable for use in steam generating heavy water reactor power stations. They receive a steam/water mixture via riser headers from the reactor core and provide by means of separators and driers steam with typically 0.5% moisture content for driving turbines. The drums are constructed as prestressed concrete pressure vessels in which the failure of one or a few of the prestressing elements does not significantly affect the overall strength of the structure. The concrete also acts as a radiation shield. (U.K.)

  15. Thermodynamic analysis of heat recovery steam generator in combined cycle power plant

    Directory of Open Access Journals (Sweden)

    Ravi Kumar Naradasu

    2007-01-01

    Full Text Available Combined cycle power plants play an important role in the present energy sector. The main challenge in designing a combined cycle power plant is proper utilization of gas turbine exhaust heat in the steam cycle in order to achieve optimum steam turbine output. Most of the combined cycle developers focused on the gas turbine output and neglected the role of the heat recovery steam generator which strongly affects the overall performance of the combined cycle power plant. The present paper is aimed at optimal utilization of the flue gas recovery heat with different heat recovery steam generator configurations of single pressure and dual pressure. The combined cycle efficiency with different heat recovery steam generator configurations have been analyzed parametrically by using first law and second law of thermodynamics. It is observed that in the dual cycle high pressure steam turbine pressure must be high and low pressure steam turbine pressure must be low for better heat recovery from heat recovery steam generator.

  16. Energetic and exergetic analysis of a steam turbine power plant in an existing phosphoric acid factory

    International Nuclear Information System (INIS)

    Hafdhi, Fathia; Khir, Tahar; Ben Yahyia, Ali; Ben Brahim, Ammar

    2015-01-01

    Highlights: • The operating mode of the factory and the power supply streams are presented. • Energetic Analysis of steam turbine power plant of an existing phosphoric acid factory. • Exergetic Analysis of each component of steam turbine power plant and the different heat recovery system. • Energy, exergy efficiency and irreversibility rates for the main components are determined. • The effect of the operating parameters on the plant performance are analyzed. - Abstract: An energetic and exergetic analysis is conducted on a Steam Turbine Power Plant of an existing Phosphoric Acid Factory. The heat recovery systems used in the different parts of the plant are also considered in the study. Mass, energy and exergy balances are established on the main compounds of the plant. A numerical code is established using EES software to perform the calculations required for the thermal and exergy plant analysis considering real variation ranges of the main operating parameters such as pressure, temperature and mass flow rate. The effects of theses parameters on the system performances are investigated. The main sources of irreversibility are the melters, followed by the heat exchangers, the steam turbine generator and the pumps. The maximum energy efficiency is obtained for the blower followed by the heat exchangers, the deaerator and the steam turbine generator. The exergy efficiency obtained for the heat exchanger, the steam turbine generator, the deaerator and the blower are 88%, 74%, 72% and 66% respectively. The effects of High Pressure steam temperature and pressure on the steam turbine generator energy and exergy efficiencies are investigated.

  17. North Anna Power Station - Unit 1: Overview of steam generator replacement project activities

    International Nuclear Information System (INIS)

    Gettler, M.W.; Bayer, R.K.; Lippard, D.W.

    1993-01-01

    The original steam generators at Virginia Electric and Power Company's (Virginia Power) North Anna Power Station (NAPS) Unit 1 have experienced corrosion-related degradation that require periodic inspection and plugging of steam generator tubes to ensure their continued safe and reliable operation. Despite improvements in secondary water chemistry, continued tube degradation in the steam generators necessitated the removal from service of approximately 20.3 percent of the tubes by plugging, (18.6, 17.3, and 25.1 for steam generators A, B, and C, respectively). Additionally, the unit power was limited to 95 % during, its last cycle of operation. Projections of industry and Virginia Power experience indicated the possibility of mid-cycle inspections and reductions in unit power. Therefore, economic considerations led to the decision to repair the steam generators (i.e., replace the steam generator lower assemblies). Three new Model 51F Steam Generator lower assembly units were ordered from Westinghouse. Virginia Power contracted Bechtel Power Corporation to provide the engineering and construction support to repair the Unit 1 steam generators. On January 4, 1993, after an extended coastdown period, North Anna Unit 1 was brought off-line and the 110 day (breaker-to-breaker) Steam Generator Replacement Project (SGRP) outage began. As of this paper, the outage is still in progress

  18. An Improved Steam Injection Model with the Consideration of Steam Override

    Directory of Open Access Journals (Sweden)

    He Congge

    2017-01-01

    Full Text Available The great difference in density between steam and liquid during wet steam injection always results in steam override, that is, steam gathers on the top of the pay zone. In this article, the equation for steam override coefficient was firstly established based on van Lookeren’s steam override theory and then radius of steam zone and hot fluid zone were derived according to a more realistic temperature distribution and an energy balance in the pay zone. On this basis, the equation for the reservoir heat efficiency with the consideration of steam override was developed. Next, predicted results of the new model were compared with these of another analytical model and CMG STARS (a mature commercial reservoir numerical simulator to verify the accuracy of the new mathematical model. Finally, based on the validated model, we analyzed the effects of injection rate, steam quality and reservoir thickness on the reservoir heat efficiency. The results show that the new model can be simplified to the classic model (Marx-Langenheim model under the condition of the steam override being not taken into account, which means the Marx-Langenheim model is corresponding to a special case of this new model. The new model is much closer to the actual situation compared to the Marx-Langenheim model because of considering steam override. Moreover, with the help of the new model, it is found that the reservoir heat efficiency is not much affected by injection rate and steam quality but significantly influenced by reservoir thickness, and to ensure that the reservoir can be heated effectively, the reservoir thickness should not be too small.

  19. Application of the coupled Relap5/Panther codes for PWR steam. Line break accident analysis

    International Nuclear Information System (INIS)

    Guisset, J.-P.; Bosso, S.; Charlier, A.; Delhaye, X.; Ergo, O.; Ouliddren, K.; Schneidesch, C.; Zhang, J.

    2001-01-01

    A dynamic coupling between the existing 1-dimensional thermal-hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER is applied via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel thermal-hydraulic analysis code COBRA 3C allows direct evaluation of the Departure from Nucleate Boiling Ratio in parallel with the coupled PANTHER/RELAP5 simulation. The coupled codes are applied to develop a Final Safety Analysis Report (FSAR) accident analysis methodology for the major Steam Line Break (SLB) accident at hot zero power in a typical three-loop pressurised water reactor. In this methodology, the uncertainties related to the plant, core thermal-hydraulic and neutronic parameters are combined in a deterministic bounding approach based on sensitivity studies. The results of coupled thermal-hydraulic and neutronic analysis of SLB are presented and discussed. It is shown that there exists an important margin in the traditional FSAR accident analysis for SLB, which can be attributed by the conservatism's introduced by de-coupling the plant sub-systems. (author)

  20. Application of the coupled Relap5/Panther codes for PWR steam. Line break accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Guisset, J.-P.; Bosso, S.; Charlier, A.; Delhaye, X.; Ergo, O.; Ouliddren, K.; Schneidesch, C.; Zhang, J. [Tractebel Energy Engineering, Brussels (Belgium)

    2001-07-01

    A dynamic coupling between the existing 1-dimensional thermal-hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER is applied via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel thermal-hydraulic analysis code COBRA 3C allows direct evaluation of the Departure from Nucleate Boiling Ratio in parallel with the coupled PANTHER/RELAP5 simulation. The coupled codes are applied to develop a Final Safety Analysis Report (FSAR) accident analysis methodology for the major Steam Line Break (SLB) accident at hot zero power in a typical three-loop pressurised water reactor. In this methodology, the uncertainties related to the plant, core thermal-hydraulic and neutronic parameters are combined in a deterministic bounding approach based on sensitivity studies. The results of coupled thermal-hydraulic and neutronic analysis of SLB are presented and discussed. It is shown that there exists an important margin in the traditional FSAR accident analysis for SLB, which can be attributed by the conservatism's introduced by de-coupling the plant sub-systems. (author)

  1. Steam generator blowdown heat exchangers degradations operational experience on EDF French NPP fleet

    International Nuclear Information System (INIS)

    Praud, M.; Doyen, F.; Wintergest, M.; Jourdain, W.; Roussillon, M.; Zidane, A.; Mayos, M.

    2015-01-01

    The main function of the Steam Generator Blowdown System (SGBS) is to purify the secondary fluid from all kinds of pollutions: corrosion products from the secondary system, consequences of raw water pollutions through condenser's leakage, potential radiochemical pollutions resulting from Primary-to-Secondary leaks. The topic of this paper is to present the main SGBS dysfunctions linked to the degradation of the tubular heat exchangers, which sometimes can lead to integrity failure, through corrosion phenomenon. The degradation mechanisms have been characterized by various visual inspections and destructive examinations performed on pulled tubes and bundles. It appears that SGBS tubes suffer two main forms of corrosion. First, for the non-regenerative heat exchangers, where external surface of tubes is exposed to intermediate fluid, alkaline corrosion under tube sheet or shell-side baffles may occur. Caustic attack results from Na 3 PO 4 decomposition by thermal or chemical process. Secondly, mainly for regenerative heat exchangers, pitting and under-deposits corrosion linked to lay-up conditions during outages. This kind of attack is the root cause of a potential 'domino effect': a steam jet from the leaking tube can induce mechanical and/or erosion on many tubes located in its vicinity. Concerning the external degradation by caustic corrosion, only design modifications and strong monitoring of the raw water inlet may able to limit the occurrence of tube perforation. The lay-up guidelines should be carefully followed to mitigate internal corrosion: a controlled atmosphere (limited humidity) and cleanliness of the tube (avoiding deposits formation on the bottom line) seem to be the main parameters

  2. Device for extracting steam or gas from the primary coolant line leading from a reactor pressure vessel to a straight through boiler or from the top primary boiler chamber of a water-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Schatz, K.

    1982-01-01

    In such a nuclear reactor, a steam or gas cushion can form when the primary system is refilled, which can cause blocking of the natural circulation or filling of the system in the area of the hot primary coolant pipe or in the top primary boiler chamber. In order to remove such a steam or gas cushion, a ventilation pipe starting from the bend of the primary coolant line is connected to the feed pipe for introducing water into the primary system. The feed pipe is designed on the principle of the vacuum pump in the area of the opening of the ventilation pipe. There is a sub-pressure in the ventilation pipe, which makes it possible to extract the steam or gas. After mixing in the area of the opening, the steam condenses or is distributed with the gas in the primary coolant. (orig.) [de

  3. Linear Dynamics Model for Steam Cooled Fast Power Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vollmer, H

    1968-04-15

    A linear analytical dynamic model is developed for steam cooled fast power reactors. All main components of such a plant are investigated on a general though relatively simple basis. The model is distributed in those parts concerning the core but lumped as to the external plant components. Coolant is considered as compressible and treated by the actual steam law. Combined use of analogue and digital computer seems most attractive.

  4. Analysis of experimental characteristics of multistage steam-jet electors of steam turbines

    Science.gov (United States)

    Aronson, K. E.; Ryabchikov, A. Yu.; Brodov, Yu. M.; Brezgin, D. V.; Zhelonkin, N. V.; Murmanskii, I. B.

    2017-02-01

    A series of questions for specification of physical gas dynamics model in flow range of steam-jet unit and ejector computation methodology, as well as functioning peculiarities of intercoolers, was formulated based on analysis of experimental characteristics of multistage team-jet steam turbines. It was established that coefficient defining position of critical cross-section of injected flow depends on characteristics of the "sound tube" zone. Speed of injected flow within this tube may exceed that of sound, and pressure jumps in work-steam decrease at the same time. Characteristics of the "sound tube" define optimal axial sizes of the ejector. According to measurement results, the part of steam condensing in the first-stage coolant constitutes 70-80% of steam amount supplied into coolant and is almost independent of air content in steam. Coolant efficiency depends on steam pressure defined by operation of steam-jet unit of ejector of the next stage after coolant of steam-jet stage, temperature, and condensing water flow. As a rule, steam entering content of steam-air mixture supplied to coolant is overheated with respect to saturation temperature of steam in the mixture. This should be taken into account during coolant computation. Long-term operation causes changes in roughness of walls of the ejector's mixing chamber. The influence of change of wall roughness on ejector characteristic is similar to the influence of reverse pressure of the steam-jet stage. Until some roughness value, injection coefficient of the ejector stage operating in superlimiting regime hardly changed. After reaching critical roughness, the ejector switches to prelimiting operating regime.

  5. Steam sterilization does not require saturated steam

    NARCIS (Netherlands)

    van Doornmalen Gomez Hoyos, J. P.C.M.; Paunovic, A.; Kopinga, K.

    2017-01-01

    The most commonly applied method to sterilize re-usable medical devices in hospitals is steam sterilization. The essential conditions for steam sterilization are derived from sterilization in water. Microbiological experiments in aqueous solutions have been used to calculate various time–temperature

  6. Experiment on the Influence Factors of Steam Distillation Rate of Crude Oil in Porous Media

    Directory of Open Access Journals (Sweden)

    Tian Guoqing

    2017-01-01

    Full Text Available To explore the influence of complexity of reservoir properties in porous media and the diversity of operating conditions on the steam distillation rate of crude oil in the process of heavy oil exploitation with steam injection, steam distillation simulation devices are used to study steam distillation rate of crude oil in porous media. Then steam distillation ratio is obtained under the condition of different core permeability, oil saturation, steam temperatures, system pressure, steam injection rates and steam distillation rates with different viscosities of crude oil. The results show that the steam distillation rate of crude oil in porous media depends mainly on the nature of the crude oil itself, for temperature and pressure are the key factors compared with the pore structure, the initial oil saturation and steam injection rate. The experimental results help estimate the amount of crude oil and the required steam in the reservoir in the steam drive process, aiming to facilitate the optimization design and operation of steam drive.

  7. Evaluation on mechanical and corrosion properties of steam generator tubing materials

    International Nuclear Information System (INIS)

    Kim, In Sup; Lee, Byong Whi; Lee, Sang Kyu; Lee, Young Ho; Kim, Jun Whan; Lee, Ju Seok; Kwon, Hyuk Sang; Kim, Su Jung

    1998-06-01

    Steam generator is one of the major components of nuclear reactor pressure boundary. It's main function os transferring heat which generated in the reactor to turbine generator through steam generator tube. In these days, steam generator tubing materials of operating plant are used Inconel 600 alloys. But according to the operation time, there are many degradation phenomena which included mechanical damage due to flow induced vibration and corrosion damage due to PWSCC, IGA/SCC and pitting etc. Recently Inconel 690 alloys are selected as new and replacement steam generator tubes for domestic nuclear power plant. But there are few study about mechanical and corrosion properties of Inconel 600 and 690. The objectives of this study is to evaluate and compare mechanical and corrosion propertied of steam generator tube materials

  8. Evaluation on mechanical and corrosion properties of steam generator tubing materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, In Sup; Lee, Byong Whi; Lee, Sang Kyu; Lee, Young Ho; Kim, Jun Whan; Lee, Ju Seok; Kwon, Hyuk Sang; Kim, Su Jung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-06-15

    Steam generator is one of the major components of nuclear reactor pressure boundary. It's main function os transferring heat which generated in the reactor to turbine generator through steam generator tube. In these days, steam generator tubing materials of operating plant are used Inconel 600 alloys. But according to the operation time, there are many degradation phenomena which included mechanical damage due to flow induced vibration and corrosion damage due to PWSCC, IGA/SCC and pitting etc. Recently Inconel 690 alloys are selected as new and replacement steam generator tubes for domestic nuclear power plant. But there are few study about mechanical and corrosion properties of Inconel 600 and 690. The objectives of this study is to evaluate and compare mechanical and corrosion propertied of steam generator tube materials.

  9. Calculation model for 16N transit time in the secondary side of steam generators

    International Nuclear Information System (INIS)

    Liu Songyu; Xu Jijun; Xu Ming

    1998-01-01

    The 16 N transit time is essential to determine the leak-rate of steam generator tubes leaks with 16 N monitoring system, which is a new technique. A model was developed for calculation 16 N transit time in the secondary side of steam generators. According to the flow characters of secondary side fluid, the transit times divide into four sectors from tube sheet to the sensor on steam line. The model assumes that 16 N is moving as vapor phase in the secondary-side. So the model for vapor velocity distribution in tube bundle is presented in detail. The 16 N transit time calculation results of this model compare with these of EDF on steam generator of Qinshan NPP

  10. Pressurizer and steam-generator behavior under PWR transient conditions

    International Nuclear Information System (INIS)

    Wahba, A.B.; Berta, V.T.; Pointner, W.

    1983-01-01

    Experiments have been conducted in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR), at the Idaho National Engineering Laboratory, in which transient phenomena arising from accident events with and without reactor scram were studied. The main purpose of the LOFT facility is to provide data for the development of computer codes for PWR transient analyses. Significant thermal-hydraulic differences have been observed between the measured and calculated results for those transients in which the pressurizer and steam generator strongly influence the dominant transient phenomena. Pressurizer and steam generator phenomena that occurred during four specific PWR transients in the LOFT facility are discussed. Two transients were accompanied by pressurizer inflow and a reduction of the heat transfer in the steam generator to a very small value. The other two transients were accompanied by pressurizer outflow while the steam generator behavior was controlled

  11. Control scheme for direct steam generation in parabolic troughs under recirculation operation mode

    Energy Technology Data Exchange (ETDEWEB)

    Valenzuela, L.; Zarza, E. [CIEMAT, Plataforma Solar de Almeria, Ctra. Senes s/n, P.O. Box 22, E-04200 Tabernas, Almeria (Spain); Berenguel, M. [Universidad de Almeria, Dpto. Lenguajes y Computacion, Ctra. Sacramento s/n, E-04120 Almeria (Spain); Camacho, E.F. [Universidad de Sevilla, Dpto. de Ingenieria de Sistemas y Automatica, Camino de los Descubrimientos s/n, E-41092 Sevilla (Spain)

    2006-01-15

    Electricity production using solar thermal energy is one of the main research areas at present in the field of renewable energies, these systems being characterised by the need of reliable control systems aimed at maintaining desired operating conditions in the face of changes in solar radiation, which is the main source of energy. A new prototype of solar system with parabolic trough collectors was implemented at the Plataforma Solar de Almeria (PSA, South-East Spain) to investigate the direct steam generation process under real solar conditions in the parabolic solar collector field of a thermal power plant prototype. This paper presents details and some results of the application of a control scheme designed and tested for the recirculation operation mode, for which the main objective is to obtain steam at constant temperature and pressure at the outlet of the solar field, so that changes produced in the inlet water conditions and/or solar radiation will only affect the amount of steam produced by the solar field. The steam quality and consequently the nominal efficiency of the plant are thus maintained. (author)

  12. Passive system with steam-water injector for emergency supply of NPP steam generators

    International Nuclear Information System (INIS)

    Il'chenko, A.G.; Strakhov, A.N.; Magnitskij, D.N.

    2009-01-01

    The calculation results of reliability indicators of emergency power supply system and emergency feed-water supply system of serial WWER-1000 unit are presented. To ensure safe water supply to steam generators during station blackout it was suggested using additional passive emergency feed-water system with a steam-water injector working on steam generators dump steam. Calculated analysis of steam-water injector operating capacity was conducted at variable parameters of steam at the entrance to injector, corresponding to various moments of time from the beginning of steam-and-water damping [ru

  13. Catalytic Steam Reforming of Bio-Oil to Hydrogen Rich Gas

    DEFF Research Database (Denmark)

    Trane-Restrup, Rasmus

    heating value and high content of oxygen, which makes it unsuited for direct utilization in engines. One prospective technology for upgrading of bio-oil is steam reforming (SR), which can be used to produce H2 for upgrading of bio-oil through hydrodeoxygenation or synthesis gas for processes like......-oil. There are two main pathways to minimize carbon deposition in steam reforming; either through optimization of catalyst formulation or through changes to the process parameters, like changes in temperature, steam to carbon ratio (S/C), or adding O2 or H2 to the feed. In this thesis both pathways have been...

  14. Two different modelling methods of the saturated steam turbine load rejection

    International Nuclear Information System (INIS)

    Negreanu, Gabriel-Paul; Oprea, Ion

    1999-01-01

    One of the most difficult operation regimes of a steam turbine is the load rejection. It happens usually when the main switchgear of the unit closes unexpectedly due to some external or internal causes. In this moment, the rotor balance collapses: the motor momentum is positive, the resistant momentum is zero and the rotation velocity increases rapidly. When this process occurs, the over-speed protection should activate the emergency stop valves and the control and intercept valves in order to stop the steam admission into the turbine. The paper presents two differential approaches of the fluid dynamic processes from the flow sections of the saturated steam turbine of the NPP, where the laws of mass and energy conservation are applied. In this manner, the 'power and speed versus time' diagrams can be drawn. The main parameters of such technical problem are the closure low of the valves, the large volume of internal cavities, the huge inertial momentum of the rotor and especially the moisture of the steam that evaporates when the pressure decreases and generates an extra power in the turbine. (authors)

  15. The Flooding Water Source Analysis following the Feed Line Break at the Compartment outside Containment for Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Song, Dong Soo [Korea Electric Power Research Institute, Daejeon (Korea, Republic of); Park, Young Chan [ACT, Daejeon (Korea, Republic of)

    2007-07-01

    The Periodic Safety Review(PSR) has been performing for the operating nuclear power plant in Korea. One of the PSR evaluation items is environmental qualification. Flooding issue for nuclear power plants designed and built in 1970 is extremely severe for main steam header compartment and main feed water line region of intermediate building and lower floor. This study presents to analyze flood level of feed water line breaks for the Westinghouse nuclear power plant. This analyses provides the mass and energy releases using the developed methodology for a break outside containment. For the analyses RETRAN-3D computer program is used.

  16. Flooding experiments with steam and water in a large diameter vertical tube

    International Nuclear Information System (INIS)

    Williams, S.N.; Solom, M.; Draznin, O.; Choutapalli, I.; Vierow, K.

    2009-01-01

    An experimental study on flooding in a large diameter tube is being conducted. In a countercurrent, two-phase flow system, flooding can be defined as the onset of flow reversal of the liquid component which results in cocurrent flow. Flooding can be perceived as a limit to two-phase countercurrent flow, meaning that pairs of liquid and gas flow rates exist that define the envelope for stable countercurrent flow for a given system. Flooding in the AP600 pressurizer surge line can affect the vessel refill rate following a small break LOCA. Analysis of hypothetical severe accidents with current simplified flooding models show that these models represent the largest uncertainty in steam generator tube creep rupture. During a hypothetical station blackout scenario without auxiliary feedwater recovery, should the hot leg become voided, the pressurizer liquid will drain to the hot leg and flooding may occur in the surge line. Experiments have been conducted in a 3-inch (76.2 mm) diameter tube with subcooled water and superheated steam as the working fluids at atmospheric pressure. Water flows down the inside of the tube as an annulus while the steam flows upward in the middle. Water flow rates vary from 3.5 to 12 GPM (0.00022 to 0.00076 m 3 /s) and the water inlet temperature is about 70degC. The steam inlet temperature is about 110degC. It was found that a larger steam flow rate was needed to achieve flooding for a lower water flow rate and for a higher water flow rate. This unique data for flooding in steam-water systems in large diameter tubes will reduce uncertainty in flooding models currently utilized in reactor safety codes. (author)

  17. Turbulence production by a steam-driven jet in a water vessel

    Energy Technology Data Exchange (ETDEWEB)

    Wissen, R.J.E. van; Schreel, K.R.A.M.; Geld, C.W.M. van der [Eindhoven Univ. of Technology (Netherlands). Dept. of Mechanical Engineering; Wieringa, J. [Unilever Research and Development, Vlaardingen (Netherlands)

    2004-04-01

    Direct steam injection is an efficient means of heating a volume of liquid. Usually the steam is injected via a nozzle, yielding a strong jet that condenses rapidly and transforms into a self-similar single phase jet. In the experiments reported in this paper, superheated steam is injected, centrally, at the bottom of a vertical, cylindrical water vessel. The resulting jet is turbulent (Re=7.9 x 10{sup 4}-18.1 x 10{sup 4} with the length scale based on the width of the jet, r{sub 1/2} and the velocity scale based on the centerline velocity, U{sub 0}). Using PIV in a vertical plane through the central axis, instantaneous velocity fields have been measured at a rate of 15 Hz. Near the inlet, the jet is mainly steam that rapidly condenses. Further downstream, the jet is essentially single phase, although some residual air is present as microscopically small bubbles. In the area directly downstream of the steam part, the ratio of r{sub 1/2} to the vessel radius R (32.5 cm) is about 1/14. The production of turbulent kinetic energy has been quantified for the main process conditions. Its dependencies on temperature, nozzle opening and inlet steam pressure have been determined. This production of energy is related to the stresses exerted on small particles in the mixture, and break-up of particles is discussed. (author)

  18. Graphite-water steam-generating reactor in the USSR

    Energy Technology Data Exchange (ETDEWEB)

    Dollezhal, N A [AN SSSR, Moscow

    1981-10-01

    One of the types of power reactor used in the USSR is the graphite-water steam-generating reactor RBMK. This produces saturated steam at a pressure of 7MPa. Reactors giving 1GWe each have been installed at the Leningrad, Kursk, Chernobyl and other power stations. Further stations using reactors of this type are being built. A description is given of the fuel element design, and of the layout of the plant. The main characteristics of RBMK reactors using fuel of rated and higher enrichment are listed.

  19. Fourth international seminar on horizontal steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Tuomisto, H. [ed.] [IVO Group, Vantaa (Finland); Purhonen, H. [ed.] [VTT, Espoo (Finland); Kouhia, V. [ed.] [Lappeenranta Univ. of Technology (Finland)

    1997-12-31

    The general objective of the International Seminars of Horizontal Steam Generator Modelling has been the improvement in understanding of realistic thermal hydraulic behaviour of the generators when performing safety analyses for VVER reactors. The main topics presented in the fourth seminar were: thermal hydraulic experiments and analyses, primary collector integrity, feedwater distributor replacement, management of primary-to-secondary leakage accidents and new developments in the VVER safety technology. The number of participants, representing designers and manufacturers of the horizontal steam generators, plant operators, engineering companies, research organizations, universities and regulatory authorities, was 70 from 10 countries.

  20. Fourth international seminar on horizontal steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Tuomisto, H [ed.; IVO Group, Vantaa (Finland); Purhonen, H [ed.; VTT, Espoo (Finland); Kouhia, V [ed.; Lappeenranta Univ. of Technology (Finland)

    1998-12-31

    The general objective of the International Seminars of Horizontal Steam Generator Modelling has been the improvement in understanding of realistic thermal hydraulic behaviour of the generators when performing safety analyses for VVER reactors. The main topics presented in the fourth seminar were: thermal hydraulic experiments and analyses, primary collector integrity, feedwater distributor replacement, management of primary-to-secondary leakage accidents and new developments in the VVER safety technology. The number of participants, representing designers and manufacturers of the horizontal steam generators, plant operators, engineering companies, research organizations, universities and regulatory authorities, was 70 from 10 countries.

  1. Fourth international seminar on horizontal steam generators

    International Nuclear Information System (INIS)

    Tuomisto, H.; Purhonen, H.; Kouhia, V.

    1997-01-01

    The general objective of the International Seminars of Horizontal Steam Generator Modelling has been the improvement in understanding of realistic thermal hydraulic behaviour of the generators when performing safety analyses for VVER reactors. The main topics presented in the fourth seminar were: thermal hydraulic experiments and analyses, primary collector integrity, feedwater distributor replacement, management of primary-to-secondary leakage accidents and new developments in the VVER safety technology. The number of participants, representing designers and manufacturers of the horizontal steam generators, plant operators, engineering companies, research organizations, universities and regulatory authorities, was 70 from 10 countries

  2. SGTR Project: Separate Effect Studies for Vertical Steam Generators

    Energy Technology Data Exchange (ETDEWEB)

    Peyres, V.; Polo, J.; Herranz, L. E.

    2003-07-01

    The SGTR project has been carried out within the fifth EURATOM Framework Programme (Contract No FIKS-CT-1999-0007). Its main objective was to provide an experimental database and to develop and/or verify models to support definition of accident management measures in the hypothetical case of a Steam Generator tube Rupture (SGTR) sequence. The project addressed both vertical and horizontal steam generator designs. This report summarises the main results obtained in the intermediate scale experimentation that addressed Western type steam generators. The specific goal of this test programme was to investigate aerosol retention in the break stage of the secondary side of a water-empty steam generator. The test matrix consisted of 12 tests that explored the influence of variables such as break type and orientation and inlet gas flow rate. This work was performed in the PECA facility of the Laboratory for Analysis of Safety Systems (LASS). Aerosol retention at the break stage of a dry steam generator was observed to be low and non-uniform. Neither break type nor orientation affected results significantly whenever gas flowrates exceeded about 100 kg/h. However, deposition patterns guillotine breaks and fish mouth ones showed remarkable differences. For flowrates above 100 kg/hm the higher the gas flow velocity, the lower the total mass depleted on tube bundle surfaces; however, at lower flowrates this trend was not maintained. An attempt to measure gas injection velocity at the break exit by Particle Image Velocity (PIV) was done but data were highly uncertain. (Author) 2 refs.

