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.

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.

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.

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.

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.

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

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

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)

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

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

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.

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

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)

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

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.

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

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)

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

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.

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

Analysis Of Feedwater Line Break Of APR1400 By MARS Code

International Nuclear Information System (INIS)

Nguyen Thi Thanh Thuy; Le Dai Dien, Hoang Minh Giang

2011-01-01

This paper will deal with analysis of Feed water Line Break problem (FWLB) of the APR 1400 NPP with initial conditions: operation at 100% of power, double-ended break area of 0.058 m 2 and the break location of the feedwater line between the check valve and the steam generator. The analysis was simulated by MARS code through two step: calculation for steady state and calculation for transient state with initial condition mentioned. Some output result were presented with explanation: sequence of events corresponding to the time of the accident, the system behavior as temperature, pressure, steam generator water levels as well as DNBR, etc. before and after the accident. (author)

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

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)

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.

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

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

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.

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

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

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

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.

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

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.

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

Experiment data of 200% recirculation pump discharge line break integral test run 961 with HPCS failure at ROSA-III and comparison with results of suction line break tests

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Tasaka, Kanji; Nakamura, Hideo; Anoda, Yoshinari; Kumamaru, Hiroshige; Murata, Hideo; Yonomoto, Taisuke; Shiba, Masayoshi

1984-03-01

This report presents the experimental data of RUN 961, a 200% double-ended break test at the recirculation pump discharge line in the ROSA-III test facility. The ROSA-III test facility is a volumetrically scaled (1/424) system of the BWR/6. The facility has the electrically heated core, the break simulator and the scaled ECCS (Emergency Core Cooling System). The MSIV (Main Steam Isolation Valve) closure and the ECCS actuation were tripped by the liquid level in the upper downcomer. The PCT (Peak Cladding Temperature) was 894 K, which was 107 K higher than a 200% pump suction line break test (RUN 926) due to the smaller depressurization rate. The effect of break location on transient LOCA phenomena was clarified by comparing the discharge and suction break tests. The whole core was quenched 71 s after LPCI actuation and the effectiveness of ECCS has been confirmed. (author)

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)

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)

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

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

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

BWR recirculation loop discharge line break LOCA tests with break areas of 50 and 100% assuming HPCS failure at ROSA-III test facility

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Tasaka, Kanji; Yonomoto, Taisuke; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Murata, Hideo; Shiba, Masayoshi; Iriko, Masanori.

1985-03-01

This report presents the experimental results of RUN 962 and RUN 963 in ROSA-III program, which are 50 and 100 % break LOCA tests at the BWR recirculation pump discharge line, respectively. The ROSA-III test facility simulates a volumetrically scaled (1/424) BWR system and has four half-length electrically heated fuel bundles, two active recirculation loops, three types of ECCSs and steam and feedwater systems. The experimental data of RUN 962 and RUN 963 were compared with those of RUN 961, a 200 % discharge line break test to study the break area effects on the transient thermal hydraulic phenomena. The least flow areas at the jet pump drive nozzles and recirculation pump discharge nozzle in the broken recirculation loop limitted the discharge flows from the pressure vessel and the depressurization rate in the 100 and 200 % break tests, whereas the least flow area at break nozzle limitted the depressurization rate in the 50 % break test. The highest PCT was observed in the 50 % break test among the three tests. (author)

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)

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

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

Analysis methods for evaluating leak-before-break in U-tube steam generators

International Nuclear Information System (INIS)

Griesbach, T.; Cipolla, R.

1985-01-01

In recent years, there has been an increased incidence of cracking in steam generator tubes. As a result, there has been increased effort in assuring that cracks in steam generator tubes will leak well in advance of significant loss in structural integrity. Demonstrating a leak-before-break condition is an integrated analysis process that utilizes several engineering disciplines, specifically, materials engineering, fracture mechanics, stress analysis, and fluid mechanics. The output from a leak-before-break assessment is typically depicted in terms of available margins against failure and measurable or detectable leak rate. In this paper, the analysis methods for performing a leak-before-break analysis for the U-tubes of a recirculating steam generator are presented. The results from generic analysis for the first row U-tubes illustrates the analysis techniques. Because of realistic input values used herein, these results also suggest that large leak rates are possible from cracks in U-bend regions, yet these cracks are small relative to their critical size for failure. Hence, orderly shutdowns can be completed prior to the point when tube bursting is of concern

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)

UPTF/TEST10B/RUN081, Steam/Water Flow Phenomena Reflood PWR Cold Leg Break LOCA

International Nuclear Information System (INIS)

1998-01-01

1 - Description of test facility: The Upper Plenum Test Facility (UPTF) is a geometrical full-scale simulation of the primary system of the four-loop 1300 MWe Siemens/KWU pressurized water reactor (PWR) at Grafenrheinfeld. The test vessel, upper plenum and its internals, downcomer, primary loops, pressurizer and surge line are replicas of the reference plant. The core, coolant pumps, steam generators and containment of a PWR are replaced by simulators which simulate the boundary and initial conditions during end-of-blowdown, refill and reflood phase following a loss-of-coolant accident (LOCA) with a hot or cold leg break. The break size and location can be simulated in the broken loop. The emergency core coolant (ECC) injection systems at the UPTF are designed to simulate the various ECC injection modes, such as hot leg, upper plenum, cold leg, downcomer or combined hot and cold leg injection of different ECC systems of German and US/Japan PWRs. Moreover, eight vent valves are mounted in the core barrel above the hot leg nozzle elevation for simulation of ABB and B and W PWRs. The UPTF primary system is divided into the investigation and simulation areas. The investigation areas, which are the exact replicas of a GPWR, consist of the upper plenum with internals, hot legs, cold legs and downcomer. The realistic thermal-hydraulic behavior in the investigation areas is assured by appropriate initial and boundary conditions of the area interface. The boundary conditions are realized by above mentioned simulators, the setup and the operation of which are based on small-scale data and mathematical models. The simulation areas include core simulator, steam generator simulators, pump simulators and containment simulator. The steam production and entrainment in a real core during a LOCA are simulated by steam and water injection through the core simulator. 2 - Description of test: Investigation of steam/water flow phenomena at the upper tie plate and in the upper plenum and

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.

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

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.

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

Thermal-hydraulic behavior on break simulation of steam generator U-tube

International Nuclear Information System (INIS)

Seul, Kwang Won; Bang, Young Seok; Lee, Sukho; Kim, Hho Jung

1995-01-01

The thermal-hydraulic behavior depending on the break simulation in a steam generator U-tube was investigated and identified the code predictability on plant responses during SGTR accident. The calculated results were compared and assessed with LSTF SB-SG-06 test data. The RELAP5/MOD3.1 code well predicted the sequence of events and the significant phenomena, such as the asymmetric loop behavior, the RCS cooldown and heat transfer by the natural circulation, and system depressurization, even though there were some differences from the experimental data. The break flowrate was found to be sensitive to the break model and affected the system behavior

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.

Skills Associated with Line Breaks in Elite Rugby Union

Directory of Open Access Journals (Sweden)

Steve den Hollander, James Brown, Michael Lambert, Paul Treu, Sharief Hendricks

2016-09-01

Full Text Available The ability of the attacking team to break through the defensive line is a key indicator of success as it creates opportunities to score tries. The aim of this study was to analyse line breaks and identify the associated skills and playing characteristics. The 2013 Super Rugby season (125 games was analysed, in which 362 line breaks were identified and coded using variables that assessed team patterns and non-contact attacking skills in the phases preceding the line break. There was an average of 3 line breaks per game, with 39% of line breaks resulting in a try. Line breaks occurred when the ball-carrier was running fast [61%, x2(4 = 25.784, p = 0.000, Cramer’s v = 0.1922, weak]. At a moderate distance, short lateral passes (19% and skip passes (15% attributed to the highest percentage of line breaks [x2(26 = 50.899, p = 0.036, Cramer’s v = 0.2484, moderate]. Faster defensive line speeds resulted in more line breaks [x2(12 = 61.703, p < 0.001, Cramer’s v = 0.3026, moderate]. Line breaks are associated with overall team success and try scoring opportunities. Awareness of the defenders line speed and depth, fast running speed when receiving the ball and quick passing between attackers to the outside backs creates line break opportunities. During training, coaches should emphasise the movement speed of the ball between attackers and manipulate the speed and distance of the defenders.

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)

Line-breaking algorithm enhancement in inverse typesetting paradigma

Directory of Open Access Journals (Sweden)

Jan Přichystal

2007-01-01

Full Text Available High quality text preparing using computer desktop publishing systems usually uses line-breaking algorithm which cannot make provision for line heights and typeset paragraph accurately when composition width, page break, line index or other object appears. This article deals with enhancing of line-breaking algorithm based on optimum-fit algorithm. This algorithm is enhanced with calculation of immediate typesetting width and thus solves problem of forced change. Line-breaking algorithm enhancement causes expansion potentialities of high-quality typesetting in cases that have not been yet covered with present typesetting systems.

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.

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)

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

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

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

Audit calculation of the limiting CESSAR feedwater-line-break transient with RELAP5/MOD1

International Nuclear Information System (INIS)

Chung, K.S.; Kennedy, M.F.; Guttmann, J.

1983-01-01

Argonne National Laboratory (ANL) performed a series of audit calculations of the limiting FLB transient presented in Appendix 15B to the CESSAR FSAR, supported by a limited number of additional calculations to investigate the sensitivity of the results (in terms of peak primary reactor system pressure) to break area and reactor trip time. The latter calculations were performed to quantify potential benefits in crediting reactor tip on low steam generator downcomer water level, which occurs earlier than the trip shown in the limiting FSAR transient, which tripped on high pressurizer pressure. These calculations were performed to verify the break spectrum results presented by C-E and to insure that C-E did indeed analyze the limiting transient. All of the ANL calculations were performed with RELAP5/MOD1 (cycle 18) using an input deck developed at ANL from CESSAR plant data provided by C-E. In this paper we compare the results and provide insight into the generic behavior of a Feedwater Line Break transient

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)

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

Evaluation of ATLAS 100% DVI Line Break Using TRACE Code

International Nuclear Information System (INIS)

Huh, Byung Gil; Bang, Young Seok; Cheong, Ae Ju; Woo, Sweng Woong

2011-01-01

ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation) is an integral effect test facility in KAERI. It had installed completely to simulate the accident for the OPR1000 and the APR1400 in 2005. After then, several tests for LBLOCA, DVI line break have been performed successfully to resolve the safety issues of the APR1400. Especially, a DVI line break is considered as another spectrum among the SBLOCAs in APR1400 because the DVI line is directly connected to the reactor vessel and the thermal hydraulic behaviors are expected to be different from those for the cold leg injection. However, there are not enough experimental data for the DVI line break. Therefore, integral effect data for the DVI line break of ATLAS is very useful and available for an improvement and validation of safety codes. For the DVI line break in ATLAS, several analyses using MARS and RELAP codes were performed in the ATLAS DSP (Domestic Standard Problem) meetings. However, TRACE code has still not used to simulate a DVI line break. TRACE code has developed as the unified code for the reactor thermal hydraulic analyses in USNRC. In this study, the 100% DVI line break in ATLAS was evaluated by TRACE code. The objectives of this study are to identify the prediction capability of TRACE code for the major thermal hydraulic phenomena of a DVI line break in ATLAS

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.

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

BWR 200 % recirculation pump suction line break LOCA tests, RUNs 942 and 943 at ROSA-III without HPCS

International Nuclear Information System (INIS)

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

1986-03-01

This report presents the experimental results of RUNs 942 and 943 in ROSA-III program, which are 200 % recirculation pump suction line break LOCA tests with assumption of HPCS failure. 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, ECCS's, and steam and feedwater systems. Effects of initial core void distribution and other fluid conditions on overall LOCA phenomena with special interest on transient core cooling phenomena were investigated by comparing the present test results with those of RUN 926, a 200 % suction line break test with standard initial fluid conditions. The initial core outlet quality was changed between 5 % and 43 %. As conclusions, (1) the initial lower core flow and higher void fraction affected significantly the core cooling conditions and resulted in earlier and higher PCT. (2) The lower plenum flashing temporarily contributed to cool down the core. (3) Flashing of remained hot water in the feedwater line affected slightly the pressure response and delayed the actuation of LPCI by 11 seconds. (4) The whole core was completely cooled down within 104 seconds after the LPCI actuation in these large break tests. (author)

An SBLOCA Test for Shutdown Cooling Line Break Using the SMART-ITL Facility

Energy Technology Data Exchange (ETDEWEB)

Bae, Hwang; Kim, Dong Eok; Ryu, Sung Uk; Shin, Yong Cheol; Ko, Yung Joo; Yi, Sung Jae; Park, Hyun Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

The objectives of SMART-ITL are to investigate and understand the integral performance of the reactor systems and components, and the thermalhydraulic phenomena occurring in the system during normal, abnormal, and emergency conditions, and to verify the system safety during various design basis events of SMART. Its height was preserved and its area and volume were scaled down to 1/49 compared with the SMART prototype plant. The SMART-ITL consists of a primary system including a reactor pressure vessel with a pressurizer, four steam generators and four main coolant pumps, a secondary system, a safety system, and an auxiliary system. The SMART was installed at KAERI and several transient tests were recently finished. In this paper, the test results for a steady-state operation and a transient of the small break loss of coolant accident (SBLOCA) are discussed. An SBLOCA test simulating the shutdown cooling line break was performed using SMART-ITL properly. All parameters were in good agreement with the target values during the steady-state operation period. The pressures and temperatures show reasonable behaviors during the SBLOCA test. SMART (System-integrated Modular Advanced ReacTor) which was designed by KAERI is an integral type reactor. The standard design approval for the SMART design was issued on July 4th of 2012 by a Korean regulatory body, the Nuclear Safety and Security Commission (NSSC). The main components including a pressurizer, steam generators, and reactor coolant pumps are installed in a reactor pressure vessel, and there are no large-size pipes. The safety systems could be simplified as an LBLOCA (Large-Break Loss of Coolant Accident) scenario is inherently excluded. An integral-effect test loop for SMART (SMART-ITL, or FESTA) was designed to simulate the integral thermal-hydraulic behavior of SMART. The SMART-ITL has been designed using a volume scaling methodology.

Break preclusion concept and its application to the EPRTM reactor

Energy Technology Data Exchange (ETDEWEB)

Chapuliot, S., E-mail: stephane.chapuliot@areva.com; Migné, C.

2014-04-01

This paper provides a synthesis of the technical basis supporting the break preclusion concept and its implementation on the Main Coolant Lines and Main Steam Lines of the EPR™ reactor. In a first step, it describes the background of the break preclusion concept, and then it details the requirements associated to its implementation in a Defense In Depth approach.In second steps, main benefits and few illustrative examples are given for the MCL.

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

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)

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

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

Evaluation of reflooding effects on an overheated boiling water reactor core in a small steam-line break accident using MAAP, MELCOR, and SCDAP/RELAP5 computer codes

International Nuclear Information System (INIS)

Lindholm, I.; Pekkarinen, E.; Sjoevall, H.

1995-01-01

Selected core reflooding situations were investigated in the case of a Finnish boiling water reactor with three severe accident analysis computer codes (MAAP, MELCOR, and SCDAP/RELAP5). The unmitigated base case accident scenario was a 10% steam-line break without water makeup to the reactor pressure vessel initially. The pumping of water to the core was started with the auxiliary feed water system when the maximum fuel cladding temperature reached 1,500 K. The auxiliary feedwater system pumps water (temperature 303 K) through the core spray spargers (core spray) on the top of the core and through feedwater nozzles to the downcomer (downcomer injection). The scope of the study was restricted to cases where the overheated core was still geometrically intact at the start of the reflooding. The following different core reflooding situations were investigated: (1) auxiliary feedwater injection to core spray (45 kg/s); (2) auxiliary feedwater injection to downcomer (45 kg/s); (3) auxiliary feedwater injection to downcomer (45 kg/s) and to core spray (45 kg/s); (4) no reflooding of the core. All the three codes predicted debris formation after the water addition, and in all MAAP and MELCOR reflooding results the core was quenched. The major difference between the code predictions was in the amount of H 2 produced, though the trends in H 2 production were similar. Additional steam production during the quenching process accelerated the oxidation in the unquenched parts of the core. This result is in accordance with several experimental observations

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

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.

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

Influence of liquid holdup in steam generator U-tubes on small break LOCA severity

International Nuclear Information System (INIS)

Leonard, M.T.; Perryman, J.L.; Johnson, G.W.

1983-01-01

The severity of small cold leg break loss-of-coolant accidents has been shown to be influenced by liquid holdup in steam generator U-tubes during pump suction loop seal formation in two experiments performed in the Semiscale Mod-2A facility. The core coolant level can be depressed lower than previously thought possible due to a positive hydrostatic head across the steam generators caused by delayed drainage of liquid from the upflow side of the U-tubes. The significance of a lower core coolant level depression is the potential for a more severe temperature excursion occurring during the coolant boiloff phase subsequent to loop seal clearing and prior to accumulator injection. Presented in this paper are the experimental data analysis and supporting computer code calculations that led to these conclusions

Evaluation of PWR response to main-steamline break with concurrent steam-generator tube rupture and small-break LOCA

International Nuclear Information System (INIS)

Laaksonen, J.T.; Sheron, B.W.

1982-12-01

In 1980, the NRC staff raised a potential safety issue involving a coincident steamline break, steam generator tube rupture, and small-break loss-of-coolant accident (LOCA). The bases for this concern were that the system response, primarily the maintenance of core cooling, was unanalyzed and the adequacy of the present guidance to operators to respond to combination LOCAs was unknown. This report discusses the staff evaluations performed to assess the system response and the adequacy of the present emergency operator guidelines. In all of the analyzed cases the primary coolant shrinkage, caused by overcooling, and the simultaneous loss of coolant can be compensated by the high pressure emergency core cooling system. The core remains covered with liquid, and the primary coolant remains subcooled, except in the vessel upper head. If the steamline break is outside the containment and cannot be isolated, the radiological consequences could be more severe than in any accident currently analyzed in a typical plant Final Safety Evaluation Report (FSAR). To decrease the risk of elevated offsite releases, an early diagnosis of the tube rupture has to be ensured. This can be done by upgrading operator instructions. The appropriate mitigating actions are in the existing instructions

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

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

Processing load imposed by line breaks in English temporal wh-questions

Directory of Open Access Journals (Sweden)

Masako Hirotani

2016-10-01

Full Text Available Prosody plays an important role in online sentence processing both explicitly and implicitly. It has been shown that prosodically packaging together parts of a sentence that are interpreted together facilitates processing of the sentence. This applies not only to explicit prosody but also implicit prosody. The present work hypothesizes that a line break in a written text induces an implicit prosodic break, which, in turn, should result in a processing bias for interpreting English wh-questions. Two experiments – one self-paced reading study and one questionnaire study – are reported. Both supported the line break hypothesis mentioned above. The results of the self-paced reading experiment showed that unambiguous wh-questions were read faster when the location of line breaks (or frame breaks matched the scope of a wh-phrase (main or embedded clause than when they did not. The questionnaire tested sentences with an ambiguous wh-phrase, one that could attach either to the main or the embedded clause. These sentences were interpreted as attaching to the main clause more often than to the embedded clause when a line break appeared after the main verb, but not when it appeared after the embedded verb.

Break size effect on the transient thermal-hydraulic behavior during the steam generator tube rupture accident

International Nuclear Information System (INIS)

Kang, K.H.; Park, H.S.; Cho, S.; Choi, N.H.; Chu, I.C.; Yun, B.J.; Kim, K.D.; Kim, Y.S.; Baek, W.P.; Choi, K.Y.

2011-01-01

In order to simulate the SGTR accident of the APR1400, integral effect tests were performed by simulating a double-ended rupture of a single and five U-tubes. Following the reactor trip, the primary system pressure decreased and the secondary system pressure increased until the MSSVs was opened to reduce the secondary system pressure. Break area affected the timings of the major events observed in the tests. Less heat transfer to the secondary side caused by earlier actuation of the safety injection pumps had more influence on the secondary pressure of the affected steam generator than the break flow. (author)

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.

SPES-2, AP600 intergral system test S01007 2 inch CL to core make-up tank pressure balance line break

Energy Technology Data Exchange (ETDEWEB)

Bacchiani, M.; Medich, C.; Rigamonti, M. [SIET S.p.A. Piacenza (Italy)] [and others

1995-09-01

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 (Ente Nazionale per l` Energia Elettrica), ENEA (Enter per le numove Technlogie, l` Energia e l` Ambient), Siet (Societa Informazioni Esperienze Termoidraulich) and ANSALDO developed an experimental program to test the integrated behaviour 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 passive and active safety systems and a main steam line break transient to demonstrate the boration 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 behaviour. Cold and hot shakedown tests have been performed on the facility to check the characteristics of the plant before starting the experimental campaign. The paper first presents a description of the SPES-2 test facility then the main results of S01007 test {open_quotes}2{close_quotes} Cold Leg (CL) to Core Make-up Tank (CMT) pressure balance line break{close_quotes} are reported and compared with predictions performed using RELAP5/mod3/80 obtained by ANSALDO through agreement with U.S.N.R.C. (U.S. Nuclear Regulatory Commission). The SPES-2 nodalization and all the calculations here presented were performed by ANSALDO and sponsored by ENEL as a part of pre-test predictions for SPES-2.

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

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

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

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

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

Water Hammer Mitigation on Postulated Pipe Break of Feed Water System

International Nuclear Information System (INIS)

Seong, Ho Je; Woo, Kab Koo; Cho, Keon Taek

2008-01-01

The Feed Water (FW) system supplies feedwater from the deaerator storage tank to the Steam Generators(S/G) at the required pressure, temperature, flow rate, and water chemistry. The part of FW system, from the S/G to Main Steam Valve House just outside the containment building wall, is designed as safety grade because of its safety function. According to design code the safety related system shall be designed to protect against dynamic effects that may results from a pipe break on high energy lines such as FW system. And the FW system should be designed to minimize blowdown volume of S/G secondary side during the postulated pipe break. Also the FW system should be designed to prevent the initiation or to minimize the effects of water hammer transients which may be induced by the pipe break. This paper shows the results of the hydrodynamic loads induced by the pipe break and the optimized design parameters to mitigate water hammer loads of FW system for Shin-Kori Nuclear Power Plant Unit 3 and 4 (SKN 3 and 4)

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

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

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

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

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)

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)

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

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

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)

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)

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

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

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.

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

Hydrogen and steam distribution following a small-break LOCA in large dry containment

Institute of Scientific and Technical Information of China (English)

DENG Jian; CAO Xuewu

2007-01-01

The hydrogen deflagration is one of the major risk contributors to threaten the integrity of the containment in a nuclear power plant, and hydrogen control in the case of severe accidents is required by nuclear regulations.Based on the large dry containment model developed with the integral severe-accident analysis tool, a small-break loss-of-coolant-accident (LOCA) without HPI, LPI, AFW and containment sprays, leading to the core degradation and large hydrogen generation, is calculated. Hydrogen and steam distribution in containment compartments is investigated. The analysis results show that significant hydrogen deflagration risk exits in the reactor coolant pump (RCP)compartment and the cavity during the early period, if no actions are taken to mitigate the effects of hydrogen accumulation.

