CFD Predictions of Severe Accident Steam Generator Flows in a 1/7th Scale Pressurized Water Reactor

2002 ◽  
Author(s):  
Christopher Boyd ◽  
Kelly Hardesty
Keyword(s):  
Steam Generator ◽  
Natural Circulation ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Steam Generator Tube ◽  
Flow Phenomena ◽  
Experimental Values ◽  
High Degree ◽  
Side Flow

Computational Fluid Dynamics (CFD) is applied to steam generator inlet plenum mixing as part of a larger plan covering steam generator tube integrity. The technique is verified by comparing predicted results with severe accident natural circulation data [1] from a 1/7th scale Westinghouse facility. This exercise demonstrates that the technique can predict the natural circulation and mixing phenomena relevant to steam generator tube integrity issues. The model includes primary side flow paths for a single hot leg and steam generator. Qualitatively, the experimentally observed flow phenomena are predicted. The paths of the natural circulation flows and the relative flow proportions are correctly predicted. Quantitatively, comparisons are made with temperatures, mass flows, and other parameters. All predictions are generally within 10% of the experimental values. Overall, there is a high degree of confidence in the CFD technique for prediction of the relevant flow phenomena associated with this type of severe accident sequence.

Assessment of RELAP/SCDAPSIM for Turbine Trip Transient in NuScale-SMR

2018 ◽  
Author(s):  
Katarzyna Skolik ◽  
Anuj Trivedi ◽  
Marina Perez-Ferragut ◽  
Chris Allison
Keyword(s):  
Natural Circulation ◽  
System Modeling ◽  
Thermal Power ◽  
Reactor Core ◽  
Operating Conditions ◽  
Severe Accident ◽  
Circulation System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Secondary Steam

The NuScale Small Modular Reactor (SMR) is an integrated Pressurized Water Reactor (iPWR) with the coolant flow based on the natural circulation. The reactor core consists of 37 fuel assemblies similar to those used in typical PWRs, but only half of their length to generate 160MW thermal power (50 MWe). Current study involves the development of a NuScale-SMR model based on its Design Certification Application (DCA) data (from NRC) using RELAP/SCDAPSIM. The turbine trip transient (TTT) was simulated and analysed. The objective was to assess this version of the code for natural circulation system modeling capabilities and also to verify the input model against the publicly available TTT results obtained using NRELAP5. This successful benchmark confirms the reliability of the thermal hydraulic model and allows authors to use it for further safety and severe accident analyses. The reactor core channels, pressurizer, riser and downcomer pipes as well as the secondary steam generator tubes and the containment were modeled with RELAP5 components. SCDAP core and control components were used for the fuel elements in the core. The final input deck achieved the steady state with the operating conditions comparable to those reported in the DCA. RELAP/SCDAPSIM predictions are found to be satisfactory and comparable to the reference study. It confirms the code code capabilities for natural circulation system transients.

Natural Circulation and Creep Damage Analyses of Severe Accident for a Typical PWR Plant

2009 ◽  
Author(s):  
Osamu Kawabata ◽  
Masao Ogino
Keyword(s):  
Steam Generator ◽  
Natural Circulation ◽  
Reactor Core ◽  
Creep Damage ◽  
Reactor Vessel ◽  
Reactor System ◽  
Severe Accident ◽  
Steam Generator Tube ◽  
Hot Gas ◽  
Core Meltdown

When the primary reactor system remain pressurized during core meltdown for a typical PWR plant, loop seals formed in the primary reactor system would lead to natural circulations in hot leg and steam generator. In this case, the hot gas released from the reactor core moves to a steam generator, and a steam generator tube would be failed with cumulative creep damage. From such phenomena, a high-pressure scenario during core meltdown may lead to large release of fission products to the environment. In the present study, natural circulation and creep damage in the primary reactor system accompanying the hot gas generation in the reactor core were discussed and the combining analysis with MELCOR and FLUENT codes were performed to examine the natural circulation behavior. For a typical 4 loop PWR plant, MELCOR code which can analyze for the severe accident progression was applied to the accident analyses from accident initiation to reactor vessel failure for the accident sequence of the main steam pipe break which is maintained at high pressure during core meltdown. In addition, using the CFD code FLUENT, fluid dynamics in the reactor vessel plenum, hot leg and steam generator of one loop were simulated with three-dimensional coordinates. And the hot gas natural circulation flow and the heat transfer to adjoining structures were analyzed using results provided by the MELCOR code as boundary conditions. The both ratios of the natural circulation flow calculated in the hot leg and the steam generator using MELCOR code and FLUENT code were obtained to be about 2 (two). And using analytical results of thermal hydraulic analysis with both codes, creep damage analysis at hottest temperature points of steam generator tube and hot leg were carried out. The results in both cases showed that a steam generator tube would be failed with creep rupture earlier than that of hot leg rupture.