  3. Steam generator replacement at Surry Power Station

    International Nuclear Information System (INIS)

    McKay, H.S.

    1982-01-01

    The purposes of the steam generator repair program at Surry Power Station were to repair the tube degradation caused by corrosion-related phenomena and to restore the integrity of the steam generators to a level equivalent to new equipment. The repair program consisted of (1) replacing the existing lower-shell assemblies with new ones and (2) adding new moisture separation equipment to the upper-shell assemblies. These tasks required that several pieces of reactor coolant piping, feedwater piping, main steam piping, and the steam generator be cut and refurbished for reinstallation after the new lower shell was in place. The safety implications and other potential effects of the repair program both during the repair work and after the unit was returned to power were part of the design basis of the repair program. The repair program has been completed on Unit 2 without any adverse effects on the health and safety of the general public or to the personnel engaged in the repair work. Before the Unit 1 repair program began, a review of work procedures and field changes for the Unit 2 repair was conducted. Several major changes were made to avoid recurrence of problems and to streamline procedures. Steam generator replacements was completed on June 1, 1981, and the unit is presently in the startup phase of the outrage

  4. HTGR steam generator development

    International Nuclear Information System (INIS)

    Schuetzenduebel, W.G.; Hunt, P.S.; Weber, M.

    1976-01-01

    More than 40 gas-cooled reactor plants have produced in excess of 400 reactor years of operating experience which have proved a reasonably high rate of gas-cooled reactor steam generator availability. The steam generators used in these reactors include single U-tube and straight-tube steam generators as well as meander type and helically wound or involute tube steam generators. It appears that modern reactors are being equipped with helically wound steam generators of the once-through type as the end product of steam generator evolution in gas-cooled reactor plants. This paper provides a general overview of gas-cooled reactor steam generator evolution and operating experience and shows how design criteria and constraints, research and development, and experience data are factored into the design/development of modern helically wound tube steam generators for the present generation of gas-cooled reactors

  5. Procedure for generating steam and steam generator for operating said procedure

    International Nuclear Information System (INIS)

    Chlique, Bernard.

    1975-01-01

    This invention concerns the generation of steam by bringing the water to be vaporised into indirect thermal exchange relation with the heating steam which condenses when passing in series, along alternate routes, through bundles of tubes immersed in a vaporising chamber. A number of steam generators working on this principle already exist. The purpose of the invention is to modify the operating method of these steam generators by means of a special disposition making it possible to build a compact unit including an additional bundle of tubes heated by the condensates collected at the outlet of each bundle through which the heating steam passes [fr

  6. Specialists' meeting on maintenance and repair of LMFBR steam generators. Summary report

    International Nuclear Information System (INIS)

    2002-01-01

    The purpose of the meeting was to review and discuss the experience accumulated in various countries on the general design philosophy of LMFBR steam generators from the view point of maintenance and repair, in-service inspection of steam generator tube bundles, identification and inspection of failed tubes and the cleaning and repairing of failed steam generators. The following main topic areas were discussed by participants: National review presentations on maintenance and repair of LMFBR steam generators - design philosophy for maintenance and repair; Research and Development work on maintenance and repair; Experience on steam generator maintenance and repair. During the meeting papers were presented by the participants on behalf of their countries and organizations. A final discussion session was held and summaries, general conclusions and recommendations were approved by consensus

  7. The chemical kinetics of the reactions of lithium with steam-air mixtures

    International Nuclear Information System (INIS)

    Barnett, D.S.; Kazimi, M.S.

    1989-04-01

    This work involved the experimental and analytical determination of the consequences of lithium fires in the presence of steam. Experiments were performed to characterize the chemical reactions of lithium with steam-nitrogen and steam-air mixtures. Models were introduced in the LITFIRE code to describe lithium fires in the presence of steam inside the containment building and plasma chamber of a hypothetical fusion reactor. The code was also equipped with the capability to determine the effects of decay heat and lithium fire on the temperature response of the reactor first wall in the event of a coolant disturbance. Forty-two kinetics experiments were performed in which a stream of steam-nitrogen or steam-air was passed over and reacted with approximately three grams of lithium heated to a predetermined temperature. The lithium reaction rates with the constituent gases were measured and characterized for a wide range of lithium temperatures and gas compositions. Experiments were performed with steam molar concentrations of 5, 15 and 30% and lithium temperatures ranging from 400 to 1100 degree C, inclusive. The LITFIRE code was modified to enable it to model the interactions of lithium with steam-air atmospheres. Results of the reaction kinetics experiments were used in the reaction model, and the heat transfer model was expanded to allow it to handle condensible atmospheres. Three groups of accidents were investigated: a spill on the containment building floor, a spill inside the reactor plasma chamber, and a spill inside the plasma chamber with steam injection to the containment building simulating a steam line break. The results were compared to dry air cases under the same conditions. 23 refs., 66 figs., 18 tabs

  8. The origin of narrowing of the Si 2p coincidence photoelectron spectroscopy main line of Si(1 0 0) surface

    International Nuclear Information System (INIS)

    Ohno, Masahide

    2011-01-01

    Highlights: → The Si 2p coincidence photoelectron spectroscopy (PES) main line of Si(1 0 0) is calculated. → The PES main line shows an asymmetric line shape change compared to the singles one. → The narrowing of the coincidence Si 2p PES main line is well reproduced. → The inherent mechanism of APECS is explained by a many-body theory. - Abstract: The Si 2p photoelectron spectroscopy (PES) main line of Si(1 0 0) surface measured in coincidence with the singles (noncoincidence) Si L 2,3 -VV Auger-electron spectroscopy (AES) elastic peak is calculated. The agreement with the experiment is good. The present work is the first many-body calculation of the experimental coincidence PES spectrum of solid surface. The narrowing of the coincidence Si 2p PES main line compared to the singles one is due to the mechanism inherent in the coincidence PES. The inherent mechanism is explained by a many-body theory by which photoemission and Auger-electron emission are treated on the same footing.

  9. Hydraulic model of the steam-lines network of the Cerro Prieto, B.C., geothermal field; Modelo hidraulico de la red de vaporductos del campo geotermico de Cerro Prieto, B.C.

    Energy Technology Data Exchange (ETDEWEB)

    Salaices, E; Garcia, A; Martinez J I; Ovando, R; Cecenas, M; Hernandez A F [Instituto de Investigaciones Electricas, Cuernavaca, Morelos (Mexico)]. E-mail: salaices@iie.org.mx; Canchola, I; Mora, O; Miranda, C; Herandez, M; Lopez, S; Murillo, I [Comision Federal de Electricidad, B.C (Mexico)

    2007-01-15

    The steam-line network of the Cerro Prieto geothermal field is composed of 184 wells, and 162 of the wells are integrated and connected by pipes. Thirteen power units, with an installed electrical capacity of 720 MW, are fed by that network. The network length is 120 km, including pipes of several diameters with branches and interconnections. The extension and complexity of the steam-line system make it difficult to analyze the transport and supply of steam to the power plants. For that it was necessary to have a tool capable of analyzing the system and the performance of the network as a whole, as well as the direction and flow volumes in each part of the system. In this paper, a hydraulic model of the Cerro Prieto steam-line network is presented. The model can determine the performance of the whole network by quantifying the pressure drops, flows and heat losses of the components. The model analyses the consequences of changes in operating conditions, steam production, maintenance activities and design (such as the integration of new wells). The model was developed using PIPEPHASE 9.0, a numeric simulator of multi-phase flow in steady state with heat transfer. It is used to model systems and pipe networks for steam- and condensate-transport. [Spanish] La red de vaporductos del campo geotermico de Cerro Prieto esta compuesta por un conjunto de 184 pozos, de los cuales 162 son pozos integrados, interconectados entre si a traves de una red de tuberias. Por medio de esta red se alimentan 13 unidades generadoras de electricidad con una capacidad total instalada de 720 MWe. La red tiene una longitud aproximada de 120 kilometros y esta compuesta por tuberias de diferentes diametros, ramales, interconexiones, etc. La complejidad y extension del sistema de vaporductos hace muy dificil el analisis del transporte y suministro de vapor a las plantas generadoras. Lo anterior creo la necesidad de contar con una herramienta que ayudara en el analisis del sistema con el fin de

  10. Limit the effects of secondary circuit water or steam piping breaks in the reactor building

    International Nuclear Information System (INIS)

    Nachev, N.

    2001-01-01

    The existing design of the WWER-1000 Model 320 does not include provisions against the local mechanical effects of pipe ruptures of the secondary system piping. This situation may lead to accidental effects beyond the design basis of the plant in case of a postulated secondary pipe rupture event. The aim of the present safety enhancement measure is to overcome this safety deficit, that means to carry out some analyses and to suggest protection measures, by which the specified design basis of the plant concerning secondary circuit design basis accidents will be assured. The systems to be considered include the main steam lines (MSL) and the main feedwater lines (MFWL) in the safety related system areas. These areas are the system portions, which are located in the reactor building (containment and room A820 outside the containment). The pipe rupture effects to be considered include the local effects, that means pipe whip impact and jet forces on the adjacent equipment and structures, as well as reaction forces due to blowdown thrust forces and pressure waves in the broken piping system. (author)

  11. Analysis and design of flow limiter used in steam generator

    International Nuclear Information System (INIS)

    Liu Shixun; Gao Yongjun

    1995-10-01

    Flow limiter is an important safety component of PWR steam generator. It can limit the blowdown rate of steam generator inventory in case of the main steam pipeline breaks, so that the rate of the primary coolant temperature reduction can be slowed down in order to prevent fuel element from burn-out. The venturi type flow limiter is analysed, its flow characteristics are delineated, physical and mathematical models defined; the detail mathematical derivation provided. The research lays down a theoretic basis for flow limiter design. The governing equations and formulas given can be directly applied to computer analysis of the flow limiter. (3 refs., 3 figs.)

  12. Simulation of the Thermal Hydraulic Processes in the Horizontal Steam Generator with the Use of the Different Interfacial Friction Correlations

    International Nuclear Information System (INIS)

    Melikhov, V.; Melikhov, O.; Parfenov, Y.; Nerovnov, A.

    2011-01-01

    The horizontal steam generator (SG) is one of specific features of Russian-type pressurized water reactors (VVERs). The main advantages of horizontal steam generator are connected with low steam loads on evaporation surface, simple separation scheme and high circulation ratio. The complex three-dimensional steam-water flows in the steam generator vessel influence significantly the processes of the steam separation, distribution, and deposition of the soluble and non soluble impurities and determine the efficiency and reliability of the steam generator operation. The 3D code for simulation of the three-dimensional steam-water flows in the steam generator could be effective tool for design and optimization of the horizontal steam generator. The results of the code calculations are determined mainly by the set of the correlations describing interaction of the steam-water mixture with the inner constructions of the SG and interfacial friction. The results obtained by 3D code STEG with the usage of the different interfacial friction correlations are presented and discussed in the paper. These results are compared with the experimental ones obtained at the experimental test facility PGV-1500 constructed for investigation of the processes in the horizontal steam generator

  13. Hydrogen-based power generation from bioethanol steam reforming

    Energy Technology Data Exchange (ETDEWEB)

    Tasnadi-Asztalos, Zs., E-mail: tazsolt@chem.ubbcluj.ro; Cormos, C. C., E-mail: cormos@chem.ubbcluj.ro; Agachi, P. S. [Babes-Bolyai University, Faculty of Chemistry and Chemical Engineering, 11 Arany Janos, Postal code: 400028, Cluj-Napoca (Romania)

    2015-12-23

    This paper is evaluating two power generation concepts based on hydrogen produced from bioethanol steam reforming at industrial scale without and with carbon capture. The power generation from bioethanol conversion is based on two important steps: hydrogen production from bioethanol catalytic steam reforming and electricity generation using a hydrogen-fuelled gas turbine. As carbon capture method to be assessed in hydrogen-based power generation from bioethanol steam reforming, the gas-liquid absorption using methyl-di-ethanol-amine (MDEA) was used. Bioethanol is a renewable energy carrier mainly produced from biomass fermentation. Steam reforming of bioethanol (SRE) provides a promising method for hydrogen and power production from renewable resources. SRE is performed at high temperatures (e.g. 800-900°C) to reduce the reforming by-products (e.g. ethane, ethene). The power generation from hydrogen was done with M701G2 gas turbine (334 MW net power output). Hydrogen was obtained through catalytic steam reforming of bioethanol without and with carbon capture. For the evaluated plant concepts the following key performance indicators were assessed: fuel consumption, gross and net power outputs, net electrical efficiency, ancillary consumptions, carbon capture rate, specific CO{sub 2} emission etc. As the results show, the power generation based on bioethanol conversion has high energy efficiency and low carbon footprint.

  14. Hydrogen-based power generation from bioethanol steam reforming

    International Nuclear Information System (INIS)

    Tasnadi-Asztalos, Zs.; Cormos, C. C.; Agachi, P. S.

    2015-01-01

    This paper is evaluating two power generation concepts based on hydrogen produced from bioethanol steam reforming at industrial scale without and with carbon capture. The power generation from bioethanol conversion is based on two important steps: hydrogen production from bioethanol catalytic steam reforming and electricity generation using a hydrogen-fuelled gas turbine. As carbon capture method to be assessed in hydrogen-based power generation from bioethanol steam reforming, the gas-liquid absorption using methyl-di-ethanol-amine (MDEA) was used. Bioethanol is a renewable energy carrier mainly produced from biomass fermentation. Steam reforming of bioethanol (SRE) provides a promising method for hydrogen and power production from renewable resources. SRE is performed at high temperatures (e.g. 800-900°C) to reduce the reforming by-products (e.g. ethane, ethene). The power generation from hydrogen was done with M701G2 gas turbine (334 MW net power output). Hydrogen was obtained through catalytic steam reforming of bioethanol without and with carbon capture. For the evaluated plant concepts the following key performance indicators were assessed: fuel consumption, gross and net power outputs, net electrical efficiency, ancillary consumptions, carbon capture rate, specific CO 2 emission etc. As the results show, the power generation based on bioethanol conversion has high energy efficiency and low carbon footprint

  15. Hydrogen-based power generation from bioethanol steam reforming

    Science.gov (United States)

    Tasnadi-Asztalos, Zs.; Cormos, C. C.; Agachi, P. S.

    2015-12-01

    This paper is evaluating two power generation concepts based on hydrogen produced from bioethanol steam reforming at industrial scale without and with carbon capture. The power generation from bioethanol conversion is based on two important steps: hydrogen production from bioethanol catalytic steam reforming and electricity generation using a hydrogen-fuelled gas turbine. As carbon capture method to be assessed in hydrogen-based power generation from bioethanol steam reforming, the gas-liquid absorption using methyl-di-ethanol-amine (MDEA) was used. Bioethanol is a renewable energy carrier mainly produced from biomass fermentation. Steam reforming of bioethanol (SRE) provides a promising method for hydrogen and power production from renewable resources. SRE is performed at high temperatures (e.g. 800-900°C) to reduce the reforming by-products (e.g. ethane, ethene). The power generation from hydrogen was done with M701G2 gas turbine (334 MW net power output). Hydrogen was obtained through catalytic steam reforming of bioethanol without and with carbon capture. For the evaluated plant concepts the following key performance indicators were assessed: fuel consumption, gross and net power outputs, net electrical efficiency, ancillary consumptions, carbon capture rate, specific CO2 emission etc. As the results show, the power generation based on bioethanol conversion has high energy efficiency and low carbon footprint.

  16. Indian Point 2 steam generator tube rupture analyses

    International Nuclear Information System (INIS)

    Dayan, A.

    1985-01-01

    Analyses were conducted with RETRAN-02 to study consequences of steam generator tube rupture (SGTR) events. The Indian Point, Unit 2, power plant (IP2, PWR) was modeled as a two asymmetric loops, consisting of 27 volumes and 37 junctions. The break section was modeled once, conservatively, as a 150% flow area opening at the wall of the steam generator cold leg plenum, and once as a 200% double-ended tube break. Results revealed 60% overprediction of breakflow rates by the traditional conservative model. Two SGTR transients were studied, one with low-pressure reactor trip and one with an earlier reactor trip via over temperature ΔT. The former is more typical to a plant with low reactor average temperature such as IP2. Transient analyses for a single tube break event over 500 seconds indicated continued primary subcooling and no need for steam line pressure relief. In addition, SGTR transients with reactor trip while the pressurizer still contains water were found to favorably reduce depressurization rates. Comparison of the conservative results with independent LOFTRAN predictions showed good agreement

  17. Water and steam sampling systems; Provtagningssystem foer vatten och aanga

    Energy Technology Data Exchange (ETDEWEB)

    Hellman, Mats

    2009-10-15

    The supervision of cycle chemistry can be divided into two parts, the sampling system and the chemical analysis. In modern steam generating plants most of the chemical analyses are carried out on-line. The detection limits of these analyzers are pushed downward to the ppt-range (parts per trillion), however the analyses are not more correct than the accuracy of the sampling system. A lot of attention has been put to the analyzers and the statistics to interpret the results but the sampling procedures has gained much less attention. This report aims to give guidance of the considerations to be made regarding sampling systems. Sampling is necessary since most analysis of interesting parameters cannot be carried out in- situ on-line in the steam cycle. Today's on-line instruments for pH, conductivity, silica etc. are designed to meet a water sample at a temperature of 10-30 deg C. This means that the sampling system has to extract a representative sample from the process, transport and cool it down to room temperature without changing the characteristics of the fluid. In the literature research work, standards and other reports can be found. Although giving similar recommendations in most aspects there are some discrepancies that may be confusing. This report covers all parts in the sampling system: Sample points and nozzles; Sample lines; Valves, regulating and on-off; Sample coolers; Temperature, pressure and flow rate control; Cooling water; and Water recovery. On-line analyzers connecting to the sampling system are not covered. This report aims to clarify what guidelines are most appropriate amongst the existing ones. The report should also give guidance to the design of the sampling system in order to achieve representative samples. In addition to this the report gives an overview of the fluid mechanics involved in sampling. The target group of this report is owners and operators of steam generators, vendors of power plant equipment, consultants working in

  18. Use of Main Loop Isolating Valves (GZZS) in WWER 440

    International Nuclear Information System (INIS)

    Stefanova, A.E.; Gencheva, R.V.; Groudev, P.P.

    2002-01-01

    This paper discusses the usage of Main Loop Isolation Valves in case of Steam Generator Tube Rupture accident in WWER440/V230. A double-ended single pipe break in SG-6 was chosen as representative. In the paper are investigated two cases. In the first one the operator isolates the affected loop by Main Loop Isolation Valves closing and after primary depressurization re-opens them to cooldown the damaged Steam Generator. The second case treats the situation, where Main Loop Isolation Valves fail to close with the necessary operator actions for managing plant recovery. RELAP5/MOD3.2 computer code has been used to simulate the Steam Generator Tube Rupture accident in WWER440 NPP model. This model was developed and validated at Institute for Nuclear Research and Nuclear Energy - Bulgarian Academy of Sciences. The results of analyses presented in this report demonstrate that in the both cases (with or without Main Loop Isolation Valves usage) the operator could bring the plant to stable and safety conditions (Authors)

  19. Analysis of fuel oil consumption in industrial steam boiler plants in Republic of Macedonia

    International Nuclear Information System (INIS)

    Armenski, Slave; Dimitrov, Konstantin; Tashevski, Done

    1999-01-01

    The steam boiler plants with heavy and light fuel oils in Republic of Macedonia are analyzed and determined. Depending of the working exit pressure, they are grouped in main industrial branches. The heat capacity and the steam production for the steam boiler plants are determined both total and separately by the different industrial branches. Depending of heat capacity and working period per year, the consumption of heavy and light oil is analyzed and determined particular for each industrial branch and total for all steam boiler plants for summer and winter period. (Author)

  20. Development and evaluation of the NSSS model with four steam lines for the LVNP using the SCDAPSIM code

    International Nuclear Information System (INIS)

    Salazar C, J.H.; Nunez C, A.; Camargo C, R.

    2005-01-01

    The present work shows the pattern of the NSSS considering the four main vapor lines as well as their evaluation. The pattern was developed by the National Commission of Nuclear Security and Safeguards (CNSNS) and it has as main objective to account with a model of the Laguna Verde Nuclear power plant (CNLV) for the simulation and analysis of transitory events where are involved some of main vapor lines, or some relief valves and safety (SRV's). The model was evaluated with data of the CNLV. In 1996 the Federal Commission of Electricity (CFE) request to the CNSNS permission to operate the Unit 2 until the first recharge, having the main vapor line 'B' isolated and operating with a level of power corresponding to a flow of total vapor of 85% of the nominal one (of 1931 MWt). The obtained values were compared with the obtained registrations of the CNLV in order to evaluate the model. Those results show relative errors inferior to 3% among the CNLV reported value and the one calculated by the SCDAPSIM code. (Author)

  1. Steam separator-superheater with drawing of a fraction of the dried steam

    International Nuclear Information System (INIS)

    Bessouat, Roger; Marjollet, Jacques.

    1976-01-01

    This invention concerns a vertical separator-superheater of the steam from a high pressure expansion turbine before it is admitted to an expansion turbine at a lower pressure, by heat exchange with steam under a greater pressure, and drawing of a fraction of the dried steam before it is superheated. Such drawing off is necessary in the heat exchange systems of light water nuclear reactors. Its purpose is to provide a separator-superheater that provides an even flow of non superheated steam and a regular distribution of the steam to be superheated to the various superheating bundles, with a significantly uniform temperature of the casing, thereby preventing thermal stresses and ensuring a minimal pressure drop. The vertical separator-superheater of the invention is divided into several vertical sections comprising as from the central area, a separation area of the steam entrained water and a superheater area and at least one other vertical section with only a separation area of the steam entrained water [fr

  2. Evaluation of steam generator U-tube integrity during PWR station blackout with secondary system depressurization

    International Nuclear Information System (INIS)

    Hidaka, Akihide; Asaka, Hideaki; Sugimoto, Jun; Ueno, Shingo; Yoshino, Takehito

    1999-12-01

    In PWR severe accidents such as station blackout, the integrity of steam generator U-tube would be threatened early at the transient among the pipes of primary system. This is due to the hot leg countercurrent natural circulation (CCNC) flow which delivers the decay heat of the core to the structures of primary system if the core temperature increases after the secondary system depressurization. From a view point of accident mitigation, this steam generator tube rupture (SGTR) is not preferable because it results in the direct release of primary coolant including fission products (FP) to the environment. Recent SCDAP/RELAP5 analyses by USNRC showed that the creep failure of pressurizer surge line which results in release of the coolant into containment would occur earlier than SGTR during the secondary system depressurization. However, the analyses did not consider the decay heat from deposited FP on the steam generator U-tube surface. In order to investigate the effect of decay heat on the steam generator U-tube integrity, the hot leg CCNC flow model used in the USNRC's calculation was, at first, validated through the analysis for JAERI's LSTF experiment. The CCNC model reproduced well the thermohydraulics observed in the LSTF experiment and thus the model is mostly reliable. An analytical study was then performed with SCDAP/RELAP5 for TMLB' sequence of Surry plant with and without secondary system depressurization. The decay heat from deposited FP was calculated by JAERI's FP aerosol behavior analysis code, ART. The ART analysis showed that relatively large amount of FPs may deposit on steam generator U-tube inlet mainly by thermophoresis. The SCDAP/RELAP5 analyses considering the FP decay heat predicted small safety margin for steam generator U-tube integrity during secondary system depressurization. Considering associated uncertainties in the analyses, the potential for SGTR cannot be ignored. Accordingly, this should be considered in the evaluation of merits

  3. Processing of biomass to Hydrocarbons – using a new catalytic steam pyrolysis route

    OpenAIRE

    Mellin, Pelle; Kantarelis, Efthymios; Yang, Weihong

    2014-01-01

    Obtaining renewable transportation fuel has been identified as one of the main challenges for a sustainable society. Catalytic pyrolysis followed by hydrotreatment has been demonstrated as one possible route for producing transportation fuels. Using steam in this process could have a number of benefits as given by our research effort. For this paper, we will show that a catalyst together with steam prolongs the activity of the catalyst by preventing coking. This means that both steam and cata...

  4. Current Status on the Development of a Double Wall Tube Steam Generator

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Ho Yun; Choi, Byoung Hae; Kim, Jong Man; Kim, Byung Ho

    2007-12-15

    A fast reactor, which uses sodium as a coolant, has a lot of merits as a next generation nuclear reactor. However, the possibility of a sodium-water reaction occurrence hinders the commercialization of this reactor. As one way to improve the reliability of a steam generator, a double-wall tube steam generator is being developed in GEN-4 program. In this report, the current state of the technical developments for a double-wall tube steam generator are reviewed and a future plan for the development of a double-wall tube steam generator is established. The current focuses of this research are an improvement of the heat transfer capability for a double-wall tube and the development of a proper leak detection method for the failure of a double-wall tube during a reactor operation. The ideal goal is an on-line leak detection of a double wall tube to prevent the sodium-water reaction. However, such a method is not developed as yet. An alternative method is being used to improve the reliability of a steam generator by performing a non-destructive test of a double wall tube during the refueling period of a reactor. In this method a straight double wall tube is employed to perform this test easily, but has a difficulty regarding an absorption of a thermal expansion of the used materials. If an on-line leak detection method is developed, the demerits of a straight double-wall tube are avoided by using a helical type double-wall tube, and the probability of a sodium-water reaction can be reduced to a level less than the design-based accident.

  5. A steam inerting system for hydrogen disposal for the Vandenberg Shuttle

    Science.gov (United States)

    Belknap, Stuart B.

    1988-01-01

    A two-year feasibility and test program to solve the problem of unburned confined hydrogen at the Vandenberg Space Launch Complex Six (SLC-6) during Space Shuttle Main Engine (SSME) firings is discussed. A novel steam inerting design was selected for development. Available sound suppression water is superheated to flash to steam at the duct entrance. Testing, analysis, and design during 1987 showed that the steam inerting system (SIS) solves the problem and meets other flight-critical system requirements. The SIS design is complete and available for installation at SLC-6 to support shuttle or derivative vehicles.

  6. Structural integrity analysis of a steam turbine

    International Nuclear Information System (INIS)

    Villagarcia, Maria P.

    1997-01-01

    One of the most critical components of a power utility is the rotor of the steam turbine. Catastrophic failures of the last decades have promoted the development of life assessment procedures for rotors. The present study requires the knowledge of operating conditions, component geometry, the properties of materials, history of the component, size, location and nature of the existing flaws. The aim of the present work is the obtention of a structural integrity analysis procedure for a steam turbine rotor, taking into account the above-mentioned parameters. In this procedure, a stress thermal analysis by finite elements is performed initially, in order to obtain the temperature and stress distribution for a subsequent analysis by fracture mechanics. The risk of a fast fracture due to flaws in the central zone of the rotor is analyzed. The procedure is applied to an operating turbine: the main steam turbine of the Atucha I nuclear power utility. (author)

  7. A non-conventional interpretation of thermal regeneration in steam cycles

    International Nuclear Information System (INIS)

    Bracco, Stefano; Damiani, Lorenzo

    2012-01-01

    Highlights: ► A better understanding of the concept of thermal regeneration in steam cycles. ► Use of a system composed by a non-regenerative cycle and several reverse cycles. ► Calculation of the heat pumps coefficients of performance. ► New interesting formulations of the regenerative cycle efficiency. -- Abstract: The paper aims to contribute to a better understanding of the thermodynamic concept of heat regeneration in steam power plants with a finite number of bleedings. A regenerative Rankine cycle is compared to a complex system (CHC – complete hybrid cycle) composed by one non-regenerative Rankine cycle (HEC – hybrid engine cycle) and more reverse cycles (RCs – reverse cycles), as many as the number of the bleedings, able to pump heat from the condenser to a series of surface feedwater heaters, disposed upstream of the steam plant boiler. The COPs (coefficients of performance) of the heat pumps are evaluated, and new interesting formulations of the efficiency of the regenerative steam cycle are proposed. In particular a steam cycle with two bleedings is analyzed, neglecting heat losses and pressure drops in the boiler and considering irreversibility only along the expansion line of the steam turbine and into the feedwater heaters. The efficiency and the work of the regenerative cycle are compared to the analogous values of the CHC cycle composed by one simple steam cycle (HEC) and two heat pump cycles (RCs), with steam as the working fluid. The two reverse cycles are considered completely reversible and raising heat from the condenser temperature to the bled steam condensing temperature. The paper shows the most significant results of the study in order to analyze the regenerative cycle and the CHC cycle in comparison with the non-regenerative Rankine cycle; in particular, the analysis is focused on the evaluation of the useful work, the heat supplied and the heat rejected for the examined cycles.

