Design and Test of a Small Turbine-Compressor Set for a Nuclear Application

1962 ◽  
Author(s):  
D. H. S. Burton ◽  
T. E. H. Beck
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Mechanical Design ◽  
Aerodynamic Performance ◽  
Test Facility ◽  
Design Problems ◽  
Open Cycle

This paper describes the design of a turbine-compressor set in the 400–500 kw power range for use in the ML-1 mobile nuclear power plant. The specification and design problems are discussed in terms of mechanical design and aerodynamic performance. The open-cycle test facility is also described and the results of open-cycle testing are surveyed.

Onset of Gas Blowthrough During Melt Expulsion From a Hole in a RPV: Numerical Analysis Using the SIMMER Code

2006 ◽  
Author(s):  
Frank Kretzschmar
Keyword(s):  
Numerical Analysis ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Severe Accident ◽  
Core Material ◽  
Test Facility ◽  
Computer Codes ◽  
Direct Containment Heating ◽  
Heating Up

In the case of a severe accident in a nuclear power plant there is a residual risk, that the Reactor Pressure Vessel (RPV) does not withstand the thermal attack of the molten core material, of which the temperature can be about 3000 K. For the analysis of the processes governing melt dispersal and heating up of the containment atmosphere of a nuclear power plant in the case of such an event, it is important to know the time of the onset of gas blowthrough during the melt expulsion through the hole in the bottom of the RPV. In the test facility DISCO-C (Dispersion of Simulant Corium-Cold) at the FZK /6/, experiments were performed to furnish data for modeling Direct Containment Heating (DCH) processes in computer codes that will be used to extrapolate these results to the reactor case. DISCO-C models the RPV, the Reactor Coolant System (RCS), cavity and the annular subcompartments of a large European reactor in a scale 1:18. The liquid type, the initial liquid mass, the type of the driving gas and the size of the hole were varied in these experiments. We present results for the onset of the gas blowthrough that were reached by numerical analysis with the Multiphase-Code SIMMER. We compare the results with the experimental results from the DISCO-C experiments and with analytical correlations, given by other authors.

Experimental Investigation of Subcooled Choking Flow in a Steam Generator Tube Crack

2016 ◽  
Author(s):  
Hung Nguyen ◽  
Mark Brown ◽  
Shripad T. Revankar ◽  
Jovica Riznic
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Test Facility ◽  
Experimental Program ◽  
Steam Generators ◽  
Steam Generator Tube ◽  
Loss Of Coolant ◽  
Steam Generator Tubes

Steam generator tubes have a history of small cracks and even ruptures, which lead to a loss of coolant from the primary side to the secondary side. These tubes have an important role in reactor safety since they serve as one of the barriers between radioactive and non-radioactive materials of a nuclear power plant. A rupture then signifies the loss of the integrity of the tube itself. Therefore, choking flow plays an integral part not only in the engineered safeguards of a nuclear power plant, but also to everyday operation. There is limited data on actual steam generators tube wall cracks. Here experiments were conducted on choked flow of subcooled water through two samples of axial cracks of steam generator tubes taken from US PWR steam generators. The purpose of the experimental program was to develop database on critical flow through actual steam generator tube cracks with subcooled liquid flow at the entrance. The knowledge of this maximum flow rate through a crack in the steam generator tubes of a pressurized water nuclear reactor will allow designers to calculate leak rates and design inventory levels accordingly while limiting losses during loss of coolant accidents. The test facility design is modular so that various steam generator tube cracks can be studied. Two sets of PWR steam generators tubes were studied whose wall thickness is 1.285 mm. Tests were carried out at stagnation pressure up to 6.89 MPa and range of subcoolings 16.2–59°C. Based on these new choking flow data, the applicability of analytical models to highlight the importance of non-equilibrium effects was examined.

Survey and Assessment of Nuclear Power Plant Control Room Operability

1980 ◽  
Vol 24 (1) ◽  
pp. 267-270
Author(s):  
Clifford C. Baker ◽  
Robin West ◽  
Kenneth M. Mallory
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Data Collection ◽  
Nuclear Power ◽  
Human Engineering ◽  
Control Room ◽  
Design Problems ◽  
Power Plant Control ◽  
Control Rooms ◽  
Plant Control