Double-Ended Break Test of an 8.5 inch Direct Vessel Injection Line using the ATLAS

International Nuclear Information System (INIS)

Choi, Ki Yong; Kang, Kyoung Ho; Kim, Bok Deuk; Kim, Yeon Sik; Min, Kyoung Ho; Park, Choon Kyoung; Park, Hyun Sik; Baek, Won Pil; Cho, Seok; Choi, Nam Hyun

2010-01-01

A thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been constructed at the Korea Atomic Energy Research Institute (KAERI). It is a 1/2 reduced height and 1/288-volume scaled test facility with respect to the APR1400, an evolutionary pressurized water reactor developed by the Korean industry. In 2008, an integral effect test for simulating a guillotine break of a DVI line of the APR1400 was carried out as the first DVI test item, named as SB-DVI-03, on May, 2008. With an improvement on the break flow measuring system, the second DVI test for a guillotine break of a DVI line of the APR1400, named as SB-DVI-08, was conducted for repeatability. The present data is the first integral effect test data of its kind for simulating a DVI line break accident. It will help in understanding the thermal hydraulic phenomena occurring during the DVI line break accident. A post-test calculation was performed with a best-estimate safety analysis code MARS 3.1 to examine its prediction capability and to identify any code deficiencies for the thermal hydraulic phenomena occurring during the DVI line break accidents. The present integral effect test data will be used to validate the current safety analysis methodology for the DVI line break accident

Comparison of the DVI line break LOCA with the equivalent cold leg break with the ATLAS facility

International Nuclear Information System (INIS)

Choi, K. Y.; Cho, S.; Kang, K. H.; Park, H. S.; Kim, Y. S.; Baek, W. P.

2010-01-01

The APR1400 (Advanced Power Reactor, 1400 MWe) adopts a DVI (Direct Vessel Injection) method for ECC (Emergency Core Cooling) water delivery rather than a conventional CLI (Cold Leg Injection) method as an advanced safety feature. The break scenario of one DVI nozzle is taken into account in the small break LOCA analysis. Transient behavior during the DVI line breaks needs to be investigated and compared with the equivalent break on the cold leg. An 8.5-inch double-ended break of one DVI nozzle was simulated with the ATLAS, and a counterpart test for the DVI break was performed at the cold leg with the equivalent break size for comparison. This comparison will contribute to enhancing a comprehensive understanding of the thermal hydraulic behavior during transients. A constructed integral effect database is also used to validate the existing conservative safety analysis methodology and to develop a best-estimate safety analysis methodology for small-break LOCAs. A post-test calculation was performed with a best-estimate safety analysis code, MARS 3.1, in order to examine its prediction capability and to identify any code deficiencies for thermal hydraulic phenomena occurring during the transient. (authors)

Characteristics of steam jet impingement on annulus

International Nuclear Information System (INIS)

Yoon, Sang H.; Kim, Won J.; Suh, Kune Y.; Song, Chul H.

2004-01-01

The steam jet impingement occurs when the steam through the cold leg from the steam generator strikes the inner reactor barrel during the reflood phase of a loss-of-coolant accident (LOCA), which is a characteristic behavior for the APR1400 (Advanced Power Reactor 1400 MWe). In the cold leg break LOCA, the steam and water flows in the downcomer are truly multidimensional. The azimuthal velocity distribution of the steam flow has an important bearing on the thermal hydraulic phenomena such as the emergency coolant water direct bypass, sweepout, steam condensation, and so forth. The investigation of jet flow is required to determine the steam path and momentum reduction rate after the impingement. For the observation of the steam behavior near the break, the computational fluid dynamic (CFD) analysis has been carried out using CFX5.6. The flow visualization and analysis demonstrate the velocity profiles of the steam flow in the annulus region for the same boundary conditions. Pursuant to the CFD results, the micro-Pitot tubes were positioned at varying angles, and corrected for their sensitivity. The experiments were carried out to directly measure the pressure differential and to visualize the flow utilizing a smoke injection method. Results from this study are slated to be applied to MARS, which is a thermal hydraulic system code for the best-estimate analysis. The current one- or two-dimensional analysis in MARS was known to distort the local flow behavior. To enhance prediction capability of MARS, it is necessary to inspect the steam path in the break flow and mechanically simulate the momentum variation. The present experimental and analytical results can locally be applied to developing the engineering models of specific and essential phenomena. (author)

Simulation of an SBLOCA Test of Shutdown Cooling System Line Break with the SMARTITL Facility using the MARS-KS Code

Energy Technology Data Exchange (ETDEWEB)

Cho, Yeon-Sik; Suh, Jae-Seung [System Engineering and Technology, Daejeon (Korea, Republic of); Bae, Hwang; Ryu, Sung-Uk; Yi, Sung-Jae; Park, Hyun-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

An LBLOCA (Large-Break Loss of Coolant Accident) was inherently eliminated in the design stage. The SMART design has a thermal power of 330MW. Its core exit temperature and pressurizer pressure are 323 .deg. C and 15MPa during normal operating conditions, respectively. An integral-effect test loop for SMART (SMARTITL), called FESTA (Facility for Experimental Simulation of Transients and Accidents), was designed to simulate the integral thermal-hydraulic behavior of SMART. The objectives of SMART-ITL are to investigate and understand the integral performance of reactor systems and components, and the thermal-hydraulic phenomena occurring in the system during normal, abnormal, and emergency conditions, and to verify the system safety during various design basis events of SMART. SMART-ITL with four steam generators and PRHRS, has an advantage for a multi-loop effect compared with VISTA-ITL with a single loop. The integral-effect test data will also be used to validate the related thermal-hydraulic models of the safety analysis code such as TASS/SMR-S which is used for a performance and accident analysis of the SMART design. In addition, a scoping analysis on the scaling difference between the standard design of SMART and the basic design of SMART-ITL was performed for an SBLOCA (Small-Break Loss of Coolant Accident) scenario using a best-estimate safety analysis code, MARS-KS. This paper introduces a comparison of an SBLOCA test of a shutdown cooling system line break using SMART-ITL with its post-test calculation using the MARS-KS code. An SBLOCA test and its post-test calculation were successfully performed using the SMART-ITL facility and MARS-KS code. The SBLOCA break is a guillotine break, and its location is on the SCS line (nozzle part of the RCP suction). The steady-state conditions were achieved to satisfy the initial test conditions presented in the test requirement and its boundary conditions were properly simulated.

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.

Research on alarm triggered fault-diagnosis expert system for U-shaped tube breaking accident of steam generators

International Nuclear Information System (INIS)

Qian Hong; Luo Jianbo; Jin Weixiao; Zhou Jinming; Wang Du

2015-01-01

According to the U-shaped tube breaking accident of steam generator (SGTR), this paper designs a fault-diagnosis expert system based on the alarm triggering. By analyzing the fault mechanism of SGTR accidents, the fault symptom is obtained. The parameters of the belief rule are set up based on the simulation experiment. The information fusion is conducted on the fault-diagnosis results from multiple expert systems to obtain the final diagnose result. The test result shows that the expert system can diagnose the SGTR accident accurately and rapidly, and provide with the operation guidance. (authors)

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)

Multiple pathways of DNA double-strand break processing in a mutant Indian muntjac cell line

International Nuclear Information System (INIS)

Bouffler, S.D.; Jha, B.; Johnson, R.T.

1990-01-01

DNA break processing is compared in the Indian muntjac cell lines, SVM and DM. The initial frequencies and resealing of X-ray generated single- and double-strand breaks are similar in the two cell lines. Inhibiting the repair of UV damage leads to greater double-strand breakage in SVM than in DM, and some of these breaks are not repaired; however, repair-associated single-strand breakage and resealing are normal. Dimethylsulfate also induces excess double-strand breakage in SVM, and these breaks are irreparable. Restricted plasmids are reconstituted correctly in SVM at approximately 30% of the frequency observed in DM. Thus SVM has a reduced capacity to repair certain types of double-strand break. This defect is not due to a DNA ligase deficiency. We conclude that DNA double-strand breaks are repaired by a variety of pathways within mammalian cells and that the structure of the break or its mode of formation determines its subsequent fate

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.

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

A nuclear power unit with a Babcock type steam generating system-analysis of the break-down in the Three Mile Island power plant

International Nuclear Information System (INIS)

Werner, A.

1980-01-01

Installations of the primary and the secondary circuits and basic automatic control and protection systems for a nuclear power unit with Babcock type vertical, once-through steam generator are described. On this background the course of the break-down in the Three Mile Island power plant at Harrisburg is presented and analysed. (author)

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)

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

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

Investigation of break location effects on thermal-hydraulics during intermediate break loss-of-coolant accident experiments at ROSA-III

International Nuclear Information System (INIS)

Koizumi, Yasuo; Tasaka, Kanji

1986-01-01

The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR/6) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. Break location effects on thermal-hydraulics during intermediate LOCAs were investigated by using four experiments at the ROSA-III, the 15 and 25 % main recirculation pump suction line break (MRPS-B) experiments, the 21 % single-ended jet pump drive line break (JPD-B) experiment and the 15 % main steam line break (MSL-B) experiment. Water injection from the high pressure core spray (HPCS) was not used in any of the experiments. Failure of ECCS actuation by the high containment pressure was also assumed in the tests. In the MRPS-B experiments, the discharge flow turned from low quality fluid to high quality fluid when the downcomer water level dropped to the main recirculation line outlet elevation, which suppressed coolant loss from the vessel and the core. In the JPD-B experiment, the jet pump drive nozzle was covered with low quality fluid and low quality fluid discharge continued even after the downcomer water level reached the jet pump suction elevation. Low quality fluid discharge ceased after the ADS actuation. It suggestes that the JPD-B LOCA has the possibility of causing larger and more severe core dryout and cladding temperature excursion than the MRPS-B LOCA. The MSL-B LOCA was characterized by mixture level swell in the downcomer and the core. The core mixture level swell resulted in the much later core dryout initiation than that in the MRPS-B LOCA, however, ECCS actuation was also delayed because of slow downcomer water level drop. (author)

FIST/6IB1, BWR/6 System Responses to Intermediate Break in Recirculation Suction Line LINE

International Nuclear Information System (INIS)

1993-01-01

1 - Description of test facility: BWR/6-218 standard plant. A full size bundle with electrically heated rods is used to simulate the reactor core. A scaling ratio of 1/624 is applied in the design of the system components. Key features of the FIST facility include: (1) Full height test vessel and internals; (2) correctly scaled fluid volume distribution; (3) simulation of ECCS, S/RV, and ADS; (4) level trip capability; (5) heated feedwater supply system, which provides the capability for steady state operation. 2 - Description of test: Test 6IB1 investigates system responses to an intermediate break in the recirculation suction line. BWR system licensing evaluations for various size recirculation break LOCA's indicates that a break size of about 0.2 sq.ft., without LPCS operation, is the highest PCT case for the intermediate break LOCA. Test 6IB1 simulates this event

Application of the leak-before-break concept to steam generator tubes

International Nuclear Information System (INIS)

Keim, E.; Kastner, W.

1994-01-01

The Leak-Before-Break (LBB) behaviour of a piping component means that the length of a crack resulting in a leak is smaller than the critical crack length and that the leak is safety detectable by a suitable monitoring system. The LBB-concept of Siemens/KWU is based on computer codes for the evaluation of critical crack lengths, crack openings, leakage areas and leakage rates, developed by Siemens/KWU. The fracture mechanics analysis supplies the input for the thermal-hydraulic analysis. The resulting leakage rate related to the crack length of a longitudinal or circumferential crack and the minimum detectable values of leakage rate and crack length lead to two criteria, which allow for the LBB-behaviour of the pipe: - the critical crack length must be larger than the crack length being safety detected by leakage monitoring systems (LMS) - the critical crack length must be larger than the crack length being safety detected by non-destructive examination (NDE). This LBB-concept is applied to steam generator (SG) tubes. Two examples, which will be presented, show that this concept is a very useful and effective tool which allows the prediction of LBB-behaviour of SG tubes. (Author)

An Overall Investigation of Direct Vessel Injection Line Break Accidents of the ATLAS Facility

International Nuclear Information System (INIS)

Kim, Yeon-Sik; Choi, Ki-Yong; Cho, Seok; Kim, Bok-Deuk

2015-01-01

For parametric evaluations of direct vessel injection (DVI) line break scenarios, the pressurizer (PZR) pressure, core collapsed water level, and peak cladding temperature were investigated between the analyses and tests. The PZR pressure was mainly dependent upon the break flow model, e.g., discharge coefficient of the Henry-Fauske critical model. The core collapsed water level and peak cladding temperature were mainly dependent on the counter-current flow limit (CCFL) option of the fuel alignment plate (FAP). The CCFL option of the cross-over leg (COL) affected the PZR pressure owing to the loop seal clearings and seemed to have little effect on the core collapsed water level. Proper C d values and applicable CCFL options were summarized. C d values seemed to be dependent on the sizes of the DVI line break. The PZR pressure was mainly dependent on the break flow model, e.g., the discharge coefficient of the Henry-Fauske critical model. The core collapsed water level and peak cladding temperature were mainly dependent on the CCFL option of the FAP. The CCFL option of the COL affected the PZR pressure owing to loop seal clearings and seemed to have little effect on the core collapsed water level. From parametric evaluations, proper C d values and applicable CCFL options were suggested. The C d values seemed to be dependent on the sizes of the DVI line break. Although there was little difference in the CCFL options of the COL, the Ku-option was the preferred one for COLs' CCFL option. The CCFL options of the FAP appeared sensitive to the core collapsed water level and peak cladding temperature. The Ku-option of the FAP tended to negatively exaggerate the core behavior and showed excessively conservative results, especially on the peak cladding temperature. For smaller breaks, e.g., 25%, NA- and Wa-options would be applicable for the FAP. However, for larger breaks, e.g., 50%, the Wa-option of the FAP was the preferred one. Comparisons between the tests and

Natural X-ray lines from the low scale supersymmetry breaking

Energy Technology Data Exchange (ETDEWEB)

Kang, Zhaofeng, E-mail: zhaofengkang@gmail.com [Center for High-Energy Physics, Peking University, Beijing 100871 (China); School of Physics, Korea Institute for Advanced Study, Seoul 130-722 (Korea, Republic of); Ko, P., E-mail: pko@kias.re.kr [School of Physics, Korea Institute for Advanced Study, Seoul 130-722 (Korea, Republic of); Li, Tianjun, E-mail: tli@itp.ac.cn [State Key Laboratory of Theoretical Physics and Kavli Institute for Theoretical Physics China (KITPC), Institute of Theoretical Physics, Chinese Academy of Sciences, Beijing 100190 (China); School of Physical Electronics, University of Electronic Science and Technology of China, Chengdu 610054 (China); Liu, Yandong, E-mail: ydliu@itp.ac.cn [State Key Laboratory of Theoretical Physics and Kavli Institute for Theoretical Physics China (KITPC), Institute of Theoretical Physics, Chinese Academy of Sciences, Beijing 100190 (China)

2015-03-06

In the supersymmetric models with low scale supersymmetry (SUSY) breaking where the gravitino mass is around keV, we show that the 3.5 keV X-ray lines can be explained naturally through several different mechanisms: (I) a keV scale dark gaugino plays the role of sterile neutrino in the presence of bilinear R-parity violation. Because the light dark gaugino obtains Majorana mass only via gravity mediation, it is a decaying warm dark matter (DM) candidate; (II) the compressed cold DM states, whose mass degeneracy is broken by gravity mediated SUSY breaking, emit such a line via the heavier one decay into the lighter one plus photon(s). A highly supersymmetric dark sector may readily provide such kind of system; (III) the light axino, whose mass again is around the gravitino mass, decays to neutrino plus gamma in the R-parity violating SUSY. Moreover, we comment on dark radiation from dark gaugino.

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)

Aerosol retention in the flooded steam generator bundle during SGTR

International Nuclear Information System (INIS)

Lind, Terttaliisa; Dehbi, Abdel; Guentay, Salih

2011-01-01

Research highlights: → High retention of aerosol particles in a steam generator bundle flooded with water. → Increasing particle inertia, i.e., particle size and velocity, increases retention. → Much higher retention of aerosol particles in the steam generator bundle flooded with water than in a dry bundle. → Much higher retention of aerosol particles in the steam generator bundle than in a bare pool. → Bare pool models have to be adapted to be applicable for flooded bundles. - Abstract: A steam generator tube rupture in a pressurized water reactor may cause accidental release of radioactive particles into the environment. Its specific significance is in its potential to bypass the containment thereby providing a direct pathway of the radioactivity from the primary circuit to the environment. Under certain severe accident scenarios, the steam generator bundle may be flooded with water. In addition, some severe accident management procedures are designed to minimize the release of radioactivity into the environment by flooding the defective steam generator secondary side with water when the steam generator has dried out. To extend our understanding of the particle retention phenomena in the flooded steam generator bundle, tests were conducted in the ARTIST and ARTIST II programs to determine the effect of different parameters on particle retention. The effects of particle type (spherical or agglomerate), particle size, gas mass flow rate, and the break submergence on particle retention were investigated. Results can be summarized as follows: increasing particle inertia was found to increase retention in the flooded bundle. Particle shape, i.e., agglomerate or spherical structure, did not affect retention significantly. Even with a very low submergence, 0.3 m above the tube break, significant aerosol retention took place underlining the importance of the jet-bundle interactions close to the tube break. Droplets were entrained from the water surface with

The structure of horizontal hydrogen-steam diffusion flames

International Nuclear Information System (INIS)

Chan, C.K.; Guerrero, A.

1997-01-01

This paper summarizes a systematic study on the stability, peak temperature and flame length of various horizontal hydrogen-steam diffusion flames in air. Results from this study are discussed in terms of their impact on hydrogen management in a nuclear containment building after a nuclear reactor accident. They show that, for a certain range of emerging hydrogen-steam compositions, a stable diffusion flame can anchor itself at the break in the primary heat transport system. The length of this flame can be up to 100 times the break diameter. This implies that creation of a stable diffusion flame at the break is a possible outcome of the deliberate ignition mitigation scheme. The high temperature and heat flux from a diffusion flame can threaten nearby equipment. However, due to the presence of steam and turbulent mixing with surrounding air, the peak temperatures of these diffusion flames are much lower than the adiabatic constant pressure combustion temperature of a stoichiometric hydrogen-air mixture. These results suggest that the threat of a diffusion flame anchored at the break may be less severe than conservative analysis would indicate. Furthermore, such a flame can remove hydrogen at the source and minimize the possibility of a global gas explosion. (author)

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

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)

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

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.

Physical data generation methodology for return-to-power steam line break analysis

Energy Technology Data Exchange (ETDEWEB)

Zee, Sung Kyun; Lee, Chung Chan; Lee, Chang Kue [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-02-01

Current methodology to generate physics data for steamline break accident analysis of CE-type nuclear plant such as Yonggwang Unit 3 is valid only if the core reactivity does not reach the criticality after shutdown. Therefore, the methodology requires tremendous amount of net scram worth, specially at the end of the cycle when moderator temperature coefficient is most negative. Therefore, we need a new methodology to obtain reasonably conservation physics data, when the reactor returns to power condition. Current methodology used ROCS which include only closed channel model. But it is well known that the closed channel model estimates the core reactivity too much negative if core flow rate is low. Therefore, a conservative methodology is presented which utilizes open channel 3D HERMITE model. Current methodology uses ROCS which include only closed channel model. But it is well known that the closed channel model estimates the core reactivity too much negative if core flow rate is low. Therefore, a conservative methodology is presented which utilizes open channel 3D HERMITE model. Return-to-power reactivity credit is produced to assist the reactivity table generated by closed channel model. Other data includes hot channel axial power shape, peaking factor and maximum quality for DNBR analysis. It also includes pin census for radiological consequence analysis. 48 figs., 22 tabs., 18 refs. (Author) .new.

Watt steam governor stability

Science.gov (United States)

Denny, Mark

2002-05-01

The physics of the fly-ball governor, introduced to regulate the speed of steam engines, is here analysed anew. The original analysis is generalized to arbitrary governor geometry. The well-known stability criterion for the linearized system breaks down for large excursions from equilibrium; we show approximately how this criterion changes.

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)

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.

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.

NPP Krsko small break LOCA analysis

International Nuclear Information System (INIS)

Mavko, B.; Petelin, S.; Peterlin, G.

1987-01-01

Parametric analysis of small break loss of coolant accident for the Krsko NPP was calculated by using RELAP5/MOD1 computer code. The model that was used in our calculations has been improved over several years and was previously tested in simulation (s) of start-up tests and known NPP Krsko transients. In our calculations we modelled automatic actions initiated by control, safety and protection systems. We also modelled the required operator actions as specified in emergency operating instructions. In small-break LOCA calculations, we varied break sizes in the cold leg. The influence of steam generator tube plugging on small break LOCA accidents was also analysed. (author)

Normal rejoining of DNA strand breaks in ataxia telangiectasia fibroblast lines after low x-ray exposure

International Nuclear Information System (INIS)

Hariharan, P.V.; Eleczko, S.; Smith, B.P.; Paterson, M.C.

1981-01-01

The alkaline elution method was used to measure the enzymatic repair of x-ray-induced DNA strand breaks in skin fibroblasts derived from human subjects afflicted with ataxia telangiectasia (AT). Monolayer cultures of normal control and AT cell lines were exposed acutely to moderately lethal (250-rad) and highly lethal (1250-rad) doses of 250-kV x rays under aerobic conditions. Upon receiving 250 rad, the control fibroblasts from a clinically normal donor rejoined all detectable single-strand breaks (plus alkali-labile bonds) within 30 to 60 min of incubation. When challenged with 1250 rad the kinetics of strand rejoining by the normal control cells were biphasic. For both exposures, no significant difference in either the rate or the extent of strand rejoining was detected between the normal cell line (GM38) and three mutant cell lines (AT2BE, AT3BI, AT4BI) belonging to the three known genetic complementation groups in AT. It would thus appear that the enhanced radiosensitivity of cultured AT cells does not stem from faulty rejoining of radiogenic DNA strand breaks

Counter-current flow limitation at hot leg pipe during reflux condensation cooling after small-break LOCA

International Nuclear Information System (INIS)

Jeong, Hae Yong; Ha, Sang Jun; Jo, Yung Jo; Jun, Hwang Yong

1999-01-01

The possibility of hot leg flooding is evaluated in case of a small-break loss-of-coolant accident in Korean Next Generation Reactor (KNGR) operating at the core power of 3983 MW normally. The vapor and liquid velocities in hot leg and steam generator tubes are calculated during reflux condensation cooling with the accident scenarios of three typical break sizes, 0.13 %, 1.02 % and 10.19 % cold leg break. The calculated results are compared with the existing flooding correlations. It is predicted that the hot leg flooding is excluded when two steam generators are available. It is also shown that the possibility of hot leg flooding under the operation with one steam generator is very low. Therefore, it can be said that the occurrence of hot leg flooding is unexpected when the reflux condensation cooling is maintained in steam generator tubes

Induction and rejoining of DNA double-strand breaks and interphase chromosome breaks after exposure to x-rays in one normal and two hypersensitive human fibroblast cell lines

Energy Technology Data Exchange (ETDEWEB)

Badie, C.; Alsbeih, G.; Malaise, E.P. [Institute Gustave-Roussy, Villejuif (France)] [and others

1995-10-01

The aim of this work was to measure simultaneously and in a quantitative manner double-strand breaks (DSBs), interphase chromosome breaks and cell lethality either immediately after irradiation, or at various times thereafter (up to 24 h), in cells of three nontransformed human fibroblast cell lines of widely different intrinsic radiosensitivity. We wished to assess initial damage, repair kinetics and residual damage at the DNA and the chromosome level, and to correlate these parameters with cell killings. We employed HF19 cells, a normal fibroblast cell line, AT2 cells, a radiosensitive cell line from a patient suffering from ataxia telangiectasia (AT), and 180BR cells, a radiosensitive cell line from a patient with no clinical symptoms of AT. AT2 and 180BR cells, in addition to being radiosensitive, also display a reduced ability to repair potentially lethal damage compared to HF19 cells. The yield of DSBs, as measured by pulsed-field gel electrophoresis, is similar in all three cell lines (slopes correspond to 1.6-1.7% Gy{sup -1} of DNA-associated radioactivity released from the gel well into the lane). In contrast, residual DSBs measured 24 h after irradiation are almost zero for HF19 cells (0.1% confidence interval=0-1.4%), but are 12.5% ({plus_minus}2.3%) and 43.8% ({plus_minus}1.2%) of those measured immediately after irradiation in HF19, AT2 and 180BR cells, respectively. Neither the initial yield of DSBs nor that of excess interphase chromosomes breaks can explain the differences in radiosensitivity between the three cell lines; however, there is a correlation between residual DSBs, rate of DSB rejoining at 24 h, residual interphase chromosome breaks on the one hand and cell survival on the other hand. 74 refs., 6 figs., 4 tabs.