  8. Steam gasification of acid-hydrolysis biomass CAHR for clean syngas production.

    Science.gov (United States)

    Chen, Guanyi; Yao, Jingang; Yang, Huijun; Yan, Beibei; Chen, Hong

    2015-03-01

    Main characteristics of gaseous product from steam gasification of acid-hydrolysis biomass CAHR have been investigated experimentally. The comparison in terms of evolution of syngas flow rate, syngas quality and apparent thermal efficiency was made between steam gasification and pyrolysis in the lab-scale apparatus. The aim of this study was to determine the effects of temperature and steam to CAHR ratio on gas quality, syngas yield and energy conversion. The results showed that syngas and energy yield were better with gasification compared to pyrolysis under identical thermal conditions. Both high gasification temperature and introduction of proper steam led to higher gas quality, higher syngas yield and higher energy conversion efficiency. However, excessive steam reduced hydrogen yield and energy conversion efficiency. The optimal value of S/B was found to be 3.3. The maximum value of energy ratio was 0.855 at 800°C with the optimal S/B value. Copyright © 2014 Elsevier Ltd. All rights reserved.

  9. Design of large reheat steam turbines for U.K. and overseas markets

    International Nuclear Information System (INIS)

    Mitchell, J.M.

    1979-01-01

    Two prototype designs of large reheat steam turbines are described, together with the technical, economic and plant design aspects that have influenced their main features. Relevant service experience is outlined and details are given of the solutions adopted to overcome the relatively few problems that were encountered. The evolution of these designs to form the current range of adaptable, pre-engineered modular designs is presented and the main features of current machines are described. A brief account is given of likely future developments in large steam turbines. (author)

  10. The testing of a steam-water separating device used for vertical steam generators

    International Nuclear Information System (INIS)

    Ding Xunshen; Cui Baoyuan; Xue Yunkui; Liu Shixun

    1989-01-01

    The air-water screening tests of a steam-water separating device used for vertical steam generators at low pressure are introduced. The article puts emphasis on the qualification test of the steam-water separating device at hot conditions in a high temperature and pressure water test rig. The performance of the comprehensive test of the steam-water separating device indicates that the humidity of the steam at the drier exit is much less than the specified amount of 0.25%

  11. Steam-generator replacement sets new marks

    International Nuclear Information System (INIS)

    Beck, R.L.

    1995-01-01

    This article describes how, in one of the most successful steam-generator replacement experiences at PWRs worldwide, the V C Summer retrofit exceeded plant goals for critical-path duration, radiation, exposure, and radwaste generation. Intensive planning and teamwork, combined with the firm support of station management and the use of mockups to prepare the work crews for activity in a radiological environment, were key factors in the record performance achieved by South Carolina Electric and Gas Co (SCE and G) in replacing three steam generators at V C Summer nuclear station. The 97-day, two-hour breaker-to-breaker replacement outage -- including an eight-day delay for repair of leak in a small-bore seal-injection line of a reactor coolant pump (unrelated to the replacement activities) -- surpassed the project goal by over one day. Moreover, the outage was only 13 hours shy of the world record held by Virginia Power Co's North Anna Unit 1

  12. Piping hydrodynamic loads for a PWR power up-rate with steam generator replacement

    International Nuclear Information System (INIS)

    Julie M Jarvis; Allen T Vieira; James M Gilmer

    2005-01-01

    Full text of publication follows: Pipe break hydrodynamic loads are calculated for various systems in a PWR for a Power Up-rate (PUR) with a Steam Generator Replacement (SGR). PUR with SGR can change the system pressures, mass flowrates and pipe routing/configuration. These changes can alter the steam generator piping steam/water hammer loads. This paper discusses the need to benchmark against the original design basis, the use of different modeling techniques, and lessons learned. Benchmarking for licensing in the United States is vital in consideration of 10CFR50.59 and other licensing and safety issues. RELAP5 and its force post-processor R5FORCE are used to model the transient loads for various piping systems such as main feedwater and blowdown systems. Other modeling applications, including the Bechtel GAFT program, are used to evaluate loadings in the main steam piping. Forces are calculated for main steam turbine stop valve closure, feedwater pipe breaks and subsequent check valve slam, and blowdown isolation valve closure. These PUR/SGR forces are compared with the original design basis forces. Modeling techniques discussed include proper valve closure modeling, sonic velocity changes due to pipe material changes, and two phase flow effects. Lessons learned based on analyses done for several PWR PUR with SGR are presented. Lessons learned from these analyses include choosing the optimal replacement piping size and routing to improve system performance without resulting in excessive piping loads. (authors)

  13. Impact of flow induced vibration acoustic loads on the design of the Laguna Verde Unit 2 steam dryer

    International Nuclear Information System (INIS)

    Forsyth, D. R.; Wellstein, L. F.; Theuret, R. C.; Han, Y.; Rajakumar, C.; Amador C, C.; Sosa F, W.

    2015-09-01

    Industry experience with Boiling Water Reactors (BWRs) has shown that increasing the steam flow through the main steam lines (MSLs) to implement an extended power up rate (EPU) may lead to amplified acoustic loads on the steam dryer, which may negatively affect the structural integrity of the component. The source of these acoustic loads has been found to be acoustic resonance of the side branches on the MSLs, specifically, coupling of the vortex shedding frequency and natural acoustic frequency of safety relief valves (SRVs). The resonance that results from this coupling can contribute significant acoustic energy into the MSL system, which may propagate upstream into the reactor pressure vessel steam dome and drive structural vibration of steam dryer components. This can lead to high-cycle fatigue issues. Lock-in between the vortex shedding frequency and SRV natural frequency, as well as the ability for acoustic energy to propagate into the MSL system, are a function of many things, including the plant operating conditions, geometry of the MSL/SRV junction, and placement of SRVs with respect to each other on the MSLs. Comision Federal de Electricidad and Westinghouse designed, fabricated, and installed acoustic side branches (ASBs) on the MSLs which effectively act in the system as an energy absorber, where the acoustic standing wave generated in the side-branch is absorbed and dissipated inside the ASB. These ASBs have been very successful in reducing the amount of acoustic energy which propagates into the steam dome. In addition, modifications to the Laguna Verde Nuclear Power Plant Unit 2 steam dryer have been completed to reduce the stress levels in critical locations in the dryer. The objective of this paper is to describe the acoustic side branch concept and the design iterative processes that were undertaken at Laguna Verde Unit 2 to achieve a steam dryer design that meets the guidelines of the American Society of Mechanical Engineers, Boiler and Pressure

  14. Impact of flow induced vibration acoustic loads on the design of the Laguna Verde Unit 2 steam dryer

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, D. R.; Wellstein, L. F.; Theuret, R. C.; Han, Y.; Rajakumar, C. [Westinghouse Electric Company LLC, Cranberry Township, PA 16066 (United States); Amador C, C.; Sosa F, W., E-mail: forsytdr@westinghouse.com [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Km 42.5 Carretera Cardel-Nautla, 91680 Alto Lucero, Veracruz (Mexico)

    2015-09-15

    Industry experience with Boiling Water Reactors (BWRs) has shown that increasing the steam flow through the main steam lines (MSLs) to implement an extended power up rate (EPU) may lead to amplified acoustic loads on the steam dryer, which may negatively affect the structural integrity of the component. The source of these acoustic loads has been found to be acoustic resonance of the side branches on the MSLs, specifically, coupling of the vortex shedding frequency and natural acoustic frequency of safety relief valves (SRVs). The resonance that results from this coupling can contribute significant acoustic energy into the MSL system, which may propagate upstream into the reactor pressure vessel steam dome and drive structural vibration of steam dryer components. This can lead to high-cycle fatigue issues. Lock-in between the vortex shedding frequency and SRV natural frequency, as well as the ability for acoustic energy to propagate into the MSL system, are a function of many things, including the plant operating conditions, geometry of the MSL/SRV junction, and placement of SRVs with respect to each other on the MSLs. Comision Federal de Electricidad and Westinghouse designed, fabricated, and installed acoustic side branches (ASBs) on the MSLs which effectively act in the system as an energy absorber, where the acoustic standing wave generated in the side-branch is absorbed and dissipated inside the ASB. These ASBs have been very successful in reducing the amount of acoustic energy which propagates into the steam dome. In addition, modifications to the Laguna Verde Nuclear Power Plant Unit 2 steam dryer have been completed to reduce the stress levels in critical locations in the dryer. The objective of this paper is to describe the acoustic side branch concept and the design iterative processes that were undertaken at Laguna Verde Unit 2 to achieve a steam dryer design that meets the guidelines of the American Society of Mechanical Engineers, Boiler and Pressure

  15. SPES-2, the full-height, full-pressure, test facility simulating the AP600 plant: Main results from the experimental campaign

    International Nuclear Information System (INIS)

    Medich, C.; Rigamonti, M.; Martinelli, R.; Tarantini, M.; Conway, L.

    1995-01-01

    The SPES-2 is a full height, full pressure experimental test facility reproducing the Westinghouse AP600 reactor with a scaling factor of 1/395. The experimental plant, designed and operated by SIET in Piacenza, consists of a full simulation of the AP600 primary core cooling system including all the passive and active safety systems. In 1992, Westinghouse, in cooperation with ENEL, ENEA, SIET and ANSALDO developed an experimental program to test the integrated behavior of the AP600 passive safety systems. The SPES-2 test matrix, concluded in November 1994, has examined the AP600 passive safety system response for a range of small break LOCAs at different locations on the primary system and on the passive system lines; single steam generator tube ruptures with both passive and active non-safety systems, and a main steam line break transient to demonstrate the capability of passive safety systems for rapid cooldown. Each of the tests has provided detailed experimental results for verification of the capability of the analysis methods to predict the integrated passive safety system behavior

  16. Strategies for steam

    International Nuclear Information System (INIS)

    Hennagir, T.

    1996-01-01

    This article is a review of worldwide developments in the steam turbine and heat recovery steam generator markets. The Far East is driving the market in HRSGs, while China is driving the market in orders placed for steam turbine prime movers. The efforts of several major suppliers are discussed, with brief technical details being provided for several projects

  17. The Effect of Condensing Steam Turbine Exhaust Hood Body Geometry on Exhaust Performance Efficiency

    Science.gov (United States)

    Gribin, V. G.; Paramonov, A. N.; Mitrokhova, O. M.

    2018-06-01

    The article presents data from combined numerical and experimental investigations of the effect that the overall dimensions of the exhaust hood of a steam turbine with an underslung condenser has on the aerodynamic losses in the hood. Owing to the properly selected minimum permissible overall dimensions of the exhaust hood, more efficient operation of this turbine component is achieved, better vibration stability of the turbine set shaft line is obtained, and lower costs are required for arranging the steam turbine plant in the turbine building. Experiments have shown that the main overall dimensions of the hood body have a determining effect on the exhaust hood flow path profile and on its aerodynamic performance. Owing to properly selected ratios between the exhaust hood body main sizes without a diffuser, a total loss coefficient equal to approximately unity has been obtained. By using an axial-radial diffuser, the energy loss can be decreased by 30-40% depending on the geometrical parameters and level of velocities in the inlet section of a hood having the optimal overall dimensions. By using the obtained results, it becomes possible to evaluate the overall dimensions necessary for achieving the maximal aerodynamic hood efficiency and, as a consequence, to obtain better technical and economic indicators of the turbine plant as a whole already at the initial stage of its designing. If a need arises to select overall dimensions smaller than their optimal values, the increase of energy loss can be estimated using the presented dependences. The cycle of investigations was carried out on the experimental setups available in the fundamental research laboratory of the Moscow Power Engineering Institute National University's Department of Steam and Gas Turbines with due regard to the operating parameters and similarity criteria.

  18. Condensation of steam

    International Nuclear Information System (INIS)

    Prisyazhniuk, V.A.

    2002-01-01

    An equation for nucleation kinetics in steam condensation has been derived, the equation taking into account the concurrent and independent functioning of two nucleation mechanisms: the homogeneous one and the heterogeneous one. The equation is a most general-purpose one and includes all the previously known condensation models as special cases. It is shown how the equation can be used in analyzing the process of steam condensation in the condenser of an industrial steam-turbine plant, and in working out new ways of raising the efficiency of the condenser, as well as of the steam-turbine plant as a whole. (orig.)

  19. Flow-induced vibration and fretting-wear specifications to ensure steam-generator and heat exchanger lifetime performance

    International Nuclear Information System (INIS)

    Janzen, V.P.; Han, Y.; Pettigrew, M.J.

    2008-01-01

    The current interest in refurbishment, life extension and new-build activity has meant a renewed emphasis on technical specifications that will ensure improved reliability and longer life. Preventing vibration and fretting-wear problems in steam generators and heat exchangers requires design specifications that bring together specific guidelines, analysis methods, requirements and appropriate performance criteria. The specifications must be firmly based on experimental data and field inspections. In addition, the specifications must be supported by theoretical analyses and fundamental scaling correlations, to cover conditions and geometries over the wide range applicable to existing components and probable future designs. The specifications are expected to evolve to meet changing industry requirements. This paper outlines the steps required to generate and support design specifications, and relates them to typical steam-generator design features and computer modeling capabilities. It also describes current issues that are driving changes to flow-induced vibration and fretting-wear specifications that can be applied to the design process for component refurbishment, replacement or new designs. These issues include recent experimental or field evidence for new excitation mechanisms, e.g., the possibility of in-plane fluidelastic instability of U-tubes, the demand for longer reactor and component lifetimes, the need for better predictions of dynamic properties and vibration response, e.g., two-phase random-turbulence excitation, and requirements to consider system 'excursions' or abnormal scenarios, e.g., a main steam line break in the case of steam generators. The paper describes steps being taken to resolve these issues. (author)

  20. Super titanium blades for advanced steam turbines

    International Nuclear Information System (INIS)

    Coulon, P.A.

    1990-01-01

    In 1986, the Alsthom Steam Turbines Department launched the manufacture of large titanium alloy blades: airfoil length of 1360 mm and overall length of 1520 mm. These blades are designed for the last-stage low pressure blading of advanced steam turbines operating at full speed (3000 rpm) and rating between 300 and 800 MW. Using titanium alloys for steam turbine exhaust stages as substitutes for chrome steels, due to their high strength/density ratio and their almost complete resistance to corrosion, makes it possible to increase the length of blades significantly and correspondingly that steam passage section (by up to 50%) with a still conservative stresses level in the rotor. Alsthom relies on 8 years of experience in the field of titanium, since as early as 1979 large titanium blades (airfoil length of 1240 mm, overall length of 1430 mm) were erected for experimental purposes on the last stage of a 900 MW unit of the Dampierre-sur-Loire power plant and now totals 45,000 operating hours without problems. The paper summarizes the main properties (chemical, mechanical and structural) recorded on very large blades and is based in particular on numerous fatigue corrosion test results to justify the use of the Ti 6 Al 4 V alloy in a specific context of micrographic structure

  1. An automated eddy current in-service inspection system for nuclear steam generator tubing

    International Nuclear Information System (INIS)

    Wells, N.S.

    1981-06-01

    A prototype steam generator in-service inspection system incorporating remotely-controlled instrumentation linked by a digital transmission line to an instrument and control trailer outside the reactor containment has been designed and manufactured and is presently undergoing field tests. The (Monel 400) steam generator tubes are scanned two at a time using absolute eddy current probes controlled by two remotely-operated probe drives at a scanning speed of 0.5 m/s. The probes are positioned on the tubesheet by a light-weight (1.5 kg) microprocessor-operated tubesheet walker mechanism. Digitized control and data signals are transmitted up to 300 m to the control trailer. There the control and analysis computers extract the relevant signal information and present it in condensed form as labelled graphics on CRT consoles for on-line visual assessment. Hard copy output is also provided for each tube scanned (one per minute). Condensed data is archived on magnetic tapes for additional off-line analysis and comparisons with other inspections

  2. Steam generator

    International Nuclear Information System (INIS)

    Fenet, J.-C.

    1980-01-01

    Steam generator particularly intended for use in the coolant system of a pressurized water reactor for vaporizing a secondary liquid, generally water, by the primary cooling liquid of the reactor and comprising special arrangements for drying the steam before it leaves the generator [fr

  3. Steam generator design for solar towers using solar salt as heat transfer fluid

    Science.gov (United States)

    González-Gómez, Pedro Ángel; Petrakopoulou, Fontina; Briongos, Javier Villa; Santana, Domingo

    2017-06-01

    Since the operation of a concentrating solar power plant depends on the intermittent character of solar energy, the steam generator is subject to daily start-ups, stops and load variations. Faster start-up and load changes increase the plant flexibility and the daily energy production. However, it involves high thermal stresses on thick-walled components. Continuous operational conditions may eventually lead to a material failure. For these reasons, it is important to evaluate the transient behavior of the proposed designs in order to assure the reliability. The aim of this work is to analyze different steam generator designs for solar power tower plants using molten salt as heat transfer fluid. A conceptual steam generator design is proposed and associated heat transfer areas and steam drum size are calculated. Then, dynamic models for the main parts of the steam generator are developed to represent its transient performance. A temperature change rate that ensures safe hot start-up conditions is studied for the molten salt. The thermal stress evolution on the steam drum is calculated as key component of the steam generator.

  4. Steam generator with perfected dryers

    International Nuclear Information System (INIS)

    Fenet, J.C.

    1987-01-01

    This steam generator has vertically superposed array of steam dryers. These dryers return the steam flow of 180 0 . The return of the water is made by draining channels to the steam production zone [fr

  5. On-line low and high frequency acoustic leak detection and location for an automated steam generator protection system

    International Nuclear Information System (INIS)

    Gaubatz, D.C.; Gluekler, E.L.

    1990-01-01

    Two on-line acoustic leak detection systems were operated and installed on a 76 MW hockey stick steam generator in the Sodium Components Test Installation (SCTI) at the Energy Technology Engineering Center (ETEC) in Southern California. The low frequency system demonstrated the capability to detect and locate leaks, both intentional and unintentional. No false alarms were issued during the two year test program even with adjacent blasting activities, pneumatic drilling, shuttle rocket engine testing nearby, scrams of the SCTI facility, thermal/hydraulic transient testing, and pump/control valve operations. For the high frequency system the capability to detect water into sodium reactions was established utilizing frequencies as high as 300 kHz. The high frequency system appeared to be sensitive to noise generated by maintenance work and system valve operations. Subsequent development work which is incomplete as of this date showed much more promise for the high frequency system. (author). 13 figs

  6. How to compute the power of a steam turbine with condensation, knowing the steam quality of saturated steam in the turbine discharge

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez Albarran, Manuel Jaime; Krever, Marcos Paulo Souza [Braskem, Sao Paulo, SP (Brazil)

    2009-07-01

    To compute the power and the thermodynamic performance in a steam turbine with condensation, it is necessary to know the quality of the steam in the turbine discharge and, information of process variables that permit to identifying with high precision the enthalpy of saturated steam. This paper proposes to install an operational device that will expand the steam from high pressure point on the shell turbine to atmosphere, both points with measures of pressure and temperature. Arranging these values on the Mollier chart, it can be know the steam quality value and with this data one can compute the enthalpy value of saturated steam. With the support of this small instrument and using the ASME correlations to determine the equilibrium temperature and knowing the discharge pressure in the inlet of surface condenser, the absolute enthalpy of the steam discharge can be computed with high precision and used to determine the power and thermodynamic efficiency of the turbine. (author)

  7. U-tube steam generator empirical model development and validation using neural networks

    International Nuclear Information System (INIS)

    Parlos, A.G.; Chong, K.T.; Atiya, A.

    1992-01-01

    Empirical modeling techniques that use model structures motivated from neural networks research have proven effective in identifying complex process dynamics. A recurrent multilayer perception (RMLP) network was developed as a nonlinear state-space model structure along with a static learning algorithm for estimating the parameter associated with it. The methods developed were demonstrated by identifying two submodels of a U-tube steam generator (UTSG), each valid around an operating power level. A significant drawback of this approach is the long off-line training times required for the development of even a simplified model of a UTSG. Subsequently, a dynamic gradient descent-based learning algorithm was developed as an accelerated alternative to train an RMLP network for use in empirical modeling of power plants. The two main advantages of this learning algorithm are its ability to consider past error gradient information for future use and the two forward passes associated with its implementation. The enhanced learning capabilities provided by the dynamic gradient descent-based learning algorithm were demonstrated via the case study of a simple steam boiler power plant. In this paper, the dynamic gradient descent-based learning algorithm is used for the development and validation of a complete UTSG empirical model

  8. Regulatory analysis for the resolution of generic issue C---8, main steam isolation valve leakage and LCS [leakage control system] failure

    International Nuclear Information System (INIS)

    Graves, C.C.

    1990-06-01

    Generic Issue C-8 deals with staff concerns about public risk because of the incidence of leak test failures reported for main steam isolation valves (MSIVs) at boiling water reactors and the limitations of the leakage control systems (LCSs) for mitigating the consequences of leakage from these valves. If the MSIV leakage is greatly in excess of the allowable value in the technical specifications, the LCS would be unavailable because of design limitations. The issue was initiated in 1983 to assess (1) the causes of MSIV leakage failures, (2) the effectiveness of the LCS and alternative mitigation paths, and (3) the need for additional regulatory action to reduce public risk. This report presents the regulatory analysis for Generic Issue C-8 and concludes that no new regulatory requirements are warranted

  9. Steam generators of Phenix: Measurement of the hydrogen concentration in sodium for detecting water leaks in the steam generator tubes

    International Nuclear Information System (INIS)

    Cambillard, E.; Lacroix, A.; Langlois, J.; Viala, J.

    1975-01-01

    The Phenix secondary circuits are provided with measurement systems of hydrogen concentration in sodium, that allow for the detection of possible water leaks in steam generators and the location of a faulty module. A measurement device consists of : a detector with nickel membranes of 0, 3 mm wall thickness, an ion pump with a 200 l/s flow rate, a quadrupole mass spectrometer and a calibrated hydrogen leak. The temperature correction is made automatically. The main tests carried out on the leak detection systems are reported. Since the first system operation (October 24, 1973), the measurements allowed us to obtain the hydrogen diffusion rates through the steam generator tube walls. (author)

  10. Vibration Analysis for Steam Dryer of APR1400 Steam Generator

    Energy Technology Data Exchange (ETDEWEB)

    Han, Sung-heum; Ko, Doyoung [KHNP CRI, Daejeon (Korea, Republic of); Cho, Minki [Doosan Heavy Industry, Changwon (Korea, Republic of)

    2016-10-15

    This paper is related to comprehensive vibration assessment program for APR1400 steam generator internals. According to U.S. Nuclear Regulatory Commission, Regulatory Guide 1.20 (Rev.3, March 2007), we conducted vibration analysis for a steam dryer as the second steam separator of steam generator internals. The vibration analysis was performed at the 100 % power operating condition as the normal operation condition. The random hydraulic loads were calculated by the computational fluid dynamics and the structural responses were predicted by power spectral density analysis for the probabilistic method. In order to meet the recently revised U.S. NRC RG 1.20 Rev.3, the CVAP against the potential adverse flow effects in APR1400 SG internals should be performed. This study conducted the vibration response analysis for the SG steam dryer as the second moisture separator at the 100% power condition, and evaluated the structural integrity. The predicted alternating stress intensities were evaluated to have more than 17.78 times fatigue margin compared to the endurance limit.

  11. Device for starting a steam generator by heating sodium in a reactor

    International Nuclear Information System (INIS)

    Nakano, Hisao.

    1975-01-01

    Object: To enhance cooperation between ventilation and steam conditions of turbine and ventilation condition relative to a superheater at the time of starting a plant using a fast breeder, and to enhance safety with respect to failure of heat transmission tubes at the time of start. Structure: In a device in which steam generated in an evaporator is fed to a high pressure turbine through a super-heater and an outlet steam of high pressure turbine is reheated by means of a re-heater and fed into a turbine on the side of low pressure to drive the turbine for power generation, opening and closing valves are mounted on outlet and inlet pipes, respectively, of the heat transmission pipe in the super heater, said outlet and inlet pipes being connected by a bypass pipe. Upstream side of the opening and closing valve on the inlet pipe and the downstream side of the opening and closing valve on the outlet pipe and connected by a bypass pipe in the re-heater and said bypass pipe in the re-heater is provided with a steam heat exchanger to be heated by steam in the outlet of the superheater, and a steam line in an auxiliary boiler is connected to the side of re-heater from the opening and closing valve on the heat transmission pipe in the re-heater. (Hanada, M.)

  12. Sourcing of Steam and Electricity for Carbon Capture Retrofits.

    Science.gov (United States)

    Supekar, Sarang D; Skerlos, Steven J

    2017-11-07

    This paper compares different steam and electricity sources for carbon capture and sequestration (CCS) retrofits of pulverized coal (PC) and natural gas combined cycle (NGCC) power plants. Analytical expressions for the thermal efficiency of these power plants are derived under 16 different CCS retrofit scenarios for the purpose of illustrating their environmental and economic characteristics. The scenarios emerge from combinations of steam and electricity sources, fuel used in each source, steam generation equipment and process details, and the extent of CO 2 capture. Comparing these scenarios reveals distinct trade-offs between thermal efficiency, net power output, levelized cost, profit, and net CO 2 reduction. Despite causing the highest loss in useful power output, bleeding steam and extracting electric power from the main power plant to meet the CCS plant's electricity and steam demand maximizes plant efficiency and profit while minimizing emissions and levelized cost when wholesale electricity prices are below 4.5 and 5.2 US¢/kWh for PC-CCS and NGCC-CCS plants, respectively. At prices higher than these higher profits for operating CCS retrofits can be obtained by meeting 100% of the CCS plant's electric power demand using an auxiliary natural gas turbine-based combined heat and power plant.

  13. Steam Digest: Volume IV

    Energy Technology Data Exchange (ETDEWEB)

    2004-07-01

    This edition of the Steam Digest is a compendium of 2003 articles on the technical and financial benefits of steam efficiency, presented by the stakeholders of the U.S. Department of Energy's BestPractices Steam effort.

  14. Steam Digest Volume IV

    Energy Technology Data Exchange (ETDEWEB)

    None

    2004-07-01

    This edition of the Steam Digest is a compendium of 2003 articles on the technical and financial benefits of steam efficiency, presented by the stakeholders of the U.S. Department of Energy's BestPractices Steam effort.

  15. Fast reactor steam generators with sodium on the tube side. Design and operational parameters

    International Nuclear Information System (INIS)

    1994-01-01

    A comparison of design and operational characteristics as well as analysis of experience gained during the long terms operation of the Micro Module Inverse Steam Generator and Module Inverse Steam Generator at BOR 60 reactor are main aims of this technical report. 20 refs, 47 figs, 14 tabs

  16. Numerical and computational aspects of the coupled three-dimensional core/ plant simulations: organization for economic cooperation and development/ U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-II. 5. TMI-1 Benchmark Performed by Different Coupled Three-Dimensional Neutronics Thermal- Hydraulic Codes

    International Nuclear Information System (INIS)

    D'Auria, F.; Galassi, G.M.; Spadoni, A.; Gago, J.L.; Grgic, D.

    2001-01-01

    A comprehensive analysis of a double-ended main-steam-line-break (MSLB) accident assumed to have occurred in the Babcock and Wilcox Three Mile Island (TMI) Unit 1 nuclear power plant (NPP) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy. The research has been carried out in cooperation with the University of Zagreb, Croatia, and with partial financial support from the European Union through a grant to one of the authors. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development Committee on the Safety of Nuclear Installations-Nuclear Science Committee PWR MSLB Benchmark. Different code versions have been adopted in the analysis. Results from the following codes (or code versions) are described in this paper: 1. RELAP5/mod 3.2.2, gamma version, coupled with the three-dimensional (3-D) neutron kinetics PARCS code; 2. RELAP5/mod 3.2.2, gamma version, coupled with the 3-D neutron kinetics QUABBOX code; 3. RELAP5/3D code coupled with the 3-D neutron kinetics NESTLE code. Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by The Pennsylvania State University in cooperation with GPU Nuclear (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission (NRC). The main challenge for the calculation was the prediction of the return to power (RTP) following the inlet of cold water into the core and one 'stuck-withdrawn' control rod. Non-realistic assumptions were proposed to augment the core power peak following scram. Zero-dimensional neutronics codes were capable of detecting the RTP after scram. However, the application of 3-D neutronics codes to the same scenario allowed the calculation of a similar value for overall core power peak but showed power increase occurrence in about one-tenth of the core volume. The results achieved in phase 1 of

  17. Apparatus for the separation of water from water-steam mixtures

    International Nuclear Information System (INIS)

    Judith, H.; Schwerdtner, O. von.

    1975-01-01

    Steam flowing from the high-pressure part of a saturated-steam turbine of nuclear power stations to the preheater or steam directly passing off to the low-pressure part contains a high amount of moisture. This is removed by a separating device in the overflow pipe working as an axial cyclon. To this end a twist generator with radially mounted guide vanes forces a twisting movement on the water-steam mixture whereby the water component is thrown towards the wall of the overflow pipe. Behind the twist generator the overflow pipe is provided with ring slots or annular gaps through which the centrifuged water gets into water collecting chambers concentrically surrounding the overflow pipe. The main water seperation results from the first annular gap through centrifugal effects. The rest is removed by steam suction through the other gaps. For steam suction purposes, i.e. in order to produce an underpressure, the collecting chambers of these gaps are connected with the overflow pipe behind the twist generators by means of a suction pipe. In order to also remove small water droplets without increasing the twist, an agglomerator is installed in the overflow pipe before the twist generator. It consists of baffle or guide plates within an elliptic intermediate piece in a bend of the overflow pipe. Therefore the flanks of the guide plates run parallel to the flow direction of the steam. (DG/PB) [de

  18. Leak detection in LMFBR steam generators during operation

    International Nuclear Information System (INIS)

    Dumm, K.