As part of an effort to develop human engineerging guidelines and a methodology for the evaluation of nuclear power plant control room operability, the Essex Corporation conducted T & E (test and evaluation) reviews of a wide sample of nuclear power plant control rooms. The objectives of these design reviews were: 1) selection, application, and development of human engineering evaluation guidelines applicable to the nuclear power industry; 2) selection and development of data collection and analysis procedures; and 3) identification of recurrent human engineering design problems in the control rooms of currently operating nuclear power plants. The present paper discusses the approach taken and the findings in item three above. Thirteen control rooms were visited, and guidelines and data collection methods under various degrees of development were applied. Following control room visits, data were analyzed according to usability, number of incidences of similar or identical operability design problems, criticality of problems with respect to both public and plant safety, and subjective assessment of operational affects due to human engineering problems in design. Results to date show that the following areas have recurrent operability design problems: layout of controls and displays according to either operational or functional use; coding of information for visual and auditory presentation; job performance aid and procedures design; communications; environmental factors such as ambient noise; violations in control and display conventions employed; use of conventions which violate population stereotypes; and failure to design within anthropometric constraints. Further work is being conducted by Essex Corporation to identify critical human engineering deficiencies in control room design and to select adequate yet cost-effective and corrective backfits.

Test Facility Design for Integrated Digital Nuclear Reactor Protection System

2010 ◽  
Author(s):  
Huasheng Xiong ◽  
Duo Li ◽  
Liangju Zhang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Verification And Validation ◽  
Protection System ◽  
Normal Operation ◽  
Test Facility ◽  
Acceptance Test ◽  
Independent Verification ◽  
Reactor Protection System

Reactor protection system is one of the most important safety systems in nuclear power plant and shall be designed with very high reliability. Digital computer-based Reactor Protection System (RPS) takes great advantages over its conventional counterpart based on analog technique and faces the issues how to effectively demonstrate and confirm the completeness and correctness of the software that performs reactor safety functions in the same time. It is commonly accepted that the essential way to solve safety software issues in a digital RPS is to pass a strict and independent Verification and Validation (V&V) process, in which integrated RPS testing play an important role to form a part of the overall system validation. Integrated RPS testing must be carried out rigorously before the system is delivered to nuclear power plant. The integrated testing are often combined with the factory acceptance test (FAT) to form a single testing activity, during which the RPS is excited by emulated static and dynamic input signals. The integration testing should simulate normal operation, anticipated operational occurrences and accident conditions, as well as anticipated faults on the inputs to the DRPS such as sensors out of range or ambiguous input readings. All safety function requirements of digital RPS should be confirmed by representative testing. The design and development of a test facility to carry out the integrated RPS testing are covered in this paper, which is merged in the research on a digital RPS engineering prototype for a nuclear power plant. The test facility is based on PXI platform and LabVIEW software development environment and its architecture design also takes into account the test functions future extensions such as hardware upgrades and software modules enhancement. The test facility provides the digital RPS with redundant, synchronized and multi-channel emulated signals that are produced to emulate all protection signals from 1E class sensors and transmitters with time varied value within their possible ranges, which would put integrated RPS testing into practice to confirm the digital RPS has fully met its predefined safety functionality requirements. The designed test facility can provide an independent verification and validation process for the research of digital RPS with scientific methods and authentic data to evaluate the RPS performance thoroughly and effectively, such as measuring threshold precision and trip response time, analyzing system statistical reliability and so on.

RCIC Turbo-Pump Scaling Through CFD and Model Testing for the Texas A&M University NHTS Facility

2018 ◽  
Author(s):  
Mohammad A. Hawila ◽  
Karen Vierow Kirkland
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Test Facility ◽  
Scaling Analysis ◽  
Cfd Analysis ◽  
Full Size ◽  
Scaling Parameters ◽  
Station Blackout ◽  
The U.S

One of the requirements for licensing a nuclear power plant in the U.S is the capability to survive and recover from a station blackout according to the U.S Nuclear Regulatory Commission (USNRC). Station blackout is the loss of all off-site and onsite power simultaneously. Therefore, experimental test facilities are being constructed and operated to test the performance of the related safety systems in a nuclear power plant. Design and construction of a test facility creates the need to perform scaling analysis to ensure proper representation of key components and phenomena of interest. One of the main outcomes of the scaling analysis is the quantitative estimation of the Similarity Level (SL), which requires derivation of dimensionless scaling parameters and prediction of appropriate input values for the scaling parameters. To study the performance of the Reactor Core Isolation Cooling (RCIC) system, the Nuclear Heat Transfer Systems (NHTS) Laboratory at Texas A&M University has constructed and is operating a RCIC test facility. This paper presents the scaling analysis with reference to a full-size RCIC system and the RCIC system turbine was used as the main component for scaling. The input parameters for dimensionless scaling parameters were obtained through experimental measurements and CFD analysis. The CFD analysis is for the ZS-1 RCIC system turbine model. The STAR-CCM+ CFD code was used in this study to create and run simulations for steady state normal and abnormal operating conditions for the NHTS-developed CAD models. The input for the dimensionless scaling parameters was estimated. Input parameters were collected both experimentally and from CFD simulations and inserted into these equations. As a result, a high degree of similarity was confirmed, with a minimum of 82% between the NHTS and full-size RCIC systems. The 82% represents the amount of transfer properties conserved between the two systems. Consequently, this high similarity level allows the NHTS RCIC system to be used to study the behavior of the full-size RCIC system under Beyond Design Bases Accident (BDBA) conditions. Future work is to study and model other components of the RCIC system such as the suppression chamber to estimate similarity levels and study their effects on behavior of the system under BDBA.