Prediction of Counter-Current Flow Limitation at Hot Leg Pipe During a Small-Break Loca

Energy Technology Data Exchange (ETDEWEB)

Jeong, H.Y. [Korea Electric Power Research Institute, Taejeon (Korea)

2001-07-01

The possibility of hot leg flooding during reflux condensation cooling after a small-break loss-of-coolant accident in a nuclear power plant is evaluated. The vapor and liquid velocities in hot leg and steam generator tubes are calculated during reflux condensation cooling with the accident scenarios of three typical break sizes, 0.13 %, 1.02 % and 10.19 % cold leg break. The effect of initial water level to counter-current flow limitation is taken into account. It is predicted that the hot leg flooding is precluded when all steam generators are available for heat removal. It is also shown the both hot leg flooding and SG flooding are possible under the operation of one steam generators. Therefore, it can be said that the occurrence of hot leg flooding under reflux condensation cooling is possible when the number of steam generators available for heat removal is limited. (author). 15 refs., 15 figs., 3 tabs.

Small break loss of coolant accident analysis of advanced PWR plant designs utilizing DVI line venturis

International Nuclear Information System (INIS)

Kemper, Robert M.; Gagnon, Andre F.; McNamee, Kevin; Cheung, Augustine C.

1995-01-01

The Westinghouse Advanced Passive and evolutionary Pressurizer Water Reactors (i.e. AP600 and APWR) incorporate direct vessel injection (DVI) of emergency core coolant as a means of minimizing the potential spilling of emergency core cooling water during a loss of coolant accident (LOCA). As a result, the most limiting small break LOCA (SBLOCA) event for these designs, with respect core inventory makeup capability, is a postulated double ended rupture of one of the DVI lines. This paper presents the results of a design optimization study that examines the installation of a venturi in the DVI line as a means of limiting the reactor coolant lost from the reactor vessel. The comparison results demonstrate that by incorporating a properly sized venturi in the DVI line, core uncovery concerns as a result of a DVI line break can be eliminated for both the AP600 and APWR plants. (author)

Steam leak detection in advance reactors via acoustics method

International Nuclear Information System (INIS)

Singh, Raj Kumar; Rao, A. Rama

2011-01-01

Highlights: → Steam leak detection system is developed to detect any leak inside the reactor vault. → The technique uses leak noise frequency spectrum for leak detection. → Testing of system and method to locate the leak is also developed and discussed in present paper. - Abstract: Prediction of LOCA (loss of coolant activity) plays very important role in safety of nuclear reactor. Coolant is responsible for heat transfer from fuel bundles. Loss of coolant is an accidental situation which requires immediate shut down of reactor. Fall in system pressure during LOCA is the trip parameter used for initiating automatic reactor shut down. However, in primary heat transport system operating in two phase regimes, detection of small break LOCA is not simple. Due to very slow leak rates, time for the fall of pressure is significantly slow. From reactor safety point of view, it is extremely important to find reliable and effective alternative for detecting slow pressure drop in case of small break LOCA. One such technique is the acoustic signal caused by LOCA in small breaks. In boiling water reactors whose primary heat transport is to be driven by natural circulation, small break LOCA detection is important. For prompt action on post small break LOCA, steam leak detection system is developed to detect any leak inside the reactor vault. The detection technique is reliable and plays a very important role in ensuring safety of the reactor. Methodology developed for steam leak detection is discussed in present paper. The methods to locate the leak is also developed and discussed in present paper which is based on analysis of the signal.

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

Temporary core liquid level depression during cold-leg small-break LOCA effect of break size and power level

International Nuclear Information System (INIS)

Koizumi, Y.; Kumamaru, H.; Mimura, Y.; Kukita, Y.; Tasaka, K.

1989-01-01

Cold-leg small break LOCA experiments (0.5-10% break) were conducted at the large scale test facility (LSTF), a volumetrically-scaled (1/48) simulator of a PWR, of the ROSA-IV Program. When a break area was less than 2.5% of the scaled cold-leg flow area, the core liquid level was temporarily further depressed to the bottom elevation of the crossover leg during the loop seal clearing early in the transient only by the manometric pressure balance since no coolant remained in the upper portion of the primary system. When the break size was larger than 5%, the core liquid level was temporarily further depressed lower than the bottom elevation of the crossover leg during the loop seal clearing since coolant remained at the upper portion of the primary system; the steam generator (SG) U-tube upflow side and the SG inlet plenum, due to counter current flow limiting by updrafting steam while the coolant drained. The amount of coolant trapped there was dependent on the vapor velocity (core power); the larger the core power, the lower the minimum core liquid level. The RELAP5/MOD2 code reasonable predicted phenomena observed in the experiments. (orig./DG)

Thermal-Hydraulic Integral Effect Test with ATLAS for an Intermediate Break Loss of Coolant Accident at a Pressurizer Surge Line

Energy Technology Data Exchange (ETDEWEB)

Kang, Kyoung Ho; Seok Cho; Park, Hyun Sik; Choi, Nam Hyun; Park, Yu Sun; Kim, Jong Rok; Bae, Byoung Uhn; Kim, Yeon Sik; Kim, Kyung Doo; Choi, Ki Yong; Song, Chul Hwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

The main objectives of this test were not only to provide physical insight into the system response of the APR1400 during the pressurizer surge line break accident but also to produce an integral effect test data to validate the SPACE code. In order to simulate a double-ended guillotine break of a pressurizer surge line in the APR1400, the IB-SUR-01R test was performed with ATLAS. The major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Despite the core was uncovered, no excursion in the cladding temperature was observed. The pressurizer surge line break can be classified as a hot leg break from a break location point of view. Compared with a cold leg break, coolability in the core may be better in case of a hot leg break due to the enhanced flow in the core region. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code. Furthermore, this data can be utilized to identify any code deficiency for an IBLOCA simulation, especially for DVI-adapted plants. Redefinition of break size for design basis accident (DBA) based on risk information is being extensively investigated due to the potential for safety benefits and unnecessary burden reduction from current LBLOCA (large break loss of coolant accident)-based ECC (Emergency Core Cooling) Acceptance Criteria. As a transition break size (TBS), the rupture of medium-size pipe is considered to be more important than ever in risk-informed regulation (RIR)-relevant safety analysis. As plants age, are up-rated, and continue to seek improved operating efficiencies, the small break and intermediate break LOCA (IBLOCA) can become a concern. In particular, IBLOCA with DVI (Direct Vessel Injection) features will be addressed to support redefinition of a design-basis LOCA. With an aim of expanding code validation to address small

Effect of pipe rupture loads inside containment in the break exclusionary piping outside containment

International Nuclear Information System (INIS)

Weiss, G.

1987-01-01

The plant design for protection against piping failures outside containment should make sure that fluid system piping in containment penetration areas are designed to meet the break exclusionary provisions contained in the BTP MEB 3-1. According to these provisions, following a piping failure (main steam line) inside containment, the part of the flued head connected to the piping outside containment, should not exceed the ASME Code stress limits for the appropriate load combinations. A finite element analysis has been performed to evaluate the stress level in this area. (orig./HP)

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.)

Experimental observation of a multi-dimensional mixing behavior of steam-water flow in the MIDAS test facility

International Nuclear Information System (INIS)

Kweon, T. S.; Yun, B. J.; Ah, D. J.; Ju, I. C.; Song, C. H.; Park, J. K.

2001-01-01

Multi-dimensional thermal-hydraulic hehavior, such as ECC (Emergency Core Cooling) bypass, ECC penetration, steam-water condensation and accumulated water level, in an annular downcomer of a PWR (Pressurized Water Reactor) reactor vessel with a DVI(Direct Vessel Injection) injection mode is presented based on the experimental observations in the MIDAS (Multi-dimensional Investigation in Downcomer Annulus Simulation) steam-water facility. From the steady-state tests to similate a late reflood phase of LBLOCA (Large Break Loss-of-Coolant Accidents), major thermal-hydraulic phenomena in the downcomer are quantified under a wide range of test conditions. Especially, isothermal lines show well multi-dimensional phenomena of phase interaction between steam and water in the annulus downcomer. Overall test results show that multi-dimensional thermal-hydraulic behaviors occur in the downcomer annulus region as expected. The MIDAS test facility is a steam-water separate effect test facility, which is 1/4.93 linearly scaled-down of a 1400 MWe PWR type of nuclear reactor, with focusing on understanding multi-dimensional thermal-hydraulic phenomena in annulus downcomer with various types of safety injection location during refill or reflood phase of a LBLOCA in PWR

Organic Contaminants and Treatment Chemicals in Steam-Water Cycles : Thermal stability, decomposition products and flow-accelerated corrosion

NARCIS (Netherlands)

Moed, D.H.

2015-01-01

Boiler feedwater and steam have to be of high purity, because of the susceptibility of the steam-water cycle to corrosion. Organic contaminants break down in boilers by hydrothermolysis, leading to the formation of organic acid anions, which are suspected to cause corrosion of steam-water cycle

Electronic cigarettes induce DNA strand breaks and cell death independently of nicotine in cell lines.

Science.gov (United States)

Yu, Vicky; Rahimy, Mehran; Korrapati, Avinaash; Xuan, Yinan; Zou, Angela E; Krishnan, Aswini R; Tsui, Tzuhan; Aguilera, Joseph A; Advani, Sunil; Crotty Alexander, Laura E; Brumund, Kevin T; Wang-Rodriguez, Jessica; Ongkeko, Weg M

2016-01-01

Evaluate the cytotoxicity and genotoxicity of short- and long-term e-cigarette vapor exposure on a panel of normal epithelial and head and neck squamous cell carcinoma (HNSCC) cell lines. HaCaT, UMSCC10B, and HN30 were treated with nicotine-containing and nicotine-free vapor extract from two popular e-cigarette brands for periods ranging from 48 h to 8 weeks. Cytotoxicity was assessed using Annexin V flow cytometric analysis, trypan blue exclusion, and clonogenic assays. Genotoxicity in the form of DNA strand breaks was quantified using the neutral comet assay and γ-H2AX immunostaining. E-cigarette-exposed cells showed significantly reduced cell viability and clonogenic survival, along with increased rates of apoptosis and necrosis, regardless of e-cigarette vapor nicotine content. They also exhibited significantly increased comet tail length and accumulation of γ-H2AX foci, demonstrating increased DNA strand breaks. E-cigarette vapor, both with and without nicotine, is cytotoxic to epithelial cell lines and is a DNA strand break-inducing agent. Further assessment of the potential carcinogenic effects of e-cigarette vapor is urgently needed. Copyright © 2015 Elsevier Ltd. All rights reserved.

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)

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

An evaluation of TRAC-PF1/MOD1 computer code performance during posttest simulations of Semiscale MOD-2C feedwater line break transients

International Nuclear Information System (INIS)

Hall, D.G.; Watkins, J.C.

1987-01-01

This report documents an evaluation of the TRAC-PF1/MOD1 reactor safety analysis computer code during computer simulations of feedwater line break transients. The experimental data base for the evaluation included the results of three bottom feedwater line break tests performed in the Semiscale Mod-2C test facility. The tests modeled 14.3% (S-FS-7), 50% (S-FS-11), and 100% (S-FS-6B) breaks. The test facility and the TRAC-PF1/MOD1 model used in the calculations are described. Evaluations of the accuracy of the calculations are presented in the form of comparisons of measured and calculated histories of selected parameters associated with the primary and secondary systems. In addition to evaluating the accuracy of the code calculations, the computational performance of the code during the simulations was assessed. A conclusion was reached that the code is capable of making feedwater line break transient calculations efficiently, but there is room for significant improvements in the simulations that were performed. Recommendations are made for follow-on investigations to determine how to improve future feedwater line break calculations and for code improvements to make the code easier to use

Scientific design of the test facility for the KNGR DVI line small break LOCA

Energy Technology Data Exchange (ETDEWEB)

Yun, Byong Jo; Park, Choon Kyung; Jun, Hyung Gil; Cho, Seok; Kwon, Tae Soon; Song, Chul Hwa; Kim, Jung Taek

1999-03-01

Scientific design of the experimental facility (OASIS) for the KNGR (Korea Next Generation Reactor) DVI line SB-LOCA simulation is carried out. Main purpose of the OASIS is to produce thermal-hydraulic data base for determining the best location of the DVI (Direct Vessel Injection) injection nozzle of the KNGR as well as verifying its design performance in view of the ECCS (Emergency Core Cooling System) effectiveness. The experimental facility is designed based on the Ishii's three-level scaling law. The facility has 1/4 height and 1/341 area scaling ratio. It corresponds to the volume scale of 1/1364. The power scaling is 1/682 and the system pressure is prototypic. The OASIS consists of a core, a downcomer, two steam generators, two pump simulators, a break simulator, a collection tank, primary piping as well as a circulation pump for initial test condition. Each component is designed based on the Ishill's global scaling and boundary flow scaling of mass, energy and momentum. In addition, local phenomena scaling is carried out for the design of major components to preserve key local phenomena in each component. Most of the key phenomena are well preserved in the OASIS. However, the local scaling analysis shows that distortions of the void fraction and mixture level can not be avoided in the core. It comes from the basic features of the Ishill's scaling law in case of the reduced-height simulation. However, it is expected that these distortions will be analyzed properly by a best estimate system analysis code. (Author). 22 refs., 20 tabs., 25 figs.

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)

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.)

Study on the behavior of moisture droplets in low pressure steam turbines

International Nuclear Information System (INIS)

Kimura, Y.; Kuramoto, Y.; Yoshida, K.; Etsu, M.

1978-01-01

Low pressure stages of fossil turbines and almost all stages of nuclear and geothermal turbines operate on wet steam. Turbine operating on wet steam have the following two disadvantages: decrease of efficiency and erosion of blades. Decrease of efficiency results from an increase in profile loss caused by water films on the blade surface; loss of steam energy in breaking up the films and accelerating moisture droplets; undercooling and condensation shocks associated with it; velocity difference between water and steam phases and consequent decelerating action of moisture droplets in the rotating blades, etc. Impingement of moisture droplets on the rotating blades also causes quick erosion of the blades. In this paper, the behavior of moisture droplets in wet steam flow is described and the correlation between their behavior and the abovementioned two disadvantages of turbines operating on wet steam is clarified. (author)

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

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)

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)

Small break critical discharge: The roles of vapor and liquid entrainment in a stratified two-phase region upstream of the break

International Nuclear Information System (INIS)

Schrock, V.E.; Revankar, S.T.; Mannheimer, R.; Wang, C.H.

1986-12-01

The main objective of this research program was to perform an experimental investigation on the phenomena of two-phase critical flow through small break from a horizontal pipe which contained a stratified two phase flow. Stagnation conditions investigated were saturated steam-water, and air-cold water at pressures ranging from 0.37 MPa to 1.07 MPa. The small breaks employed were cylindrical tubes of diameters 3.96 mm, 6.32 mm, and 10.1 mm with sharp-edged entrance. For breaks resulting from a small hole in a primary coolant pipe or in a small pipe, a sharp-edged orifice or a sharp-edged tube can be the approximation

Multi-dimensional Mixing Behavior of Steam-Water Flow in a Downcomer Annulus during LBLOCA Reflood Phase with a DVI Injection Mode

International Nuclear Information System (INIS)

Kwon, T.S.; Yun, B.J.; Euh, D.J.; Chu, I.C.; Song, C.H.

2002-01-01

Multi-dimensional thermal-hydraulic behavior in the downcomer annulus of a pressurized water reactor vessel with a Direct Vessel Injection (DVI) mode is presented based on the experimental observation in the MIDAS (Multi-dimensional Investigation in Downcomer Annulus Simulation) steam-water test facility. From the steady-state test results to simulate the late reflood phase of a Large Break Loss-of-Coolant Accidents(LBLOCA), isothermal lines show the multidimensional phenomena of a phasic interaction between steam and water in the downcomer annulus very well. MIDAS is a steam-water separate effect test facility, which is 1/4.93 linearly scaled-down of 1400 MWe PWR type of a nuclear reactor, focused on understanding multi-dimensional thermalhydraulic phenomena in downcomer annulus with various types of safety injection during the refill or reflood phase of a LBLOCA. The initial and the boundary conditions are scaled from the pre-test analysis based on the preliminary calculation using the TRAC code. The superheated steam with a superheating degree of 80 K at a given downcomer pressure of 180 kPa is injected equally through three intact cold legs into the downcomer. (authors)

ATHLET calculations of the pressurizer surge line break (PH-SLB test) at the PMK-2 test facility

International Nuclear Information System (INIS)

Krepper, E.; Schaefer, F.

2000-01-01

At the Hungarian integral test facility PMK-2 a pressurizer surge line break experiment (PH-SLB test) was carried out with the PHARE 4.2.6b project. The primary objective of the test was to provide experimental data for a surge line break transient at VVER-440 reactors with reduced injection from the emergency core cooling systems (ECC). At the Institute of Safety Research calculations of the experiment were performed with the thermohydraulic computer code ATHLET, which was developed by GRS (Gesellschaft fuer Anlagen- und Reaktorsicherheit) mbH. In the context of the PHARE 4.2.6b project the Institute of Safety Research has also supplied the void fraction measurement system for the PMK-2 test facility and was involved in the evaluation of the experimental results. (orig.)

Steam generator tube rupture effects on a LOCA

International Nuclear Information System (INIS)

LaChance, J.L.

1979-01-01

A problem currently experienced in commercial operating pressurized water reactors (PWR) in the United States is the degradation of steam generator tubes. Safety questions have arisen concerning the effect of these degraded tubes rupturing during a postulated loss-of-coolant accident (LOCA). To determine the effect of a small number of tube ruptures on the behavior of a large PWR during a postulated LOCA, a series of computer simulations was performed. The primary concern of the study was to determine whether a small number (10 or less of steam generator tubes rupturing at the beginning surface temperatures. Additional reflood analyses were performed to determine the system behavior when from 10 to 60 tubes rupture at the beginning of core reflood. The FLOOD4 code was selected as being the most applicable code for use in this study after an extensive analysis of the capabilities of existing codes to perform simulations of a LOCA with concurrent steam generator tube ruptures. The results of the study indicate that the rupturing of 10 or less steam generator tubes in any of the steam generators during a 200% cold leg break will not result in a significant increase in the peak cladding temperature. However, because of the vaporization of the steam generator secondary water in the primary side of the steam generator, a significant increase in the core pressure occurs which retards the reflooding process

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

Loss of Coolant Accident Simulation for the Top-Slot break at Cold Leg Focusing on the Loop Seal Reformation under Long Term Cooling with the ATLAS

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong Rok; Park, Yu Sun; Bae, Byoung Uhn; Choi, Nam Hyun; Kang, Kyoung Ho; Choi, Ki Yong [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In the present paper, loss of coolant accident for the top-slot break at cold leg was simulated with the ATLAS, which is a thermal-hydraulic integral effect test facility for evolutionary pressurized water reactors (PWRs) of an advanced power reactor of 1400 MWe (APR1400). The simulation was focused on the loop seal reformation under long term cooling condition. During a certain class of Loss of Coolant Accident (LOCA) in a PWR like an advanced power reactor of 1400 MWe (APR1400), the steam volume in the reactor vessel upper plenum and/or upper head may continue expanding until steam blows liquid out of the intermediate leg (U-shaped pump suction cold leg), called loop seal clearing (LSC), opening a path for the steam to be relieved from the break. Prediction of the LSC phenomena is difficult because they are varies for many parameters, which are break location, type, size, etc. This LSC is the major factor that affects the coolant inventory in the small break LOCA (SBLOCA) or intermediate break LOCA (IBLOCA). There is an issue about the loop seal reformation that liquid refills intermediate leg and blocks the steam path after LSC. During the SBLOCA or IBLOCA, the Emergency Core Cooling System (ECCS) is operated. For long term of the top slot small or intermediate break at cold leg, the primary steam condensation by SG heat transfer or SIP, SIT water flooding (reverse flow to loop seal) make loop seal reformation possibly. The primary pressure increase at the top core region due to the steam release blockage by loop seal reformation. And then core level decreases and partial core uncover may occur. The loss of coolant accident for the top-slot break at cold leg was simulated with the ATLAS. The loop seal clearing and loop seal reformation were occurred repeatedly.

Two-dimensional modeling of water spray cooling in superheated steam

Directory of Open Access Journals (Sweden)

Ebrahimian Vahid

2008-01-01

Full Text Available Spray cooling of the superheated steam occurs with the interaction of many complex physical processes, such as initial droplet formation, collision, coalescence, secondary break up, evaporation, turbulence generation, and modulation, as well as turbulent mixing, heat, mass and momentum transfer in a highly non-uniform two-phase environment. While it is extremely difficult to systematically study particular effects in this complex interaction in a well defined physical experiment, the interaction is well suited for numerical studies based on advanced detailed models of all the processes involved. This paper presents results of such a numerical experiment. Cooling of the superheated steam can be applied in order to decrease the temperature of superheated steam in power plants. By spraying the cooling water into the superheated steam, the temperature of the superheated steam can be controlled. In this work, water spray cooling was modeled to investigate the influences of the droplet size, injected velocity, the pressure and velocity of the superheated steam on the evaporation of the cooling water. The results show that by increasing the diameter of the droplets, the pressure and velocity of the superheated steam, the amount of evaporation of cooling water increases. .