    1978-01-01

    This paper deals with the following four main aspects: requirement on the leak detection of the SNR-300 steam generators; the hydrogen detector of SNR-300; remarks on the disadvantage of impurity detectors; and the first approach to acoustic leak detection systems

  19. Leak detection in LMFBR steam generators during operation

    Energy Technology Data Exchange (ETDEWEB)

    Dumm, K [INTERATOM, Bergisch Gladbach (Germany)

    1978-10-01

    This paper deals with the following four main aspects: requirement on the leak detection of the SNR-300 steam generators; the hydrogen detector of SNR-300; remarks on the disadvantage of impurity detectors; and the first approach to acoustic leak detection systems.

  20. Selling steam

    International Nuclear Information System (INIS)

    Zimmer, M.J.; Goodwin, L.M.

    1991-01-01

    This article addresses the importance of steam sales contract is in financing cogeneration facilities. The topics of the article include the Public Utility Regulatory Policies Act provisions and how they affect the marketing of steam from qualifying facilities, the independent power producers market shift, and qualifying facility's benefits

  1. Wet steam wetness measurement in a 10 MW steam turbine

    Directory of Open Access Journals (Sweden)

    Kolovratník Michal

    2014-03-01

    Full Text Available The aim of this paper is to introduce a new design of the extinction probes developed for wet steam wetness measurement in steam turbines. This new generation of small sized extinction probes was developed at CTU in Prague. A data processing technique is presented together with yielded examples of the wetness distribution along the last blade of a 10MW steam turbine. The experimental measurement was done in cooperation with Doosan Škoda Power s.r.o.

  2. A nodalization study of steam separator in real time simulation

    International Nuclear Information System (INIS)

    Horugshyang, Lein; Luh, R.T.J.; Zen-Yow, Wang

    1999-01-01

    The motive of this paper is to investigate the influence of steam separator nodalization on reactor thermohydraulics in terms of stability and level response. Three different nodalizations of steam separator are studied by using THEATRE and REMARK Code in a BWR simulator. The first nodalization is the traditional one with two nodes for steam separator. In this nodalization, the steam separation is modeled in the outer node, i.e., upper downcomer. Separated steam enters the Steen dome node and the liquid goes to the feedwater node. The second nodalization is similar to the first one with the steam separation modeled in the inner node. There is one additional junction connecting steam dome node and the inner node. The liquid fallback junction connects the inner node and feedwater node. The third nodalization is a combination of the former two with an integrated node for steam separator. Boundary conditions in this study are provided by a simplified feedwater and main steam driver. For comparison purpose, three tests including full power steady state initialisation, recirculation pumps runback and reactor scram are conducted. Major parameters such as reactor pressure, reactor level, void fractions, neutronic power and junction flows are recorded for analysis. Test results clearly show that the first nodalization is stable for steady state initialisation. However it has too responsive level performance in core flow reduction transients. The second nodalization is the closest representation of real plant structure, but not the performance. Test results show that an instability occurs in the separator region for both steady state initialisation and transients. This instability is caused by an unbalanced momentum in the dual loop configuration. The magnitude of the oscillation reduces as the power decreases. No superiority to the other nodalizations is shown in the test results. The third nodalization shows both stability and responsiveness in the tests. (author)

  3. Pressure drop, steam content and turbulent cross exchange in water/steam flows

    International Nuclear Information System (INIS)

    Teichel, H.

    1978-01-01

    For describing the behaviour of two-phase flows of water and steam with the help of calculating patterns, a number of empirical correlations are required. - In this article, correlations for the friction pressure drop in water/steam flows are compared, as well as for the steam mass and the volumetric steam content with each other and with the test results on simple geometries. As the mutual effect between cooling chanels plays an important part at the longitudinal flow through bar bundles, the appertaining equations are evaluated, in addition. (orig.) 891 HP [de

  4. Steam generator assessment for sustainable power plant operation

    International Nuclear Information System (INIS)

    Drexler, Andreas; Fandrich, Joerg; Ramminger, Ute; Montaner-Garcia, Violeta

    2012-09-01

    Water and steam serve in the water-steam cycle as the energy transport and work media. These fluids shall not affect, through corrosion processes on the construction materials and their consequences, undisturbed plant operation. The main objectives of the steam water cycle chemistry consequently are: - The metal release rates of the structural materials shall be minimal - The probability of selective / localized forms of corrosion shall be minimal. - The deposition of corrosion products on heat transfer surfaces shall be minimized. - The formation of aggressive media, particularly local aggressive environments under deposits, shall be avoided. These objectives are especially important for the steam generators (SGs) because their condition is a key factor for plant performance, high plant availability, life time extension and is important to NPP safety. The major opponent to that is corrosion and fouling of the heating tubes. Effective ways of counteracting all degradation problems and thus of improving the SG performance are to keep SGs in clean conditions or if necessary to plan cleaning measures such as mechanical tube sheet lancing or chemical cleaning. Based on more than 40 years of experience in steam-water cycle water chemistry treatment AREVA developed an overall methodology assessing the steam generator cleanliness condition by evaluating all available operational and inspection data together. In order to gain a complete picture all relevant water chemistry data (e.g. corrosion product mass balances, impurity ingress), inspection data (e.g. visual inspections and tube sheet lancing results) and thermal performance data (e.g. heat transfer calculations) are evaluated, structured and indexed using the AREVA Fouling Index Tool Box. This Fouling Index Tool Box is more than a database or statistical approach for assessment of plant chemistry data. Furthermore the AREVA's approach combines manufacturer's experience with plant data and operates with an

  5. Fuqing nuclear power of nuclear steam turbine generating unit No.1 at the implementation and feedback

    International Nuclear Information System (INIS)

    Cao Yuhua; Xiao Bo; He Liu; Huang Min

    2014-01-01

    The article introduces the Fuqing nuclear power of nuclear steam turbine generating unit no.l purpose, range of experience, experiment preparation, implementation, feedback and response. Turn of nuclear steam turbo-generator set flush, using the main reactor coolant pump and regulator of the heat generated by the electric heating element and the total heat capacity in secondary circuit of reactor coolant system (steam generator secondary side) of saturated steam turbine rushed to 1500 RPM, Fuqing nuclear power of nuclear steam turbine generating unit no.1 implementation of the performance of the inspection of steam turbine and its auxiliary system, through the test problems found in the clean up in time, the nuclear steam sweep turn smooth realization has accumulated experience. At the same time, Fuqing nuclear power of nuclear steam turbine generating unit no.1 at turn is half speed steam turbine generator non-nuclear turn at the first, with its smooth realization of other nuclear power steam turbine generator set in the field of non-nuclear turn play a reference role. (authors)

  6. On line chemical analyzers for high purity steam and water, applied to steam power plants; Analizadores quimicos en linea para agua y vapor de alta pureza, aplicados a centrales termoelectricas

    Energy Technology Data Exchange (ETDEWEB)

    Diaz Perez, Ruth [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1990-12-31

    This article presents a general overview of the advances in the subject of on line analyzers of chemical parameters for high purity water and steam and specifies which ones are commercially available. Also are mentioned besides, the criteria nowadays applied for the selection of the sites for sample grabbing and the analysis that is necessary to perform in each point, depending on the power plant type and the treatment administered (phosphates-Ph coordinated or AVT treatment). [Espanol] El articulo presenta un panorama general de los avances que en materia de analizadores de parametros quimicos en linea para agua y vapor de alta pureza, y especifica cuales estan disponibles en forma comercial. Se citan, ademas los criterios que se aplican actualmente para seleccionar los puntos de toma de muestra y los analisis que es necesario efectuar en cada punto, dependiendo del tipo de central y del tratamiento que se le administre (fosfatos-pH coordinado o tratamiento AVT).

  7. On line chemical analyzers for high purity steam and water, applied to steam power plants; Analizadores quimicos en linea para agua y vapor de alta pureza, aplicados a centrales termoelectricas

    Energy Technology Data Exchange (ETDEWEB)

    Diaz Perez, Ruth [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1989-12-31

    This article presents a general overview of the advances in the subject of on line analyzers of chemical parameters for high purity water and steam and specifies which ones are commercially available. Also are mentioned besides, the criteria nowadays applied for the selection of the sites for sample grabbing and the analysis that is necessary to perform in each point, depending on the power plant type and the treatment administered (phosphates-Ph coordinated or AVT treatment). [Espanol] El articulo presenta un panorama general de los avances que en materia de analizadores de parametros quimicos en linea para agua y vapor de alta pureza, y especifica cuales estan disponibles en forma comercial. Se citan, ademas los criterios que se aplican actualmente para seleccionar los puntos de toma de muestra y los analisis que es necesario efectuar en cada punto, dependiendo del tipo de central y del tratamiento que se le administre (fosfatos-pH coordinado o tratamiento AVT).

  8. Arrangement for separating water and steam in a through boiler

    International Nuclear Information System (INIS)

    Wittchow, E.

    1979-01-01

    An axial cyclone in the form of a spin generator with radial nozzles and a flow controller are installed directly in the overflow line between high and low pressure parts of a turbine. The water is centrifuged sideways via outlet slots into a water collecting chamber. The nozzles themselves are hollow and form the support for the flow controller. The extracted steam is taken via the nozzles to the flow controller and returned from this to the overflow line. (DG) 891 HP/DG 892 MB [de

  9. Experience in integration of on-line control systems for turbine plants at the Ignalina and Zaporozhe NPPs

    International Nuclear Information System (INIS)

    Gapunin, A.Ya.; Guchshin, V.I.; Kozlov, N.Yu.; Kravtsov, V.F.; Rasskazov, I.Eh.; Fridman, P.Yu.

    1989-01-01

    The main stages of designing and improvement of turbine on-line control systems (TOCS) are considered. TOCS is mainly designed for the optimization of NPP equipment operation modes taking into account power system requirements, as well as reactor and steam generator capabilities. TOCS of the Ignalina (K750-65/1300 turbine) and Zaporozhe (K1000-60/1500-2) NPPs are the systems of the third generation and are produced in the form of decentralized complex of software and hardware on the basis of the Ehlektronika S5-2200 computer. 3 refs.; 2 figs

  10. Modeling bubble condenser containment with computer code COCOSYS: post-test calculations of the main steam line break experiment at ELECTROGORSK BC V-213 test facility

    International Nuclear Information System (INIS)

    Lola, I.; Gromov, G.; Gumenyuk, D.; Pustovit, V.; Sholomitsky, S.; Wolff, H.; Arndt, S.; Blinkov, V.; Osokin, G.; Melikhov, O.; Melikhov, V.; Sokoline, A.

    2005-01-01

    Containment of the WWER-440 Model 213 nuclear power plant features a Bubble Condenser, a complex passive pressure suppression system, intended to limit pressure rise in the containment during accidents. Due to lack of experimental evidence of its successful operation in the original design documentation, the performance of this system under accidents with ruptures of large high-energy pipes of the primary and secondary sides remains a known safety concern for this containment type. Therefore, a number of research and analytical studies have been conducted by the countries operating WWER-440 reactors and their Western partners in the recent years to verify Bubble Condenser operation under accident conditions. Comprehensive experimental research studies at the Electrogorsk BC V-213 test facility, commissioned in 1999 in Electrogorsk Research and Engineering Centre (EREC), constitute essential part of these efforts. Nowadays this is the only operating large-scale facility enabling integral tests on investigation of the Bubble Condenser performance. Several large international research projects, conducted at this facility in 1999-2003, have covered a spectrum of pipe break accidents. These experiments have substantially improved understanding of the overall system performance and thermal hydraulic phenomena in the Bubble Condenser Containment, and provided valuable information for validating containment codes against experimental results. One of the recent experiments, denoted as SLB-G02, has simulated steam line break. The results of this experiment are of especial value for the engineers working in the area of computer code application for WWER-440 containment analyses, giving an opportunity to verify validity of the code predictions and identify possibilities for model improvement. This paper describes the results of the post-test calculations of the SLB-G02 experiment, conducted as a joint effort of GRS, Germany and Ukrainian technical support organizations for

  11. Steam Digest 2001

    Energy Technology Data Exchange (ETDEWEB)

    2002-01-01

    Steam Digest 2001 chronicles BestPractices Program's contributions to the industrial trade press for 2001, and presents articles that cover technical, financial and managerial aspects of steam optimization.

  12. ORTURB, HTGR Steam Turbine Dynamic for FSV Reactor

    International Nuclear Information System (INIS)

    Conklin, J.C.

    2001-01-01

    1 - Description of program or function: ORTURB was written specifically to calculate the dynamic behavior of the Fort St. Vrain (FSV) High- Temperature Gas-Cooled Reactor (HTGR) steam turbines. The program is divided into three main parts: the driver subroutine; turbine subroutines to calculate the pressure-flow balance of the high-, intermediate-, and low-pressure turbines; and feedwater heater subroutines. 2 - Method of solution: The program uses a relationship derived for ideal gas flow in an iterative fashion that minimizes computational time to determine the pressure and flow in the FSV steam turbines as a function of plant transient operating conditions. An important computer modeling characteristic, unique to FSV, is that the high-pressure turbine exhaust steam is used to drive the reactor core coolant circulators prior to entering the reheater. A feedwater heater dynamic simulation model utilizing seven state variables for each of the five heaters is included in the ORTURB computer simulation of the regenerative Rankine cycle steam turbines. The seven temperature differential equations are solved at each time- step using a matrix exponential method. 3 - Restrictions on the complexity of the problem: The turbine shaft is assumed to rotate at a constant (rated) speed of 3600 rpm. Energy and mass storage of steam in the high-, intermediate-, and low-pressure turbines is assumed to be negligible. These limitations exclude the use of ORTURB during a turbine transient such as startup from zero power or very low turbine flows

  13. Brazilian pediatric research groups, lines of research, and main areas of activity

    Directory of Open Access Journals (Sweden)

    Priscila H.A. Oliveira

    2015-06-01

    Full Text Available OBJECTIVES: The Brazilian scientific production in the pediatrics field has been increasing significantly. It is important to identify the distribution and activity of these groups in the country and the main study areas, contributing with data for better resource allocation by institutions. METHODS: An active research was conducted in the National Council of Technological and Scientific Development (Conselho Nacional de Desenvolvimento Científico e Tecnológico [CNPq] website, using as filters the macro area of the research group (Health Sciences, the area (Medicine, and descriptors related to pediatrics. Research lines and main area of pediatric research groups were classified according to the subject predominantly studied by each group. The scientific production of the leader of the pediatric research group between 2011 and 2014 was also analyzed. RESULTS: Most pediatric research groups in Brazil have more than five years of activity and are concentrated in the Southeast and South regions of the country; São Paulo, Rio Grande do Sul, and Minas Gerais are the states with most groups. Of the 132 specific pediatric research groups analyzed, 14.4% have lines of research in multiple areas and 11.4% in child and adolescent health. Among the 585 lines of research of these groups, the most prevalent areas were: oncology, infectious diseases, epidemiology, and gastroenterology. CONCLUSIONS: The pediatric research groups in Brazil have relevant scientific production, including works published in international publications, and are concentrated in regions with higher socioeconomic index. Most groups registered in CNPq started their activity in the last five years (46%, reflecting the recent growth of scientific production in this area.

  14. Status of Siemens steam generator design and measures to assure continuous long-term reliable operation

    International Nuclear Information System (INIS)

    Hoch, G.

    1999-01-01

    Operating pressurized water reactors with U-tube steam generators have encountered difficulties with either one or a combination of inadequate material selection, poor design or manufacturing and an insufficient water chemistry control which resulted in excessive tube degradation. In contrast to the above mentioned problems, steam generators from Siemens/KWU are proving by operating experience that all measures undertaken at the design stage as well as during the operating and maintenance phase were effective enough to counteract any tube corrosion phenomena or other steam generator related problem. An Integrated Service Concept has been developed, applied and wherever necessary improved in order to ensure reliable steam generator operation. The performance of the steam generators is updated continuously, evaluated and implemented in lifetime databases. The main indicator for steam generator integrity are the results of the eddy current testing of the steam generator tubes. Tubes with indications are rated with lifetime threshold values and if necessary plugged, based on individual assessment criteria.(author)

  15. Current experience and a new modeling on water hammer due to steam condensation in PWR secondary system

    International Nuclear Information System (INIS)

    Kawanishi, K.; Kasahara, J.; Ueno, T.; Suzuta, T.

    1998-01-01

    There have been possibilities to occur water hammer in pipelines of turbine system for nuclear or fossil fuel power plants. According to the NUREG report, approximately 150 events have been reported since 1969, we also have an experience recently. Water hammer occurs due to sudden steam condensation with pressure pulse. This kind of pressure pulses has been made by alternative producing and condensing of steam slug in the pipe and its frequency relates subcooling and pipe structures. This paper presents our current experience on water hammer with some experimental studies. The present experiment has been performed to obtain the data base for evaluating the pressure pulses. The test pipe was horizontal tubes with dead end connected to vertical tube which simulating drain line in PWR secondary system. The main results are shown as follows; Magnitude of pressure pulse depends drain velocity and initial subcooling. Pipe structure effects on the frequency and continual time of water hammer phenomenon. A new modeling for quantitative explanation of the phenomena is also presented

  16. A three-dimensional laboratory steam injection model allowing in situ saturation measurements. [Comparing steam injection and steam foam injection with nitrogen and without nitrogen

    Energy Technology Data Exchange (ETDEWEB)

    Demiral, B.M.R.; Pettit, P.A.; Castanier, L.M.; Brigham, W.E.

    1992-08-01

    The CT imaging technique together with temperature and pressure measurements were used to follow the steam propagation during steam and steam foam injection experiments in a three dimensional laboratory steam injection model. The advantages and disadvantages of different geometries were examined to find out which could best represent radial and gravity override flows and also fit the dimensions of the scanning field of the CT scanner. During experiments, steam was injected continuously at a constant rate into the water saturated model and CT scans were taken at six different cross sections of the model. Pressure and temperature data were collected with time at three different levels in the model. During steam injection experiments, the saturations obtained by CT matched well with the temperature data. That is, the steam override as observed by temperature data was also clearly seen on the CT pictures. During the runs where foam was present, the saturation distributions obtained from CT pictures showed a piston like displacement. However, the temperature distributions were different depending on the type of steam foam process used. The results clearly show that the pressure/temperature data alone are not sufficient to study steam foam in the presence of non-condensible gas.

  17. Fluid distribution network and steam generators and method for nuclear power plant training simulator

    International Nuclear Information System (INIS)

    Alliston, W.H.; Johnson, S.J.; Mutafelija, B.A.

    1975-01-01

    A description is given of a training simulator for the real-time dynamic operation of a nuclear power plant which utilizes apparatus that includes control consoles having manual and automatic devices corresponding to simulated plant components and indicating devices for monitoring physical values in the simulated plant. A digital computer configuration is connected to the control consoles to calculate the dynamic real-time simulated operation of the plant in accordance with the simulated plant components to provide output data including data for operating the control console indicating devices. In the method and system for simulating a fluid distribution network of the power plant, such as that which includes, for example, a main steam system which distributes steam from steam generators to high pressure turbine steam reheaters, steam dump valves, and feedwater heaters, the simultaneous solution of linearized non-linear algebraic equations is used to calculate all the flows throughout the simulated system. A plurality of parallel connected steam generators that supply steam to the system are simulated individually, and include the simulation of shrink-swell characteristics

  18. An energy signature scheme for steam trap assessment and flow rate estimation using pipe-induced acoustic measurements

    Science.gov (United States)

    Olama, Mohammed M.; Allgood, Glenn O.; Kuruganti, Teja P.; Lake, Joe E.

    2012-06-01

    The US Congress has passed legislation dictating that all government agencies establish a plan and process for improving energy efficiencies at their sites. In response to this legislation, Oak Ridge National Laboratory (ORNL) has recently conducted a pilot study to explore the deployment of a wireless sensor system for a real-time measurement-based energy efficiency optimization framework within the steam distribution system within the ORNL campus. We make assessments on the real-time status of the distribution system by observing the state measurements of acoustic sensors mounted on the steam pipes/traps/valves. In this paper, we describe a spectral-based energy signature scheme that interprets acoustic vibration sensor data to estimate steam flow rates and assess steam traps health status. Experimental results show that the energy signature scheme has the potential to identify different steam trap health status and it has sufficient sensitivity to estimate steam flow rate. Moreover, results indicate a nearly quadratic relationship over the test region between the overall energy signature factor and flow rate in the pipe. The analysis based on estimated steam flow and steam trap status helps generate alerts that enable operators and maintenance personnel to take remedial action. The goal is to achieve significant energy-saving in steam lines by monitoring and acting on leaking steam pipes/traps/valves.

  19. Modelling the horizontal steam generator with APROS

    Energy Technology Data Exchange (ETDEWEB)

    Ylijoki, J. [VTT Energy, Espoo (Finland); Palsinajaervi, C.; Porkholm, K. [IVO International Ltd, Vantaa (Finland)

    1995-12-31

    In this paper the capability of the five- and six-equation models of the simulation code APROS to simulate the behaviour of the horizontal steam generator is discussed. Different nodalizations are used in the modelling and the results of the stationary state runs are compared. Exactly the same nodalizations have been created for the five- and six-equation models. The main simulation results studied in this paper are void fraction and mass flow distributions in the secondary side of the steam generator. It was found that quite a large number of simulation volumes is required to simulate the distributions with a reasonable accuracy. The simulation results of the different models are presented and their validity is discussed. (orig.). 4 refs.

  20. Modelling the horizontal steam generator with APROS

    Energy Technology Data Exchange (ETDEWEB)

    Ylijoki, J [VTT Energy, Espoo (Finland); Palsinajaervi, C; Porkholm, K [IVO International Ltd, Vantaa (Finland)

    1996-12-31

    In this paper the capability of the five- and six-equation models of the simulation code APROS to simulate the behaviour of the horizontal steam generator is discussed. Different nodalizations are used in the modelling and the results of the stationary state runs are compared. Exactly the same nodalizations have been created for the five- and six-equation models. The main simulation results studied in this paper are void fraction and mass flow distributions in the secondary side of the steam generator. It was found that quite a large number of simulation volumes is required to simulate the distributions with a reasonable accuracy. The simulation results of the different models are presented and their validity is discussed. (orig.). 4 refs.

  1. Dynamic Modeling of Steam Condenser and Design of PI Controller Based on Grey Wolf Optimizer

    OpenAIRE

    Shu-Xia Li; Jie-Sheng Wang

    2015-01-01

    Shell-and-tube condenser is a heat exchanger for cooling steam with high temperature and pressure, which is one of the main kinds of heat exchange equipment in thermal, nuclear, and marine power plant. Based on the lumped parameter modeling method, the dynamic mathematical model of the simplified steam condenser is established. Then, the pressure PI control system of steam condenser based on the Matlab/Simulink simulation platform is designed. In order to obtain better performance, a new meta...

  2. Control device for steam turbine

    International Nuclear Information System (INIS)

    Hoshi, Hiroyuki.

    1993-01-01

    A power load imbalance detection circuit detects a power load imbalance when a load variation coefficient is large and output-load deviation is great. Then, it self-holds and causes a timer to start counting up and releases the self-holding after the elapse of a certain period of time. Upon load separation caused by system accidents, the power load imbalance detection circuit operates along with the increase of turbine rpm, to operate the control valve abrupt closing circuit and a bypassing value abrupt opening circuit. Then, self-holding of the power load imbalance detection circuit is released and, subsequently, a steam control value and a bypass valve are controlled by a control valve flow rate demand signal and a bypass flow rate demand signal determined by an entire main steam flow rate signal and a speed/load control signal. Accordingly, the turbine rpm is settled to about a rated rpm. This enables to avoid reactor shutdown upon occurrence of load interruption. (I.N.)

  3. Sodium and steam leak simulation studies for fluidized bed steam generators

    International Nuclear Information System (INIS)

    Keeton, A.R.; Vaux, W.G.; Lee, P.K.; Witkowski, R.E.

    1976-01-01

    An experimental program is described which was conducted to study the effects of sodium or steam leaking into an operating fluidized bed of metal or ceramic particles at 680 to 800 0 K. This effort was part of the early development studies for a fluidized-bed steam generator concept using helium as the fluidizing gas. Test results indicated that steam and small sodium leaks had no effect on the quality of fluidization, heat transfer coefficient, temperature distribution, or fluidizing gas pressure drop across the bed. Large sodium leaks, however, immediately upset the operation of the fluidized bed. Both steam and sodium leaks were detected positively and rapidly at an early stage of a leak by instruments specifically selected to accomplish this

  4. Solar energy for steam generation in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    De Carvalho, Jr, A V; Orlando, A DeF; Magnoli, D

    1979-05-01

    Steam generation is a solar energy application that has not been frequently studied in Brazil, even though for example, about 10% of the national primary energy demand is utilized for processing heat generation in the range of 100 to 125/sup 0/C. On the other hand, substitution of automotive gasoline by ethanol, for instance, has received much greater attention even though primary energy demand for process heat generation in the range of 100 to 125/sup 0/C is of the same order of magnitude than for total automotive gasoline production. Generation of low-temperature steam is analyzed in this article using distributed systems of solar collectors. Main results of daily performance simulation of single flat-plate collectors and concentrating collectors are presented for 20/sup 0/S latitude, equinox, in clear days. Flat plate collectors considered are of the aluminum roll-bond absorber type, selective surface single or double glazing. Considering feedwater at 20/sup 0/C, saturated steam at 120/sup 0/C and an annual solar utilization factor of 50%, a total collector area of about 3,000 m/sup 2/ is necessary for the 10 ton/day plant, without energy storage. A fuel-oil back-up system is employed to complement the solar steam production, when necessary. Preliminary economic evaluation indicates that, although the case-study shows today a long payback period relative to subsidized fuel oil in the domestic market (over 20 years in the city of Rio de Janeiro), solar steam systems may be feasible in the medium term due to projected increase of fuel oil price in Brazil.

  5. Calculation Method of Steam Generator Level for swelling and shrinking effects in YGN 1/2 Simulator

    International Nuclear Information System (INIS)

    Hwang, Do Hyun; Seo, In Yong; Park, Weon Seo; Suh, Jae Seung

    2007-01-01

    In August 2006, the development of new simulator for YGN 1/2 Simulator was completed. The NSSS (Nuclear Steam Supply System) T/H(Thermal- Hydraulic) module in this simulator was developed with ARTS code based on RETRAN, which is a best estimate thermal-hydraulic code designed to analyze several operational transients by EPRI(Electric Power Research Institute). RETRAN, however, has some limitations in real-time calculation capability and its robustness to be used in the simulator for some transient conditions. To overcome these limitations, its robustness and real-time calculation capability have been improved with simplifications and removing of discontinuities of the physical correlations of the RETRAN code. And some supplements are also developed to extend its simulation scope of the ARTS code. In comparison to KNPEC(Kori Nuclear Power Education Center) no.2 simulator, the simulator based on Younggwang Unit 1 developed in the year 2001, the ARTS code was upgraded that it extended its calculating region to the steam line and common header before turbine while it had calculated to the steam generator exit before steam line in KNPEC no.2 simulator. Consequently, the number of volume and fill/normal junction in ARTS nodalization increased to 109 and 174 from 62 and 125, respectively

  6. DEMONSTRATION BULLETIN STEAM ENHANCED REMEDIATION STEAM TECH ENVIRONMENTAL SERVICES, INC.

    Science.gov (United States)

    Steam Enhanced Remediation is a process in which steam is injected into the subsurface to recover volatile and semivolatile organic contaminants. It has been applied successfully to recover contaminants from soil and aquifers and at a fractured granite site. This SITE demonstra...

  7. Evaluation of yield and forage quality in main and ratoon crops of different sorghum lines

    Directory of Open Access Journals (Sweden)

    K.S. Vinutha

    2017-01-01

    Full Text Available Improving the yield and quality of sorghum (Sorghum bicolor forage for livestock feeding is a major breeding objective, because of sorghum’s inherently high biomass accumulation, high productivity per unit water utilized and its ability to produce a ratoon crop after harvesting of the plant crop. Newly bred sorghum lines, including 36 lines falling in 5 different categories, i.e. 12 experimental dual-purpose lines, 6 germplasm accessions from the ICRISAT collection, 11 commercial varieties and hybrids, 6 forage varieties and 1 bmr mutant line, were evaluated in terms of fodder yield, quality and ratooning ability. The main crop produced more dry biomass (P<0.05 at 80 days after planting (mean 22.87 t DM/ha; range 17.32‒33.82 t DM/ha than the ratoon crop (mean 8.47 t DM/ha; range 3.2‒17.42 t DM/ha after a further 80 days of growth. Mean nitrogen concentration in forage did not differ greatly between main and ratoon crops (2.56 vs. 2.40%, respectively but there was wide variation between lines (2.06‒2.89%. The line N 610 recorded highest N percentage of 2.89%, followed by SSG 59 3 (2.86% and SX 17 (2.81%. Highest acid detergent fiber % was recorded by ICSV 12008 (42.1%, closely followed by CO 31 and IS 34638 (40.0%. The least acid detergent lignin % was observed in MLSH-296 Gold (3.59%, ICSV 700 (3.75% and ICSSH 28 (3.83%. Metabolizable energy concentration was highest in N 610, Phule Yashodha and SX 17 (mean 8.34 MJ/kg DM, while in vitro organic matter digestibility ranged from 52.5 to 62.6%. The main crop contained much higher mean concentrations of the cyanogenic glycoside, dhurrin, than the ratoon (639 vs. 233 ppm, respectively with ranges of 38 to 2,298 ppm and 7 to 767 ppm, respectively. There was no significant correlation between dhurrin concentration and dry biomass yield so breeding and selection for low dhurrin concentrations should not jeopardize yields. Hence, breeding for sorghum can target simultaneously both quality and

  8. Operating experience with an on-line vibration control system for PWR main coolant pumps

    International Nuclear Information System (INIS)

    Runkel, J.; Stegemann, D.; Vortriede, A.