MARS-KS Code Analysis of the Pressure Wave Propagation Test 0 Performed at the PMK-2 Test Facility

2014 ◽  
Author(s):  
Xin-Guo Yu ◽  
Ki-Yong Choi ◽  
Chul-Hwa Song ◽  
Istvan Trosztel ◽  
Ivan Toth ◽  
...  
Keyword(s):  
Sensitivity Analysis ◽  
Wave Propagation ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Wave ◽  
Nuclear Power ◽  
Initial Conditions ◽  
Test Facility ◽  
Plant Safety ◽  
Nuclear Power Plant Safety

The pressure waves might be expected in the nuclear reactor systems due to sudden rupture of a pipe, quick opening or closure of a system valve. If generated, they can result in large mechanical loads on the RPV internal structures and pipelines, threating their integrity. This kind of phenomena is an important issue and a limiting accident case for the nuclear power plant safety, which requires extensive analysis to ensure the nuclear power plant safety. To study these phenomena, four PWP (Pressure Wave Propagation) tests have been performed in the PMK-2 test facility in MTA EK. In addition, these tests have been used to assess the capability of the MARS-KS code in simulating the PWP phenomena. Then, an input model representing the PMK-2 test facility was developed to simulate the tests. The MARS-KS simulation results are then compared with the test results. The comparison shows that the MARS code can well simulate the PWP frequencies and the initial pressure peaks as well. After the qualified assessment, the MARS-KS code is then deployed to conduct the sensitivity analysis on the effect of the break size, break time, coolant initial conditions on the PWP phenomena. The sensitivity analysis on the break sizes shows that the pressure wave amplitude is relevant to the break times: the shorter the break opening time is, the faster the pressure. The sensitivity analysis on the break sizes shows that the larger the break size is, the higher the pressure peak is.

Design and analysis of a thermal hydraulic integral test facility for Bushehr nuclear power plant

2009 ◽  
Vol 51 (3) ◽  
pp. 443-450 ◽  
Author(s):  
Toraj Khoshnevis ◽  
Jalil Jafari ◽  
Mostafa Sohrabpour
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Test Facility ◽  
Integral Test

Scaling Application of a Hot Leg SBLOCA Scenario to a Nuclear Power Plant

2014 ◽  
Author(s):  
Andrea Querol ◽  
Sergio Gallardo ◽  
Gumersindo Verdú
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Large Scale ◽  
Injection System ◽  
Test Facility ◽  
Experimental Conditions ◽  
Integral Test ◽  
Pressurized Water ◽  
Choked Flow

Several experimental facilities, such as the Large Scale Test Facility (LSTF) of the Japan Atomic Energy Agency (JAEA), have been built to reproduce some accidental scenarios because full-scale testing is usually impossible to perform. One of the objectives of these Integral Test Facilities (ITFs) is to obtain measured data to be compared to simulations in order to test the capability of the thermalhydraulic codes to reproduce experimental conditions. The applicability of these experimental results to a full-size power plant system depends on the scaling criteria adopted. The present paper is focused on the simulation and the scaling of the Test 1-2 in the frame of the OECD/NEA ROSA Project to a Nuclear Power Plant (NPP). This test simulates a hot leg 1% Small Break Loss-Of-Coolant Accident (SBLOCA) in a Pressurized Water Reactor (PWR) under the actuation of High Pressure Injection (HPI) system and Accumulator Injection System (AIS). A scaled-up NPP TRACE5 input has been developed from a LSTF TRACE5 model validated by authors in previous works. The scaled-up model has been developed conserving the power-to-volume scaling ratios of LSTF components, initial and boundary conditions. Lengths and diameters of hot legs have been scaled from LSTF model trying to conserve Froude number. A comparison between both TRACE5 models (LSTF and scaled-up NPP) is performed (system pressures, discharged inventory and collapsed liquid levels). Special TRACE5 models such as Choked flow model and OFFTAKE model have been tested. A 3D VESSEL component has been tested in comparison to 1D TEE component to simulate the hot leg where the SBLOCA is located and varying the break orientation (downwards and upwards). Finally, a sensitivity analysis has been made to determine the effect of the break size in the SBLOCA range.

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