Special small-break applications with TRAC

International Nuclear Information System (INIS)

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

1981-01-01

Input models for the Transient Reactor Analysis Code (TRAC) are described and applications of these models to reactor transients involving small breaks in the primary coolant pressure boundary are demonstrated. The operation of the primary overpressure protection system (relief and safety valves) and the thermal-hydraulic response of the reactor to these transients are obtained from numerical simulations. Also, the effects of steam generator recirculation, steam generator tube rupture, Emergency Core Cooling (ECC) injection and reactivity feedback on the course and consequences of these transients are investigated. These models allow reliable predictions of accident signatures that can help determine the adequacy of equipment and procedures at nuclear power plants to prevent and to control severe accidents

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)

PACTEL: Experiments on the behaviour of the new horizontal steam generator

Energy Technology Data Exchange (ETDEWEB)

Kouhia, J.; Riikonen, V.; Purhonen, H. [VTT Energy, Lappeenranta (Finland)

1995-12-31

Experiments were performed to study the behaviour of the PACTEL facility, a medium scale integral test loop simulating VVER 440 pressurized water reactors. The study focused on the operation of the new horizontal steam generator model installed in PACTEL. Three experiments were conducted: a small-break test to observe the steam generator behaviour over a range of primary coolant inventories, a hot leg loop seal experiment to study the cyclic behaviour of a loop seal and a loss of secondary side feedwater test to examine the effect of uncovered tubes in the steam generator. A reverse flow was observed in the lower part of the U-tube bundle of the steam generator during natural circulation. The flow reversal point dropped when the tubes uncovered, during secondary inventory reduction. (orig.). 5 refs.

PACTEL: Experiments on the behaviour of the new horizontal steam generator

Energy Technology Data Exchange (ETDEWEB)

Kouhia, J; Riikonen, V; Purhonen, H [VTT Energy, Lappeenranta (Finland)

1996-12-31

Experiments were performed to study the behaviour of the PACTEL facility, a medium scale integral test loop simulating VVER 440 pressurized water reactors. The study focused on the operation of the new horizontal steam generator model installed in PACTEL. Three experiments were conducted: a small-break test to observe the steam generator behaviour over a range of primary coolant inventories, a hot leg loop seal experiment to study the cyclic behaviour of a loop seal and a loss of secondary side feedwater test to examine the effect of uncovered tubes in the steam generator. A reverse flow was observed in the lower part of the U-tube bundle of the steam generator during natural circulation. The flow reversal point dropped when the tubes uncovered, during secondary inventory reduction. (orig.). 5 refs.

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

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)

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)

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

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

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

Modelling of WWER-1000 steam generators by REALP5/MOD3.2 code

Energy Technology Data Exchange (ETDEWEB)

D`Auria, F.; Galassi, G.M. [Univ. of Pisa (Italy); Frogheri, M. [Univ. of Genova (Italy)

1997-12-31

The presentation summarises the results of best estimate calculations carried out with reference to the WWER-1000 Nuclear Power Plant, utilizing a qualified nodalization set-up for the Relap5/Mod3.2 code. The nodalization development has been based on the data of the Kozloduy Bulgarian Plant. The geometry of the steam generator imposed drastic changes in noding philosophy with respect to what is suitable for the U-tubes steam generators. For the secondary side a symmetry axis was chosen to separate (in the nodalization) the hot and the cold sides of the tubes. In this way the secondary side of the steam generators was divided into three zones: (a) the hot zone including the hot collector and the hot l/2 parts of the tubes; (b) the cold zone including the cold collector and the cold 1/2 parts of the tubes; (c) the downcomer region, where down flow is assumed. As a consequence of above in the primary side more nodes are placed on the hot side of the tubes. Steady state and transient qualification has been achieved, considering the criteria proposed at the University of Pisa, utilizing plant transient data from the Kozloduy and the Ukrainian Zaporosche Plants. The results of the application of the qualified WWER-1000 Relap5/Mod3.2 nodalization to various transients including large break LOCA, small break LOCA and steam generator tube rupture, together with a sensitivity analysis on the steam generators, are reported in the presentation. Emphasis is given to the prediction of the steam generators performances. 23 refs.

Modelling of WWER-1000 steam generators by REALP5/MOD3.2 code

Energy Technology Data Exchange (ETDEWEB)

D` Auria, F; Galassi, G M [Univ. of Pisa (Italy); Frogheri, M [Univ. of Genova (Italy)

1998-12-31

The presentation summarises the results of best estimate calculations carried out with reference to the WWER-1000 Nuclear Power Plant, utilizing a qualified nodalization set-up for the Relap5/Mod3.2 code. The nodalization development has been based on the data of the Kozloduy Bulgarian Plant. The geometry of the steam generator imposed drastic changes in noding philosophy with respect to what is suitable for the U-tubes steam generators. For the secondary side a symmetry axis was chosen to separate (in the nodalization) the hot and the cold sides of the tubes. In this way the secondary side of the steam generators was divided into three zones: (a) the hot zone including the hot collector and the hot l/2 parts of the tubes; (b) the cold zone including the cold collector and the cold 1/2 parts of the tubes; (c) the downcomer region, where down flow is assumed. As a consequence of above in the primary side more nodes are placed on the hot side of the tubes. Steady state and transient qualification has been achieved, considering the criteria proposed at the University of Pisa, utilizing plant transient data from the Kozloduy and the Ukrainian Zaporosche Plants. The results of the application of the qualified WWER-1000 Relap5/Mod3.2 nodalization to various transients including large break LOCA, small break LOCA and steam generator tube rupture, together with a sensitivity analysis on the steam generators, are reported in the presentation. Emphasis is given to the prediction of the steam generators performances. 23 refs.

Investigation of Steam Flow Behavior During Horizontal Injection into Vertical Annulus

International Nuclear Information System (INIS)

Yoon, Sang H.; Kim, Won J.; Ku, Ja H.; Suh, Kune Y.; Song, Chul H.

2004-01-01

Qualification of uncertainty margins for accidents, which are classified as the design basis accidents, requires thermal hydraulic codes and related code models with an enhanced level of sophistication. In a cold leg break accident, the flow in downcomer is multidimensional and the velocity distribution of the steam flow in downcomer serves as a good example. For observation of the flow behavior near the break, experiments are performed to measure the velocity of the steam flow in a vessel scaled down from the APR1400 (Advanced Power Reactor 1400 MWe). In this case, the steam has a quality approaching unity and thus is dealt with as a single-phase gas. The velocity of the steam flow is measured by micro-Pitot tubes arranged horizontally and vertically around the outer shell of the 1/20 scaled-down test vessel OMEGA (Optimized Multidimensional Experiment Geometric Apparatus). A commercial computational fluid dynamics code yields analytic results of multidimensional flow motion in the complex annular passage with flow obstacles. CFX is run with well-defined boundary conditions to obtain velocity profiles of the steam flow in the annular downcomer. Results of CFX shed light on the experimental setup as to fixing the location and angle of the micro-Pitot tubes, and correcting the sensitivity of the micro- Pitot tubes, for instance. This study aims to improve the multidimensional capability of the MARS code, which is based on RELAP5 and COBRA-IV, in predicting the multiphase flow behavior in the reactor downcomer. MARS is currently based on one- and two-dimensional flow analyses, which tends to distort total flow due to misrepresentation of the local phenomena. It is thus necessary to scrutinize the steam flow path and mechanistically model the momentum variation. These experimental and analytical results can locally be applied to developing the models of specific forms and essential phenomena treated in MARS. (authors)

Sensitivity of break-flow-partition on the containment pressure and temperature

International Nuclear Information System (INIS)

Kwon, Young Min; Song, Jin Ho; Lee, Sang Yong

1994-01-01

For the case of RCS blowdown into the vapor region of a containment at low pressure, the blowdown mixture will start to boil at the containment pressure and liquid will separate from the flow near the break location. The flashed steam is added to the containment atmosphere and liquid is falled to the sump. Analytically, the break flow can be divided into steam and liquid in a number of ways. Discussed in this study is three partition models and Instantaneous Mixing(IM) Model employed in different containment analysis computer codes. IM model is employed in the CONTRANS code developed by ABB-CE for containment thermodynamic analysis. The various partition models were applied to the double ended discharge leg slot break (DEDLS) LOCA which is containment design base accident (CDBA) for Ulchin 3 and 4 PSAR. It was shown that IM model is the most conservative for containment design pressure analysis. Results of the CONTRANS analyses are compared with those of UCN PSAR for which CONTEMPT-LT code was used

Particle deposition modeling in the secondary side of a steam generator bundle model

Energy Technology Data Exchange (ETDEWEB)

Mukin, Roman, E-mail: roman.mukin@psi.ch; Dehbi, Abdel, E-mail: abdel.dehbi@psi.ch

2016-04-01

A steam generator (SG) tube rupture (SGTR) model is studied in this paper. This model based on a experimental facility called Aerosol Trapping In a Steam Generator (ARTIST), which is a model of a scaled steam generator tube bundle consisting of 270 tubes and a guillotine tube to address aerosol deposition phenomena on two different scales: near the tube break, where the gas velocities and turbulence are very intensive, and far away from the break, where the flow velocities are three orders of magnitude lower. Owing to complexity of the flow, 3D simulations with highly resolved computational mesh near the break were done. First, the flow inside an isolated tube with a guillotine tube break has been studied in the framework of Reynolds Averaged Navier Stokes (RANS) approach. The next part is devoted to the simulation of an inclined gas jet entering the SG tube bundle via the guillotine tube breach with more advanced CFD tools. In particular, Detached Eddy Simulation (DES) and RANS are applied to tackle the wide range of flow scales. Flow field velocity comparison showed that DES results are reproducing wavy structure of the flow field in far field from the break observed in experiment. Particle transport and deposition is modelled by Lagrangian continuous random walk (CRW) model, which has been developed and validated previously. It is found that the DES combined with the CRW to supply fluctuating velocity components predicts deposition rates that are generally within the scatter of the measured data. Monodisperse, spherical SiO{sub 2} particles with AMMD = 1.4 μm were used as aerosol particles in simulations. To be economically feasible, the computations were made with the open source CFD code OpenFOAM. Comparison of the calculated flow with the experimental axial velocity distribution data at different vertical levels has been performed.

Integral effect test and code analysis on the cooling performance of the PAFS (passive auxiliary feedwater system) during an FLB (feedwater line break) accident

International Nuclear Information System (INIS)

Bae, Byoung-Uhn; Kim, Seok; Park, Yu-Sun; Kang, Kyoung-Ho

2014-01-01

Highlights: • This study focuses on the experimental validation of the operational performance of the PAFS (passive auxiliary feedwater system). • A transient simulation of the FLB (feedwater line break) in the integral effect test facility, ATLAS-PAFS, was performed to investigate thermal hydraulic behavior during the PAFS actuation. • The test result confirmed that the APR+ has the capability of coping with the FLB scenario by adopting the PAFS and proper set-points for its operation. • The experimental result was utilized to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. - Abstract: APR+ (Advanced Power Reactor Plus), which is a GEN-III+ nuclear power plant developed in Korea, adopts PAFS (passive auxiliary feedwater system) as an advanced safety feature. The PAFS can completely replace an active auxiliary feedwater system by cooling down the secondary side of steam generators with a natural convection mechanism. This study focuses on experimental and analytical investigation for cooling and operational performance of the PAFS during an FLB (feedwater line break) transient with an integral effect test facility, ATLAS-PAFS. To realistically simulate the FLB accident of the APR+, the three-level scaling methodology was taken into account to design the test facility and determine the test condition. From the test result, the PAFS was actuated to successfully cool down the decay heat of the reactor core by the condensation heat transfer at the PCHX (passive condensation heat exchanger), and thus it could be confirmed that the APR+ has the capability of coping with a FLB scenario by adopting the PAFS and proper set-points for its operation. This integral effect test data were used to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. The code analysis result proved that it could reasonably predict the FLB transient including the actuation of the PAFS and the natural convection

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

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)

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.

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

Single failure effects of reactor coolant system large bore hydraulic snubbers for Korean Standard Nuclear Power Plant

International Nuclear Information System (INIS)

Choi, T.S.; Park, S.H.; Sung, K.K.; Kim, T.W.; Jheon, J.H.

1996-01-01

A potential snubber single failure is one of the safety significances identified in General Safety Issue 113 for Large Bore Hydraulic Snubber (LBHS) dynamic qualification. This paper investigates dynamic structural effects of single failures of the steam generator and reactor coolant pump snubbers in Korean Standard Nuclear Power Plant by performing the time history dynamic analyses for the reactor coolant system under seismic and postulated pipe break events. The seismic input motions considered are the synthesized ground time histories conforming to SRP 3.7.1, and he postulated pipe break input loadings result from steam generator main seam line and feedwater line pipe breaks which govern pipe breaks remaining after applying LBB to the main coolant line and primary side ranch lines equal to and greater than 12 inch nominal pipe size

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

A feasibility study on the extended cycle from the point of view of Non-LOCA safety analysis

International Nuclear Information System (INIS)

Lee, Chul Sin; Kim, Hee Chul; Kim, Jeong Jin; Baek, Seung Su; Lee, Byung Il; Jeon, Tae Hyun; Lee, Jeong Chan

1996-06-01

Extended cycle operation has many advantages as compared with standard cycle operation (12 months) from the viewpoint of operators' exposure to radiation dose, man-power for maintenance and the production of nuclear waste. And it is more economic than the standard cycle operation. If the extended cycle operation is adopted in the CE type nuclear plant, the effective management and enhancement of operation capability could be expected. A feasibility study on the extended cycle operation should be performed. The purpose of this technical report is to perform safety analysis using the design data for Yonggwang Nuclear Plant (YGN) 3 and 4 extended cycle operation and to evaluate qualitatively and quantitatively the safety related design basis events to verify the satisfaction of the safety criteria. Boron dilution and steam line break accidents were found to be most influenced by the change of physics data due to the adaptation of the extended cycle operation. For boron dilution accident, source range monitor ratio of 3.07 for YGN 3 cycle 2 was decreased to 2.92. The 3D reactivity feedback effect due to the local heatup in the vicinity of stuck CEA was credited in the analysis of steam line break. No return-to-power occurred for the steam line break with offsite power available and return-to-power occurred for the steam line break with loss of offsite power. For the steam line break with loss of offsite power, the safety margin was preserved with respect to fuel performance (DNB and LHGR) despite the return-to-power. 6 tabs., 10 figs., 5 refs. (Author) .new

SPACE code simulation of ATLAS DVI line break accident test (SB DVI 08 Test)

Energy Technology Data Exchange (ETDEWEB)

Lim, Sang Gyu [KHNP, Daejeon (Korea, Republic of)

2012-10-15

APR1400 has adopted new safety design features which are 4 mechanically independent DVI (Direct Vessel Injection) systems and fluidic device in the safety injection tanks (SITs). Hence, DVI line break accident has to be evaluated as one of the small break LOCA (SBLOCA) to ensure the safety of APR1400. KAERI has been performed for DVI line break test (SB DVI 08) using ATLAS (Advanced Thermal Hydraulic Test Loop for Accident Simulation) facility which is an integral effect test facility for APR1400. The test result shows that the core collapsed water level decreased before a loop seal clearance, so that a core uncover occurred. At this time, the peak cladding temperature (PCT) is rapidly increased even though the emergency core cooling (ECC) water is injected from safety injection pump (SIP). This test result is useful for supporting safety analysis using thermal hydraulic safety analysis code and increases the understanding of SBLOCA phenomena in APR1400. The SBLOCA evaluation methodology for APR1400 is now being developed using SPACE code. The object of the development of this methodology is to set up a conservative evaluation methodology in accordance with appendix K of 10 CFR 50. ATLAS SB DVI 08 test is selected for the evaluation of SBLOCA methodology using SPACE code. Before applying the conservative models and correlations, benchmark calculation of the test is performed with the best estimate models and correlations to verify SPACE code capability. This paper deals with benchmark calculations results of ATLAS SB DVI 08 test. Calculation results of the major hydraulics variables are compared with measured data. Finally, this paper carries out the SPACE code performances for simulating the integral effect test of SBLOCA.

Methods for calculating the speed-up characteristics of steam-water turbines

International Nuclear Information System (INIS)

Golovach, E.A.

1981-01-01

The methods of approximate and specified calculations of speed- up characteristics of steam-water turbines are considered. The specified non-linear method takes into account change of thermal efficiency, heat drop and losses in the turbine as well as vacuum break-up the condenser. Speed-up characteristics of the K-1000-60-1500 turbine are presented. The calculational results obtained by the non-linear method are compared with the calculations conducted by the approximate linearized method. Differences in the frequency speed up of the turbine rotor rotation calculated by the two methods constitute only 0.5-2.0%. That is why it is necessary to take into account in the specified calculations first of all the most important factors following the rotor speed- up in the following consequence: valve shift of the high pressure cylinder (HPC); steam volume in front of the HPC; shift of the valves behind the separator-steam superheater (SSS); steam volumes and moisture boiling in the SSS; steam consumption for regenerating heating of feed water, steam volumes at the intermediate elements of the turbine, losses in the turbine, heat drop and thermal efficiency [ru

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)

DNA double strand break repair in a radioresistant cell line

International Nuclear Information System (INIS)

Koval, T.M.; Kazmar, E.R.

1987-01-01

TN-368 lepidopteran insect cells are on the order of 100 times more resistant to the lethal effects of ionizing radiation than cultured mammalian cells. DNA double strand breaks (DSB) are believed by many to be the critical molecular lesion leading to cell death. The authors therefore measured the rejoining of DSB in TN-368 and V79 Chinese hamster cells. Cells were irradiated on ice with /sup 137/Cs γ rays at a dose rate of 2.5 Gy/min, incubated for various periods of time, and assayed for DNA DSB using the method of neutral elution. The kinetics of DSB rejoining following a dose of 90.2 Gy are similar for both cell lines. Approximately 80% of the DSB are rejoined in both lines by 1 hr postirradiation. However, no further rejoining occurs in the TN-368 cells through at least 6 hr postirradiation, whereas 90% of the DSB are rejoined in the V79 cells by 2 hr postirradiation. Other studies (from 22.6 to 226 Gy) demonstrate that the amount of rejoining of DSB varies inversely with dose for the V79 cells but remains constant for the TN-368 cells. These findings do not support the hypothesis that unrejoined DNA DSB represent the major lesion resulting in cell death

Recirculation pump suction line 2.8% break integral test at ROSA-III with HPCS failure, RUN 984

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Anoda, Yoshinari; Tasaka, Kanji; Kumamaru, Hiroshige; Nakamura, Hideo; Yonomoto, Taisuke; Murata, Hideo; Shiba, Masayoshi

1984-06-01

This report presents the experimental data of 2.8% suction line break test RUN 984 at ROSA-III, which was conducted as one of counterpart tests to FIST program sponsored by GE, EPRI and USNRC. The similarity study between the ROSA-III and FIST tests is on the way. The report also presents the information on the ROSA-III test facility, experiment results and the effects of the ADS flow rate and the MSIV trip level comparing with the previously conducted ROSA-III small break tests, RUNs 920 and 922. Major conclusions obtained are as follows. (1) Change of the MSIV trip level from L2 to L1 gives delay of MSIV closure and longer actuation of pressure control system in a small break LOCA. (2) Larger ADS flow gives faster depressurization rate and earlier ECCS actuation, which results in shorter fuel rod dryout period and lower PCT. (author)

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.

Multicomponent seismic reservoir characterization of a steam-assisted gravity drainage (SAGD) heavy oil project, Athabasca oil sands, Alberta

Science.gov (United States)

Schiltz, Kelsey Kristine

Steam-assisted gravity drainage (SAGD) is an in situ heavy oil recovery method involving the injection of steam in horizontal wells. Time-lapse seismic analysis over a SAGD project in the Athabasca oil sands deposit of Alberta reveals that the SAGD steam chamber has not developed uniformly. Core data confirm the presence of low permeability shale bodies within the reservoir. These shales can act as barriers and baffles to steam and limit production by prohibiting steam from accessing the full extent of the reservoir. Seismic data can be used to identify these shale breaks prior to siting new SAGD well pairs in order to optimize field development. To identify shale breaks in the study area, three types of seismic inversion and a probabilistic neural network prediction were performed. The predictive value of each result was evaluated by comparing the position of interpreted shales with the boundaries of the steam chamber determined through time-lapse analysis. The P-impedance result from post-stack inversion did not contain enough detail to be able to predict the vertical boundaries of the steam chamber but did show some predictive value in a spatial sense. P-impedance from pre-stack inversion exhibited some meaningful correlations with the steam chamber but was misleading in many crucial areas, particularly the lower reservoir. Density estimated through the application of a probabilistic neural network (PNN) trained using both PP and PS attributes identified shales most accurately. The interpreted shales from this result exhibit a strong relationship with the boundaries of the steam chamber, leading to the conclusion that the PNN method can be used to make predictions about steam chamber growth. In this study, reservoir characterization incorporating multicomponent seismic data demonstrated a high predictive value and could be useful in evaluating future well placement.

Post accidental small breaks analysis

International Nuclear Information System (INIS)

Depond, G.; Gandrille, J.

1980-04-01

EDF ordered to FRAMATOME by 1977 to complete post accidental long term studies on 'First Contrat-Programme' reactors, in order to demonstrate the safety criteria long term compliance, to get information on NSSS behaviour and to improve the post accidental procedures. Convenient analytical models were needed and EDF and FRAMATOME respectively developped the AXEL and FRARELAP codes. The main results of these studies is that for the smallest breaks, it is possible to manually undertake cooling and pressure reducing actions by dumping the steam generators secondary side in order to meet the RHR operating specifications and perform long term cooling through this system. A specific small breaks procedure was written on this basis. The EDF and FRAMATOME codes are continuously improved; the results of a French set of separate effects experiments will be incorporated as well as integral system verification

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)

Theoretical investigations into the influence of the position of a breaking line on the tensile failure of flat, round, bevel-edged tablets using finite element methodology (FEM) and its practical relevance for industrial tablet strength testing.

Science.gov (United States)

Podczeck, Fridrun; Newton, J Michael; Fromme, Paul

2014-12-30

Flat, round tablets may have a breaking ("score") line. Pharmacopoeial tablet breaking load tests are diametral in their design, and industrially used breaking load testers often have automatic tablet feeding systems, which position the tablets between the loading platens of the machine with the breaking lines in random orientation to the applied load. The aim of this work was to ascertain the influence of the position of the breaking line in a diametral compression test using finite element methodology (FEM) and to compare the theoretical results with practical findings using commercially produced bevel-edged, scored tablets. Breaking line test positions at an angle of 0°, 22.5°, 45°, 67.5° and 90° relative to the loading plane were studied. FEM results obtained for fully elastic and elasto-plastic tablets were fairly similar, but they highlighted large differences in stress distributions depending on the position of the breaking line. The stress values at failure were predicted to be similar for tablets tested at an angle of 45° or above, whereas at lower test angles the predicted breaking loads were up to three times larger. The stress distributions suggested that not all breaking line angles would result in clean tensile failure. Practical results, however, did not confirm the differences in the predicted breaking loads, but they confirmed differences in the way tablets broke. The results suggest that it is not advisable to convert breaking loads obtained on scored tablets into tablet tensile strength values, and comparisons between different tablets or batches should carefully consider the orientation of the breaking line with respect to the loading plane, as the failure mechanisms appear to vary. Copyright © 2014 Elsevier B.V. All rights reserved.