    1996-01-01

    The main circulation pumps are key components of nuclear power plants with pressurized water reactors, because the availability of the main circulation pumps has a direct influence on the availability and electrical output of the entire plant. The on-line automatic vibration control system ASMAS was developed for early failure detection during the normal operation of the main circulation pumps in order to avoid unexpected outages and to establish the possibility of preventive maintenance of the pumps. This system is permanently and successfully operating in three German 1300 MW el NPP's with PWR and has been successfully tested in a 350 MW el NPP with a PWR. (orig.)

  9. Operating experience with an on-line vibration control system for PWR main coolant pumps

    International Nuclear Information System (INIS)

    Runkel, J.; Stegemann, D.; Vortriede, A.

    1998-01-01

    The main circulation pumps are key components of nuclear power plants with pressurized water reactors (PWRs), because the availability of the main circulation pumps has a direct influence on the availability and electrical output of the entire plant. The on-line automatic vibration control system ASMAS was developed for early failure detection during the normal operation of the main circulation pumps in order to avoid unexpected outages and to establish the possibility of preventive maintenance of the pumps. This system is permanently and successfully operating in three German 1300 MW e1 NPP's with PWR and has been successfully tested in a 350 MW e1 NPP with a PWR. (orig.)

  10. Preliminary condensation pool experiments with steam using DN80 and DN100 blowdown pipes

    International Nuclear Information System (INIS)

    Laine, J.; Puustinen, M.

    2004-03-01

    The report summarizes the results of the preliminary steam blowdown experiments. Altogether eight experiment series, each consisting of several steam blows, were carried out in autumn 2003 with a scaled-down condensation pool test rig designed and constructed at Lappeenranta University of Technology. The main purpose of the experiments was to evaluate the capabilities of the test rig and the needs for measurement and visualization devices. The experiments showed that a high-speed video camera is essential for visual observation due to the rapid condensation of steam bubbles. Furthermore, the maximum measurement frequency of the current combination of instrumentation and data acquisition system is inadequate for the actual steam tests in 2004. (au)

  11. A balanced strategy in managing steam generator thermal performance

    International Nuclear Information System (INIS)

    Hu, M. H.; Nelson, P. R.

    2009-01-01

    This paper presents a balanced strategy in managing thermal performance of steam generator designed to deliver rated megawatt thermal (MWt) and megawatt electric (MWe) power without loss with some amount of thermal margin. A steam generator (SG) is a boiling heat exchanger whose thermal performance may degrade because of steam pressure loss. In other words, steam pressure loss is an indicator of thermal performance degradation. Steam pressure loss is mainly a result of either 1) tube scale induced poor boiling or 2) tube plugging historically resulting from tubing corrosion, wear due to flow induced tube vibration or loose parts impact. Thermal performance degradation was historically due to tube plugging but more recently it is due to poor boiling caused by more bad than good constituents of feedwater impurities. The whole SG industry still concentrates solely on maintenance programs towards preventing causes for tube plugging and yet almost no programs on maintaining adequate boiling of fouled tubes. There can be an acceptable amount of tube scale that provides excellent boiling capacity without tubing corrosion, as operational experience has repeatedly demonstrated. Therefore, future maintenance has to come up balanced programs for allocating limited resources in both maintaining good boiling capacity and preventing tube plugging. This paper discusses also thermal performance degradation due to feedwater impurity induced blockage of tube support plate and thus subsequent water level oscillations, and how to mitigate them. This paper provides a predictive management of tube scale for maintaining adequate steam pressure and stable water level without loss in MWt/MWe or recovering from steam pressure loss or water level oscillations. This paper offers a balanced strategy in managing SG thermal performance to fulfill its mission. Such a strategy is even more important in view of the industry trend in pursuing extended power uprate as high as 20 percent

  12. Analysis of a main steam isolation value closure anticipated transient without scram in a boiling water reactor

    International Nuclear Information System (INIS)

    Liaw, T.J.; Pan, C.; Chen, G.S.

    1989-01-01

    Anticipated transient without scram (ATWS) could be a major accident sequence with possible core melt and containment damage in a boiling water reactor (BWR). The behavior of a BWR/6 during a main steam isolation valve closure ATWS is investigated using the best-estimate computer program, RETRAN-02. The effects of both makeup coolant and boron injection on the reactor behavior are studied. It is found that the BWR/6 behaves similarly to the BWR/2 and BWR/4. Without boron injection and makeup coolant, the reactor loses its coolant inventory very quickly and the reactor power drops rapidly to ∼ 16% of rated power due to negative void reactivity. With coolant makeup from the high-pressure core spray and the reactor core isolation cooling systems, the rector reaches a quasi-steady-state condition after an initially rapidly changing transient. The dome pressure, downcomer water level, and core power oscillate around a mean value; the average core power is ∼ 15%, which is approximately equal to the power needed to heat and evaporate the subcooled makeup coolant. Lower boron concentrations in the core tend to complicate reactor behavior due to the combination of two competing phenomena: the negative boron reactivity and the positive reactivity caused by a void collapse

  13. Mechanical design of the hot steam headers of the THTR-300 steam generators

    International Nuclear Information System (INIS)

    Blumer, U.; Stumpf, M.

    1988-01-01

    The high pressure steam headers of the THTR steam generators have been subject to special attention during the design phase due to the following reasons: these components are the pressure retaining parts with the heaviest wall thickness in the region of the steam generators; they therefore are sensitive to thermal transient conditions; they are operated in the elevated temperature regime, where creep effects cannot be neglected; there is almost no service experience from fossil steam generators with this type of material (Alloy 800). Safety consideration therefore have been rather extensive and have focussed on two main areas which will be treated in this paper: 1. Analytical investigations on the cyclic material behaviour under all specified operating conditions, taking into account the non-elastic response of the material. 2. Limitation of the consequences of a header rupture by installation of heavy whip restraints. Elastic-plastic-creep analyses: The analyses were performed in different stages and are explained in the corresponding order: Evaluation of the critical location on the header and establishment of a simplified model of a nozzle region for further analysis. Preliminary thermal analyses of all specified transient conditions on simplified procedures, in order to establish a severity ranking of the conditions. Establishment of representative loading blocks. Evaluation of the material properties for thermal and structural, especially non-elastic behaviour. Detailed thermal analyses. Detailed structural analyses of the non-elastic cyclic response. Extrapolation for all cycles and assessment of the results by design codes. Discussion of the results. Header whip restraint design: In addition to the above analysis efforts, heavy whip restraints were provided to assure limitation of the effects of a header failure. This pager shows the measures that were taken to restrain the movement in case of longitudinal and transverse breaks: The anti-whip designs are

  14. Steam generators, turbines, and condensers. Volume six

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    Volume six covers steam generators (How steam is generated, steam generation in a PWR, vertical U-tube steam generators, once-through steam generators, how much steam do steam generators make?), turbines (basic turbine principles, impulse turbines, reaction turbines, turbine stages, turbine arrangements, turbine steam flow, steam admission to turbines, turbine seals and supports, turbine oil system, generators), and condensers (need for condensers, basic condenser principles, condenser arrangements, heat transfer in condensers, air removal from condensers, circulating water system, heat loss to the circulating water system, factors affecting condenser performance, condenser auxiliaries)

  15. Horizontal steam generator thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Ubra, O. [SKODA Praha Company, Prague (Czechoslovakia); Doubek, M. [Czech Technical Univ., Prague (Czechoslovakia)

    1995-09-01

    Horizontal steam generators are typical components of nuclear power plants with pressure water reactor type VVER. Thermal-hydraulic behavior of horizontal steam generators is very different from the vertical U-tube steam generator, which has been extensively studied for several years. To contribute to the understanding of the horizontal steam generator thermal-hydraulics a computer program for 3-D steady state analysis of the PGV-1000 steam generator has been developed. By means of this computer program, a detailed thermal-hydraulic and thermodynamic study of the horizontal steam generator PGV-1000 has been carried out and a set of important steam generator characteristics has been obtained. The 3-D distribution of the void fraction and 3-D level profile as functions of load and secondary side pressure have been investigated and secondary side volumes and masses as functions of load and pressure have been evaluated. Some of the interesting results of calculations are presented in the paper.

  16. Enhancement of enzymatic saccharification of Eucalyptus globulus: steam explosion versus steam treatment.

    Science.gov (United States)

    Martin-Sampedro, Raquel; Revilla, Esteban; Villar, Juan C; Eugenio, Maria E

    2014-09-01

    Steam explosion and steam pre-treatment have proved capable of enhancing enzymatic saccharification of lignocellulosic materials. However, until now, these methods had not been compared under the same operational conditions and using the same raw material. Both pre-treatments lead to increased yields in the saccharification of Eucalyptus globulus; but results have been better with steam pre-treatments, despite the more accessible surface of exploded samples. The reason for this finding could be enzymatic inhibition: steam explosion causes a more extensive extraction of hemicelluloses and releases a greater amount of degradation products which can inhibit enzymatic action. Enzymatic inhibition is also dependent on the amount and chemical structure of lignin, which was also a contributing factor to the lower enzymatic yields obtained with the most severe pre-treatment. Thus, the highest yields (46.7% glucose and 73.4% xylose yields) were obtained after two cycle of steam treatment, of 5 and 3 min, at 183°C. Copyright © 2014 Elsevier Ltd. All rights reserved.

  17. Analysis of ex-vessel steam explosion with MC3D

    International Nuclear Information System (INIS)

    Leskovar, M.; Mavko, B.

    2007-01-01

    An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. In the paper, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which was developed for the simulation of fuel-coolant interactions. A comprehensive parametric study was performed varying the location of the melt release (central, left and right side melt pour), the cavity water subcooling, the primary system overpressure at vessel failure and the triggering time for explosion calculations. The main purpose of the study was to determine the most challenging ex-vessel steam explosion cases in a typical pressurized water reactor and to estimate the expected pressure loadings on the cavity walls. The performed analysis shows that for some ex-vessel steam explosion scenarios significantly higher pressure loads are predicted than obtained in the OECD programme SERENA Phase 1. (author)

  18. Steam and sodium leak simulation in a fluidized-bed steam generator

    International Nuclear Information System (INIS)

    Vaux, W.G.; Keeton, A.R.; Keairns, D.L.

    1977-01-01

    A fluidized-bed steam generator for the liquid metal fast breeder reactor enhances plant availability and minimizes the probability of a water/sodium reaction. An experimental test program was conceived to assess design criteria and fluidized-bed operation under conditions of water, steam, and sodium leaks. Sodium, steam, and water were leaked into helium-fluidized beds of metal and ceramic particles at 900 F. Test results show the effects of leaks on the heat transfer coefficient, quality of fluidization, leak detection, and cleanup procedures

  19. Steam generator life-management, reliability, maintenance and refurbishment

    International Nuclear Information System (INIS)

    Spekkens, P.

    2012-01-01

    importance, operating chemistry and system cleanliness often do not receive the attention or understanding at the front lines that they deserve. Operate Clean - Build Clean is addressed in a number of papers at this conference B. 'Heat Exchangers, Valves and the Multitude of Other Off-Reactor Components': Notwithstanding the importance of fuel channels and feeders, the handling of these other components is also critical to a successful refurbishment. A number of papers at this conference address this. C. 'Effectively Engaging Service-Providers: 'Effectively Engaging' being the key words here. D. 'The Fundamentals of Thermal Hydraulic Design and Functional Architecture': This includes the details and expertise of fluid mechanics; operating chemistry; detailed design characteristics, etc. - getting the details right in these areas that are critical to achieving technical success or a satisfactory schedule and budget outcome. Plenary Presentations and the Lead Papers in the various sessions lay these items out more fully. Selected aspects of these themes as they relate to steam generators will be explored in this plenary paper. (author)

  20. Influence of the loop design of the feedwater- and steam quality in a power plant with pressurized water reactor

    International Nuclear Information System (INIS)

    Bennert, J.; Becher, L.

    1977-01-01

    At nuclear power plants with pressurized water reactors, condensate occurs on the high pressure part of the water-steam circuit, caused by the operation with low steam parameters. The behaviour of the electrolytes which entered into the circuit (solubility, distribution in water and/or steam) shows that these electrolytes (salts) are to be found mainly in the condensate. The insinuated electrolytes are reconcentrated during the common arrangements with 'Small Circuit' - consisting of steam generator, high pressure turbine, water separator, feedwater vessel, and have a negative influence on the feedwater - boiler water - and the steam quality. Remedy is possible by modified arrangements, during which these electrolyte-containing condensates will be treated and traced back into the main circuit. Nevertheless that the efficiency decrease is insignificant and additional efforts are necessary, a change over to these arrangements is recommendable, due to the fact that the feedwater quality, the boiler water quality, the steam quality in front of the turbine, and finally also the operational safety, as well as the availability will be improved. (orig.) [de

  1. Design of SMART steam generator cassette

    International Nuclear Information System (INIS)

    Kim, Y. W.; Kim, J. I.; Jang, M. H.

    2001-01-01

    Basic design development for the steam generator to be installed in the integral reactor SMART has been performed. Optimization of the steam generator shape, determination of the basic dimension and confirmation of the structural strength have been carried out. Individual steam generator cassette can be replaced in the optimized design concept of steam generator. Shape design of the steam generator cassette has been done on the computer based on 3-D CAE strategy. The structural integrity of the developed steam generator was investigated by performing the dynamic analysis for the steam generator cassette, flow induced vibration analysis for the tube bundle, and the thermo-mechanical analysis for the module header and tube. As for the manufacturing of steam generator, the numerical and the experimental simulation have been carried to control the amount of spring back and to eliminate residual stress. SMART steam generator cassette was developed by a sequential research of the aforementioned activities

  2. Ethanol production from lignocellulosic materials. Fermentation and on-line analysis

    Energy Technology Data Exchange (ETDEWEB)

    Olsson, L.

    1994-04-01

    The fermentation performance of bacteria, yeast and fungi was investigated in lignocellulosic hydrolysates with the aim of finding microorganisms which both withstand the inhibitors and that have the ability to ferment pentoses. Firstly, the performance of Saccharomyces cidri, Saccharomyces cerevisiae, Lactobacillus brevis, Lactococcus lactis ssp lactis, Escherichia coli and Zymomonas mobilis was investigated in spent sulphite liquor and enzymatic hydrolysate of steam-pretreated willow. Secondly, the performance of natural and recombinant E. coli, Pichia stipitis, recombinant S. cerevisiae, S. cerevisiae in combination with xylose isomerase and Fusarium oxysporum was investigated in a xylose-rich acid hydrolysate of corn cob. Recombinant E. coli was the best alternative for fermentation of lignocellulosic hydrolysates, giving both high yields and productivities. The main drawback was that detoxification was necessary. The kinetics of the fermentation with recombinant E. coli KO11 was investigated in the condensate of steam-pretreated willow. A cost analysis of the ethanol production from willow was made, which predicted an ethanol production cost of 3.9 SEK/l for the pentose fermentation. The detoxification cost constituted 22% of this cost. The monitoring of three monosaccharides and ethanol in lignocellulosic hydro lysates is described. The monosaccharides were determined using immobilized pyranose oxidase in an on-line amperometric analyser. Immobilization and characterization of pyranose oxidase from Phanerochaete chrysosporium is also described. The ethanol was monitored on-line using a micro dialysis probe as an in situ sampling device. The dialysate components were then separated in a column liquid chromatographic system and the ethanol was selectively detected by an amperometric alcohol bio sensor. The determinations with on-line analysis methods agreed well with off-line methods. 248 refs, 4 figs, 12 tabs

  3. Steam generators - problems and prognosis

    International Nuclear Information System (INIS)

    Tapping, R.L.

    1997-05-01

    Steam-generator problems, largely a consequence of corrosion and fouling, have resulted in increased inspection requirements and more regulatory attention to steam-generator integrity. In addition, utilities have had to develop steam-generator life-management strategies, including cleaning and replacement, to achieve design life. This paper summarizes the pertinent data to 1993/1994, and presents an overview of current steam-generator management practices. (author)

  4. Steam generators: critical components in nuclear steam supply systems

    Energy Technology Data Exchange (ETDEWEB)

    Stevens-Guille, P D

    1974-02-28

    Steam generators are critical components in power reactors. Even small internal leaks result in costly shutdowns for repair. Surveys show that leaks have affected one half of all water-cooled reactors in the world with steam generators. CANDU reactors have demonstrated the highest reliability. However, AECL is actively evolving new technology in design, manufacture, inspection and operation to maintain reliability. (auth)

  5. VVER-1000 main steam line break analysis using the coupled code system DYN3D/ATHLET

    International Nuclear Information System (INIS)

    Kliem, Soeren; Hoehne, Thomas; Rohde, Ulrich; Weiss, Frank-Peter; Kozmenkov, Yaroslav

    2008-01-01

    Calculations using the coupled code system DYN3D/ATHLET were performed in the frame of the OECD/NEA MSLB benchmark for a VVER-1000 reactor. The coolant mixing inside the reactor pressure vessel was treated using a validated empirical mixing model implemented into the DYN3D/ATHLET code. Using very conservative boundary conditions (reduced scram worth, two stuck rods, running MCP throughout the whole transient) a return-to-power was predicted. For the assessment of the empirical mixing model a time dependent calculation using the computational fluid dynamics code CFX-10 was performed. For that analysis, a detailed model of the reactor pressure vessel consisting of the inlets nozzles, downcomer, lower plenum and a part of the core and having 4.67 million unstructured tetra cell elements was used. For the considered case with running main coolant pumps, this calculation shows a sector formation at the core inlet with a certain amount of mixing at the edges of the sector. A core calculation using these CFX results as boundary conditions predicted also a return-to-power with a maximum value being about 200 MW lower than in the coupled code calculation. This variation calculation confirms the applicability of the empirical mixing model. The comparison shows also, that in this way results with a reasonable degree of conservatism can be obtained. (authors)

  6. Steam power plant

    International Nuclear Information System (INIS)

    Campbell, J.W.E.

    1981-01-01

    This invention relates to power plant forced flow boilers operating with water letdown. The letdown water is arranged to deliver heat to partly expanded steam passing through a steam reheater connected between two stages of the prime mover. (U.K.)

  7. Modelling of steam condensation in the primary flow channel of a gas-heated steam generator

    International Nuclear Information System (INIS)

    Kawamura, H.; Meister, G.

    1982-10-01

    A new simulation code has been developed for the analysis of steam ingress accidents in high temperatures reactors which evaluates the heat transfer in a steam generator headed by a mixture of helium and water steam. Special emphasis is laid on the analysis of steam condensation in the primary circuit of the steam generator. The code takes wall and bulk condensation into account. A new method is proposed to describe the entrainment of water droplets in the primary gas flow. Some typical results are given. Steam condensation in the primary channel may have a significant effect on temperature distributions. The effect on the heat transferred by the steam generator, however, is found to be not so prominent as might be expected. The reason is discussed. A simplified code will also be described, which gives results with reasonable accuracy within much shorter execution times. This code may be used as a program module in a program simulating the total primary circuit of a high temperature reactor. (orig.) [de

  8. Steam turbines for the future

    International Nuclear Information System (INIS)

    Trassl, W.

    1988-01-01

    Approximately 75% of the electrical energy produced in the world is generated in power plants with steam turbines (fossil and nuclear). Although gas turbines are increasingly applied in combined cycle power plants, not much will change in this matter in the future. As far as the steam parameters and the maximum unit output are concerned, a certain consolidation was noted during the past decades. The standard of development and mathematical penetration of the various steam turbine components is very high today and is applied in the entire field: For saturated steam turbines in nuclear power plants and for steam turbines without reheat, with reheat and with double reheat in fossil-fired power plants and for steam turbines with and without reheat in combined cycle power plants. (orig.) [de

  9. The Invisibility of Steam

    Science.gov (United States)

    Greenslade, Thomas B., Jr.

    2014-01-01

    Almost everyone "knows" that steam is visible. After all, one can see the cloud of white issuing from the spout of a boiling tea kettle. In reality, steam is the gaseous phase of water and is invisible. What you see is light scattered from the tiny droplets of water that are the result of the condensation of the steam as its temperature…

  10. Deliberate ignition of hydrogen-air-steam mixtures in condensing steam environments

    International Nuclear Information System (INIS)

    Blanchat, T.K.; Stamps, D.W.

    1997-05-01

    Large scale experiments were performed to determine the effectiveness of thermal glow plug igniters to burn hydrogen in a condensing steam environment due to the presence of water sprays. The experiments were designed to determine if a detonation or accelerated flame could occur in a hydrogen-air-steam mixture which was initially nonflammable due to steam dilution but was rendered flammable by rapid steam condensation due to water sprays. Eleven Hydrogen Igniter Tests were conducted in the test vessel. The vessel was instrumented with pressure transducers, thermocouple rakes, gas grab sample bottles, hydrogen microsensors, and cameras. The vessel contained two prototypic engineered systems: (1) a deliberate hydrogen ignition system and (2) a water spray system. Experiments were conducted under conditions scaled to be nearly prototypic of those expected in Advanced Light Water Reactors (such as the Combustion Engineering (CE) System 80+), with prototypic spray drop diameter, spray mass flux, steam condensation rates, hydrogen injection flow rates, and using the actual proposed plant igniters. The lack of any significant pressure increase during the majority of the burn and condensation events signified that localized, benign hydrogen deflagration(s) occurred with no significant pressure load on the containment vessel. Igniter location did not appear to be a factor in the open geometry. Initially stratified tests with a stoichiometric mixture in the top showed that the water spray effectively mixes the initially stratified atmosphere prior to the deflagration event. All tests demonstrated that thermal glow plugs ignite hydrogen-air-steam mixtures under conditions with water sprays near the flammability limits previously determined for hydrogen-air-steam mixtures under quiescent conditions. This report describes these experiments, gives experimental results, and provides interpretation of the results. 12 refs., 127 figs., 16 tabs

  11. Optimal design of marine steam turbine

    International Nuclear Information System (INIS)

    Liu Chengyang; Yan Changqi; Wang Jianjun

    2012-01-01

    The marine steam turbine is one of the key equipment in marine power plant, and it tends to using high power steam turbine, which makes the steam turbine to be heavier and larger, it causes difficulties to the design and arrangement of the steam turbine, and the marine maneuverability is seriously influenced. Therefore, it is necessary to apply optimization techniques to the design of the steam turbine in order to achieve the minimum weight or volume by means of finding the optimum combination of design parameters. The math model of the marine steam turbine design calculation was established. The sensitivities of condenser pressure, power ratio of HP turbine with LP turbine, and the ratio of diameter with height at the end stage of LP turbine, which influence the weight of the marine steam turbine, were analyzed. The optimal design of the marine steam turbine, aiming at the weight minimization while satisfying the structure and performance constraints, was carried out with the hybrid particle swarm optimization algorithm. The results show that, steam turbine weight is reduced by 3.13% with the optimization scheme. Finally, the optimization results were analyzed, and the steam turbine optimization design direction was indicated. (authors)

  12. Steam-water separator

    International Nuclear Information System (INIS)

    Modrak, T.M.; Curtis, R.W.

    1978-01-01

    A two-stage steam-water separating device is introduced, where the second stage is made as a cyclone separator. The water separated here is collected in the first stage of the inner tube and is returned to the steam raising unit. (TK) [de

  13. EPRI steam generator programs

    International Nuclear Information System (INIS)

    Martel, L.J.; Passell, T.O.; Bryant, P.E.C.; Rentler, R.M.

    1977-01-01

    The paper describes the current overall EPRI steam generator program plan and some of the ongoing projects. Because of the recent occurrence of a corrosion phenomenon called ''denting,'' which has affected a number of operating utilities, an expanded program plan is being developed which addresses the broad and urgent needs required to achieve improved steam generator reliability. The goal of improved steam generator reliability will require advances in various technologies and also a management philosophy that encourages conscientious efforts to apply the improved technologies to the design, procurement, and operation of plant systems and components that affect the full life reliability of steam generators

  14. Questions raised in developing fast reactor steam generator designs

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, P A; Hayden, O

    1975-07-01

    The most important component in the achievement of satisfactory LMFBR reliability is the steam generator. When the failure statistics of other nuclear steam generators and the implications of a sodium water reaction are considered, there is some cause for concern. It is apparent that considerable improvement in technology is necessary and until more experience on operating plant is available a conservative design approach must be taken. Many solutions have been proposed, varying from forced circulation straight tube modular to large single vessel once through helical designs. The paper poses what are considered to be the main questions which arise when making a choice of fast reactor steam generator type and tube configuration. The aim is to promote discussion amongst the assembled experts on their relative design approaches and the importance placed upon the various factors in reaching our common goal of ensuring the success of the LMFBR in its essential role of conserving world energy resources. (author)

  15. Questions raised in developing fast reactor steam generator designs

    International Nuclear Information System (INIS)

    Taylor, P.A.; Hayden, O.

    1975-01-01

    The most important component in the achievement of satisfactory LMFBR reliability is the steam generator. When the failure statistics of other nuclear steam generators and the implications of a sodium water reaction are considered, there is some cause for concern. It is apparent that considerable improvement in technology is necessary and until more experience on operating plant is available a conservative design approach must be taken. Many solutions have been proposed, varying from forced circulation straight tube modular to large single vessel once through helical designs. The paper poses what are considered to be the main questions which arise when making a choice of fast reactor steam generator type and tube configuration. The aim is to promote discussion amongst the assembled experts on their relative design approaches and the importance placed upon the various factors in reaching our common goal of ensuring the success of the LMFBR in its essential role of conserving world energy resources. (author)

  16. Control concepts for direct steam generation in parabolic troughs

    Energy Technology Data Exchange (ETDEWEB)

    Valenzuela, Loreto; Zarza, Eduardo [CIEMAT, Plataforma Solar de Almeria, Tabernas (Almeria) (Spain); Berenguel, Manuel [Universidad de Almeria, Dept. de Lenguajes y Computacion, Almeria (Spain); Camacho, Eduardo F. [Universidad de Sevilla, Dept. de Ingenieria de Sistemas y Automatica, Sevilla (Spain)

    2005-02-01

    A new prototype parabolic-trough collector system was erected at the Plataforma Solar de Almeria (PSA) (1996-1998) to investigate direct steam generation (DSG) in a solar thermal power plant under real solar conditions. The system has been under evaluation for efficiency, cost, control and other parameters since 1999. The main objective of the control system is to obtain steam at constant temperature and pressure at the solar field outlet, so that changes in inlet water conditions and/or in solar radiation affect the amount of steam, but not its quality or the nominal plant efficiency. This paper presents control schemes designed and tested for two operating modes, 'Recirculation', for which a proportional-integral-derivative (PI/PID) control functions scheme has been implemented, and 'Once-through', requiring more complex control strategies, for which the scheme is based on proportional-integral (PI), feedforward and cascade control. Experimental results of both operation modes are discussed. (Author)

  17. Exergy Analysis of a Subcritical Reheat Steam Power Plant with Regression Modeling and Optimization

    Directory of Open Access Journals (Sweden)

    MUHIB ALI RAJPER

    2016-07-01

    Full Text Available In this paper, exergy analysis of a 210 MW SPP (Steam Power Plant is performed. Firstly, the plant is modeled and validated, followed by a parametric study to show the effects of various operating parameters on the performance parameters. The net power output, energy efficiency, and exergy efficiency are taken as the performance parameters, while the condenser pressure, main steam pressure, bled steam pressures, main steam temperature, and reheat steam temperature isnominated as the operating parameters. Moreover, multiple polynomial regression models are developed to correlate each performance parameter with the operating parameters. The performance is then optimizedby using Direct-searchmethod. According to the results, the net power output, energy efficiency, and exergy efficiency are calculated as 186.5 MW, 31.37 and 30.41%, respectively under normal operating conditions as a base case. The condenser is a major contributor towards the energy loss, followed by the boiler, whereas the highest irreversibilities occur in the boiler and turbine. According to the parametric study, variation in the operating parameters greatly influences the performance parameters. The regression models have appeared to be a good estimator of the performance parameters. The optimum net power output, energy efficiency and exergy efficiency are obtained as 227.6 MW, 37.4 and 36.4, respectively, which have been calculated along with optimal values of selected operating parameters.

  18. In-service inspection of nuclear reactor vessels and steam generators. Results and evolution of the technics

    International Nuclear Information System (INIS)

    Rapin, Michel; Saglio, Robert.

    1978-01-01

    Methods and original technics have been developed by the CEA for inspection of the primary coolant circuit of PWR. Multifrequency Eddy currents for inspection of steam generators tubes gudgeons and bolts; focussed ultrasonics to test all the welds of the reactor vessel and its cover of mixed welds of tanks and steam generators, pressurizer welds and gudgeons from the inside; gamma radiography of vessel mixed welds, televisual examination of the stainless steel lining of the reactor vessel and its cover. Use of these technics is made with specific automatic machines designed either for inspection of steam generator tubes or for complete inspection of the vessel. Several reactors were inspected with these devices [fr

  19. STEAM DALAM PEMBUATAN PAKAN UNTUK KOMODITAS AKUAKULTUR

    Directory of Open Access Journals (Sweden)

    Sukarman Sukarman

    2010-12-01

    Full Text Available Kualitas fisik pakan (pelet untuk hewan akuakultur sangat penting, karena akan dimasukkan ke dalam air dan diharapkan tidak banyak mencemari lingkungan. Salah satu faktor yang berpengaruh dalam menjaga kualitas fisik pakan adalah penambahan dan pengaturan steam pada saat proses pembuatan pelet. Steam adalah aliran gas yang dihasilkan oleh air pada saat mendidih. Steam dibagi menjadi 3 jenis yaitu steam basah, saturated steam, dan superheated steam. Steam yang digunakan dalam proses pembuatan pelet adalah saturated steam. Pengaruh penambahan steam pada kualitas pelet bisa mencapai 20%. Penambahan steam dengan jumlah dan kualitas yang tepat akan menghasilkan pelet berkualitas. Sedangkan jika pengaturan dan penambahannya tidak tepat, maka kualitas fisik pelet akan rendah dan kemungkinan bisa merusak kandungan nutrisi seperti vitamin dan protein. Penambahan steam yang benar bisa dilakukan di dalam kondisioner dengan mengatur retention time, sudut kemiringan paddle conditioner, kecepatan putaran bearing dan menjaga kualitas steam dari mesin boiler sampai dengan kondisioner.

  20. Synthesis and optimization of steam system networks. 2. Multiple steam levels

    CSIR Research Space (South Africa)

    Price, T

    2010-08-01

    Full Text Available The use of steam in heat exchanger networks (HENs) can be reduced by the application of heat integration with the intention of debottlenecking the steam boiler and indirectly reducing the water requirement [Coetzee and Majozi. Ind. Eng. Chem. Res...

  1. Analysis of steam-generator tube-rupture events combined with auxiliary-feedwater control-system failure for Three Mile Island-Unit 1 and Zion-Unit 1 pressurized water reactors

    International Nuclear Information System (INIS)

    Nassersharif, B.

    1986-01-01

    A steam-generator tube-rupture (SGTR) event combined with loss of all offsite alternating-current power and failure of the auxiliary-feedwater (AFW) control system has been investigated for the Three Mile Island-Unit 1 (TMI-1) and Zion-Unit 1 (Zion-1) pressurized water reactors. The Transient Reactor Analysis Code was used to simulate the accident sequence for each plant. The objectives of the study were to predict the plant transient response with respect to tube-rupture flow termination, extent of steam generator overfill, and thermal-hydraulic conditions in the steam lines. Two transient cases were calculated: (1) a TMI-1 SGTR and runaway-AFW transient, and (2) a Zion-1 SGTR and runaway-AFW transient. Operator actions terminated the tube-rupture flow by 1342 s (22.4 min) and 1440 s (24.0 min) for TMI-1 and Zion-1, respectively, but AFW injection was continued. The damaged steam generator (DSG) overfilled by 1273 s (21.2 min) for the TMI-1 calculation and by 1604 s (26.7 min) for the Zion-1 calculation. The DSG steam lines were completely filled by 1500 s (25 min) and 2000 s (33.3 min) for TMI-1 and Zion-1, respectively. The maximum subcooling in the steam lines was approx.63 K (approx.113 0 F) for TMI-1 and approx.44 K (approx.80 0 F) for Zion-1

  2. Steam generator tubing NDE performance

    International Nuclear Information System (INIS)

    Henry, G.; Welty, C.S. Jr.

    1997-01-01

    Steam generator (SG) non-destructive examination (NDE) is a fundamental element in the broader SG in-service inspection (ISI) process, a cornerstone in the management of PWR steam generators. Based on objective performance measures (tube leak forced outages and SG-related capacity factor loss), ISI performance has shown a continually improving trend over the years. Performance of the NDE element is a function of the fundamental capability of the technique, and the ability of the analysis portion of the process in field implementation of the technique. The technology continues to improve in several areas, e.g. system sensitivity, data collection rates, probe/coil design, and data analysis software. With these improvements comes the attendant requirement for qualification of the technique on the damage form(s) to which it will be applied, and for training and qualification of the data analysis element of the ISI process on the field implementation of the technique. The introduction of data transfer via fiber optic line allows for remote data acquisition and analysis, thus improving the efficiency of analysis for a limited pool of data analysts. This paper provides an overview of the current status of SG NDE, and identifies several important issues to be addressed

  3. Modeling, Simulation and Optimization of Hydrogen Production Process from Glycerol using Steam Reforming

    International Nuclear Information System (INIS)

    Park, Jeongpil; Cho, Sunghyun; Kim, Tae-Ok; Shin, Dongil; Lee, Seunghwan; Moon, Dong Ju

    2014-01-01

    For improved sustainability of the biorefinery industry, biorefinery-byproduct glycerol is being investigated as an alternate source for hydrogen production. This research designs and optimizes a hydrogen-production process for small hydrogen stations using steam reforming of purified glycerol as the main reaction, replacing existing processes relying on steam methane reforming. Modeling, simulation and optimization using a commercial process simulator are performed for the proposed hydrogen production process from glycerol. The mixture of glycerol and steam are used for making syngas in the reforming process. Then hydrogen are produced from carbon monoxide and steam through the water-gas shift reaction. Finally, hydrogen is separated from carbon dioxide using PSA. This study shows higher yield than former U.S.. DOE and Linde studies. Economic evaluations are performed for optimal planning of constructing domestic hydrogen energy infrastructure based on the proposed glycerol-based hydrogen station

  4. Steam generator development in France for the Super Phenix project; Generateurs de vapeur developpes en France pour Super Phenix

    Energy Technology Data Exchange (ETDEWEB)

    Robin, M G

    1975-07-01

    'Steam Generator Development for Super Phenix Project'. The development program of steam generators studied by Fives-Cail Babcock and Stein Industrie Companies, jointly with CEA end EDF, for the Super Phenix 1200 MWe Fast Breeder Power Plant, is presented. The main characteristics of both sodium heated steam generators are emphasized and experimental studies related to their key features are reported. (author)

  5. Main heat transfer components for SNR-300

    International Nuclear Information System (INIS)

    De Haas Van Dorsser, A.H.

    1976-01-01

    Early in the joint German-Belgium-Dutch fast breeder programme it was decided that all main components should be tested, if possible at full scale, before fabrication of the actual SNR-300 components. Descriptions are given of the results of testing, and subsequent modifications, of the pumps, intermediate heat exchangers, and steam generators. A full scale model of the primary pump, free surface vertical shaft centrifugal type, was constructed and tested in the 5000 cubic metres per hour pump test facility erected at Bensberg. A 70 MW model of an intermediate heat exchanger, straight tube type with floating head, was tested in the 50 MW steam generator test station at Hengelo. Also tested in the Hengelo facility was an almost full scale straight tube 50 MW steam generator and subsequently a 50 MW helical tube evaporator. The latter tests were of more than 3000 h operation and resulted in minor changes in design and manufacturing operation. (U.K.)

  6. Steam purity in PWRs

    International Nuclear Information System (INIS)

    Hopkinson, J.

    1982-01-01

    Impurities enter the secondary loop of the PWR through both makeup water from lake or well and cooling-water leaks in the condenser. These impurities can be carried to the steam generator, where they cause corrosion deposits to form. Corrosion products in steam are swept further through the system and become concentrated at the point in the low-pressure turbine where steam begins to condense. Several plants have effectively reduced impurities, and therefore corrosion, by installing a demineralizer for the makeup water, a resin-bed system to clean condensed steam from the condenser, and a deaerator to remove oxygen from the water and so lower the risk of system metal oxidation. 5 references, 1 figure

  7. Wet-steam erosion of steam turbine disks and shafts

    International Nuclear Information System (INIS)

    Averkina, N. V.; Zheleznyak, I. V.; Kachuriner, Yu. Ya.; Nosovitskii, I. A.; Orlik, V. G.; Shishkin, V. I.

    2011-01-01

    A study of wet-steam erosion of the disks and the rotor bosses or housings of turbines in thermal and nuclear power plants shows that the rate of wear does not depend on the diagrammed degree of moisture, but is determined by moisture condensing on the surfaces of the diaphragms and steam inlet components. Renovating the diaphragm seals as an assembly with condensate removal provides a manifold reduction in the erosion.

  8. PWR steam-side chemistry follow program. Final report

    International Nuclear Information System (INIS)

    Economy, G.; Wootten, M.J.

    1982-08-01

    Denting is the diametral reduction of Inconel 600 tubes at the support plates of several nuclear steam generators during operation. This tube deformation was shown to be caused by the expansion forces resulting from formation of voluminous oxide generated by corrosion of the carbon steel support plate. The denting phenomenon and potential methods for its inhibition were investigated in the laboratory using isothermal capsules and heated crevice (heat transfer) testing at temperatures relevant to operating steam generators. Both test methods simulate the tube-support plate geometry and materials. Mathematical models were developed to predict the denting behavior of isothermal capsules. Three additives were investigated as potential denting inhibitors. Soaking with calcium hydroxide or sodium phosphate under heat transfer conditions did not inhibit denting when the environment was recontaminated, in contrast to similar soaks in boric acid which inhibited denting for two to four weeks following recontamination. When the boric acid soak was followed by on-line boric acid treatment, inhibition was maintained

  9. High-efficiency condenser of steam from a steam-gas mixture

    Science.gov (United States)

    Milman, O. O.; Krylov, V. S.; Ptakhin, A. V.; Kondratev, A. V.; Yankov, G. G.

    2017-12-01

    The design of a module for a high-efficiency condenser of steam with a high content (up to 15%) of noncondensable gases (NCGs) with a nearly constant steam-gas mixture (SGM) velocity during the condensation of steam has been developed. This module provides the possibility to estimate the operational efficiency of six condenser zones during the motion of steam from the inlet to the SGM suction point. Some results of the experimental tests of the pilot high-efficiency condenser module are presented. The dependence of the average heat transfer coefficient k¯ on the volumetric NCG concentration v¯ has been derived. It is shown that the high-efficiency condenser module can provide a moderate decrease in k¯ from 4400-4600 to 2600-2800 W/(m2 K) at v¯ ≈ 0.5-9.0%. The heat transfer coefficient distribution over different module zones at a heat duty close to its nominal value has been obtained. From this distribution, it can be seen that the average heat transfer coefficient decreases to 2600 W/(m2 K) at an NCG concentration v¯ = 7.5%, but the first condenser sections ( 1- 3) retain high values of k¯ at a level of no lower than 3200 W/(m2 K), and the last sections operate less well, having k¯ at a level of 1700 W/(m2 K). The dependence of the average heat transfer coefficient on the water velocity in condenser tubes has been obtained at a nearly nominal duty such that the extrapolation of this dependence to the water velocity of 2 m/s may be expected to give k¯ = 5000 W/(m2 K) for relatively pure steam, but an increase in k¯ at v¯ = 8% will be smaller. The effect of the gas removal device characteristic on the operation of the high-efficiency condenser module is described. The design developed for the steam condenser of a gas-turbine plant with a power of 25 MW, a steam flow rate of 40.2 t/h, and a CO2 concentration of up to 12% with consideration for the results of performed studies is presented.

  10. Exergoeconomic analysis of small-scale biomass steam cogeneration

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez Sotomonte, Cesar Adolfo; Lora, Electo Eduardo Silva [Universidade Federal de Itajuba, MG (Brazil)], e-mails: c.rodriguez32@unifei.edu.br, electo@unifei.edu.br; Venturini, Osvaldo Jose; Escobar, Jose Carlos [Universidad Federal de Itajuba, MG (Brazil)], e-mail: osvaldo@unifei.edu.br

    2010-07-01

    The principal objective of this work is to develop a calculation process, based on the second law of thermodynamics, for evaluating the thermoeconomic potential of a small steam cogeneration plant using waste from pulp processing and/or sawmills as fuel. Four different configurations are presented and assessed. The exergetic efficiency of the cycles that use condensing turbines is found to be around 11%, which has almost 3 percent higher efficiency than cycles with back pressure turbines. The thermoeconomic equations used in this paper estimated the production costs varying the fuel price. The main results show that present cost of technologies in a small-scale steam cycle cogeneration do not justify the implementation of more efficient systems for biomass prices less than 100 R$/t. (author)

  11. Commissioning and maintenance experience on mechanical equipment in steam generators of captive power plant at HWP, Manuguru (Paper No. 5.3)

    International Nuclear Information System (INIS)

    Bhatnagar, R.; Sinha, Ashok; Mohan Rao, A.C.

    1992-01-01

    Heavy Water Project (Manuguru) is having a captive power plant to cater to the demands of steam and power for the main plant. During the commissioning and initial run of the steam generators and their auxiliaries, teething/initial problems were encountered in nearly all the equipment of the steam generators. This paper briefly describes some of the major problems faced during the commissioning of the steam generators. (author). 4 figs

  12. Steam injection for heavy oil recovery: Modeling of wellbore heat efficiency and analysis of steam injection performance

    International Nuclear Information System (INIS)

    Gu, Hao; Cheng, Linsong; Huang, Shijun; Li, Bokai; Shen, Fei; Fang, Wenchao; Hu, Changhao

    2015-01-01

    Highlights: • A comprehensive mathematical model was established to estimate wellbore heat efficiency of steam injection wells. • A simplified approach of predicting steam pressure in wellbores was proposed. • High wellhead injection rate and wellhead steam quality can improve wellbore heat efficiency. • High wellbore heat efficiency does not necessarily mean good performance of heavy oil recovery. • Using excellent insulation materials is a good way to save water and fuels. - Abstract: The aims of this work are to present a comprehensive mathematical model for estimating wellbore heat efficiency and to analyze performance of steam injection for heavy oil recovery. In this paper, we firstly introduce steam injection process briefly. Secondly, a simplified approach of predicting steam pressure in wellbores is presented and a complete expression for steam quality is derived. More importantly, both direct and indirect methods are adopted to determine the wellbore heat efficiency. Then, the mathematical model is solved using an iterative technique. After the model is validated with measured field data, we study the effects of wellhead injection rate and wellhead steam quality on steam injection performance reflected in wellbores. Next, taking cyclic steam stimulation as an example, we analyze steam injection performance reflected in reservoirs with numerical reservoir simulation method. Finally, the significant role of improving wellbore heat efficiency in saving water and fuels is discussed in detail. The results indicate that we can improve the wellbore heat efficiency by enhancing wellhead injection rate or steam quality. However, high wellbore heat efficiency does not necessarily mean satisfactory steam injection performance reflected in reservoirs or good performance of heavy oil recovery. Moreover, the paper shows that using excellent insulation materials is a good way to save water and fuels due to enhancement of wellbore heat efficiency

  13. STEAM INJECTION INTO FRACTURED LIMESTONE AT LORING AIR FORCE BASE

    Science.gov (United States)

    A research project on steam injection for the remediation of spent chlorinated solvents from fractured limestone was recently undertaken at the former Loring AFB in Limestone, ME. Participants in the project include the Maine Department of Environmental Protection, EPA Region I,...

  14. Bed Agglomeration During the Steam Gasification of a High Lignin Corn Stover Simultaneous Saccharification and Fermentation (SSF) Digester Residue

    Energy Technology Data Exchange (ETDEWEB)

    Howe, Daniel T.; Taasevigen, Danny J.; Gerber, Mark A.; Gray, Michel J.; Fernandez, Carlos A.; Saraf, Laxmikant; Garcia-Perez, Manuel; Wolcott, Michael P.

    2015-11-13

    This research investigates the bed agglomeration phenomena during the steam gasification of a high lignin residue produced from the simultaneous saccharification and fermentation (SSF) of corn stover in a bubbling fluidized bed. The studies were conducted at 895°C using alumina as bed material. Biomass was fed at 1.5 kg/hr, while steam was fed to give a velocity equal to 2.5 times the minimum fluidization velocity, with a steam/carbon ratio of 0.9. The pelletized feedstock was co-fed with a cooling nitrogen stream to mitigate feed line plugging issues. Tar production was high at 50.3 g/Nm3, and the fraction of C10+ compounds was greater than that seen in the gasification of traditional lignocellulosic feedstocks. Carbon closures over 94 % were achieved for all experiments. Bed agglomeration was found to be problematic, indicated by pressure drop increases observed below the bed and upstream of the feed line. Two size categories of solids were recovered from the reactor, +60 mesh and -60 mesh. After a 2.75-hour experiment, 61.7 wt % was recovered as -60 mesh particles and 38.2 wt% of the recovered reactor solids were +60 mesh. A sizeable percentage, 31.8 wt%, was +20 mesh. The -60 mesh particles were mainly formed by the initial bed material (Al2O3). Almost 50 wt. % of the + 20 mesh particles was found to be formed by organics. The unreacted carbon remaining in the reactor resulted in a low conversion rate to product gas. ICP-AES, SEM, SEM-EDS, and XRD confirmed that the large agglomerates (+ 20 mesh) were not encapsulated bed material but rather un-gasified feedstock pellets with sand particles attached to it.

  15. Strategy for assessment of WWER steam generator tube integrity. Report prepared within the framework of the coordinated research project on verification of WWER steam generator tube integrity

    International Nuclear Information System (INIS)

    2007-12-01

    improve structural integrity assessment of steam generators of WWER-440/1000 NPPs. This TECDOC describes the main achievements of the CRP, that is, a proven approach to steam generator integrity assessment which consists in three critical elements: degradation assessment, condition monitoring, and operational assessment. This approach can provide assurance that the steam generators will continue to satisfy the appropriate performance criteria

  16. Hybrid preheat/recirculating steam generator

    International Nuclear Information System (INIS)

    Lilly, G.P.

    1985-01-01

    The patent describes a hybrid preheat/recirculating steam generator for nuclear power plants. The steam generator utilizes recirculated liquid to preheat incoming liquid. In addition, the steam generator incorporates a divider so as to limit the amount of recirculating water mixed with the feedwater. (U.K.)

  17. High temperature technological heat exchangers and steam generators with helical coil assembly tube bundle

    International Nuclear Information System (INIS)

    Korotaev, O.J.; Mizonov, N.V.; Nikolaevsky, V.B.; Nazarov, E.K.

    1990-01-01

    Analysis of thermal hydraulics characteristics of nuclear steam generators with different tube bundle arrangements and waste heat boilers for ammonia production units was performed on the basis of operating experience results and research and development data. The present report involves the obtained information. The estimations of steam generator performances and repair-ability are given. The significant temperature profile of the primary and secondary coolant flows are attributed to all steam generator designs. The intermediate mixing is found to be an effective means of temperature profile overcoming. At present the only means to provide an effective mixing in heat exchangers of the following types: straight tubes, field tubes, platen tubes and multibank helical coil tubes (with complicated bend distribution along their length) are section arrangements in series in conjunction with forced and natural mixing in connecting lines. Development of the unificated system from mini helical coil assemblies allows to design and manufacture heat exchangers and steam generators within the wide range of operating conditions without additional expenses on the research and development work

  18. Functionalized Graphene Enables Highly Efficient Solar Thermal Steam Generation.

    Science.gov (United States)

    Yang, Junlong; Pang, Yunsong; Huang, Weixin; Shaw, Scott K; Schiffbauer, Jarrod; Pillers, Michelle Anne; Mu, Xin; Luo, Shirui; Zhang, Teng; Huang, Yajiang; Li, Guangxian; Ptasinska, Sylwia; Lieberman, Marya; Luo, Tengfei

    2017-06-27

    The ability to efficiently utilize solar thermal energy to enable liquid-to-vapor phase transition has great technological implications for a wide variety of applications, such as water treatment and chemical fractionation. Here, we demonstrate that functionalizing graphene using hydrophilic groups can greatly enhance the solar thermal steam generation efficiency. Our results show that specially functionalized graphene can improve the overall solar-to-vapor efficiency from 38% to 48% at one sun conditions compared to chemically reduced graphene oxide. Our experiments show that such an improvement is a surface effect mainly attributed to the more hydrophilic feature of functionalized graphene, which influences the water meniscus profile at the vapor-liquid interface due to capillary effect. This will lead to thinner water films close to the three-phase contact line, where the water surface temperature is higher since the resistance of thinner water film is smaller, leading to more efficient evaporation. This strategy of functionalizing graphene to make it more hydrophilic can be potentially integrated with the existing macroscopic heat isolation strategies to further improve the overall solar-to-vapor conversion efficiency.

  19. STEAM by Design

    Science.gov (United States)

    Keane, Linda; Keane, Mark

    2016-01-01

    We live in a designed world. STEAM by Design presents a transdisciplinary approach to learning that challenges young minds with the task of making a better world. Learning today, like life, is dynamic, connected and engaging. STEAM (Science, Technology, Environment, Engineering, Art, and Math) teaching and learning integrates information in…

  20. Application of Computer Simulation to Identify Erosion Resistance of Materials of Wet-steam Turbine Blades

    Science.gov (United States)

    Korostelyov, D. A.; Dergachyov, K. V.

    2017-10-01

    A problem of identifying the efficiency of using materials, coatings, linings and solderings of wet-steam turbine rotor blades by means of computer simulation is considered. Numerical experiments to define erosion resistance of materials of wet-steam turbine blades are described. Kinetic curves for erosion area and weight of the worn rotor blade material of turbines K-300-240 LMP and atomic icebreaker “Lenin” have been defined. The conclusion about the effectiveness of using different erosion-resistant materials and protection configuration of rotor blades is also made.

  1. Cost analysis for application of solidified waste fission product canisters in U.S. Army steam plants

    International Nuclear Information System (INIS)

    Sande, W.E.; Bjorklund, W.J.; Brooks, N.A.

    1977-04-01

    The main objectives of the present study are to design steam plants using projected waste fission product canister characteristics, to analyze the overall impact and cost/benefit to the nuclear fuel cycle associated with these plants, and to develop plans for this application if the cost analysis so warrants it. The construction and operation of a steam plant fueled with waste fission product canisters would require the involvement and cooperation of various government agencies and private industry; thus the philosophies of these groups were studied. These philosophies are discussed, followed by a forecast of canister supply, canister characteristics, and strategies for Army canister use. Another section describes the safety and licensing of these steam plants since this affects design and capital costs. The discussion of steam plant design includes boiler concepts, boiler heat transfer, canister temperature distributions, steam plant size, and steam plant operation. Also, canister transportation is discussed since this influences operating costs. Details of economics of Army steam plants are provided including steam plant capital costs, operating costs, fuel reprocessor savings due to Army canister storage, and overall economics. Recommendations are made in the final section

  2. Determination of stresses caused by fluctuation of acoustic load in the steam dryers of a BWR; Determinacion de esfuerzos originados por fluctuacion de carga acustica en los secadores de vapor de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Centeno P, J.; Quezada G, S.; Prieto G, A.; Vazquez R, A.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico); Nunez C, A., E-mail: javcuami26@hotmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2014-10-15

    The extended power up-rate (EPU) in a nuclear power plant cause various problems in BWR components also in the steam system. This due to increased steam flow generated in the reactor and is conveyed to the turbine by the four main steam lines (MSL). One of the most serious problems is the generation of acoustic pressure loads in the metal structure of the steam dryer which eventually leads to fatigue failure and even the appearance of cracks, and in turn it causes loose parts that are entrained by the steam and transported in the MSL. This problem is due to the fluctuation of load acoustics caused by the union of the safety or relief valves (SRV) with the MSL, spreading through these to reach the reactor pressure vessel (RPV) where the effect of resonance of the acoustic wave is amplified and impacts directly in the supporting structure of the steam dryer, skirt and the panels where the mixture liquid-steam is dried, by centrifugation effect and runoff of liquid water. Efforts in the steam dryer operating conditions of EPU for two cases will be analyzed in this work, the first is before the installation of Acoustic Side Branch (ASB), and in the second case we consider the installation of said ASB in the standpipes of SRV. The analysis was performed with numerical experiments on a platform for computational fluid dynamics with virtual geometries previously designed based on the actual components of the reactor and steam system. The model to study is delimited by the top of the RPV, the steam dryer and a section of each of the four MSL with ten standpipes of SRV. With the obtained data and considering the mechanical-structural properties of the steam dryer material, we can evaluate the mechanical resistance to impacts by acoustic pressure load and its possible deformation or cracking. (Author)

  3. Steam generators secondary side chemical cleaning at Point Lepreau using the Siemens high temperature process

    International Nuclear Information System (INIS)

    Verma, K.; MacNeil, C.; Odar, S.; Kuhnke, K.

    1997-01-01

    This paper describes the chemical cleaning of the four steam generators at the Point Lepreau facility, which was accomplished as a part of a normal service outage. The steam generators had been in service for twelve years. Sludge samples showed the main elements were Fe, P and Na, with minor amounts of Ca, Mg, Mn, Cr, Zn, Cl, Cu, Ni, Ti, Si, and Pb, 90% in the form of Magnetite, substantial phosphate, and trace amounts of silicates. The steam generators were experiencing partial blockage of broached holes in the TSPs, and corrosion on tube ODs in the form of pitting and wastage. In addition heat transfer was clearly deteriorating. More than 1000 kg of magnetite and 124 kg of salts were removed from the four steam generators

  4. Steam Pressure-Reducing Station Safety and Energy Efficiency Improvement Project

    Energy Technology Data Exchange (ETDEWEB)

    Lower, Mark D [ORNL; Christopher, Timothy W [ORNL; Oland, C Barry [ORNL

    2011-06-01

    The Facilities and Operations (F&O) Directorate is sponsoring a continuous process improvement (CPI) program. Its purpose is to stimulate, promote, and sustain a culture of improvement throughout all levels of the organization. The CPI program ensures that a scientific and repeatable process exists for improving the delivery of F&O products and services in support of Oak Ridge National Laboratory (ORNL) Management Systems. Strategic objectives of the CPI program include achieving excellence in laboratory operations in the areas of safety, health, and the environment. Identifying and promoting opportunities for achieving the following critical outcomes are important business goals of the CPI program: improved safety performance; process focused on consumer needs; modern and secure campus; flexibility to respond to changing laboratory needs; bench strength for the future; and elimination of legacy issues. The Steam Pressure-Reducing Station (SPRS) Safety and Energy Efficiency Improvement Project, which is under the CPI program, focuses on maintaining and upgrading SPRSs that are part of the ORNL steam distribution network. This steam pipe network transports steam produced at the ORNL steam plant to many buildings in the main campus site. The SPRS Safety and Energy Efficiency Improvement Project promotes excellence in laboratory operations by (1) improving personnel safety, (2) decreasing fuel consumption through improved steam system energy efficiency, and (3) achieving compliance with applicable worker health and safety requirements. The SPRS Safety and Energy Efficiency Improvement Project being performed by F&O is helping ORNL improve both energy efficiency and worker safety by modifying, maintaining, and repairing SPRSs. Since work began in 2006, numerous energy-wasting steam leaks have been eliminated, heat losses from uninsulated steam pipe surfaces have been reduced, and deficient pressure retaining components have been replaced. These improvements helped ORNL

  5. The decommissioning of the BR3 steam generator

    International Nuclear Information System (INIS)

    Denissen, L.

    2006-01-01

    A steam generator is a crucial component in a PWR (Pressurized Water Reactor). It is the crossing between the primary, contaminated, circuit and the secondary water-steam circuit. The heat from the primary reactor coolant loop is transferred to the secondary side in thousands of small tubes. Due to several problems in the material of those tubes, like SCC (Stress Corrosion Cracking), insufficient control in water chemistry, which can be the cause of tube leakage, more and more steam generators are replaced today. Only in Belgium, already 17 of them are replaced. The old 300 ton heavy SGs are stored at the 2 nuclear power plants of Doel and Tihange . While it was foreseen in the BR3 strategy to dismantle the steam generator (only 30 ton), we took the opportunity to search for a complete package in the decommissioning of a steam generator. The complete management consists of a decontamination of the primary side followed by the complete dismantling. The first step, the decontamination with MEDOC (water box + tube bundle) has already been achieved in 2002. It has led to an important DF (Decontamination Factor) between 100 and 1000 and an important dose rate reduction. This hard chemical decontamination process has been described earlier in the scientific report 2002 (The BR3 steam generator decontamination with the MEDOC process - M. Ponnet). The second step, the complete dismantling of the SG has been executed in 2005. With the BR3 SG, the main goal was to dismantle it in a safe way and to free release a maximum of material. We've used two cutting tools to perform the job: A HPWJC (High Pressure Water Jet Cutting) tool in combination with a hydraulic robot and a water cooled diamond cable. The last technique was done in close collaboration with the external company Husqvarna. It was important to have an idea of the performance, the efficiency of the cable and the quantity and the type of secondary waste

  6. The effect of small specimen volume on the deformation of Zircaloy-4 PWR cladding under mainly convective cooling by steam

    International Nuclear Information System (INIS)

    Oakden, M.M.; Reynolds, A.E.