Safety Evaluation for IHTS Integrity due to the Steam Generator Sodium-Water Reaction Event in the PGSFR

Energy Technology Data Exchange (ETDEWEB)

Ahn, Sang-Jun; Lee, Kwi Lim; Ha, Kwi-Seok; Lee, Seung Won; Jeong, Taekyeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

In this paper, the integrity of the IHTS and SG by the SWR event are evaluated using the SWAAMII code. A sodium has a chemical characteristics to rigorously react the water or steam and produce the high pressure waves and high temperature reaction heat. It has an excellent characteristics as a reactor coolant. But, there is an event to be considered in the sodium cooled fast reactor design. The Sodium-Water Reaction (SWR) event can be occurred by the water or steam leaks due to the break of the steam generator tubes. The propagated high pressure waves threathen the structural integrity of the affected Intermediate Heat Transport System (IHTS) and steam generator. If the IHTS pipes are failed, the sodium of the IHTS can be released to the containment building. To the peak pressure point of view, it is performed to evaluate the integrity of the major components due to the SWR event in the SG. The generated peak pressures due to the five SG tubes simultaneous break event are within the range of the design pressure for the SG, IHX and IHTS including the related pipes.

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

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

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

Analysis of containment pressure and temperature changes following loss of coolant accident (LOCA)

International Nuclear Information System (INIS)

Nguyen Van Thai; Kieu Ngoc Dung

2015-01-01

This paper present a preliminary thermal-hydraulics analysis of AP1000 containment following loss of coolant accident events such as double-end cold line break (DECLB) or main steam line break (MSLB) using MELCOR code. A break of this type will produce a rapid depressurization of the reactor pressure vessel (primary system) and release initially high pressure water into the containment followed by a much smaller release of highly superheated steam. The high pressure liquid water will flash and rapidly pressurize the containment building. The performance of passive containment cooling system for steam removal by condensation on large steel containment structure is a major contributing process, controlling the pressure and temperature maximum reached during the accident event. The results are analyzed, discussed and compared with the similar work done by Sandia National Laboratories. (author)

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

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

Differences in heavy-ion-induced DNA double-strand breaks in a mouse DNA repair-deficient mutant cell line (SL3-147) before and after chromatin proteolysis

International Nuclear Information System (INIS)

Murakami, Masahiro; Eguchi-Kasai, Kiyomi; Sato, Koki; Minohara, Shinichi; Kanai, Tatsuaki; Yatagai, Fumio.

1995-01-01

DNA double-strand breaks induced by X- or neon beam-irradiation in a DNA double-strand break-repair-deficient mutant cell line (SL3-147) were examined. The increase in the number of DNA double-strand breaks was dose-depend after irradiation with X-rays and neon beams and was enhanced by chromatin-proteolysis treatment before irradiation. These results suggest that the induction of DNA double-strand breaks by ionizing radiation, including heavy-ions, is influenced by the chromatin structure. (author)

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)

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.

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

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)

ISP-50 Specifications for a Direct Vessel Injection Line Break Test with the ATLAS

International Nuclear Information System (INIS)

Choi, Ki Yong; Baek, Won Pil; Kim, Yeon Sik; Park, Hyun Sik; Cho, Seok; Kang, Kyoung Ho; Choi, Nam Hyun; Min, Kyoung Ho

2009-06-01

An OECD/NEA International Standard Problem Exercise (ISP) focussing on a DVI line break simulation result with the ATLAS was approved by the NEA Committee on the Safety of Nuclear Installation (CSNI) meeting in December 2008 and was numbered by ISP-50. The ISP-50 program will be operated by an operating agency, KAERI for three years starting from the physical year 2009. Fourteen international organizations confirmed their participation in the ISP-50, including NRC (USA), JAEA, JNES (Japan), GRS (Germany), KFKI-AEKI (Hungary), EDO Gidropress (Russia), VTT, Fortum (Finland), NRI (Czech Republic), Univ. of Pisa (Italy), KINS, KNF, KOPEC, and KAERI (Korea). In addition, KTH in Sweden and HSE in UK are considering late participation. Recently, NPIC and CIAE in China hope to join the ISP-50. As for the safety analysis codes, nine codes are expected to be used for the ISP-50: MARS-3D, RELAP5- 3D, RELAP5, TRACE, CATHARE, APROS, ATHELET, TRAP, and KORSAR. It is the first ISP exercise in Korea in which a domestic test facility is utilized by international nuclear society and this exercise will contribute to extending our physical understanding on thermal hydraulic phenomena during the DVI line break accidents and to verifying the best-estimate thermal-hydraulic safety analysis codes. This report was prepared to define technical specifications of the ISP-50 exercise according the guideline provided by OECD/CSNI. It includes general objectives, phases, deliverables to participants, parameters required for comparison and the time table

Re-evaluation of superheat conditions postulated in NRC Information Notice 84-90

International Nuclear Information System (INIS)

Alsammarae, A.; Kruger, D.; Beutel, D.; Spisak, M.

1994-01-01

Information Notice 84-90, ''Main Steam Line Break Effect on Environmental Qualification of Equipment,'' describes a potential problem regarding existing plant analyses and Equipment Qualification (EQ) related to a postulated Main Steam Line Break (MSLB) with releases of superheated stream. This notice states that certain methodologies for computing mass and energy releases for a postulated MSLB did not account for heat transfer from the steam generator tube bundles if they were uncovered. Due to this potential change in the original environmental analysis, the EQ of various components may not consider the thermal environment which could result from superheated steam. Subsequent technical assessments may determine that the existing qualification basis for equipment/components does not envelop the postulated superheat condition. Corrective actions need to be taken to demonstrate that the affected equipment is qualified

Hydrogen distribution in a containment with a high-velocity hydrogen-steam source

International Nuclear Information System (INIS)

Bloom, G.R.; Muhlestein, L.D.; Postma, A.K.; Claybrook, S.W.

1982-09-01

Hydrogen mixing and distribution tests are reported for a modeled high velocity hydrogen-steam release from a postulated small pipe break or release from a pressurizer relief tank rupture disk into the lower compartment of an Ice Condenser Plant. The tests, which in most cases used helium as a simulant for hydrogen, demonstrated that the lower compartment gas was well mixed for both hydrogen release conditions used. The gas concentration differences between any spatial locations were less than 3 volume percent during the hydrogen/steam release period and were reduced to less than 0.5 volume percent within 20 minutes after termination of the hydrogen source. The high velocity hydrogen/steam jet provided the dominant mixing mechanism; however, natural convection and forced air recirculation played important roles in providing a well mixed atmosphere following termination of the hydrogen source. 5 figures, 4 tables

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.

Quick look report for semiscale MOD-2C Test S-FS-11

International Nuclear Information System (INIS)

Plessinger, M.P.

1985-11-01

Results of a preliminary analysis of the fifth test performed in the Semiscale MOD-2C Steam Generator Feedwater and Steam Line Break (FS) experiment series are presented. Test S-FS-11 simulated a pressurized water reactor transient initiated by a 50% break in a steam generator bottom feedwater line downstream of the check valve. With the exception of primary pressure, the initial conditions represented the initial conditions used for the C-E System 80 Final Safety Analysis Report (FSAR) Appendix 15B calculations. The transient included an initial 600 s period in which only automatic plant protection systems responded to the initiating event. This period was followed by a series of operator actions necessary to stabilize the plant followed by break isolation and affected loop steam generator refill with auxiliary feedwater. The test results provided a measured evaluation of the effectiveness of the automatic responses in minimizing primary system overpressurization and operator actions in stabilizing the plant. Test data also provided a basis for comparison with other tests in the series of the effects of break size on primary overpressurization and primary-to-secondary heat transfer. 64 figs

Design of a partial inter-tube lancing system actuated by hydraulic power for type F model steam generator in nuclear power plant

International Nuclear Information System (INIS)

Kim, S. T.; Jeong, W. T.

2008-01-01

The sludge grown up in steam generators of nuclear power plants shortens the life-cycle of steam generators and reduces the output of power plants. So KHNP(Korea Hydro and Nuclear Power), the only nuclear power utility in Korea, removes it periodically using a steam generator lancing system during the outage of plants for an overhaul. KEPRI(Korea Electric Power Research Institute) has developed lancing systems with high pressured water nozzle for steam generators of nuclear power plants since 2001. In this paper, the design of a partial inter-tube lancing system for model F type steam generators will be described. The system is actuated without a DC motor inner steam generators because the motors in a steam generator make a trouble from high intensity of radioactivity as a break down

Architecture-dependent distribution of Mesopores in steamed Zeolite crystals as visualized by FIB-SEM Tomography

NARCIS (Netherlands)

Karwacki, L.|info:eu-repo/dai/nl/304824283; de Winter, D.A.M.|info:eu-repo/dai/nl/304838616; Aramburo, L.R.; Lebbink, M.N.|info:eu-repo/dai/nl/304834246; Post, J.A.|info:eu-repo/dai/nl/074120050; Drury, M.R.|info:eu-repo/dai/nl/304829315; Weckhuysen, B.M.|info:eu-repo/dai/nl/285484397

2011-01-01

Break on through: Steaming-induced mesopores of individual ZSM-5 crystals were studied by a combination of focused ion beam (FIB) and scanning electron microscopy (SEM) tomography (see picture). In this manner, quantitative insight into the width, length, morphology, and distribution of mesopores

Reliability analysis of the containment spray system of Angra-1 : the injection phase

International Nuclear Information System (INIS)

Gibelli, S.M.O.; Oliveira, L.F.S. de.

1981-12-01

The system studied is projected to perform two basic functions : to reduce the pressure and temperature in the containment after a LOCA (loss of coolant accident), to break the main steam line or the main feed line in the containment after a LOCA (loss of coolant accident), to break the main steam line or the main feed line in the containment and to remove the fission products, mainly the iodine of the containment atmosphere. The spray system was analyzed concerning the probability of non-acomplishment of both functions at the same time; therefore the failure of the components of the chemical aditions subsystem are included in the failure tree shown here. (E.G.) [pt

Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

OpenAIRE

Hwang Bae; Dong Eok Kim; Sung-Uk Ryu; Sung-Jae Yi; Hyun-Sik Park

2017-01-01

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are s...

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)

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.

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.

Best estimate small break LOCA analysis for KNGR SIS optimization

International Nuclear Information System (INIS)

Song, JIn Ho; Lim, Hong Sik; Bae, Kyoo Hwan; Lee, Joon

1996-01-01

The KNGR has an advanced ECCS design feature which employs four mechanically-separated SI trains where each train consisting of one HPSI pump and one SIT injects ECC water directly into the reactor vessel downcomer annulus. To demonstrate that the KNGR ECCS design features meet the EPRI ALWR requirements of no core uncovery for a break of up to 6 inch diameter, small break LOCA cases with various break sizes were analyzed using a best-estimate analytical procedure. Two kinds of break locations are considered: cold leg and DVI line breaks. It was observed that the KNGR ECCS design can tolerate a cold leg break of up to 10 inches with no core uncovery. However, since DVI line break with 6 inch diameter undergoes slight core uncovery, further investigation is required for KNGR SIS optimization

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

ROSA-V/LSTF vessel top head LOCA tests SB-PV-07 and SB-PV-08 with break sizes of 1.0 and 0.1% and operator recovery actions for core cooling

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

2010-02-01

A series of break size parameter tests (SB-PV-07 and SB-PV-08) were conducted at the Large Scale Test Facility (LSTF) of ROSA-V Program by simulating a vessel top small break loss-of-coolant accident (SBLOCA) at a pressurized water reactor (PWR). Typical phenomena to the vessel top break LOCA and effectiveness of operator recovery actions on core cooling were studied under an assumption of total failure of high pressure injection (HPI) system. The LSTF simulates a 4-loop 3423 MWt PWR by a full-height, full-pressure and 1/48 volume scaling two-loop system. Typical phenomena of vessel top break LOCA are clarified for the cases with break sizes of 1.0 and 0.1% cold leg break equivalent. The results from a 0.5% top break LOCA test (SB-PV-02) in the early ROSA-IV Program was referred during discussion. Operator actions of HPI recovery in the 1.0% top break test and steam generator (SG) depressurization in the 0.1% top break test were initiated when temperature at core exit thermocouple (CET) reached 623 K during core boil-off. Both operator actions resulted in immediate recovery of core cooling. Based on the obtained data, several thermal-hydraulic phenomena were discussed further such as relations between vessel top head water level and steam discharge at the break, and between coolant mass inventory transient and core heat-up and quench behavior, and CET performances to detect core heat-up under influences of three-dimensional (3D) steam flows in the core and core exit. (author)

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)

CITADEL: a computer code for the analysis of iodine behavior in steam generator tube rupture accidents

International Nuclear Information System (INIS)

1982-04-01

The computer code CITADEL was written to analyze iodine behavior during steam generator tube rupture accidents. The code models the transport and deposition of iodine from its point of escape at the steam generator primary break until its release to the environment. This report provides a brief description of the code including its input requirements and the nature and form of its output. A user's guide describing the manner in which the input data are required to be set up to run the code is also provided

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.

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

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)

The study on water ingress mass in the steam generator heat-exchange tube rupture accident of modular high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Wang Yan; Shi Lei; Li Fu; Zheng Yanhua

2012-01-01

The steam generator heat-exchange tube rupture (SGTR) accident is an important and particular accident which will result in water ingress to the primary loop of reactor. Water ingress will result in chemical reaction of graphite fuel and structure with water, which may cause overpressure due to generation of explosive gaseous in large quantity. The study on the water ingress accident is significant for the verification of the inherent characteristics of high temperature gas-cooled reactor. The previous research shows that the amount of water ingress mass is the dominant key factor on the severity of the accident consequence. The 200 MWe high temperature gas-cooled reactor (HTR-PM), which is the first modular pebble-bed high temperature gas-cooled reactor in China designed by the Institute of Nuclear and New Energy Technology of Tsinghua University, is selected to be analyzed in this paper. The different DBA accident scenarios of double-ended break of single heat-exchange tube are simulated respectively by the thermal-hydraulic analysis code RETRAN-02. The results show the water ingress mass through the broken heat-exchange tube is related to the break location. The amount of water ingress mass is affected obviously by the capacity of the emptier system. With the balance of safety and economical efficiency, the amount of water ingress mass from the secondary side of steam generator into the primary coolant loop will be reduced by increasing properly the diameter of the draining lines. (authors)

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

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

Modelling of Zircaloy-steam-oxidation under severe fuel damage conditions

International Nuclear Information System (INIS)

Malang, S.; Neitzel, H.J.

1983-01-01

Small break loss-of-coolant accidents and special transients in an LWR, in combination with loss of required safety systems, may lead to an uncovered core for an extended period of time. As a consequence, the cladding temperature could rise up to the melting point due to the decay heat, resulting in severely damaged fuel rods. During heat-up the claddings oxidize due to oxygen uptake from the steam atmosphere in the core. The modeling and assessment of the Zircaloy-steam oxidation under such conditions is important, mainly for two reasons: The oxidation of the cladding influences the temperature transients due to the exothermic heat of reaction; the amount of liquified fuel depends on the oxide layer thickness and the oxygen content of the remaining Zircaloy metal when the melting point is reached. (author)

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.

Estimation of break location and size for loss of coolant accidents using neural networks

International Nuclear Information System (INIS)

Na, Man Gyun; Shin, Sun Ho; Jung, Dong Won; Kim, Soong Pyung; Jeong, Ji Hwan; Lee, Byung Chul

2004-01-01

In this work, a probabilistic neural network (PNN) that has been applied well to the classification problems is used in order to identify the break locations of loss of coolant accidents (LOCA) such as hot-leg, cold-leg and steam generator tubes. Also, a fuzzy neural network (FNN) is designed to estimate the break size. The inputs to PNN and FNN are time-integrated values obtained by integrating measurement signals during a short time interval after reactor scram. An automatic structure constructor for the fuzzy neural network automatically selects the input variables from the time-integrated values of many measured signals, and optimizes the number of rules and its related parameters. It is verified that the proposed algorithm identifies very well the break locations of LOCAs and also, estimate their break size accurately

Performance Analysis of AP1000 Passive Systems during Direct Vessel Injection (DVI Line Break

Directory of Open Access Journals (Sweden)

A.S. Ekariansyah

2016-08-01

Full Text Available Generation II Nuclear Power Plants (NPPs have a design weakness as shown by the Fukushima accident. Therefore, Generation III+ NPPs are developed with focus on improvements of fuel technology and thermal efficiency, standardized design, and the use of passive safety system. One type of Generation III+ NPP is the AP1000 that is a pressurized water reactor (PWR type that has received the final design acceptance from US-NRC and is already under construction at several sites in China as of 2015. The aim of this study is to investigate the behavior and performance of the passive safety system in the AP1000 and to verify the safety margin during the direct vessel injection (DVI line break as selected event. This event was simulated using RELAP5/SCDAP/Mod3.4 as a best-estimate code developed for transient simulation of light water reactors during postulated accidents. This event is also described in the AP1000 design control document as one of several postulated accidents simulated using the NOTRUMP code. The results obtained from RELAP5 calculation was then compared with the results of simulations using the NOTRUMP code. The results show relatively good agreements in terms of time sequences and characteristics of some injected flow from the passive safety system. The simulation results show that the break of one of the two available DVI lines can be mitigated by the injected coolant flowing, which is operated effectively by gravity and density difference in the cooling system and does not lead to core uncovery. Despite the substantial effort to obtain an apropriate AP1000 model due to lack of detailed geometrical data, the present model can be used as a platform model for other initiating event considered in the AP1000 accident analysis.

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.)

Computational simulation of chromosome breaks in human liver

International Nuclear Information System (INIS)

Yang Jianshe; Li Wenjian; Jin Xiaodong

2006-01-01

An easy method was established for computing chromosome breaks in cells exposed to heavily charged particles. The cell chromosome break value by 12 C +6 ions was theoretically calculated, and was tested with experimental data of chromosome breaks by using a premature chromosome condensation technique. The theoretical chromosome break value agreed well with the experimental data. The higher relative biological effectiveness of the heavy ions was closely correlated to its physical characteristics. In addition, the chromosome break value can be predicted off line. (authors)

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.)

Quantum Monte Carlo detection of SU(2 symmetry breaking in the participation entropies of line subsystems

Directory of Open Access Journals (Sweden)

David J. Luitz, Nicolas Laflorencie

2017-03-01

Full Text Available Using quantum Monte Carlo simulations, we compute the participation (Shannon-R\\'enyi entropies for groundstate wave functions of Heisenberg antiferromagnets for one-dimensional (line subsystems of length $L$ embedded in two-dimensional ($L\\times L$ square lattices. We also study the line entropy at finite temperature, i.e. of the diagonal elements of the density matrix, for three-dimensional ($L\\times L\\times L$ cubic lattices. The breaking of SU(2 symmetry is clearly captured by a universal logarithmic scaling term $l_q\\ln L$ in the R\\'enyi entropies, in good agreement with the recent field-theory results of Misguish, Pasquier and Oshikawa [arXiv:1607.02465]. We also study the dependence of the log prefactor $l_q$ on the R\\'enyi index $q$ for which a transition is detected at $q_c\\simeq 1$.

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)

Correlation between cell survival and DNA single-strand break repair proficiency in the Chinese hamster ovary cell lines AA8 and EM9 irradiated with 365-nm ultraviolet-A radiation

Energy Technology Data Exchange (ETDEWEB)

Churchill, M.E.; Peak, J.G.; Peak, M.J. (Argonne National Lab., IL (USA))

1991-02-01

Cell survival parameters and the induction and repair of DNA single-strand breaks were measured in two Chinese hamster ovary cell lines after irradiation with monochromatic UVA radiation of wavelength 365 nm. The radiosensitive mutant cell line EM9 is known to repair ionizing-radiation-induced single-strand breaks (SSB) more slowly than the parent line AA8. EM9 was determined to be 1.7-fold more sensitive to killing by 365-nm radiation than AA8 at the 10% survival level, and EM9 had a smaller shoulder region on the survival curve ({alpha} = 1.76) than AA8 ({alpha} = 0.62). No significant differences were found between the cell lines in the initial yields of SSB induced either by {gamma}-radiation (as determined by alkaline sucrose gradient sedimentation) or by 365-nm UVA (as determined by alkaline elution). For measurement of initial SSB, cells were irradiated at 0.5{sup o}C to minimize DNA repair processes. Rejoining of 365-nm induced SSB was measured by irradiating cells at 0.5{sup o}C, allowing them to repair at 37{sup o}C in full culture medium, and then quantitating the remaining SSB by alkaline elution. The repair of these breaks followed biphasic kinetics in both cell lines. EM9 repaired the breaks more slowly (T{sub 1/2} values of 1.3 and 61.3 min) than did AA8 (T{sub 1/2} values of 0.9 and 53.3 min), and EM9 also left more breaks unrepaired 90 min after irradiation (24% vs 8% for AA8). Thus, the sensitivity of EM9 to 365-nm radiation correlated with its deficiency in repairing DNA lesions revealed as SSB in alkaline elution. These results suggest that DNA may be a critical target in 365-nm induced cellular lethality and that the ability of AA8 and EM9 cells to repair DNA strand breaks may be related to their ability to survive 365-nm radiation. (author).

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

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.

A simple evaluation of containment integrity against ex-vessel steam explosion

International Nuclear Information System (INIS)

Nishiura, Hiroshi

2000-01-01

The guideline for consideration to severe accidents on containment design for next-generation LWR was published in 1999. In order to verify the validity of future containment designs, we have developed a method of assessing for the containment integrity against ex-vessel steam explosion. First, we conducted a simple evaluation on an Advanced PWR. The strength of the reactor cavity wall was assumed to be equivalent to the total strain energy which would accumulate by the time one reinforcing bar element would first reach the failure strain in FEM analyses. As a result, the strength was evaluated to be about 72 MJ. The explosion energy was assumed to be a function of the mass of the dropping melted core and the conversion ratio. Assuming the conversion ratio of 1%, it was estimated that the explosion energy would amount to about 1 MJ if the melt mass corresponds to the break of one instrumentation guide tube penetration, and about 40 MJ if the mass corresponds to the simultaneous break of all penetrations. Therefore, it is expected that the explosion energy would be less than the wall strength; thus, the containment integrity would be maintained even if an ex-vessel steam explosion were to occur. (author)

Single Tube Rupture at Cold Side of Steam Generator Simulation Test Report Using the ATLAS

International Nuclear Information System (INIS)

Kang, Kyoung Ho; Park, Hyun Sik; Cho, Seok

2010-12-01

In this study, a postulated SGTR event of the APR1400 was experimentally investigated with the ATLAS. In order to simulate a double-ended rupture of a single U-tube in the APR1400, the SGTR-CL-01 test was performed with the ATLAS. The main objectives of this test were not only to provide a physical insight into the system response of the APR1400 during the SGTR but also to produce integral effect experimental data to validate the safety analysis code. In the present report, major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Following the reactor trip induced by high steam generator level (HSGL) signal, the primary system pressure decreased and the secondary system pressure increased until the MSSVs was opened. The MSSVs repeated on and off status depending on the secondary system pressure during the whole test period. Due to the break flow, the collapsed water level of the affected steam generator showed milder decrease than that of the intact steam generator. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code. Furthermore, this data can be utilized to identify any code deficiency for a SGTR simulation, especially for DVI-adapted plants

ANALISIS KEJADIAN STEAM GENERATOR TUBE RUPTURE (SGTR BERDASARKAN SKENARIO MIHAMA UNIT 2

Directory of Open Access Journals (Sweden)

Andi Sofrany Ekariansyah

2015-03-01

Japanese standard PWR to be simulated using RELAP5/SCDAP/Mod3.2 thermal-hydraulic code. The purpose is to compare consequences resulted if this accident is occurred on the Japanese standard PWR. Parameter compared are break mass flow, fluctuation of primary and secondary pressure, and fluctuation of pressurizer level. The simulation result shown that the difference in the time duration from the initiation of rupture up to the leak termination, which takes place in shorter duration on the standard Japanese PWR. It is also shown that the total amount of the primary coolant leaked through the break nozzle to the secondary system that calculated is bigger than on the Mihama unit 2. The character of break mass flow, fluctuation of the primary system and level of pressurizer is slightly different in the beginning of the event, but is in similar trend in the end of event as the break flow is terminated. The simulation result also shows the necessity of operator action to manually isolate the auxiliary feedwater system in the affected steam generator, to actuate the main steam relief valves in the intact steam generator, and to actuate the auxiliary spray and power operated relief valve on pressurizer to anticipate the event as part of the emergency operating procedures. Keywords: SGTR, Mihama Unit 2,standard Japanese PWR

Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

Energy Technology Data Exchange (ETDEWEB)

Bae, Hwang; Ryu, Sung Uk; Yi, Sung Jae; Park, Hyun Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Dong Eok [Dept. of Precision Mechanical Engineering, Kyungpook National University, Sangju (Korea, Republic of)

2017-08-15

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

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...