    1983-01-01

    Creep rupture tests were performed in flowing steam on single rod specimens of 17x17 type PWR Zircaloy-4 cladding 460 mm long. They were tested at temperatures between 640 deg.C and 985 deg.C with internal pressures in the range 1.00-9.65 MPa (gauge) (145-1400 lb/in 2 ). The internal free volume was limited to 2.9 ml. Axially extended 'carrot' shaped deformations were produced, the range of temperatures and pressures over which these occurred was found to be similar to that observed in previous tests conducted with a much larger free volume. The main result of limiting the internal volume was that straining of the specimens was accompanied by a more rapid drop in the internal pressure than occurred previously, and which reduced the extent of the deformation compared with that seen in the earlier work. However, diametral strains in excess of 33% were observed which would result in mechanical interaction of neighbouring bulges if this occurred in a multi-rod array. (author)

  7. Preliminary condensation pool experiments with steam using DN80 and DN100 blowdown pipes[VIDEO CAMERAS

    Energy Technology Data Exchange (ETDEWEB)

    Laine, J.; Puustinen, M. [Lappeenranta University of Technology (Finland)

    2004-03-01

    The report summarizes the results of the preliminary steam blowdown experiments. Altogether eight experiment series, each consisting of several steam blows, were carried out in autumn 2003 with a scaled-down condensation pool test rig designed and constructed at Lappeenranta University of Technology. The main purpose of the experiments was to evaluate the capabilities of the test rig and the needs for measurement and visualization devices. The experiments showed that a high-speed video camera is essential for visual observation due to the rapid condensation of steam bubbles. Furthermore, the maximum measurement frequency of the current combination of instrumentation and data acquisition system is inadequate for the actual steam tests in 2004. (au)

  8. Facility to separate water and steam

    International Nuclear Information System (INIS)

    Loesel, G.

    1977-01-01

    The water/steam mixture from the pressure vessel e.g. of a BWR is separated by means of centrifugal separators untilizing the natural separation of steam. The steam is supplied to a steam drying vessel and the water to a water collecting tank. These vessels may be combined to a common vessel or connected through additional pipes. From the water collecting tank, arranged below the steam dryer, a feedwater pipe runs back to the pressure vessel. By construction out of individual components cleaning, decontamination, and operating control are essentially simplified. (RW) 891 RW [de

  9. Microbial effect of steam vacuum pasteurisation implemented after slaughtering and dressing of sheep and lamb.

    Science.gov (United States)

    Hassan, Ammar Ali; Skjerve, Eystein; Bergh, Claus; Nesbakken, Truls

    2015-01-01

    The main objective of the study was to assess the effect of steam vacuum pasteurisation on carcass contamination with focus on Escherichia coli, Enterobacteriaceae and total plate count (TPC). Additionally, the effect of an additional tryptone soy agar (TSA) step for resuscitation of Enterobacteriaceae after steam vacuum pasteurisation was investigated. Steam vacuum pasteurisation was applied at a temperature of >82°C for a duration of 10s on sheep and lamb carcasses (n=120). Samples were taken immediately: i) after trimming just before the use of steam vacuum and ii) after use of steam vacuum. Nordic Committee on Food Analysis methods were used in microbial analyses. The differences in log reduction were found significant for all of the three microorganisms (psteam vacuum pasteurisation was higher in samples where TSA+violet red bile glucose agar (VRBGA) was used compared to samples where only VRBGA was used (pSteam vacuum pasteurisation was found efficient in reducing the total count, read as TPC, as well as the level of E. coli and Enterobacteriaceae. Copyright © 2014 Elsevier Ltd. All rights reserved.

  10. Drying system for steam generators, particularly for steam generators of nuclear power stations

    International Nuclear Information System (INIS)

    Lavalerie, Claude; Borrel, Christian.

    1982-01-01

    A drying system is described which allows for modular construction and which provides a significant available exchange area in a reduced volume. All the drying elements are identical and are distributed according to a ternay circular symmetry and are placed radially and associated to steam guiding facilities which alternately provide around the axis of revolution an output volume of dry steam from one element and an input volume of wet steam in the following element [fr

  11. Influencia de los parámetros de proceso, en el comportamiento del parámetro vibración absoluta en turbinas de vapor. // Influence of process parameters of the behaviour of absolute vibration parameters in steam turbine.

    Directory of Open Access Journals (Sweden)

    F. de la Torre Silva

    2008-01-01

    Full Text Available Este trabajo muestra el estudio realizado en turbinas de vapor de pequeña capacidad de generación de las centralestermoeléctricas cubanas, relacionadas con la influencia estadística entre parámetros de proceso de la turbina, respecto alparámetro vibración absoluta, medidos en las chumaceras de la turbina. Se utilizan las bases de datos de los sistemas demonitoreado on-line de la turbina. Se exponen las relaciones existentes entre los principales parámetros seleccionados paraeste estudio.Palabras claves: Monitorado on-line, Monitorado de la condición, Parámetros de proceso, Vibraciónabsoluta, Turbinas de vapor.______________________________________________________________________________AbstractThis work show the estudy in small steam turbine capacity of generation in cubans thermoelectric power station to relative withstatistical influence between steam process parameters regarding absolute vibration parameter, in the steam housing measurements.Themselves use database as steam on-line monitoring systems. Expose the existing relation between the principal selection parametersfor this study.Key Words: On.line monitoring, Condition monitoring, Process parameters, Absolute vibration, SteamTurbine.

  12. Modelling horizontal steam generator with ATHLET. Verification of different nodalization schemes and implementation of verified constitutive equations

    Energy Technology Data Exchange (ETDEWEB)

    Beliaev, J.; Trunov, N.; Tschekin, I. [OKB Gidropress (Russian Federation); Luther, W. [GRS Garching (Germany); Spolitak, S. [RNC-KI (Russian Federation)

    1995-12-31

    Currently the ATHLET code is widely applied for modelling of several Power Plants of WWER type with horizontal steam generators. A main drawback of all these applications is the insufficient verification of the models for the steam generator. This paper presents the nodalization schemes for the secondary side of the steam generator, the results of stationary calculations, and preliminary comparisons to experimental data. The consideration of circulation in the water inventory of the secondary side is proved to be necessary. (orig.). 3 refs.

  13. Modelling horizontal steam generator with ATHLET. Verification of different nodalization schemes and implementation of verified constitutive equations

    Energy Technology Data Exchange (ETDEWEB)

    Beliaev, J; Trunov, N; Tschekin, I [OKB Gidropress (Russian Federation); Luther, W [GRS Garching (Germany); Spolitak, S [RNC-KI (Russian Federation)

    1996-12-31

    Currently the ATHLET code is widely applied for modelling of several Power Plants of WWER type with horizontal steam generators. A main drawback of all these applications is the insufficient verification of the models for the steam generator. This paper presents the nodalization schemes for the secondary side of the steam generator, the results of stationary calculations, and preliminary comparisons to experimental data. The consideration of circulation in the water inventory of the secondary side is proved to be necessary. (orig.). 3 refs.

  14. Strategic management of steam generators

    International Nuclear Information System (INIS)

    Hernalsteen, P.; Berthe, J.

    1991-01-01

    This paper addresses the general approach followed in Belgium for managing any kind of generic defect affecting a Steam Generator tubebundle. This involves the successive steps of: problem detection, dedicated sample monitoring, implementation of preventive methods, development of specific plugging criteria, dedicated 100% inspection, implementation of repair methods, adjusted sample monitoring and repair versus replacement strategy. These steps are illustrated by the particular case of Primary Water Stress Corrosion Cracking in tube roll transitions, which is presently the main problem for two Belgian units Doele-3 and Tihange-2. (author)

  15. Hard sludge formation in modern steam generators of nuclear power plants. Formation, risks and mitigation

    International Nuclear Information System (INIS)

    Strohmer, F.

    2014-01-01

    In recent years modern steam generators with triangular pitch tube bundle geometry have experienced damage caused by hard sludge formation on top of the tube sheet and denting. The effect can lead to a limitation of the modern steam generators’ lifetime. The current publication shows reasons for the generation of hard sludge formation. Moreover, it describes the risk arising from hard sludge formation for the concerned steam generators and the mitigation of the problem. The main factors contributing to the formation of hard sludge are: the amount of corrosion product ingress into the steam generators, hard sludge formation favouring impurities and, skipped maintenance applications during outages. The main damaging mechanism of denting that can arise under certain secondary side conditions from hard sludge is explained. For steam generator tube denting, aggressive, oxidizing conditions have to be established in crevices beneath the hard sludge piles. Severely dented tubes are sensitive toward outer diameter stress corrosion cracking (ODSCC). The denting and ODSCC mechanism is explained. In addition, a proactive long-term maintenance strategy to avoid the formation of hard sludge piles will be shown. The strategy is based on a reduction of the corrosion product ingress into the steam generator's secondary side, and on the regular removal of deposits from the tube sheet and from the entire upper bundle area by latest mechanical cleaning methods. For hard deposits - formed either by silicates or long term hardened corrosion products, which, in the past, could not be removed by chemical or mechanical means - a new, simple, mechanical cleaning method is presented. This method can be used during the normal time frame of an outage and allows the restart of the unit with clean steam generator tube sheets. This mitigates the tendency to form hard sludge and denting in the long term. (author)

  16. Possibilities of the metallurgical base in the manufacture of tubes for nuclear power plant steam generators

    International Nuclear Information System (INIS)

    Prnka, T.; Walder, V.; Dolenek, J.

    Current possibilities are briefly summarized of metallurgy in the manufacture of high-quality tubes for nuclear power plant steam generators, mainly for fast reactor power plants. Discussed are steel making possibilities, semi-finished product and tube forming with special regard to 2.25Cr1MoNiNb steel problems, heat treatment, finishing, and testing. Necessary equipment and technology for the production of steam generator tubes are less common in the existing practice and are demanding on investment; their introduction, however, is inevitable for securing quality production of steam generator tubes. (Kr)

  17. Methodology for energy diagnosis in distribution steam lines; Metodologia para diagnostico de energia en lineas de distribucion de vapor

    Energy Technology Data Exchange (ETDEWEB)

    Almanza, M; Ambriz, J J; Romero P, H [Universidad Autonoma Metropolitana Iztapalapa, Mexico, D. F. (Mexico)

    1993-12-31

    This paper shows a methodology that results of great advantage in the development of the energy analysis of an industrial facility that utilizes steam as a mean of energy transport and where the steam operated equipment is physically located in a remote place, away from the generation site. Mention is made here of the equipment characteristics that can be used for such purpose, the most important parameters to identify in a rapid and efficient way the faults presented in the steam distribution system in the industrial plants and the formats that contribute to keep the corresponding records for efficiently maintain in operation the steam pipeline net in conjunction with the involved accessories. [Espanol] En el presenta trabajo se muestra una metodologia que resulta de gran utilidad en el desarrollo del analisis energetico de una planta industrial, en la cual se emplee el vapor como medio de transporte de la energia y en donde el equipo que utiliza el vapor se encuentre fisicamente en un lugar apartado de la zona de generacion. Aqui se mencionan las caracteristicas del equipo que se puede utilizar para dicho diagnostico, los parametros de mayor importancia para identificar en forma rapida y eficiente las fallas que se presentan en el sistema de distribucion de vapor en las plantas industriales y los formatos que contribuyen a llevar los registros correspondientes para mantener operando eficientemente la red de vapor, en conjunto con los accesorios que en ella se involucran.

  18. Methodology for energy diagnosis in distribution steam lines; Metodologia para diagnostico de energia en lineas de distribucion de vapor

    Energy Technology Data Exchange (ETDEWEB)

    Almanza, M.; Ambriz, J. J.; Romero P, H. [Universidad Autonoma Metropolitana Iztapalapa, Mexico, D. F. (Mexico)

    1992-12-31

    This paper shows a methodology that results of great advantage in the development of the energy analysis of an industrial facility that utilizes steam as a mean of energy transport and where the steam operated equipment is physically located in a remote place, away from the generation site. Mention is made here of the equipment characteristics that can be used for such purpose, the most important parameters to identify in a rapid and efficient way the faults presented in the steam distribution system in the industrial plants and the formats that contribute to keep the corresponding records for efficiently maintain in operation the steam pipeline net in conjunction with the involved accessories. [Espanol] En el presenta trabajo se muestra una metodologia que resulta de gran utilidad en el desarrollo del analisis energetico de una planta industrial, en la cual se emplee el vapor como medio de transporte de la energia y en donde el equipo que utiliza el vapor se encuentre fisicamente en un lugar apartado de la zona de generacion. Aqui se mencionan las caracteristicas del equipo que se puede utilizar para dicho diagnostico, los parametros de mayor importancia para identificar en forma rapida y eficiente las fallas que se presentan en el sistema de distribucion de vapor en las plantas industriales y los formatos que contribuyen a llevar los registros correspondientes para mantener operando eficientemente la red de vapor, en conjunto con los accesorios que en ella se involucran.

  19. French steam generator design developments

    International Nuclear Information System (INIS)

    Ginier, R.; Campan, J.L.; Pontier, M.; Leridon, A.; Remond, A.; Castello, G.; Holcblat, A.; Paurobally, H.

    1986-01-01

    From the outset of the French nuclear power program, a significant R and D effort has been invested in improvement of the design and operation of Pressurized Water Reactors including a special committment to improving steam generators. The steam generator enhancement program has spawned a wide variety of specific R and D resources, e.g., low temperature hydraulic models for investigation of areas with single-phase flow, and freon-filled models for simulation of areas of steam generators experiencing two-phase flow (tube bundles and moisture separators). For the moisture separators, a large scale research program using freon-filled models and highly sophisticated instrumentation was used. Tests at reactor sites during startup of both 900 MWe and 1300 MWe have been used to validate the assumptions made on the basis of loop tests. These tests also demonstrated the validity of using freon to simulate two-phase flow conditions. The wealth of knowledge accumulated by the steam generator R and D program has been used to develop a new design of steam generators for the N4 plants. The current R and D effort is aimed at qualifying the N4 steam generator model and developing more comprehensive models. One prong of the R and D effort is the Megeve program. Megeve is a 25 MW steam generator which simulates operating conditions of the N4 model. The other prong is Clotaire, a freon-filled steam generator model which will be used to qualify thermal/hydraulic design codes used for multidimensional calculations for design of tube bundles

  20. Will the Steam Coal Price Rebound under the New Economy Normalcy in China?

    Directory of Open Access Journals (Sweden)

    Xiaopeng Guo

    2016-09-01

    Full Text Available The steam coal price in China has been continuously decreasing since the second half of 2012. Constant low price of coal will accelerate the development of thermal power, cause more serious air pollution problems, and bring adverse influence to China’s energy reformation in the future. Therefore, analyzing the factors underlying the phenomenon of the decreasing steam coal price is significant. In this study, we first qualitatively analyze five main factors, namely, economy, supply, demand, substitutes, and port stocks. On the basis of the relationships among these five factors, we obtain the causality diagram and the system flow diagram of coal price for further quantitative research. Then, we conduct an empirical analysis using the system dynamics (SD method and determine the simulated price from 2012 to 2017. Finally, we discuss the running results and come to the conclusion that the steam coal price will continue to decrease under the combined actions of the five main factors and it will not rebound in the near future.

  1. Influence of surface topography on the surface durability of steam oxidised sintered iron

    Directory of Open Access Journals (Sweden)

    José Daniel Biasoli de Mello

    2005-06-01

    Full Text Available Durability of surfaces has been reported as the main factor affecting tribological behavior of steam oxidised sintered iron. The presence of surface pores and their negative influence on load bearing capacity, suggest that surface topography might play an important role on the durability of the oxide layer. In this paper, the influence of compaction pressure and powder grade on surface topography, and as a consequence, its effect on the tribological behavior of steam oxidised sintered iron has been analysed. Specimens prepared from atomised iron powders with different sizes were compacted using 4 different pressures, sintered, and then subjected to steam treatment. Tribological characterisation was carried out in a reciprocating sliding wear test. Although the processing parameters affected the surface topography to a considerable extent, the main influence may be attributed to powder grade. A strong influence of surface topography on the durability distance, evaluated in terms of the evolution of contact resistance with total sliding distance, has been highlighted. Surfaces which were smoother and had high load-carrying capacity were always associated with a higher durability distance.

  2. Ex-Vessel Steam Explosion Analysis of Central Melt Pour Scenario

    International Nuclear Information System (INIS)

    Ursic, M.; Leskovar, M.

    2008-01-01

    gained at 2 bar primary system overpressure and water sub-cooling of 50 degree C. The purpose of the work presented in the paper is to perform a detailed analysis of the central melt pour scenario at 2 bar primary system overpressure and water sub-cooling of 50 degree C, where the strongest steam explosion occurred. The steam explosion analysis will be performed with the computer code MC3D, which is build mainly for the simulation of the complex FCI phenomena. The MC3D is being developed by IRSN, France. The reasons for the obtained so high pressure loads during the steam explosion in the selected scenario will be highlighted and discussed in details.(author)

  3. Exergy Steam Drying and Energy Integration

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Prem; Muenter, Claes (Exergy Engineering and Consulting, SE-417 55 Goeteborg (Sweden)). e-mail: verma@exergyse.com

    2008-10-15

    Exergy Steam Drying technology has existed for past 28 years and many new applications have been developed during this period. But during past few years the real benefits have been exploited in connection with bio-fuel production and energy integration. The steam dryer consists of a closed loop system, where the product is conveyed by superheated and pressurised carrier steam. The carrier steam is generated by the water vapours from the product being dried, and is indirectly superheated by another higher temperature energy source such as steam, flue gas, thermal oil etc. Besides the superior heat transfer advantages of using pressurised steam as a drying medium, the energy recovery is efficient and simple as the recovered energy (80-90%) is available in the form of steam. In some applications the product quality is significantly improved. Examples presented in this paper: Bio-Combine for pellets production: Through integration of the Exergy Steam Dryer for wood with a combined heat and power (CHP) plant, together with HP steam turbine, the excess carrier steam can be utilised for district heating and/or electrical power production in a condensing turbine. Bio-ethanol production: Both for first and second generation of ethanol can the Exergy process be integrated for treatment of raw material and by-products. Exergy Steam Dryer can dry the distillers dark grains and solubles (DDGS), wood, bagasse and lignin. Bio-diesel production: Oil containing seeds and fruits can be treated in order to improve both the quality of oil and animal feed protein, thus minimizing further oil processing costs and increasing the sales revenues. Sewage sludge as bio-mass: Municipal sewage sludge can be considered as a renewable bio-fuel. By drying and incineration, the combustion heat value of the sludge is sufficient for the drying process, generation of electrical energy and production of district heat. Keywords; Exergy, bio-fuel, bio-mass, pellets, bio-ethanol, biodiesel, bio

  4. Evolution of design of steam generator for sodium cooled reactors

    International Nuclear Information System (INIS)

    Chetal, S.C.; Vaidyanathan

    1997-01-01

    The first sodium cooled reactor was the experimental breeder reactor (EBR-I) in usa which was commissioned in 1951 and was incidentally the first nuclear reactor to generate electrical energy. This was followed by fast breeder reactors in USSR, UK, france, USA, japan, germany and India. The use of sodium as a coolant is due to its low moderation which helps in breeding fissile fuel from fertile materials and also its high heat transfer coefficient at comparatively low velocities. The good heat transfer properties introduce thermal stresses when there are rapid changes in the sodium temperatures. Also sodium has a chemical affinity with air and water. The steam generators for sodium cooled reactors have to allow for these novel conditions and in addition, unlike other components. Choices have to be made whether it is a recirculation type as in most fossil plants or an once through unit, the power rating, shape of the tube (straight, helical, U-tube), materials (Ferritic or austenitic), with free level of sodium or not, sodium on tube side or shell side and so on. With higher pressures and steam temperatures reheating steam after partial expansion in the turbine becomes essential as in conventional turbines. For this purpose the choice of reheating fluid viz sodium or live main steam has to be made. This paper traces the evolution of steam generator designs in the different sodium cooled reactors (chronologically) and the operation experience. 16 figs., 1 tab

  5. Steam jet ejectors are examined automatically

    International Nuclear Information System (INIS)

    Lardiere, C.

    2013-01-01

    Steam jet ejectors are used in the nuclear industry particularly for the transfer of radioactive fluids. Their working is based on the Venturi effect and the conservation of energy. A steam ejector can be considered as a thermodynamical pump without mobile parts. The Descote enterprise manufactures a broad range of steam jet ejectors and the characterization and testing of the steam ejectors was made manually and empirically so far. A new test bench has been designed, the tests are led automatically and allow a more accurate characterization and optimization of the steam jet ejectors. (A.C.)

  6. Steam generator tube integrity program

    International Nuclear Information System (INIS)

    Dierks, D.R.; Shack, W.J.; Muscara, J.

    1996-01-01

    A new research program on steam generator tubing degradation is being sponsored by the U.S. Nuclear Regulatory Commission (NRC) at Argonne National Laboratory. This program is intended to support a performance-based steam generator tube integrity rule. Critical areas addressed by the program include evaluation of the processes used for the in-service inspection of steam generator tubes and recommendations for improving the reliability and accuracy of inspections; validation and improvement of correlations for evaluating integrity and leakage of degraded steam generator tubes, and validation and improvement of correlations and models for predicting degradation in steam generator tubes as aging occurs. The studies will focus on mill-annealed Alloy 600 tubing, however, tests will also be performed on replacement materials such as thermally-treated Alloy 600 or 690. An overview of the technical work planned for the program is given

  7. Some engineering aspects of the steam generator system for the United States LMFBR demonstration plant

    International Nuclear Information System (INIS)

    Tippets, F.E.

    1975-01-01

    This paper describes the main design features of the steam generator system for the Clinch River Breeder Reactor Plant and the engineering approach being employed for some of the critical elements of this system, including in particular the sodium-steam/water boundary, the efforts to have this boundary be of highest integrity, and the system features to safely accommodate any failure of the boundary. (author)

  8. Some engineering aspects of the steam generator system for the United States LMFBR demonstration plant

    Energy Technology Data Exchange (ETDEWEB)

    Tippets, F E

    1975-07-01

    This paper describes the main design features of the steam generator system for the Clinch River Breeder Reactor Plant and the engineering approach being employed for some of the critical elements of this system, including in particular the sodium-steam/water boundary, the efforts to have this boundary be of highest integrity, and the system features to safely accommodate any failure of the boundary. (author)

  9. ACQUA97 - a Fortran subroutine to evaluate the IAPWS-IF97 equations of state for water and steam

    International Nuclear Information System (INIS)

    Veloso, Marcelo A.; Dias, Marcio S.; Fortini, Maria Auxiliadora

    2007-01-01

    Fundamental to subroutine ACQUA97 is the IAPWS-IF97 formulation for the thermodynamic properties of water and steam adopted by the International Association for the Properties of Water and Steam (IAPWS) in 1997. This new formulation is designed specifically for industrial applications, and replaces the previous industrial formulation, IFC-67, that has formed the basis of steam tables used in many areas of steam power industry throughout the world since the late 1960's. ACQUA97 has been programmed to compute the thermodynamic properties of water and steam (pressure, temperature, specific volume, specific entropy, specific enthalpy, and internal specific energy) and their main first partial derivatives for several combinations of two independent variables. One of the independent variable is either pressure or temperature. Vapor-liquid saturation properties are calculated at specified temperature or pressure. Transport properties (dynamic viscosity and thermal conductivity) and vapor-liquid surface tension are also calculated with formulations adopted by IAPWS. Any of the above mentioned thermophysical properties can be computed by this single subroutine from a simple main program supplied by the user. ACQUA97 might be very useful to those who deal with design and evaluation of thermal power plants. (author)

  10. Surry Power Station secondary water chemistry improvement since steam generator replacement and the unit two feedwater pipe rupture

    International Nuclear Information System (INIS)

    Swindell, E.T.

    1988-01-01

    Surry Power Station has two Westinghouse-designed three-loop PWRs of 811 MWe design rating. The start of commercial operation was in July, 1972 in No.1 plant, and May, 1973 in No.2 plant. Both plants began the operation using controlled phosphate chemistry for the steam generators. In 1975, both plants were converted to all volatile treatment on the secondary side due to the tube wall thinning corrosion in the steam generators, which was associated with the phosphate sludge that was building up on the tube sheets and created acidic condition. Thereafter, condenser and air leakage and steam generator denting occurred, and after the operation of 8 years 2 month of No.1 plant and 5 years 9 months of No.2 plant, the steam generators were replaced. A major plant improvement program was designed and implemented from 1979 to 1980. The improvement in new steam generators, the modification for preventing corrosion, the addition of a steam generator blowdown recovery system, the reconstruction of condensers, the installation of full flow, deep bed condensate polishers, the addition of Dionex 8,000 on-line ion chromatograph, a long term maintenance agreement with Westinghouse and so on are reported. (Kako, I.)

  11. Numerical and computational aspects of the coupled three-dimensional core/ plant simulations: organization for economic cooperation and development/ U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-II. 3. Analysis of the OECD TMI-1 Main-Steam- Line-Break Benchmark Accident Using the Coupled RELAP5/PANTHER Codes

    International Nuclear Information System (INIS)

    Schneidesch, C.R.; Guisset, J.P.; Zhang, J.; Bryce, P.; Parkes, M.

    2001-01-01

    The RELAP5 best-estimate thermal-hydraulic system code has been coupled with the PANTHER three-dimensional (3-D) neutron kinetics code via the TALINK dynamic data exchange control and processing tool. The coupled RELAP5/PANTHER code package is being qualified and will be used at British Energy (BE) and Tractebel Energy Engineering (TEE), independently, to analyze pressurized water reactor (PWR) transients where strong core-system interactions occur. The Organization for Economic Cooperation and Development/Nuclear Energy Agency PWR Main-Steam-Line-Break (MSLB) Benchmark problem was performed to demonstrate the capability of the coupled code package to simulate such transients, and this paper reports the BE and TEE contributions. In the first exercise, a point-kinetics (PK) calculation is performed using the RELAP5 code. Two solutions have been derived for the PK case. The first corresponds to scenario, 1 where calculations are carried out using the original (BE) rod worth and where no significant return to power (RTP) occurs. The second corresponds to scenario 2 with arbitrarily reduced rod worth in order to obtain RTP (and was not part of the 'official' results). The results, as illustrated in Fig. 1, show that the thermalhydraulic system response and rod worth are essential in determining the core response. The second exercise consists of a 3-D neutron kinetics transient calculation driven by best-estimate time-dependent core inlet conditions on a 18 T and H zones basis derived from TRAC-PF1/MOD2 (PSU), again analyzing two scenarios of different rod worths. Two sets of PANTHER solutions were submitted for exercise 2. The first solution uses a spatial discretization of one node per assembly and 24 core axial layers for both flux and T and H mesh. The second is characterized by spatial refinement (2 x 2 nodes per assembly, 48 core layers for flux, and T and H calculation), time refinement (half-size time steps), and an increased radial discretization for solution

  12. Steam generator water lancing

    International Nuclear Information System (INIS)

    Kamler, F.; Schneider, W.

    1992-01-01

    The tubesheet and tube support plate deposits in CANDU steam generators are notable for their hardness. Also notable is the wide variety of steam generator access situations. Because of the sludge hardness and the difficulty of the access, traditional water lancing processes which directed jets from the central tube free lane or from the periphery of the bundle have proven unsuitable. This has led to the need for some very unique inter tube water lancing devices which could direct powerful water jets directly onto the deposits. This type of process was applied to the upper broached plates of the Bruce A steam generators, which had become severely blocked. It has since been applied to various other steam generator situations. This paper describes the flexlance equipment development, qualification, and performance in the various CANDU applications. 4 refs., 2 tabs., 7 figs

  13. Engineering report for interim solids removal modifications of the Steam Plant Wastewater Treatment Facility

    International Nuclear Information System (INIS)

    1995-04-01

    The Steam Plant Wastewater Treatment Facility (SPWTF) treats wastewater from the Y-12 Plant coal yard, steam plant, and water demineralizer facility. The facility is required to comply with National Pollutant Discharge Elimination System (NPDES) standards prior to discharge to East Fork Poplar Creek (EFPC). The existing facility was designed to meet Best Available Technology (BAT) standards and has been in operation since 1988. The SPWTF has had intermittent violations of the NPDES permit primarily due to difficulties in complying with the limit for total iron of 1.0 ppM. A FY-1997 Line Item project, SPWTF Upgrades, is planned to improve the capabilities of the SPWTF to eliminate non-compliances with the permit limits. The intent of the Interim Solids Removal Modification project is to improve the SPWTF effluent quality and to provide pilot treatment data to assist in the design and implementation of the SPWTF Upgrades Line Item Project

  14. Advanced steam power plant concepts with optimized life-cycle costs: A new approach for maximum customer benefit

    Energy Technology Data Exchange (ETDEWEB)

    Seiter, C.