Nitrosative stress induces DNA strand breaks but not caspase mediated apoptosis in a lung cancer cell line

Directory of Open Access Journals (Sweden)

Bentz Brandon G

2004-12-01

Full Text Available Abstract Background Key steps crucial to the process of tumor progression are genomic instability and escape from apoptosis. Nitric oxide and its interrelated reactive intermediates (collectively denoted as NOX have been implicated in DNA damage and mutational events leading to cancer development, while also being implicated in the inhibition of apoptosis through S-nitrosation of key apoptotic enzymes. The purpose of this study was to explore the interrelationship between NOX-mediated DNA strand breaks (DSBs and apoptosis in cultured tumor cell lines. Methods Two well-characterized cell lines were exposed to increasing concentrations of exogenous NOX via donor compounds. Production of NOX was quantified by the Greiss reaction and spectrophotometery, and confirmed by nitrotyrosine immunostaining. DSBs were measured by the alkaline single-cell gel electrophoresis assay (the COMET assay, and correlated with cell viability by the MTT assay. Apoptosis was analyzed both by TUNEL staining and Annexin V/propidium iodine FACS. Finally, caspase enzymatic activity was measured using an in-vitro fluorogenic caspase assay. Results Increases in DNA strand breaks in our tumor cells, but not in control fibroblasts, correlated with the concentration as well as rate of release of exogenously administered NOX. This increase in DSBs did not correlate with an increase in cell death or apoptosis in our tumor cell line. Finally, this lack of apoptosis was found to correlate with inhibition of caspase activity upon exposure to thiol- but not NONOate-based NOX donor compounds. Conclusions Genotoxicity appears to be highly interrelated with both the concentration and kinetic delivery of NOX. Moreover, alterations in cell apoptosis can be seen as a consequence of the explicit mechanisms of NOX delivery. These findings lend credence to the hypothesis that NOX may play an important role in tumor progression, and underscores potential pitfalls which should be considered when

Sizewell B

International Nuclear Information System (INIS)

1983-07-01

The safety of pressure circuit components is discussed with respect to coolant loop piping, coolant pumps, pressuriser, steam generator channel head and shell, primary component supports and restraints, accumulators, main steam line no-break zone, reactor internals and core, and valves including main steam isolating valves. Outstanding issues are identified and a programme of additional work is discussed. (U.K.)

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.)

Evaluation of Flood Level under Main Feedwater Line Break Accident using GOTHIC Computer Code and Analytical Calculation by ANSI 56.11

International Nuclear Information System (INIS)

Kim, Keon Yeop; Park, Jae Won; Jeon, Woo Jae

2016-01-01

The design basis internal flooding is caused by postulated pipe ruptures or component failures. The flooding can cause failure of safety-related equipment and affect the integrity of the structure. Though large diameter pipe rupture is significant in flooding analysis, split breaks should also be considered with consideration of a spectrum of pipe break size and power level. The pipe rupture analysis should be based on the most severe single active failure. For enveloping spectrum of pipe break condition, flood relief paths are necessary and passive flood protection without operating action, basically, shall be applied. In this study, the evaluation of flood level in case of Main Feedwater Line Break (MFLB) was performed by using GOTHIC computer program and hand calculation. The flooding analyses were performed by hand calculation and GOTHIC analysis for an assumed MFLB condition. The calculated flood levels were 0.823m and 0.691m for hand calculation and GOTHIC analysis, respectively. In comparison to the GOTHIC analysis, hand calculation showed conservative results. However, in actual flood protection design, margin for uncertainty shall be considered, in order to reflect the outflow reducing effect due to vortex and intake of air

Evaluation of Flood Level under Main Feedwater Line Break Accident using GOTHIC Computer Code and Analytical Calculation by ANSI 56.11

Energy Technology Data Exchange (ETDEWEB)

Kim, Keon Yeop; Park, Jae Won; Jeon, Woo Jae [FNC Technology Co., Yongin (Korea, Republic of)

2016-10-15

The design basis internal flooding is caused by postulated pipe ruptures or component failures. The flooding can cause failure of safety-related equipment and affect the integrity of the structure. Though large diameter pipe rupture is significant in flooding analysis, split breaks should also be considered with consideration of a spectrum of pipe break size and power level. The pipe rupture analysis should be based on the most severe single active failure. For enveloping spectrum of pipe break condition, flood relief paths are necessary and passive flood protection without operating action, basically, shall be applied. In this study, the evaluation of flood level in case of Main Feedwater Line Break (MFLB) was performed by using GOTHIC computer program and hand calculation. The flooding analyses were performed by hand calculation and GOTHIC analysis for an assumed MFLB condition. The calculated flood levels were 0.823m and 0.691m for hand calculation and GOTHIC analysis, respectively. In comparison to the GOTHIC analysis, hand calculation showed conservative results. However, in actual flood protection design, margin for uncertainty shall be considered, in order to reflect the outflow reducing effect due to vortex and intake of air.

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.

Quick Look Report for Semiscale MOD-2C Test S-FS-2

International Nuclear Information System (INIS)

Boucher, T.J.; Chen, T.H.

1985-01-01

Results of a preliminary analysis of the first test performed in the Semiscale MOD-2C Steam Generator Feedwater and Steam Line Break (FS) experiment series are presented. Test S-FS-2 simulated a pressurized water reactor transient initiated by a double-ended offset shear of a steam generator main steam line upstream of the flow restrictor. Initial conditions represented normal ''hot-standby'' operation. The transient included an initial 600-s period in which only automatic plant protection systems responded to the initiating event. This period was followed by a series of operator actions necessary to stabilize the plant at conditions required to allow a natural circulation cooldown. The test results provided a measured evaluation of the effectiveness of the automatic responses in minimizing primary system overcooling and operator actions in stabilizing the plant. Test data also provided a basis for comparison with other tests in the series of the effects of break size on primary overcooling and primary-to-secondary heat transfer. 57 figs., 3 tabs

Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

Directory of Open Access Journals (Sweden)

Hwang Bae

2017-08-01

Full Text Available Three small-break loss-of-coolant accident (SBLOCA tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor, i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

Calculation of relative tube/tube support plate displacements in steam generators under accident condition loads using non-linear dynamic analysis methodologies

International Nuclear Information System (INIS)

Smith, R.E.; Waisman, R.; Hu, M.H.; Frick, T.M.

1995-01-01

A non-linear analysis has been performed to determine relative motions between tubes and tube support plates (TSP) during a steam line break (SLB) event for steam generators. The SLB event results in blowdown of steam and water out of the steam generator. The fluid blowdown generates pressure drops across the TSPS, resulting in out-of-plane motion. The SLB induced pressure loads are calculated with a computer program that uses a drift-flux modeling of the two-phase flow. In order to determine the relative tube/TSP motions, a nonlinear dynamic time-history analysis is performed using a structural model that considers all of the significant component members relative to the tube support system. The dynamic response of the structure to the pressure loads is calculated using a special purpose computer program. This program links the various substructures at common degrees of freedom into a combined mass and stiffness matrix. The program accounts for structural non-linearities, including potential tube and TSP interaction at any given tube position. The program also accounts for structural damping as part of the dynamic response. Incorporating all of the above effects, the equations of motion are solved to give TSP displacements at the reduced set of DOF. Using the displacement results from the dynamic analysis, plate stresses are then calculated using the detailed component models. Displacements form the dynamic analysis are imposed as boundary conditions at the DOF locations, and the finite element program then solves for the overall distorted geometry. Calculations are also performed to assure that assumptions regarding elastic response of the various structural members and support points are valid

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%

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)

Coincident steam generator tube rupture and stuck-open safety relief valve carryover tests: MB-2 steam generator transient response test program

International Nuclear Information System (INIS)

Garbett, K.; Mendler, O.J.; Gardner, G.C.; Garnsey, R.; Young, M.Y.

1987-03-01

In PWR steam generator tube rupture (SGTR) faults, a direct pathway for the release of radioactive fission products can exist if there is a coincident stuck-open safety relief valve (SORV) or if the safety relief valve is cycled. In addition to the release of fission products from the bulk steam generator water by moisture carryover, there exists the possibility that some primary coolant may be released without having first mixed with the bulk water - a process called primary coolant bypassing. The MB-2 Phase II test program was designed specifically to identify the processes for droplet carryover during SGTR faults and to provide data of sufficient accuracy for use in developing physical models and computer codes to describe activity release. The test program consisted of sixteen separate tests designed to cover a range of steady-state and transient fault conditions. These included a full SGTR/SORV transient simulation, two SGTR overfill tests, ten steady-state SGTR tests at water levels ranging from very low levels in the bundle up to those when the dryer was flooded, and three moisture carryover tests without SGTR. In these tests the influence of break location and the effect of bypassing the dryer were also studied. In a final test the behavior with respect to aerosol particles in a dry steam generator, appropriate to a severe accident fault, was investigated

ARI3SG: Aerosol retention in the secondary side of a steam generator. Part II: Model validation and uncertainty analysis

International Nuclear Information System (INIS)

Lopez, Claudia; Herranz, Luis E.

2012-01-01

Highlights: ► Validation of a model (ARI3SG) for the aerosol retention in the break stage of a steam generator under SGTR conditions. ► Interpretation of the experimental SGTR and CAAT data by using the ARI3SG model. ► Assessment of the epistemic and stochastic uncertainties effect on the ARI3SG results. - Abstract: A large body of data has been gathered in the last decade through the EU-SGTR, ARTIST and ARTIST 2 projects for aerosol retention in the steam generator during SGTR severe accident sequences. At the same time the attempt to extend the analytical capability has resulted in models that need to be validated. The ARI3SG is one of such developments and it has been built to estimate the aerosol retention in the break stage of a “dry” steam generator. This paper assesses the ARI3SG predictability by comparing its estimates to open data and by analyzing the effect of associated uncertainties. Datamodel comparison has been shown to be satisfactory and highlight the potential use of an ARI3SG-like formulation in system codes.

Attentuation effect of sprinklers on a superheated environment using COMPACT

International Nuclear Information System (INIS)

Wills, M.E.

1991-01-01

As a result of a high-energy line break (HELB) in the auxiliary building of a pressurized water reactor, superheated steam is released through a 15.2-cm (6-in.) break in the main stream line. A sophisticated analytical tool is needed to simulate the effects of natural circulation airflow in the building. The COMPACT computer code was used for this study not only because of its ability to model natural circulation airflow but also because it is a recognized and approved tool for outside containment steam-line break applications. Two cases were analyzed in this study. The first case did not model the Appendix R fire protection sprinklers in the break node, while the second case did. Previous conservative analyses have shown that four sprinklers would be available, each providing a flow rate of ∼15 gal/min. A fusible link must fail in order to actuate the sprinklers, and conservative calculations show that the failure time would be 113 s from event initiation. Since the sprinkler headers are already full of water, there is no additional delay time for sprinkler flow to the break compartment

Assessment of RELAP5/MOD3.1 with the LSTF SB-SG-06 experiment simulating a steam generator tube rupture transient

International Nuclear Information System (INIS)

Seul, K.W.; Bang, Y.S.; Lee, S.; Kim, H.J.

1996-09-01

The objective of the present work is to identify the predictability of RELAP5/MOD3.1 regarding thermal-hydraulic behavior during a steam generator tube rupture (SGTR). To evaluate the computed results, LSTF SB-SG-06 test data simulating the SGTR that occurred at the Mihama Unit 2 in 1991 are used. Also, some sensitivity studies of the code change in RELAP5, the break simulation model, and the break valve discharge coefficient are performed. The calculation results indicate that the RELAP5/MOD3.1 code predicted well the sequence of events and the major phenomena during the transient, such as the asymmetric loop behavior, reactor coolant system (RCS) 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 (SG) relief valve, and so on. However, there are some differences from the experimental data in the number of the relief valve cycling in the affected SG, and the flow regime of the hot leg with the pressurizer, and the break flow rates. Finally, the calculation also indicates that the coolant in the core could remain in a subcooled state as a result of the heat transfer caused by the natural circulation flow even if the reactor coolant pumps (RCPs) turned off and that the affected SG could be properly isolated to minimize the radiological release after the SGTR

Assessment of RELAP5/MOD3.1 with the LSTF SB-SG-06 experiment simulating a steam generator tube rupture transient

Energy Technology Data Exchange (ETDEWEB)

Seul, K.W.; Bang, Y.S.; Lee, S.; Kim, H.J. [Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)

1996-09-01

The objective of the present work is to identify the predictability of RELAP5/MOD3.1 regarding thermal-hydraulic behavior during a steam generator tube rupture (SGTR). To evaluate the computed results, LSTF SB-SG-06 test data simulating the SGTR that occurred at the Mihama Unit 2 in 1991 are used. Also, some sensitivity studies of the code change in RELAP5, the break simulation model, and the break valve discharge coefficient are performed. The calculation results indicate that the RELAP5/MOD3.1 code predicted well the sequence of events and the major phenomena during the transient, such as the asymmetric loop behavior, reactor coolant system (RCS) 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 (SG) relief valve, and so on. However, there are some differences from the experimental data in the number of the relief valve cycling in the affected SG, and the flow regime of the hot leg with the pressurizer, and the break flow rates. Finally, the calculation also indicates that the coolant in the core could remain in a subcooled state as a result of the heat transfer caused by the natural circulation flow even if the reactor coolant pumps (RCPs) turned off and that the affected SG could be properly isolated to minimize the radiological release after the SGTR.

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.

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.

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.

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

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.)

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

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)

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

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.

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

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

Uncertainty analysis for probabilistic steam generators tube rupture in LBB applications

Energy Technology Data Exchange (ETDEWEB)

Durbec, V.; Pitner, P.; Pages, D. [Electricite de France, 78 - Chatou (France). Research and Development Div.; Riffard, T. [Electricite de France, 69 - Villeurbanne (France). Engineering and Construction Div.; Flesch, B. [Electricite de France, 92 - Paris la Defense (France). Generation and Transmission Div.

1997-10-01

Steam Generators (SG) of Pressurized Water Reactors have experienced world wide various types of tube degradations, mainly from stress corrosion cracking; because of this damage, primary-secondary leakage or tube rupture can occur. Safety against the risk of tube rupture is achieved through a combination of periodic in-service inspections (eddy current testing), surveillance of leaks during operation (leak before break concept) and tube plugging. In order to optimize the tube bundle SG maintenance, Electricite de France has developed a specific software named COMPROMIS. The model, based on probabilistic fracture mechanics makes it possible to quantify the influence of in service inspections and maintenance work on the risk of a SG Tube Rupture (SGTR), taking all significant parameters into account as random variables (initial defect size distribution, reliability of non-destructive examinations, crack initiation and propagation, critical sizes, leak before risk of break, etc...). This paper focuses on the leak rate calculation module and presents a sensitivity study of the influence of the leak before break on the conditional failure probability. (author) 8 refs.

Uncertainty analysis for probabilistic steam generators tube rupture in LBB applications

International Nuclear Information System (INIS)

Durbec, V.; Pitner, P.; Pages, D.; Riffard, T.; Flesch, B.

1997-10-01

Steam Generators (SG) of Pressurized Water Reactors have experienced world wide various types of tube degradations, mainly from stress corrosion cracking; because of this damage, primary-secondary leakage or tube rupture can occur. Safety against the risk of tube rupture is achieved through a combination of periodic in-service inspections (eddy current testing), surveillance of leaks during operation (leak before break concept) and tube plugging. In order to optimize the tube bundle SG maintenance, Electricite de France has developed a specific software named COMPROMIS. The model, based on probabilistic fracture mechanics makes it possible to quantify the influence of in service inspections and maintenance work on the risk of a SG Tube Rupture (SGTR), taking all significant parameters into account as random variables (initial defect size distribution, reliability of non-destructive examinations, crack initiation and propagation, critical sizes, leak before risk of break, etc...). This paper focuses on the leak rate calculation module and presents a sensitivity study of the influence of the leak before break on the conditional failure probability. (author)

Major Achievements and Prospect of the ATLAS Integral Effect Tests

International Nuclear Information System (INIS)

Choi, K.; Kim, Y.; Song, C.; Baek, W.

2012-01-01

A large-scale thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been operated by KAERI. The reference plant of ATLAS is the APR1400 (Advanced Power Reactor, 1400 MWe). Since 2007, an extensive series of experimental works were successfully carried out, including large break loss of coolant accident tests, small break loss of coolant accident tests at various break locations, steam generator tube rupture tests, feed line break tests, and steam line break tests. These tests contributed toward an understanding of the unique thermal-hydraulic behavior, resolving the safety-related concerns and providing validation data for evaluation of the safety analysis codes and methodology for the advanced pressurized water reactor, APR1400. Major discoveries and lessons found in the past integral effect tests are summarized in this paper. As the demand for integral effect tests is on the rise due to the active national nuclear R and D program in Korea, the future prospects of the application of the ATLAS facility are also discussed.

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.

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

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

Design and manufacture of steam generators for replacement

International Nuclear Information System (INIS)

Hirano, Hiroshi; Kuri, Syuhei

1995-01-01

The basic specification of the steam generators for replacement as heat exchangers (the pressure, temperature, flow rate and thermal output on primary and secondary sides) is set same as that of steam generators before replacement, but the latest design reflecting the operation experience obtained so far and taking the countermeasures for preventing heating tube damage in it is adopted, such as the heating tubes made of TT 690 alloy, the tube support plates with four-lobe shape tube holes made of stainless steel, the stainless steel rest fittings of three in one set and so on. After the heating tube break accident in Mihama No. 2 plant, the quality control was further strengthened. The comparison of the specifications of the steam generators of respective plants before and after the replacement is shown. The main objective of improving steam generators is the heightening of the reliability of heating tubes against intergranular attack and primary water stress corrosion cracking. The improvements of heating tube material, tube support plate material, secondary side heat flow, the shape of tube holes of tube support plates, the method of expanding heating tubes, and vibration-controlling fittings are explained. As to the manufacturing procedure and quality control, the manufacture of raw materials, the build-up welding of tube plates, the manufacture of lower half shell plates, the tube hole making of support plates, the insection of outer cylinder, flow rate distribution plate. Support plates and heating tubes, the sealing welding and expanding of heating tubes, the fixing of rest fittings, the manufacture and fixing of water chamber cover, the manufacture of upper half shell, the fixing of parts inside it, the final joint and inspection are described. (K.I.)

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

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

Frequency and distribution of leakages in steam generators of gas-cooled reactors

International Nuclear Information System (INIS)

Bongratz, R.; Breitbach, G.; Wolters, J.

1988-01-01

In gas cooled reactors with graphitic primary circuit structures - such as HTR, AGR or Magnox - the water ingress is an event of great safety concern. Water or steam entering the primary circuit react with the hot graphite and carbon-oxide and hydrogen are produced. As the most important initiating event a leak in a steam generator must be taken into account. From the safety point of view as well as for availability reasons it is necessary to construct reliable boilers. Thus the occurrence of a boiler leak should be a rare event. In the context of a probabilistic safety study for an HTR-Project much effort was invested to get information about the frequency and the size distribution of tube failures in steam generators of gas cooled reactors. The main data base was the boiler tube failure statistics of United Kingdom gas cooled reactors. The data were selected and applied to a modern HTR steam generator design. A review of the data showed that the failure frequency is not connected with the load level (pressures, temperatures) or with the geometric size of the heating surface of the boiler. Design, construction, fabrication, examination and operation conditions have the greatest influence an the failure frequency but they are practically not to be quantified. The typical leak develops from smallest size. By erosion effects of the entering water or steam it is enlarged to perhaps some mm 2 , then usually it is detected by moisture monitors. Sudden tube breaks were not reported in the investigated period. As a rule boiler leaks in gas cooled reactors are much more, rare then leaks in steam generators of light water reactors and fossil fired boilers. (author)

Probability of a steam generator tube rupture due to the presence of axial through wall cracks

International Nuclear Information System (INIS)

Mavko, B.; Cizelj, L.

1991-01-01

Using the Leak-Before-Break (LBB) approach to define tube plugging criteria a possibility to operate with through wall crack(s) in steam generator tubes may be considered. This fact may imply an increase in tube rupture probability. Improved examination techniques (in addition to the 100% tube examination) have been developed and introduced to counterbalance the associated risk. However no estimates of the amount of total increase or decrease of risk due to the introduction of LBB have been made. A scheme to predict this change of risk is proposed in the paper, based on probabilistic fracture mechanics analysis of axial cracks combined with available data of steam generator tube nondestructive examination reliability. (author)

Modelling and testing the molecular mechanism of radiation-induced chromatid breaks

International Nuclear Information System (INIS)

Bryant, P. E.

2001-01-01

Chromatid breaks induced by ionizing radiation in the G2 phase of the cell cycle are considered as markers of individual human radiosensitivity and may indicate the presence of low-penetrance genes regulating susceptibility to breast and other cancers). Together with our own study of Scottish (Tayside) breast cancer patients and a group of normal controls these studies show an overall 10-fold variation in chromatid break frequency (the parameter defining individual chromosomal 'radiosensitivity'). Thus, an understanding of the mechanisms and genes involved in determining these widely different responses should help to clarify the reasons for individual radiosensitivity and may lead us to identify key genes involved in cancer susceptibility. The presence of colour-switches at around 16% of chromatid break points (detected in harlequin-stained chromosomes) indicates that this type of chromatid break is formed by a chromatin rearrangement involving one or more large chromatin domains of the order of 3 - 5 Mbp, possibly representing transcription 'factories'. The signal model of chromatid breaks assumes that all chromatid breaks are the result of chromatin rearrangements, and that the initiating DNA double-strand break (dsb) is itself not involved in the rearrangement but signals its presence (possibly via ATM protein or DNAPK) leading to the initiation of the chromatin rearrangement. Experimental evidence from radiosensitive cell lines (e.g. human AT and hamster irs2) and with the nucleoside analogue araA (9-β-D-arabinofuranosyladenine) demonstrates the lack of correspondence between the rejoining kinetics of dsb and that of disappearance of chromatid breaks, thus supporting the signal model. Coupled with the linear induction of chromatid breaks with dose in both human and rodent cell lines of various types, and the production of chromatid breaks by single dsb in genetically engineered cell lines the classical 'breakage-first' model of chromatid breaks is no longer

Iris small break loca phenomena identification and ranking table (PIRT)

International Nuclear Information System (INIS)

Larson, T.K.; Moody, F.J.; Wilson, G.E.; Brown, W.L.; Frepoli, C.; Hartz, J.; Woods, B.G.; Oriani, L.

2007-01-01

The international reactor innovative and secure (IRIS) is a modular pressurized water reactor with an integral configuration (all primary system components - reactor core, internals, pumps, steam generators, pressurizer, and control rod drive mechanisms - are inside the reactor vessel). The IRIS plant conceptual design was completed in 2001 and the preliminary design is currently underway. The pre-application licensing process with the United States Nuclear Regulatory Commission (USNRC) started in October 2002. The first line of defense in IRIS is to eliminate event initiators that could potentially lead to core damage. If it is not possible to eliminate certain accidents altogether, then the design inherently reduces their consequences and/or decreases their probability of occurring. One of the most obvious advantages of the IRIS Safety-by-Design TM approach is the elimination of large break loss-of-coolant accidents (LBLOCAs), since no large primary penetrations of the reactor vessel or large loop piping exist. While the IRIS Safety-by-Design TM approach is a logical step in the effort to produce advanced reactors, the desired advances in safety must still be demonstrated in the licensing arena. With the elimination of LBLOCA, an important next consideration is to show the IRIS design fulfills the promise of increased safety also for small break LOCAs (SBLOCAs). Accordingly, the SBLOCA phenomena identification and ranking table (PIRT) project was established. The primary objective of the IRIS SBLOCA PIRT project was to identify the relative importance of phenomena in the IRIS response to SBLOCAs. This relative importance, coupled with the current relative state of knowledge for the phenomena, provides a framework for the planning of the continued experimental and analytical efforts. To satisfy the SBLOCA PIRT project objectives, Westinghouse organized an expert panel whose members were carefully selected to insure that the PIRT results reflect internationally

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

Steam Turbine Control Valve Stiction Effect on Power System Stability

International Nuclear Information System (INIS)

Halimi, B.

2010-01-01

four stages. The steam chest is represented by the first stage. The three other stages represent either reheaters or crossover piping. The shaft model which represents a four-mass system is coupled to the mass in a synchronous machine (generator) for a total of five masses. To analyze the oscillation in power system, an IEEE benchmark is used as reference scenario. In this scenario, a single generator is connected to infinite bus via a transmission line

Preliminary Analysis of Severe Accident Progression Initiated from Small Break LOCA of a SMART Reactor

International Nuclear Information System (INIS)

Jin, Young Ho; Park, Jong Hwa; Kim, Dong Ha; Cho, Seong Won

2010-01-01

SMART (System integrated Modular Advanced ReacTor), is under the development at Korea Atomic Energy Research Institute (KAERI). SMART is an integral type pressurized water reactor which contains a pressurizer, 4 reactor coolant pumps (RCPs), and 8 steam generator cassettes(S/Gs) in a single reactor vessel. This reactor has substantially enhanced its safety with an integral layout of its major components, 4 trains of safety injection systems (SISs), and an adoption of 4 trains of passive residual heat removal systems (PRHRS) instead of an active auxiliary feedwater system . The thermal power is 330 MWth. During the conceptual design stage, a preliminary PSA was performed. PSA results identified that a small break loss of coolant accident (SLOCA) with all safety injections unavailable is one of important severe core damage sequences. Clear understanding of this sequence helps in the developing accident mitigation strategies. MIDAS/SMR computer code is used to simulate the severe accident progression initiated from a small break LOCA in SMART reactor. This code has capability to model a helical steam generator which is adopted in SMART reactor. The important accident progression results for SMART reactor are then compared with the typical pressurized water reactor (PWR) result

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

Aerosol trapping in steam generator (artist): an investigation of aerosol and iodine behaviour in the secondary side of a steam generator

International Nuclear Information System (INIS)

Guentay, S.; Birchley, J.; Suckow, D.; Dehbi, A.

2000-01-01

Incidents such as a steam generator tube rupture (SGTR) with stuck-open relief valve are important accident sequences for analysis by virtue of the open path for release of radioactivity which ensues. The release may be mitigated by deposition of fission products on the steam generator (SG) tubes and other structures, or by scrubbing in the secondary coolant. The absence of empirical data, the complexity of the geometry and controlling processes, however, make the retention difficult to quantify and its full import is typically not taken into account in risk assessment studies. The ARTIST experimental programme at PSI will simulate the flow and retention of aerosol-borne fission products in the SG secondary, and thus provide a unique database to support safety assessments and analytical models. Scaling of the break flow represents a particular challenge since the aerosol retention processes operate at contrasting length scales. Preliminary calculations have identified a baseline set of conditions, and confirmed the feasibility of the rig design and scaling principles. Flexibility of the rig layout enables simulations to be performed for a range of SG designs, accident situations and accident management philosophies. (authors)

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.)

Steam reforming of bio-oil from coconut shell pyrolysis over Fe/olivine catalyst

International Nuclear Information System (INIS)

Quan, Cui; Xu, Shaoping; Zhou, Congcong

2017-01-01

Highlights: • Steam reforming of the actual bio-oil was investigated with Fe/olivine catalyst. • Most of phenols in bio-oil were converted into gas products. • A carbon conversion of 97.2% was obtained under optimized conditions. - Abstract: Catalytic steam reforming of coconut shell pyrolysis bio-oil over Fe/olivine catalyst was conducted in a fixed-bed quartz reactor. The effects of calcination temperature, iron loading, reaction temperature, steam to carbon ratio (S/C), bio-oil weight hourly space velocity (W b HSV) on gas composition and carbon conversion were investigated. The results indicate that Fe/olivine has good activity for steam reforming of bio-oil, the couple Fe 2+/3+ /Fe 2+ may be sufficiently efficient for C–C, C–O and C–H breaking. After steam reforming, most of the phenolics in pyrolysis oil are converted into light molecular compounds such as H 2 , CO, CO 2 , and CH 4 . The H 2 concentration and carbon conversion were enhanced by increasing reaction temperature from 750 to 800 °C and the S/C from 1.5 to 2, but decreased with increasing calcination temperature. In the W b HSV range of 0.5–0.6, the hydrogen concentration decreased obviously, whereas it decreased slightly by further increasing W b HSV. The highest hydrogen concentration of 47.6 vol% was obtained among the catalysts tested, and the best carbon conversion was 97.2% over 10% Fe/olivine catalyst under the reforming conditions of temperature = 800 °C, W b HSV = 0.5, S/C = 2.

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

Nuclear power plant status diagnostics using artificial neural networks

International Nuclear Information System (INIS)

Bartlett, E.B.; Uhrig, R.E.

1991-01-01

In this work, the nuclear power plant operating status recognition issue is investigated using artificial neural networks (ANNs). The objective is to train an ANN to classify nuclear power plant accident conditions and to assess the potential of future work in the area of plant diagnostics with ANNS. To this end, an ANN was trained to recognize normal operating conditions as well as potentially unsafe conditions based on nuclear power plant training simulator generated accident scenarios. These scenarios include; hot and cold leg loss of coolant, control rod ejection, loss of offsite power, main steam line break, main feedwater line break and steam generator tube leak accidents. Findings show that ANNs can be used to diagnose and classify nuclear power plant conditions with good results

A study on optimization of environmental qualification envelope for Kori 3 and 4 NPP

International Nuclear Information System (INIS)

Song, Dong Soo; Byun, Choong Sup; Jo, Jong Young

2009-01-01

The purpose of this study is to present the reevaluation of the Main Steam Line Break (MSLB) applied Boron Injection Tank (BIT) removal and to optimize the environmental qualification (EQ) temperature envelope with thermal lag analysis and liquid entrainment method. BIT alleviates the reactor power excursion during Main Steam Line Break (MSLB) accident. Thermal lag analysis methods by NUREG-0588 is using four times condensing heat transfer coefficient on the passive heat sink surface, the forced convection heat transfer coefficient whenever the condensing is not occurring and during blowdown stage. And the entrainment model is that the all of the break regions within the secondary side are represented by non-homogeneous vapor volumes in which the liquid and steam are uniformly mixed throughout. These methods are focused on making higher the surface temperature of the safety equipment. For the analysis, amount of released mass and energy is calculated using the LOFTRAN code and containment temperature is predicted by CONTEMPT-LT 28 code. These two codes are used to for safety analysis. In accordance with the analysis result, a plant specific EQ test envelope was proposed for Kori 3 and 4 NPP

A study on optimization of environmental qualification envelope for Kori 3 and 4 NPP

Energy Technology Data Exchange (ETDEWEB)

Song, Dong Soo; Byun, Choong Sup; Jo, Jong Young [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

2009-07-01

The purpose of this study is to present the reevaluation of the Main Steam Line Break (MSLB) applied Boron Injection Tank (BIT) removal and to optimize the environmental qualification (EQ) temperature envelope with thermal lag analysis and liquid entrainment method. BIT alleviates the reactor power excursion during Main Steam Line Break (MSLB) accident. Thermal lag analysis methods by NUREG-0588 is using four times condensing heat transfer coefficient on the passive heat sink surface, the forced convection heat transfer coefficient whenever the condensing is not occurring and during blowdown stage. And the entrainment model is that the all of the break regions within the secondary side are represented by non-homogeneous vapor volumes in which the liquid and steam are uniformly mixed throughout. These methods are focused on making higher the surface temperature of the safety equipment. For the analysis, amount of released mass and energy is calculated using the LOFTRAN code and containment temperature is predicted by CONTEMPT-LT 28 code. These two codes are used to for safety analysis. In accordance with the analysis result, a plant specific EQ test envelope was proposed for Kori 3 and 4 NPP.

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

Corruption analysis and life span forecast research on sylphon bellows of nuclear-powered steam system of ship

International Nuclear Information System (INIS)

Song Chao; Chen Lisheng; Song Meicun; Wang Wei

2012-01-01

The fracture of the corrugated pipe has a dad effect to the operation of reactor which can cause the small-break LOCA. The corrosion is the key reason of the fracture. On the base of the analysis on corruption reason of the sylphon bellows and combine the characteristic of the limited sample point of it, the Grey theory was used in the assessment of corrosion life span of corrugated pipe in nuclear steam system. Through applying the GM (1, 1) model in inferring the discipline of corrosion quantity and combining traditional statistical method, the corrosion life of steam pipe was evaluated. It indicates that the model is precise, simple and the result is reliable. (authors)

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

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.

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

Prediction of loop seal formation and clearing during small break loss of coolant accident

International Nuclear Information System (INIS)

Lee, Suk Ho; Kim, Hho Jung

1992-01-01

Behavior of loop seal formation and clearing during small break loss of coolant accident is investigated using the RELAP5/MOD2 and /MOD3 codes with the test of SB-CL-18 of the LSTF(Large Scale Test Facility). The present study examines the thermal-hydraulic mechanisms responsible for early core uncovery includeing the manometric effect due to an asymmetric coolant holdup in the steam generator upflow and downflow side. The analysis with the RELAP5/ MOD2 demonstrates the main phenomena occuring in the depressurization transient including the loop seal formation and clearing with sufficient accuracy. Nevertheless, several differences regarding the evolution of phenomena and their timing have been pointed out in the base calculations. The RELAP5/MOD3 predicts overall phenomena, particularly the steam generator liquid holdup better than the RELAP5/MOD2. The nodalization study in the components of the steam generator U-tubes and the cross-over legs with the RELAP5/MOD3 results in good prediction of the loop seal clearing phenomena and their timing. (Author)

Analysis of ATLAS FLB-EC6 Experiment using SPACE Code

International Nuclear Information System (INIS)

Lee, Donghyuk; Kim, Yohan; Kim, Seyun

2013-01-01

The new code is named SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). As a part of code validation effort, simulation of ATLAS FLB(Feedwater Line Break) experiment using SPACE code has been performed. The FLB-EC6 experiment is economizer break of a main feedwater line. The calculated results using the SPACE code are compared with those from the experiment. The ATLAS FLB-EC6 experiment, which is economizer feedwater line break, was simulated using the SPACE code. The calculated results were compared with those from the experiment. The comparisons of break flow rate and steam generator water level show good agreement with the experiment. The SPACE code is capable of predicting physical phenomena occurring during ATLAS FLB-EC6 experiment

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.

Physical and morphological properties of z ~ 3 Lyman break galaxies: dependence on Lyα line emission

Science.gov (United States)

Pentericci, L.; Grazian, A.; Scarlata, C.; Fontana, A.; Castellano, M.; Giallongo, E.; Vanzella, E.

2010-05-01

Aims: We investigate the physical and morphological properties of Lyman break galaxies (LBGs) at redshift ~2.5 to ~3.5, to determine if and how they depend on the nature and strength of the Lyα emission. Methods: We selected U-dropout galaxies from the z-detected GOODS-MUSIC catalog by adapting the classical Lyman break criteria on the GOODS filter set. We kept only those galaxies with spectroscopic confirmation, mainly from VIMOS and FORS public observations. Using the full multi-wavelength 14-bands information (U to IRAC), we determined the physical properties of the galaxies through a standard spectral energy distribution fitting procedure with the updated Charlot & Bruzual (2009) templates. We also added other relevant observations of the GOODS field, i.e. the 24 μm observations from Spitzer/MIPS and the 2 MSec Chandra X-ray observations. Finally, using non parametric diagnostics (Gini, Concentration, Asymmetry, M20 and ellipticity), we characterized the rest-frame UV morphologies of the galaxies. We then analyzed how these physical and morphological properties correlate with the presence of the Lyα emission line in the optical spectra. Results: We find that unlike at higher redshift, the dependence of physical properties on the Lyα line is milder: galaxies without Lyα in emission tend to be more massive and dustier than the rest of the sample, but all other parameters, ages, star formation rates (SFR), X-ray emission and UV morphology do not depend strongly on the presence of the Lyα emission. A simple scenario where all LBGs have intrinsically high Lyα emission, but where the dust and neutral hydrogen content (which shapes the final appearance of the Lyα) depend on the mass of the galaxies, is able to reproduce the majority of the observed properties at z˜3. Some modification might be needed to account for the observed evolution of these properties with cosmic epoch, which is also discussed.

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%

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

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

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.)

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

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

Grundremmingen nuclear power plant: The accident that came from the cold

International Nuclear Information System (INIS)

Anon.

1977-01-01

The accident of January 13, 1977, is shortly described. The power plant was cut off from the network by an extreme temperature drop and high humidity, which caused a break in a number of porcelain insulators of two HV lines. The response of the turbine speed control was delayed, and there was a sudden pressure drop in the primary steam line. Weakly radioactive steam and water were released into the containment without polluting the environment, since all safety systems responded to the accident. (HP) [de

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

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

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)

Performance of R + D works to study the behavior during blowdown of LWRs

International Nuclear Information System (INIS)

Kaffanke, E.

1980-03-01

The aim of the RS 16B experiments is to study thermohydraulic behavior during blowdown of pressure vessels (with internals) of BWRS or PWR-steam generators. A small scale vessel will be used for simulation and measurement. Three series of experiments with varying paramters will be carried out in order to gain insight into flow behavior and loadings on internals caused by steam and feedwater line breaks. The experiments, besides supplying experimental results, will also serve to verify mathematical codes that have been developed to describe such highly complex phenomena. The contributions was as follows: 1) Planning and design of models to simulate BWR-internals. 2) Pre-test calculation of the effects of pressure differences on these models. 3) Post-test calculations upon completion of experiments. Since the beginning of the RS 16B-project in 1972, blowdown effects due to steam and feedwater line breaks have been calculated. A version of the short time computer program LAMB was used for the predictions. This program is an accepted code for the licensing procedure where blowdown analysis of BWR and PWR-steam generator is involved. (orig./HP) [de

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.

ARTIST: An International Project Investigating Aerosol Retention in a Ruptured Steam Generator

International Nuclear Information System (INIS)

Guentay, S.; Dehbi, A.; Suckow, D.; Birchley, J.

2002-01-01

Steam generator tube ruptures (SGTR) with a concurrent stuck open safety relief valve are counted among the risk dominant accident sequences because of the potential for radioactive products to bypass the containment. Owing to the absence of relevant empirical data and the complexity of the geometry and controlling processes, the aerosol removal in the steam generator (SG) tubes and in the secondary side is not well understood. Therefore, little or no credit is usually taken for aerosol retention due to natural processes in the various components of a SG. To help reduce the uncertainties associated with fission product release following an SGTR sequence, the Paul Scherrer Institut has initiated an international experimental project to be performed in the ARTIST (AeRosol Trapping In a Steam generaTor) facility in the time period from 2002 to 2007. The ARTIST test section is a scaled model of a real SG, and is comprised of a 264-tube bundle with a maximum height of 3.8 m, as well as one full-size droplet separator and one full-size steam dryer. The ARTIST facility is capable of producing soluble and insoluble aerosols and entrain them at sonic gas flow rates (up to 0.25 kg/s, thus matching comparable values predicted by the codes. In addition, aerosols can be generated at prototypical concentrations (up to 5 g/m 3 ) and sizes (0.2-5 mm AMMD). State of the art instrumentation is used (Low-pressure impactors, photometers, on-line particle sizer, online droplet sizer, etc.). The ARTIST project will simulate the flow and retention of aerosol-borne fission products in the SG, and provide a unique database to support safety assessments and analytical models. The project is foreseen in seven phases: 1) Aerosol retention in the tube under dry secondary side conditions, 2) Aerosol retention in the near field close to break under dry conditions, 3) Aerosol retention in the bundle far field under dry conditions, 4) Aerosol retention in the separator and dryer under dry

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.)

Field experience of cleanliness inspection for secondary-side in PWR steam generators

International Nuclear Information System (INIS)

Ding Xunshen

1997-05-01

The mechanical cleaning and TV inspection technology for secondary-side in steam generators of Daya Bay Nuclear Power Plant has been used for preventing the heat transfer tubes from damage caused by residues in steam generator and foreign objects. A lancing has been used for steam generators. The high-pressure jet sent from the central lane to the inside of the bundle has two objectives: (a) the mechanical energy in the jet breaks up the deposit, and (b) the particles are then carried to the periphery of the generator, where they are collected by a water circulation system. The TV inspection consists of the inspection before lancing and after lancing. The former includes the inspection of outer tube lane and central tube lane, and the extraction of foreign objects; the latter includes the inspection of inter tube area, outer tube lane and central tube lane, and the extraction of foreign objects. Video cassette are visualized by a specialist who is qualified to realize a map representative of the cleanliness state of the tubesheet, and to judge if it is or not acceptable. The TV inspection obtained a cleanliness conclusion according to acceptance criteria. So, it is important to work out cleanliness acceptance criteria suit for every operation stages. The site practices shown that after lancing, although hard deposit existed and height of hard deposit in some local place exceeded 5 mm, but the sludge remaining on tubesheet was less. If we can conduct periodic lancing to steam generator, the increase rate of hard deposit will be small. (7 refs., 12 figs.)

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.

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.

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

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.

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

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

Experimental investigation on the aerosol retention during severe accident SGTR sequences:Influence of the shape of tube break

International Nuclear Information System (INIS)

Delgado Tardaguila, R.; Herranz, L. E.

2011-01-01

This paper presents the experimental integral test, CIEMAT Aerosol Artist Test 2 (CAAT2), with particular emphasis on the effect of the breach shape on aerosol retention in the break stage of a dry steam generator. A set of 5 experiments under different conditions has been conducted with fish-mouth breach. Additionally, comparisons have been set to the previous experimental campaign (CAAT tests).

TLTA/6431, Two-Loop-Test-Apparatus, BWR/6 Simulator, Small-Break LOCA

International Nuclear Information System (INIS)

1992-01-01

1 - Description of test facility: The Two-Loop-Test-Apparatus (TLTA) is a 1:624 volume scaled BWR/6 simulator. It was the predecessor of the better-scaled FIST facility. The facility is capable of full BWR system pressure and has a simulated core with a full size 8 x 8, full power single bundle of indirect electrically heated rods. All major BWR systems are simulated including lower plenum, guide tube, core region (bundle and bypass), upper plenum, steam separator, steam dome, annular downcomer, recirculation loops and ECC injection systems. The fundamental scaling consideration was to achieve real-time response. A number of the scaling compromises present in TLTA were corrected in the FIST configuration. These compromises include a number of regional volumes and component elevations. 2 - Description of test: 64.45 sqcm small break LOCA with activation of the full emergency core cooling system, but without activation of the automatic decompression system

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

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).

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).

PWR steam generators tube integrity: plugging criteria for PWSCC in roll transition zone

International Nuclear Information System (INIS)

Mattar Neto, Miguel; Cruz, Julio R.B.

1999-01-01

One of the most important causes for tube plugging in PWR (Pressurized Water Reactor) steam generators is the degradation mechanism called Primary Water Stress Corrosion Cracking (PWSCC) in roll transition zone (RTZ) near the tubesheet, mainly for Alloy 600 tubes. To avoid an excessive tube plugging, alternative criteria have been developed based on an approach that consists in withdrawing from service any tube containing a defect for which there is a high probability of a critical size under accident conditions to be reached during next operation cycle. Predictions of the number of tubes to be plugged can be done aiming at preventive maintenance and tube repair, and even a steam generator replacement, without a large and non-planned plant outage. This work presents important aspects related to tube plugging criteria for PWSCC in RTZ based on the risk of break after a leak detection. Calculations of allowable crack length and allowable leak rate for a particular situation are also shown. (author)

A 3D Photothermal Structure toward Improved Energy Efficiency in Solar Steam Generation

KAUST Repository

Shi, Yusuf

2018-04-18

Summary The energy efficiency in solar steam generation by 2D photothermal materials has approached its limit. In this work, we fabricated 3D cylindrical cup-shaped structures of mixed metal oxide as solar evaporator, and the 3D structure led to a high energy efficiency close to 100% under one-sun illumination due to the capability of the cup wall to recover the diffuse reflectance and thermal radiation heat loss from the 2D cup bottom. Additional heat was gained from the ambient air when the 3D structure was exposed under one-sun illumination, leading to an extremely high steam generation rate of 2.04 kg m−2 h−1. The 3D structure has a high thermal stability and shows great promise in practical applications including domestic wastewater volume reduction and seawater desalination. The results of this work inspire further research efforts to use 3D photothermal structures to break through the energy efficiency limit of 2D photothermal materials.

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

ATLAS Cold Leg Top Slot Break Analysis using RELAP5

Energy Technology Data Exchange (ETDEWEB)

Sung, Haejung; Lee, Sang Ik; Park, Ju-Hyun; Choi, Tong-Soo [KEPCO NF, Daejeon (Korea, Republic of)

2016-10-15

U.S. Nuclear Regulatory Commission (US-NRC) has been reviewing the design certification application for APR1400 submitted by Korea Electric Power Corporation (KEPCO). The main concern about cold leg top slot break is that cladding temperature might be increased by core uncover due to four loop seal reformation following flooding of safety injection water. An integral effect test for cold leg top slot break was performed by KAERI (Korea Atomic Energy Research Institute) using ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation), which is a scaled down experimental facility for APR1400. In this study, RELAP5/MOD3.3/Patch04 is assessed by experimental result of ATLAS cold leg top slot break. Also, thermal hydraulic phenomena by four loop seals reformation is observed by RELAP5 result. The RELAP5/MOD3.3/Patch04 is assessed by the experimental result of ATLAS cold leg top slot break. The top slot break is described by offtake model, and the mass flow rate is fairly well estimated. The RELAP5 well predicts the correlation between general trend and four loop seal reformation. The pressure of the core region and the cladding temperature tends to increase during four loop seal reformation due to steam path blockage on four loop seals. It is presumed that the code cannot estimate two phase phenomena by loop seal clearing as same as experiments. In terms of cladding temperature, loop seal reformation due to loop seal elevation of APR1400 does not need to be the issue, since the void fraction at the active top core is maintained over 0.4.