    1998-07-01

    The use of coal power generation applications is currently enjoying a renaissance. New highly efficient and cost-effective plant concepts together with environmental protection technologies are the main factors in this development. In addition, coal is available on the world market at attractive prices and in many places it is more readily available than gas. At the economical leading edge, standard power plant concepts have been developed to meet the requirements of emerging power markets. These concepts incorporate the high technological state-of-the-art and are designed to achieve lowest life-cycle costs. Low capital cost, fuel costs and operating costs in combination with shortest lead times are the main assets that make these plants attractive especially for IPPs and Developers. Other aspects of these comprehensive concepts include turnkey construction and the willingness to participate in BOO/BOT projects. One of the various examples of such a concept, the 2 x 610-MW Paiton Private Power Project Phase II in Indonesia, is described in this paper. At the technological leading edge, Siemens has always made a major contribution and was pacemaker for new developments in steam power plant technology. Modern coal-fired steam power plants use computer-optimized process and plant design as well as advanced materials, and achieve efficiencies exceeding 45%. One excellent example of this high technology is the world's largest lignite-fired steam power plant Schwarze Pumpe in Germany, which is equipped with two 800 MW Siemens steam turbine generators with supercritical steam parameters. The world's largest 50-Hz single-shaft turbine generator with supercritical steam parameters rated at 1025 MW for the Niederaussem lignite-fired steam power plant in Germany is a further example of the sophisticated Siemens steam turbine technology and sets a new benchmark in this field.

  15. Safety Picks up "STEAM"

    Science.gov (United States)

    Roy, Ken

    2016-01-01

    This column shares safety information for the classroom. STEAM subjects--science, technology, engineering, art, and mathematics--are essential for fostering students' 21st-century skills. STEAM promotes critical-thinking skills, including analysis, assessment, categorization, classification, interpretation, justification, and prediction, and are…

  16. Steam generator tube extraction

    International Nuclear Information System (INIS)

    Delorme, H.

    1985-05-01

    To enable tube examination on steam generators in service, Framatome has now developed a process for removing sections of steam generator tubes. Tube sections can be removed without being damaged for treating the tube section expanded in the tube sheet

  17. Steam jet mill-a prospective solution to industrial exhaust steam and solid waste.

    Science.gov (United States)

    Zhang, Mingxing; Chen, Haiyan

    2018-04-20

    Bulk industrial solid wastes occupy a lot of our resources and release large amounts of toxic and hazardous substances to the surrounding environment, demanding innovative strategies for grinding, classification, collection, and recycling for economically ultrafine powder. A new technology for grinding, classification, collection, and recycling solid waste is proposed, using the superheated steam produced from the industrial exhaust steam to disperse, grind, classify, and collect the industrial solid waste. A large-scale steam jet mill was designed to operate at an inlet steam temperature 230-300 °C and an inlet pressure of 0.2-0.6 MPa. A kind of industrial solid waste fluidized-bed combustion ashes was used to grinding tests at different steam temperatures and inlet pressures. The total process for grinding, classification, and collection is drying. Two kinds of particle sizes are obtained. One particle size is d 50  = 4.785 μm, and another particle size is d 50  = 8.999 μm. For particle size d 50  = 8.999 μm, the inlet temperature is 296 °C and an inlet pressure is 0.54 MPa for the grinding chamber. The steam flow is 21.7 t/h. The yield of superfine powder is 73 t/h. The power consumption is 3.76 kW h/t. The obtained superfine powder meets the national standard S95 slag. On the basis of these results, a reproducible and sustainable industrial ecological protocol using steam produced by industrial exhaust heat coupled to solid waste recycling is proposed, providing an efficient, large-scale, low-cost, promising, and green method for both solid waste recovery and industrial exhaust heat reutilization.

  18. The study of reactions influencing the biomass steam gasification process

    Energy Technology Data Exchange (ETDEWEB)

    C. Franco; F. Pinto; I. Gulyurtlu; I. Cabrita [INETI-DEECA, Lisbon (Portugal)

    2003-05-01

    Steam gasification studies were carried out in an atmospheric fluidised bed. The gasifier was operated over a temperature range of 700 900{sup o}C whilst varying a steam/biomass ratio from 0.4 to 0.85 w/w. Three types of forestry biomass were studied: Pinus pinaster (softwood), Eucalyptus globulus and holm-oak (hardwood). The energy conversion, gas composition, higher heating value and gas yields were determined and correlated with temperature, steam/biomass ratio, and species of biomass used. The results obtained seemed to suggest that the operating conditions were optimised for a gasification temperature around 830{sup o}C and a steam/biomass ratio of 0.6 0.7 w/w, because a gas richer in hydrogen and poorer in hydrocarbons and tars was produced. These conditions also favoured greater energy and carbon conversions, as well the gas yield. The main objective of the present work was to determine what reactions were dominant within the operation limits of experimental parameters studied and what was the effect of biomass type on the gasification process. As biomass wastes usually have a problem of availability because of seasonal variations, this work analysed the possibility of replacing one biomass species by another, without altering the gas quality obtained. 19 refs., 8 figs. 2 tabs.

  19. Efficient steam generation by inexpensive narrow gap evaporation device for solar applications.

    Science.gov (United States)

    Morciano, Matteo; Fasano, Matteo; Salomov, Uktam; Ventola, Luigi; Chiavazzo, Eliodoro; Asinari, Pietro

    2017-09-20

    Technologies for solar steam generation with high performance can help solving critical societal issues such as water desalination or sterilization, especially in developing countries. Very recently, we have witnessed a rapidly growing interest in the scientific community proposing sunlight absorbers for direct conversion of liquid water into steam. While those solutions can possibly be of interest from the perspective of the involved novel materials, in this study we intend to demonstrate that efficient steam generation by solar source is mainly due to a combination of efficient solar absorption, capillary water feeding and narrow gap evaporation process, which can also be achieved through common materials. To this end, we report both numerical and experimental evidence that advanced nano-structured materials are not strictly necessary for performing sunlight driven water-to-vapor conversion at high efficiency (i.e. ≥85%) and relatively low optical concentration (≈10 suns). Coherently with the principles of frugal innovation, those results unveil that solar steam generation for desalination or sterilization purposes may be efficiently obtained by a clever selection and assembly of widespread and inexpensive materials.

  20. 49 CFR 229.105 - Steam generator number.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Steam generator number. 229.105 Section 229.105..., DEPARTMENT OF TRANSPORTATION RAILROAD LOCOMOTIVE SAFETY STANDARDS Safety Requirements Steam Generators § 229.105 Steam generator number. An identification number shall be marked on the steam generator's...

  1. One-dimensional simulation for attemperator based on commissioning data of coal-fired steam power plant

    International Nuclear Information System (INIS)

    Cho, Baekhyun; Choi, Geunwon; Uruno, Yumi; Kim, Hyunseo; Chung, Jaewon; Kim, Hyojun; Lee, Kihyun

    2017-01-01

    Highlights: • An attemperator is a device to spray water into the superheated steam. • The evaporation was analyzed using the enthalpy balance from the commissioning data. • The spray atomization and its concurrent evaporation in an attemperator were physically modeled. • A simple one-dimensional simulation was conducted to verify the commissioning results. - Abstract: An attemperator is a device that is used to spray water into the superheated steam between the primary, platen, and final superheaters and the reheat lines. The goal of the attemperator is to control the temperature of the superheated steam in accordance with desired turbine-inlet temperature during both steady-state and transient operation. Because the thermowell installed at the attemperator outlet is tied back to the feedback control of the spray water, the spray water should evaporate ahead of the thermowell for accurate control of the steam temperature. In this work, the completion of the evaporation ahead of the thermowell was analyzed using the enthalpy balance from the start-up commissioning data of an 800-MW coal-fired steam power plant. In addition, the phenomena of the spray atomization and its concurrent evaporation in an attemperator were physically modeled, and a simple one-dimensional simulation was conducted to verify the analysis of the commissioning data.

  2. Identification of leaky steam generators by iodine mapping technique and development of tools for cutting of tubes of steam generators of Indian PHWRS

    International Nuclear Information System (INIS)

    Subba Rao, D.

    2006-01-01

    Kakrapar Atomic Power Station (2X220 MWe) located in Mandvi Taluka of Surat District in the state of Gujarat is the fifth Nuclear Power Station of the country. It has got an excellent record in the field of operation, safety, public awareness and emergency preparedness. KAPS Unit -1 achieved first criticality in Sep-1992 and was declared for commercial operation in may-1993. KAPS Unit -2 achieved first criticality in Jan-1995 and was declared for commercial operation in Sep-1995. So far station has generated about 30 billion units.Unit-1 achieved 98.4% and was graded as a world's No.1 in year 2002 amongst all CANDU type reactors. KAPS Unit -1 has made another record of operating continuously for more than 300 days in Indian PHWR s operating history. This paper mainly deals with the Indian PHWRs Steam Generators (SG) tube leaks, leaky steam generator identification by Iodine mapping, and development of special tool for cutting, removal and plugging of leaky tubes. These Steam Generators are designed by M/s Kraft Werke Union (KWU) of Siemens Group, West Germany, and Manufactured by M/s ENSA, SPAIN for Unit- 1 and by M/s MAN-GHH, Germany for Unit- 2. First time in October-2002 one of the Steam Generators of Unit-1 developed tube leak. To identify leaky Steam Generator, KAPS has established a method of Iodine mapping. With that the leaky SG was identified in very short time and corrective actions were taken immediately. Total three tube leaks (two in SG-4 of Unit-1 and one in SG-1 of unit-2) were experienced in both Units'. Following observations were made on SG tubes failure: All failures were in cold leg side; All Failures / deterioration locations were in front of main feed water nozzle; All Failures / deterioration locations were observed to be just above tube support plate (TSP) number 4 or 5; Deterioration ( i.e. wall thinning) observed from OD side and these tubes were adjacent to failed tubes; In all the three incidents, failed / deteriorated tubes were

  3. Experimental and Numerical Investigations of Surge Extension on a Centrifugal Compressor with Vaned Diffuser Using Steam Injection

    Directory of Open Access Journals (Sweden)

    Chuang Gao

    2017-01-01

    Full Text Available This paper presents the first report on surge extension with steam injection through both experimental and numerical simulation. The experimental section covers the test facility, instrumentation, and prestall modes comparison with and without steam injection. It is found that surge extension is not in proportion to injected steam. There exists an upper bound above which deteriorates the margin. Injection of less than 1% of the designed mass flow can bring about over 10% margin improvement. Test results also indicated that steam injection not only damps out prestall waves, but also changes prestall modes and traveling direction. At 90% speed, injection changed the prestall mode from spike to modal, while at 80% speed line, it made the forward traveling wave become backward. Through numerical simulation, location and number of injectors, molecular weight, and temperature of injected gas are modified to explore their influences on surge margin. Similar to the test results, there exists an upper bound for the amount of steam injected. The flow field investigation indicates that this bound is caused by the early trigger of flow collapse due to the injected steam which is similar to the tip leakage flow spillage caused spike stall in axial compressors.

  4. Technical basis for the CANDU steam generator tube fitness-for-service guidelines

    International Nuclear Information System (INIS)

    Kozluk, M.J.; Scarth, D.A.; Graham, D.B.

    2002-01-01

    Active degradation mechanisms in steam generators and preheaters in Canadian CANDU T M generating stations are managed through Steam Generator Programs that incorporate tube inspection, maintenance (cleaning), fitness-for-service assessment, and preventative plugging as part of the overall steam generator management strategy. Steam generator and preheater tubes are inspected in accordance with the CSA Standard CAN/CSA-N285.4-94[l]. When a detected flaw indication does not satisfy the criteria of acceptance by examination, CSA-N285.4-94 permits a fitness-for-service assessment to determine acceptability. In 1999 Ontario Power Generation issued, for trial use, fitness-for-service guidelines for steam generator and preheater tubes in CANDU nuclear power plants. The main objectives of the Fitness-for-Service Guidelines are to provide reasonable assurance that tube structural integrity is maintained, and to provide reasonable assurance that there are adequate margins between estimated accumulated dose and applicable site dose limits. The Fitness-for-Service Guidelines are intended to provide industry-standard acceptance criteria and evaluation procedures for assessing the condition of steam generator and preheater tubes in terms of tube structural integrity, operational leak rate, and consequential leakage during an upset or abnormal event. This paper describes the technical basis for the minimum required safety factors specified in Table IC-1 of the Fitness-for-Service Guidelines and for the flaw models used to develop the flaw stability requirements in the nonmandatory, Appendix C of the Fitness-for-Service Guidelines. (author)

  5. Steam-water separator

    International Nuclear Information System (INIS)

    Modrak, T.M.; Curtis, R.W.

    1978-01-01

    The steam-water separator connected downstream of a steam generator consists of a vertical centrifugal separator with swirl blades between two concentric pipes and a cyclone separator located above. The water separated in the cyclone separator is collected in the inner tube of the centrifugal separator which is closed at the bottom. This design allows the overall height of the separator to be reduced. (DG) [de

  6. Erosion corrosion in wet steam

    International Nuclear Information System (INIS)

    Tavast, J.

    1988-03-01

    The effect of different remedies against erosion corrosion in wet steam has been studied in Barsebaeck 1. Accessible steam systems were inspected in 1984, 1985 and 1986. The effect of hydrogen peroxide injection of the transport of corrosion products in the condensate and feed water systems has also been followed through chemical analyses. The most important results of the project are: - Low alloy chromium steels with a chromium content of 1-2% have shown excellent resistance to erosion corrosion in wet steam. - A thermally sprayed coating has shown good resistance to erosion corrosion in wet steam. In a few areas with restricted accessibility minor attacks have been found. A thermally sprayed aluminium oxide coating has given poor results. - Large areas in the moisture separator/reheater and in steam extraction no. 3 have been passivated by injection of 20 ppb hydrogen peroxide to the high pressure steam. In other inspected systems no significant effect was found. Measurements of the wall thickness in steam extraction no. 3 showed a reduced rate of attack. - The injection of 20 ppb hydrogen peroxide has not resulted in any significant reduction of the iron level result is contrary to that of earlier tests. An increase to 40 ppb resulted in a slight decrease of the iron level. - None of the feared disadvantages with hydrogen peroxide injection has been observed. The chromium and cobalt levels did not increase during the injection. Neither did the lifetime of the precoat condensate filters decrease. (author)

  7. Latest information about development of gas-fueled steam convection ovens; Ou en est la mise au point du four a gaz a convection de vapeur au Japon

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, T. [Tokyo-gas co., Ltd. (Japan)

    2000-07-01

    In Japan, gas has been the main source of heat for heating-type culinary equipment in the food service industry, however, this situation has been recently changing. One reason for this is the introduction of electric steam convection ovens. To promote the use of gas steam convection ovens that can compete with electric appliances, Tokyo Gas has conducted the following development projects. Firstly, we set development targets for gas appliances to out-perform the best electric appliances, and were able to develop appliance products that met the targets. Secondly, in order to develop new markets for gas appliances, we worked on the development of compact appliances with a comparatively low initial cost, launching the world's smallest product in this category in October 1998. Thirdly, in order to make gas appliances as widely used here as they are in Europe and America, we developed appliances with costs cut by 30%, and in October 1999 we launched the cheapest gas steam convection oven in the domestic market. We plan to continue providing technological expertise to domestic manufacturers, enhancing our line-up with top performance gas appliances at even lower cost. (author)

  8. Reducing scram frequency by modifying/eliminating steam generator low-low level reactor trip setpoint for Maanshan nuclear power plant

    International Nuclear Information System (INIS)

    Yuann, R.Y.; Chiang, S.C.; Hsiue, J.K.; Chen, P.C.

    1987-01-01

    The feasibility of modification/elimination of steam generator low-low level reactor trip setpoint is evaluated by using RETRAN-02 code for the purpose of reducing scram frequency in Maanshan 3-loop pressurized water reactor. The ANS Condition II event loss of normal feedwater and condition IV event feedwater system line break are the basis for steam generator low-low level reactor trip setpoint sensitivity analysis, including various initial reactor power levels, reactivity feedback coefficients, and system functions assumptions etc., have been performed for the two basis events with steam generator low-low level reactor trip setpoint at 0% narrow range and without this trip respectively. The feasibility of modifying/eliminating current steam generator low-low level reactor trip setpoint is then determined based on whether the analysis results meet with the ANS Condition II and IV acceptance criteria or not

  9. Effects of non-condensable gas on the condensation of steam

    International Nuclear Information System (INIS)

    Jackson, J.D.; An, P.; Reinert, A.; Ahmadinejad, M.

    2000-01-01

    The experimental work reported here was undertaken with the aim of extending the database currently available on the condensation of steam in the presence of non-condensable gases and thereby improving the empirical input to thermal-hydraulic codes which might be used for design and safety assessment of advanced water-cooled nuclear reactors. Heat was removed from flowing mixtures of steam and air in a test section by means of a water-cooled condensing plate. The test facility constructed for the study incorporates a degassing unit which supplies water to a boiler. This delivers steam steadily to a mixing chamber where it joins with a flow of preheated air. The mixture of steam and air is supplied to the bottom of a cylindrical test section in which it flows upwards over a double sided condensing plate which can be vertical, inclined or horizontal, The rate at which heat is removed by cooling water flowing through internal passages in the plate can de determined calorimetrically knowing the flow rate of the water and its temperature rise. After commissioning experiments had shown that reliable measurements of condensation heat transfer rate could be made using the test facility, a programme of development work followed in the course of which three different designs of condensing plate were evaluated in turn. The version eventually used in the main programme of experiments which followed was made from copper. However, its surfaces were coated with a thin layer of nickel and then with one of chromium. It was found that such a surface consistently promoted dropwise condensation and showed no signs of deterioration after lengthy periods of use. The rate of heat removal from pure steam and from mixtures of steam and air in varying proportions was measured as a function of plate sub-cooling for a variety of plate orientations. (author)

  10. Steam generator reliability improvement project

    International Nuclear Information System (INIS)

    Blomgren, J.C.; Green, S.J.

    1987-01-01

    Upon successful completion of its research and development technology transfer program, the Electric Power Research Institute's Steam Generator Owners Group (SGOG II) will disband in December 1986 and be replaced in January 1987 by a successor project, the Steam Generator Reliability Project (SGRP). The new project, funded in the EPRI base program, will continue the emphasis on reliability and life extension that was carried forward by SGOG II. The objectives of SGOG II have been met. Causes and remedies have been identified for tubing corrosion problems, such as stress corrosion cracking and pitting, and steam generator technology has been improved in areas such as tube wear prediction and nondestructive evaluation (NDE). These actions have led to improved reliability of steam generators. Now the owners want to continue with a centrally managed program that builds on what has been learned. The goal is to continue to improve steam generator reliability and solve small problems before they become large problems

  11. Steam generator reliability improvement project

    International Nuclear Information System (INIS)

    Blomgren, J.C.; Green, S.J.

    1987-01-01

    Upon successful completion of its research and development technology transfer program, the Electric Power Research Institute's (EPRI's) Steam Generator Owners Group (SGOG II) will disband in December 1986, and be replaced in January 1987, by a successor project, the Steam Generator Reliability Project (SGRP). The new project, funded in the EPRI base program, will continue to emphasize reliability and life extension, which were carried forward by SGOG II. The objectives of SGOG II have been met. Causes and remedies have been identified for tubing corrosion problems such as stress corrosion cracking and pitting, and steam generator technology has been improved in areas such as tube wear prediction and nondestructive evaluation. These actions have led to improved reliability of steam generators. Now the owners want to continue with a centrally managed program that builds on what has been learned. The goal is to continue to improve steam generator reliability and to solve small problems before they become large problems

  12. A drier unit for steam separators

    International Nuclear Information System (INIS)

    Peyrelongue, J.-P.

    1973-01-01

    Description is given of a drier unit adapted to equip a water separator mounted in a unit for treating a wet steam fed from a high pressure enclosure, so as to dry and contingently superheat said steam prior to injecting same into a turbine low pressure stage. This drier unit is constituted by at least a stack of separating sheets maintained in parallel relationship and at a slight angle with respect to the horizontal so as to allow the water provided by wet steam to flow toward a channel communicating with a manifold, and by means for guiding the steam between the sheets and evenly distributing it. This can be applied to steam turbines in nuclear power stations [fr

  13. Steam explosion studies review

    International Nuclear Information System (INIS)

    Hwang, Moon Kyu; Kim, Hee Dong

    1999-03-01

    When a cold liquid is brought into contact with a molten material with a temperature significantly higher than the liquid boiling point, an explosive interaction due to sudden fragmentation of the melt and rapid evaporation of the liquid may take place. This phenomenon is referred to as a steam explosion or vapor explosion. Depending upon the amount of the melt and the liquid involved, the mechanical energy released during a vapor explosion can be large enough to cause serious destruction. In hypothetical severe accidents which involve fuel melt down, subsequent interactions between the molten fuel and coolant may cause steam explosion. This process has been studied by many investigators in an effort to assess the likelihood of containment failure which leads to large scale release of radioactive materials to the environment. In an effort to understand the phenomenology of steam explosion, extensive studies has been performed so far. The report presents both experimental and analytical studies on steam explosion. As for the experimental studies, both small scale tests which involve usually less than 20 g of high temperature melt and medium/large scale tests which more than 1 kg of melt is used are reviewed. For the modelling part of steam explosions, mechanistic modelling as well as thermodynamic modelling is reviewed. (author)

  14. Water regime of steam power plants

    International Nuclear Information System (INIS)

    Oesz, Janos

    2011-01-01

    The water regime of water-steam thermal power plants (secondary side of pressurized water reactors (PWR); fossil-fired thermal power plants - referred to as steam power plants) has changed in the past 30 years, due to a shift from water chemistry to water regime approach. The article summarizes measures (that have been realised by chemists of NPP Paks) on which the secondary side of NPP Paks has become a high purity water-steam power plant and by which the water chemistry stress corrosion risk of heat transfer tubes in the VVER-440 steam generators was minimized. The measures can also be applied to the water regime of fossil-fired thermal power plants with super- and subcritical steam pressure. Based on the reliability analogue of PWR steam generators, water regime can be defined as the harmony of construction, material(s) and water chemistry, which needs to be provided in not only the steam generators (boiler) but in each heat exchanger of steam power plant: - Construction determines the processes of flow, heat and mass transfer and their local inequalities; - Material(s) determines the minimal rate of general corrosion and the sensitivity for local corrosion damage; - Water chemistry influences the general corrosion of material(s) and the corrosion products transport, as well as the formation of local corrosion environment. (orig.)

  15. Citric pectin. Effect of steam haulage in the extraction and of different drying methods

    International Nuclear Information System (INIS)

    Estrada E, Amparo; Lopez O, Beatriz

    1998-01-01

    It is carried out this investigation with the purpose of determining the effects of steam haulage and of two drying methods in extracted pectin of orange shell, using the residuals of the machine processor of the Caldas region, that it uses the system In Line in the obtaining of the juice for the effect. The essential oils were eliminated the pericarp with haulage of steam, the pectin was extracted with solution to 50% of citric acid and hexametaphosphate of sodium to 0.2%. For shell weight the extraction temperature was of 80 applied during one hour and the pH of 3.4. The pectin was characterized extracted by means of determinations of humidity, ashy, nitrogen, methoxyl percentage, anhidrogalacturonic acid, pH, and acidity. Their behavior is also evaluated by means of determinations that are carried out to the gel with them prepared as its hardness, time and low temperature. They use a totally aleatory statistical design in factorial arrangement 3x2, three treatments: T1 flavedo and albedo with haulage of steam, T2 albedo with steam haulage and T3 albedo without steam haulage and two drying methods, with hot air and with microwaves. The results were analyzed by means of ANDEVA and tests of comparison of Tuckey and Scheff at a level of probability of 95%. They concluded that the steam haulage doesn't have significant influence in the characteristics of the obtained product while, the drying for microwaves influenced significantly in a positive way in the characteristics of the extracted pectin

  16. Simulating Porous Magnetite Layer Deposited on Alloy 690TT Steam Generator Tubes.

    Science.gov (United States)

    Jeon, Soon-Hyeok; Son, Yeong-Ho; Choi, Won-Ik; Song, Geun Dong; Hur, Do Haeng

    2018-01-02

    In nuclear power plants, the main corrosion product that is deposited on the outside of steam generator tubes is porous magnetite. The objective of this study was to simulate porous magnetite that is deposited on thermally treated (TT) Alloy 690 steam generator tubes. A magnetite layer was electrodeposited on an Alloy 690TT substrate in an Fe(III)-triethanolamine solution. After electrodeposition, the dense magnetite layer was immersed to simulate porous magnetite deposits in alkaline solution for 50 days at room temperature. The dense morphology of the magnetite layer was changed to a porous structure by reductive dissolution reaction. The simulated porous magnetite layer was compared with flakes of steam generator tubes, which were collected from the secondary water system of a real nuclear power plant during sludge lancing. Possible nuclear research applications using simulated porous magnetite specimens are also proposed.

  17. Code Assessment of SPACE 2.19 using LSTF Steam Generator Tube Rupture Test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minhee; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The SPACE is a best estimated two-phase three-field thermal-hydraulic analysis code used to analyze the safety and performance of pressurized water reactors. As a result of the development, the 2.19 version of the code was released through the successive various verification and validation works. The present work is on the line of expanding the work by Kim et al. In this study, results produced by the SPACE 2.19 code were compared with the experimental data from JAERI's LSTF Test Run LSTF SB-SG-06 experiment simulating a Steam Generator Tube Rupture (SGTR) transient. In order to identify the predictability of SPACE 2.19, the LSTF steam generator tube rupture test was simulated. To evaluate the computed results, LSTF SB-SG-06 test data simulating the SGTR and the RELAP5/ MOD3.1 are used. The calculation results indicate that the SPACE 2.19 code predicted well the sequence of events and the major phenomena during the transient, such as the asymmetric loop behavior, reactor coolant system cooldown and heat transfer by natural circulation, the primary and secondary system depressurization by the pressurizer auxiliary spray and the steam dump using the intact loop steam generator relief valve.

  18. The deformation of zircaloy PWR cladding with low internal pressures, under mainly convective cooling by steam

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.; Reynolds, A.E.

    1981-01-01

    The deformation behaviour is reported of specimens of Zircaloy PWR fuel cladding when directly heated in flowing steam. The range of internal pressures studied was 0.69-2.07 MPa; this extended earlier studies using higher pressures. The specimens were ramped and then held at a steady test temperature until rupture or until 600 seconds had elapsed. Under these conditions it was found that extended deformation occurred with pressures down to 1 MPa at temperatures up to 900 deg C. At lower pressures and higher temperatures there was no large extended deformation; this is believed to result from the effects of oxidation

  19. Ferritic steels for French LMFBR steam generators

    International Nuclear Information System (INIS)

    Aubert, M.; Mathieu, B.; Petrequin, P.

    1983-06-01

    Austenitic stainless steels have been widely used in many components of the French LMFBR. Up to now, ferritic steels have not been considered for these components, mainly due to their relatively low creep properties. Some ferritic steels are usable when the maximum temperatures in service do not exceed about 530 0 C. It is the case of the steam generators of the Phenix plant, where the exchange tubes of the evaporator are made of 2,25% Cr-1% Mo steel, stabilized or not by addition of niobium. These ferritic alloys have worked successfully since the first steam production in October 1973. For the SuperPhenix power plant, an ''all austenitic stainless alloy'' apparatus has been chosen. However, for the future, ferritic alloys offer potential for use as alternative materials in the evaporators: low alloys steels type 2,25% Cr-1% Mo (exchange tubes, tube-sheets, shells), or at higher chromium content type 9% Cr-2% Mo NbV (exchange tubes) or 12M Cr-1% Mo-V (tube-sheets). Most of these steels have already an industrial background, and are widely used in similar applications. The various potential applications of these steels are reviewed with regards to the French LMFBR steam generators, indicating that some points need an effort of clarification, for instance the properties of the heterogeneous ferritic/austenitic weldments

  20. Future development of large steam turbines

    International Nuclear Information System (INIS)

    Chevance, A.

    1975-01-01

    An attempt is made to forecast the future of the large steam turbines till 1985. Three parameters affect the development of large turbines: 1) unit output; and a 2000 to 2500MW output may be scheduled; 2) steam quality: and two steam qualities may be considered: medium pressure saturated or slightly overheated steam (light water, heavy water); light enthalpie drop, high pressure steam, high temperature; high enthalpic drop; and 3) the quality of cooling supply. The largest range to be considered might be: open system cooling for sea-sites; humid tower cooling and dry tower cooling. Bi-fluid cooling cycles should be also mentioned. From the study of these influencing factors, it appears that the constructor, for an output of about 2500MW should have at his disposal the followings: two construction technologies for inlet parts and for high and intermediate pressure parts corresponding to both steam qualities; exhaust sections suitable for the different qualities of cooling supply. The two construction technologies with the two steam qualities already exist and involve no major developments. But, the exhaust section sets the question of rotational speed [fr