The common project for completion of Bubbler Condenser Qualification (Bohunice, Mochovce, Dukovany and Paks NPPs)

International Nuclear Information System (INIS)

Jaroslav, H.; Pavol, B.

2003-01-01

Described is the common project for completion of bubbler condenser qualification for nuclear power plants in Bohunice, Mochovice, Dukovany and Paks. Functionality of the bubbler condenser was elaborated during the simulation of the main steam line brake, medium break and small break LOCA. On this basis the appropriate operation of bubbler condenser containment under accident conditions can be positively confirmed

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.

Simulation of the SPE-4 small-break loss-of-coolant accident

International Nuclear Information System (INIS)

Cebull, P.; Hassan, Y.A.

1993-01-01

A small-break loss of coolant accident (SBLOCA) conducted at the PMK-2 integral test facility was analyzed using RELAP5/MOD3. 1. The experiment simulated a 7.4% break in the cold leg of a VVER-440/213-type nuclear power plant as part of the International Atomic Energy Agency's Fourth Standard Problem Exercise (SPE-4). The VVER design differs from pressurized water reactors (PWRS) of western origin, primarily in its use of horizontal steam generators, hot- and cold-leg loop seals, and safety injection tanks. Because of these differences, it will exhibit somewhat different transient behavior than most PWRS. The PMK-2 test facility, located at the KFKI Atomic Energy Research Institute (AEKI), is a scale model of the Paks nuclear power plant in Hungary with scaling factors of 1:2070 in power and volume and 1:1 in elevation. Primarily used to study SBLOCAs and natural circulation behavior of VVER reactors, it has been used in three previous SPEs

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.

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.

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

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

Evaluating the break flow for the 100% DVI line break accident of ATLAS using the RELAP5/MOD3.3 code

Energy Technology Data Exchange (ETDEWEB)

Lee, Suk Ho; You, Sung Chang; Kim, Han Gon [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

2010-10-15

An integral effect test database for major design basis accidents using the Advanced Test Loop for Accident Simulation (ATLAS) facility has been compiled by the Korea Atomic Energy Research Institute (KAERI). In order to effectively utilize the database, the Domestic Standard Problem (DSP) exercise was proposed and launched in 2009. As the first DSP exercise, scenario involving a 100% break of the DVI nozzle was determined by considering its technical importance including such phenomena as the break flow, loop seal clearing. The first DSP exercise was performed in an open calculation environment. Thus, integral effect test data were opened to the participants prior to code calculations. Ten domestic organizations including members of nuclear industry, a research institute, and universities participated in the DSP exercise using various best-estimate safety analysis codes and finally presented their code prediction results, comparing them to the experimental data. This paper presents the analysis results performed by NETEC as one of the first DSP exercise participants. This analysis focuses on the break flow phenomena and modeling

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...

Large eddy simulation of breaking waves

DEFF Research Database (Denmark)

Christensen, Erik Damgaard; Deigaard, Rolf

2001-01-01

A numerical model is used to simulate wave breaking, the large scale water motions and turbulence induced by the breaking process. The model consists of a free surface model using the surface markers method combined with a three-dimensional model that solves the flow equations. The turbulence....... The incoming waves are specified by a flux boundary condition. The waves are approaching in the shore-normal direction and are breaking on a plane, constant slope beach. The first few wave periods are simulated by a two-dimensional model in the vertical plane normal to the beach line. The model describes...... the steepening and the overturning of the wave. At a given instant, the model domain is extended to three dimensions, and the two-dimensional flow field develops spontaneously three-dimensional flow features with turbulent eddies. After a few wave periods, stationary (periodic) conditions are achieved...

Thermodynamic evaluation of hydrogen production via bioethanol steam reforming

Energy Technology Data Exchange (ETDEWEB)

Tasnadi-Asztalos, Zsolt; Cormos, Ana-Maria; Imre-Lucaci, Árpád; Cormos, Călin C. [Babes-Bolyai University, Faculty of Chemistry and Chemical Engineering, Arany Janos 11, RO-400028, Cluj-Napoca (Romania)

2013-11-13

In this article, a thermodynamic analysis for bioethanol steam reforming for hydrogen production is presented. Bioethanol is a newly proposed renewable energy carrier mainly produced from biomass fermentation. Reforming of bioethanol provides a promising method for hydrogen production from renewable resources. Steam reforming of ethanol (SRE) takes place under the action of a metal catalyst capable of breaking C-C bonds into smaller molecules. A large domain for the water/bioethanol molar ratio as well as the temperature and average pressure has been used in the present work. The interval of investigated temperature was 100-800°C, the pressure was in the range of 1-10 bar and the molar ratio was between 3-25. The variations of gaseous species concentration e.g. H{sub 2}, CO, CO{sub 2}, CH{sub 4} were analyzed. The concentrations of the main products (H{sub 2} and CO) at lower temperature are smaller than the ones at higher temperature due to by-products formation (methane, carbon dioxide, acetylene etc.). The concentration of H2 obtained in the process using high molar ratio (>20) is higher than the one at small molar ratio (near stoichiometric). When the pressure is increased the hydrogen concentration decreases. The results were compared with literature data for validation purposes.

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)

Application of a generalized interface module to the coupling of PARCS with both RELAP5 and TRAC-M

Energy Technology Data Exchange (ETDEWEB)

Barber, D.A.; Wang, W. [SCIENTECH, Inc. (United States); Miller, R.M.; Downar, T.J. [Purdue Univ., West Lafayette, IN (United States); Joo, H.G. [Korean Atomic Energy Research Inst., Seoul (Korea, Republic of); Mousseau, V.A. [Los Alamos National Lab., NM (United States); Ebert, D.E. [Nuclear Regulatory Commission, Washington, DC (United States)

1999-04-01

In an effort to more easily assess various combinations of 3-D neutronic/thermal-hydraulic codes, the USNRC has sponsored the development of a generalized interface module for the coupling of any thermal-hydraulics code to any spatial kinetics code. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the Parallel Virtual Machine (PVM) software to manage inter-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCS, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for an OECD/NEA main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated; nonetheless, the capabilities of the coupled code are presented for the OECD/NEA main steam line break benchmark problem.

Application of a generalized interface module to the coupling of PARCS with both RELAPS and TRAC-M

International Nuclear Information System (INIS)

Barber, D.A.; Wang, W.; Miller, R.M.; Downar, T.J.; Joo, H.G.; Mousseau, V.A.; Ebert, D.E.

1999-01-01

In an effort to more easily assess various combinations of 3-D neutronic/thermal-hydraulic codes, the USNRC has sponsored the development of a generalized interface module for the coupling of any thermal-hydraulics code to any spatial kinetics code. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the Parallel Virtual Machine (PVM) software to manage inter-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCS, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for an OECD/NEA main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated; nonetheless, the capabilities of the coupled code are presented for the OECD/NEA main steam line break benchmark problem

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.

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.

Leak-before-break behaviour of nuclear piping systems

International Nuclear Information System (INIS)

Bartholome, G.; Wellein, R.

1992-01-01

The general concept for break preclusion of nuclear piping systems in the FRG consists of two main prerequisites: Basic safety; independent redundancies. The leak-before-break behaviour is open of these redundancies and will be verified by fracture mechanics. The following items have to be evaluated: The growth of detected and postulated defects must be negligible in one life time of the plant; the growth behaviour beyond design (i.e. multiple load collectives are taken into account) leads to a stable leak; This leakage of the piping must be detected by an adequate leak detection system long before the critical defect size is reached. The fracture mechanics calculations concerning growth and instability of the relevant defects and corresponding leakage areas are described in more detail. The leak-before-break behaviour is shown for two examples of nuclear piping systems in pressurized water reactors: main coolant line of SIEMENS-PWR 1300 MW (ferritic material, diameter 800 mm); surge line of Russian WWER 440 (austenitic material, diameter 250 mm). The main results are given taking into account the relevant leak detection possibilities. (author). 9 refs, 9 figs

ROSA-III 100 % break integral test Run 914

International Nuclear Information System (INIS)

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

1987-05-01

This report presents the experimental data of RUN 914 conducted at the ROSA-III test facility. The facility is a volumetrically scaled (1/424) simulator for a BWR/6 with the electrically heated core, the break simulator and the scaled ECCS (emergency core cooling system). RUN 914 was a 100% split break test at the recirculation pump suction line with an assumption of HPCS diesel generator failure and conducted as one of the break area parameter tests. A peak cladding temperature (PCT) of 851 K was reached at 130 s after the break during the reflooding phase. Whole core was completely quenched by ECCS, and the effectiveness of ECCS was confirmed. The primary test results of RUN 914 are compared in this report with those of RUN 926, which was a 200 % double-ended break test. The initiation of core dryout in RUN 914 was almost the same as that in RUN 926. Duration of core dryourt was, however, longer in RUN 914 because of later actuation of ECCSs. PCT in RUN 914 was 67 K higher than that in RUN 926. (author)

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

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)

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

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)

Core design with respect to the safety concept

International Nuclear Information System (INIS)

Kollmar, W.

1981-01-01

In the present paper the following topics are dealt with: Principles of reactor core design and optimization, fuel management and safety concept for higher cycles and results of risk analyses (e.g. rod ejection, steam line break etc.) (RW)

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)

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.)

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

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

Application of probabilistic fracture mechanics to optimize the maintenance of PWR steam generator tubes

International Nuclear Information System (INIS)

Pitner, P.; Riffard, T.

1993-09-01

This paper describes the COMPROMIS code developed by Electricite de France (EDF) to optimize the tube bundle maintenance of steam generators (SG). The model, based on probabilistic fracture mechanics, makes it possible to quantify the influence of in-service inspections and maintenance work on the risk of an SG tube rupture, taking all significant parameters into account as random variables (initial defect size distribution, reliability of nondestructive detection and sizing, crack initiation and propagation, critical sizes, leak before risk of break, etc). (authors). 14 figs., 4 tabs., 12 refs

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…

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

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.

Polymeric dispersants for control of steam generator fouling

International Nuclear Information System (INIS)

Balakrishnan, P.V.; Klimas, S.J.; Lepine, L.; Turner, C.W.

1999-05-01

Fouling of steam generators by corrosion products from the feedtrain leads to loss of heat-transfer efficiency, disturbances in thermalhydraulics, and potential corrosion problems resulting from the development of sites for localized accumulation of aggressive chemicals. This report summarizes studies of the use of polymeric dispersants for the control of fouling, which were conducted at the Chalk River Laboratories. High-temperature settling studies on magnetite suspensions were performed to screen available generic dispersants, and the dispersants were ranked in terms of their dispersion efficiency; polyacrylic acid (PAA) and the phosphonate, HEDP, were ranked as the most efficient. Polyacrylic acid was considered more suitable than HEDP for nuclear steam generators, and more emphasis was given to the former in these studies. The dispersants had no effect on the particle deposition rates under single-phase forced-convective flow, but did reduce the deposition rates under flow-boiling conditions. The extent to which the deposition rates were reduced increased in proportion to the dispersant concentration. Preliminary corrosion tests indicated that pitting or general corrosion of steam generator tube materials in the presence of PAA was negligible. Corrosion of carbon steel, although higher in a magnetite-packed crevice under heat flux than in bulk water, was lower in the presence of PAA than in its absence. Some impurities (e.g., sulphate, sodium) were observed in commercially available PAA products at small, though significant concentrations, making these products unacceptable for use in nuclear plants. However, the PAA could be purified by ion exchange. Preliminary experiments, to assess the thermal stability of PAA at steam generator operating temperature, showed the polymer to break down in deaerated solutions and under argon cover to give hydrogen and carbon dioxide as the two major products in the gas phase and variable concentrations of acetate and formate

Evaluation of tube rupture simulation test (TRUST-1) for FBR steam generators

Energy Technology Data Exchange (ETDEWEB)

Hayashida, Yoshihiko; Hamada, Hirotsugu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1996-06-01

The intermediate water leak in an FBR Steam Generator (SG) causes a high temperature and corrosive sodium-water reaction jet. In such cases, it is necessary to evaluate the wastage and overheating rupture behavior of heat transfer tubes. Especially, in the large SG that aims at high temperature of sodium and high temperature/pressure of water, the establishment of the rational evaluation method is important. In this paper, as a basic experiment to make clear the phenomenon of overheating rupture, tests and analysis of Tube Rupture Simulation Test-1 (TRUST-1) were conducted. TRUST-1 simulates the overheating rupture of the tube made of Mod.9Cr-1Mo steel by nitrogen gas pressurization and quick induction heating. The result of TRUST-1 are as follows: (1) The breaking strength predicted by the internal pressure is larger than the tensile strength of the tube material. (2) The margin of the breaking strength from the tensile strength of the tube material has a tendency of decreasing with the heating rate, especially in the lower temperature region. (3) Using an theoretical formula that is deduced from the steady creep model and appropriate experimental coefficients that are determined by the test data, the breaking strength can be reasonably evaluated. (author)

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)

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

Full-scale mark II CRT program data report no. 7

International Nuclear Information System (INIS)

Kukita, Yutaka; Namatame, Ken; Yamamoto, Nobuo; Takeshita, Isao; Shiba, Masayoshi

1980-03-01

The Full-Scale Mark II CRT (Containment Response Test) Program was initiated in 1977 to provide a data base for evaluation of the LOCA hydrodynamic loads in the BWR Mark II pressure suppression system. The test facility is 1/18 in volume and has a wetwell which is a full-scale replica of one 20 0 -sector of that of a reference Mark II. This report documents test data obtained from TEST 3102, which is a large (200 mm) water break test performed on June 29, 1979. The test was performed for a supernominal break area (approx. 160% of the scaled break area for a postulated double-ended break in the recirculation line) to obtain vent steam mass fluxes ranging up to the maximum design value of the actual plants. Before the initiation of the test more than 90% of the drywell air was replaced by steam and transferred into wetwell airspace to reduce the air content in the vent flow during the test. Because of this pre-test treatment (so-called prepurging) the test data obtained for high- and intermediate-steam-flux condensation oscillations are believed to be to a high degree on conservative side. (author)

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

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

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)

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

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.

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...

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

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

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.)

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

Five Tubes Rupture at Cold Side of Steam Generator Simulation Test Report Using the ATLAS

International Nuclear Information System (INIS)

Kang, Kyoung Ho; Park, Hyun Sik; Cho, Seok

2010-12-01

In this study, a postulated SGTR event of the APR1400 was experimentally investigated with the ATLAS. In order to simulate a double-ended rupture of five U-tubes in the APR1400, the SGTR-CL-02 test was performed with the ATLAS. The main objectives of this test were not only to provide a physical insight into the system response of the APR1400 during the SGTR but also to produce integral effect experimental data to validate the safety analysis code. In the present report, major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Compared to the case of a single U-tube rupture test, opening frequency of the MSSVs in the intact steam generator (SG-2) was highly reduced after 500 seconds in the present SGTR-CL-02 test. Large discharge of the primary inventory resulted in rapid depressurization of the primary system and consequently early injection of the SIP. Supply of cold ECC water by the SIPs reduced the energy transfer to the secondary side compared with the single U-tube rupture case. Less heat transfer to the secondary side had more influence on the secondary pressure of the affected steam generator than the break flow. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code

Five Tubes Rupture at Hot Side of Steam Generator Simulation Test Report Using the ATLAS

International Nuclear Information System (INIS)

Kang, Kyoung Ho; Park, Hyun Sik; Cho Seok

2010-12-01

In this study, a postulated SGTR event of the APR1400 was experimentally investigated with the ATLAS. In order to simulate a double-ended rupture of five U-tubes in the APR1400, the SGTR-HL-05 test was performed with the ATLAS. The main objectives of this test were not only to provide a physical insight into the system response of the APR1400 during the SGTR but also to produce integral effect experimental data to validate the SPACE code. In the present report, major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. On the contrary to the case of a single U-tube rupture test, the MSSV of the intact steam generator was not opened any more after 1500 seconds in the present SGTR-HL-05 test. Large discharge of the primary inventory resulted in rapid depressurization of the primary system and consequently early injection of the SIP. Supply of cold ECC water by the SIPs reduced the energy transfer to the secondary side compared with the single U-tube rupture case. Less heat transfer to the secondary side had more influence on the secondary pressure of the affected steam generator than the break flow. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code

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.

Elevated Levels of DNA Strand Breaks Induced by a Base Analog in the Human Cell Line with the P32T ITPA Variant

Science.gov (United States)

Waisertreiger, Irina S.-R.; Menezes, Miriam R.; Randazzo, James; Pavlov, Youri I.

2010-01-01

Base analogs are powerful antimetabolites and dangerous mutagens generated endogenously by oxidative stress, inflammation, and aberrant nucleotide biosynthesis. Human inosine triphosphate pyrophosphatase (ITPA) hydrolyzes triphosphates of noncanonical purine bases (i.e., ITP, dITP, XTP, dXTP, or their mimic: 6-hydroxyaminopurine (HAP) deoxynucleoside triphosphate) and thus regulates nucleotide pools and protects cells from DNA damage. We demonstrate that the model purine base analog HAP induces DNA breaks in human cells and leads to elevation of levels of ITPA. A human polymorphic allele of the ITPA, 94C->A encodes for the enzyme with a P32T amino-acid change and leads to accumulation of nonhydrolyzed ITP. The polymorphism has been associated with adverse reaction to purine base-analog drugs. The level of both spontaneous and HAP-induced DNA breaks is elevated in the cell line with the ITPA P32T variant. The results suggested that human ITPA plays a pivotal role in the protection of DNA from noncanonical purine base analogs. PMID:20936128

Elevated Levels of DNA Strand Breaks Induced by a Base Analog in the Human Cell Line with the P32T ITPA Variant

Directory of Open Access Journals (Sweden)

Irina S.-R. Waisertreiger

2010-01-01

Full Text Available Base analogs are powerful antimetabolites and dangerous mutagens generated endogenously by oxidative stress, inflammation, and aberrant nucleotide biosynthesis. Human inosine triphosphate pyrophosphatase (ITPA hydrolyzes triphosphates of noncanonical purine bases (i.e., ITP, dITP, XTP, dXTP, or their mimic: 6-hydroxyaminopurine (HAP deoxynucleoside triphosphate and thus regulates nucleotide pools and protects cells from DNA damage. We demonstrate that the model purine base analog HAP induces DNA breaks in human cells and leads to elevation of levels of ITPA. A human polymorphic allele of the ITPA, 94C->A encodes for the enzyme with a P32T amino-acid change and leads to accumulation of nonhydrolyzed ITP. The polymorphism has been associated with adverse reaction to purine base-analog drugs. The level of both spontaneous and HAP-induced DNA breaks is elevated in the cell line with the ITPA P32T variant. The results suggested that human ITPA plays a pivotal role in the protection of DNA from noncanonical purine base analogs.

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

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.

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

Reactive Pad-Steam Dyeing of Cotton Fabric Modified with Cationic P(St-BA-VBT Nanospheres

Directory of Open Access Journals (Sweden)

Kuanjun Fang

2018-05-01

Full Text Available The Poly[Styrene-Butyl acrylate-(P-vinylbenzyl trimethyl ammonium chloride] P(St-BA-VBT nanospheres with N+(CH33 functional groups were successfully prepared and applied to modify cotton fabrics using a pad-dry process. The obtained cationic cotton fabrics were dyed with pad-steam dyeing with reactive dye. The results show that the appropriate concentration of nanospheres was 4 g/L. The sodium carbonate of 25 g/L and steaming time of 3 min were suitable for dyeing cationic cotton with 25 g/L of C.I. Reactive Blue 222. The color strength and dye fixation rates of dyed cationic cotton fabrics increased by 39.4% and 14.3% compared with untreated fabrics. Moreover, sodium carbonate and steaming time were reduced by 37.5% and 40%, respectively. The rubbing and washing fastness of dyed fabrics were equal or higher 3 and 4–5 grades, respectively. Scanning electron microscopy (SEM images revealed that the P(St-BA-VBT nanospheres randomly distributed and did not form a continuous film on the cationic cotton fiber surfaces. The X-ray photoelectron spectroscopy (XPS analysis further demonstrated the presence of cationic nanospheres on the fiber surfaces. The cationic modification did not affect the breaking strength of cotton fabrics.

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)

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.)

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

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…

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

Construction of reactor and outline of breaking state

International Nuclear Information System (INIS)

Imanaka, Tetsuji

1980-01-01

The Mihama No. 1 reactor of Kansai Electric Power Co., Inc., is the first power-generating PWR in Japan, and it commenced the commercial operation in November, 1970. In June, 1972, the leak of steam generator tubes occurred, and the reactor was stopped for about a half year. In the second periodic inspection in March, 1973, the accident of broken fuel rods was discovered, but it was generally published in December, 1976. In August, 1974, the leak of steam generator tubes occurred again, and since then, the reactor was stopped for a long period. From October, 1978, the reactor has entered the test operation called cycling operation. The Kyoto University Reactor Research Institute lent the cask for transporting the broken pieces of the fuel rods to Kansai Electric Power Co., and obtained the investigation report of Japan Atomic Energy Research Institute, Based on the contract. The accident investigation group of the KURRI has pursued the causes of the accident and examined the propriety of the countermeasures. The outline of the construction of the reactor is described. The upper part of two fuel rods broke and fell on the baffle supporting plate. The broken fuel assembly was named C-34. About a half of fuel pellets and the cladding tube of several tens cm have not yet recovered. The state of break and the presumption of the causes are reported. (Kako, I.)

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.

ROSA-III 50 % break integral test RUN 916

International Nuclear Information System (INIS)

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

1985-08-01

This report presents the experimental data of RUN 916 conducted at the ROSA-III test facility. The facility is a volumetrically scaled (1/424) simulator for a BWR/6 with the electrically heated core, the break simulator and the scaled ECCS(emergency core cooling system). RUN 916 was a 50 % split break test at the recirculation pump suction line with an assumption of HPCS diegel generator failure and conducted as one of the break area parameter tests. A peak cladding temperature (PCT) of 917 K was reached at 190 s after the break during the reflooding phase. Whole core was completely quenched by ECCS, and the effectiveness of ECCS was confermed. The primary test results of RUN 916 are compared in this report with those of RUN 926, which was a 200 % double-ended break test. The initiation of core dryout in RUN 916 was later than that in RUN 926 because of the smaller discharge flow rate. Duration of core dryourt was, however, longer in RUN 916 because of later actuation of ECCSs. PCT in RUN 916 was 133 K higher than that in RUN 926. (author)

Track 3: growth of nuclear technology and research 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-I. 5. Analyses of the OECD MSLB Benchmark with the Codes DYN3D and DYN3D/ATHLET