Development of Steam Generator Model Using RELAP5 to Simulate Two-Phase Natural Circulation for Pressurized Water Reactor

2014 ◽  
Author(s):  
Shinya Miyata ◽  
Satoru Kamohara ◽  
Wataru Sakuma ◽  
Hiroaki Nishi
Keyword(s):  
Steam Generator ◽  
Natural Circulation ◽  
Injection System ◽  
Pressurized Water Reactor ◽  
Two Phase ◽  
Pressurized Water ◽  
Water Reactor ◽  
Loss Of Coolant ◽  
Mass Inventory ◽  
Core Cooling

In typical pressurized water reactor (PWR), to cope with beyond design basis events such as station black out (SBO) or small break loss of coolant accident with safety injection system failure, injection from accumulator sustains core cooling by compensating for loss of coolant. Core cooling is sustained by single- or two-phase natural circulation or reflux condensation depending on primary coolant mass inventory. Behavior of the natural circulation in PWR has been investigated in the facilities such as Large Scale Test Facility (LSTF) which is a full-height and full-pressure and thermal-hydraulic simulator of typical four-loop PWR. Two steady-state natural circulation tests were conducted in LSTF at both high and low pressure. These two tests were conducted changing the primary mass inventory as a test parameter, while keeping the other parameters such as core power, steam generator (SG) pressure, and steam generator water level as they are. Mitsubishi Heavy Industries (MHI) plans new natural circulation tests to cover wider range of core power and pressure as test-matrix (including the previous LSTF tests) to validate applicability of the model in wider range of core power and pressure conditions including the SBO conditions. In this paper, the previous LSTF natural circulation tests are reviewed and the new test plan will be described. Additionally, MHI also started a feasibility study to improve the steam generator tube and inlet/outlet plenum model using the M-RELAP5 code [4]. Newly developed model gives reasonable agreement with the previous LSTF tests and applies to the new test conditions. The feasibility findings will also be described in this paper.

The PSI Artist Project: Aerosol Retention and Accident Management Issues Following a Steam Generator Tube Rupture

2002 ◽  
Author(s):  
Salih Gu¨ntay ◽  
Abdel Dehbi ◽  
Detlef Suckow ◽  
Jon Birchley
Keyword(s):  
Steam Generator ◽  
Tube Bundle ◽  
Severe Accident ◽  
Pressurized Water ◽  
Steam Generator Tube ◽  
Full Size ◽  
Management Actions ◽  
Accident Management ◽  
Dry Conditions ◽  
Secondary Side

Steam generator tube rupture (SGTR) incidents, such as those, which occurred in various operating pressurized, water reactors in the past, are serious operational concerns and remain among the most risk-dominant events. Although considerable efforts have been spent to understand tube degradation processes, develop improved modes of operation, and take preventative and corrective measures, SGTR incidents cannot be completely ruled out. Under certain conditions, high releases of radionuclides to the environment are possible during design basis accidents (DBA) and severe accidents. The severe accident codes’ models for aerosol retention in the secondary side of a steam generator (SG) have not been assessed against any experimental data, which means that the uncertainties in the source term following an unisolated SGTR concurrent with a severe accident are not currently quantified. The accident management (AM) procedures aim at avoiding or minimizing the release of fission products from the SG. The enhanced retention of activity within the SG defines the effectiveness of the accident management actions for the specific hardware characteristics and accident conditions of concern. A sound database on aerosol retention due to natural processes in the SG is not available, nor is an assessment of the effect of management actions on these processes. Hence, the effectiveness of the AM in SGTR events is not presently known. To help reduce uncertainties relating to SGTR issues, an experimental project, ARTIST (AeRosol Trapping In a Steam generaTor), has been initiated at the Paul Scherrer Institut to address aerosol and droplet retention in the various parts of the SG. The test section is comprised of a scaled-down tube bundle, a full-size separator and a full-size dryer unit. The project will study phenomena at the separate effect and integral levels and address AM issues in seven distinct phases: Aerosol retention in 1) the broken tube under dry secondary side conditions, 2) the near field close to break under dry conditions, 3) the bundle far-field under dry conditions, 4) the separator and dryer under dry conditions, 5) the bundle section under wet conditions, 6) droplet retention in the separator and dryer sections and 7) the overall SG (integral tests). Prototypical test parameters are selected to cover the range of conditions expected in severe accident as well as DBA scenarios. This paper summarizes the relevant issues and introduces the ARTIST facility and the provisional test program which will run between 2003 and 2007.

Research on Flow Characteristics of Once-Through Steam Generator in Rolling Motion

2014 ◽  
Author(s):  
Ren Li ◽  
Min-jun Peng ◽  
Geng-lei Xia ◽  
Yuan-li Sun ◽  
Ya-nan Zhao
Keyword(s):  
Steam Generator ◽  
Natural Circulation ◽  
Flow Characteristics ◽  
Rolling Motion ◽  
Pressurized Water Reactor ◽  
Two Phase ◽  
Compact Structure ◽  
Pressurized Water ◽  
Drift Flux Model ◽  
Drift Flux

Owing to its compact structure, the once-through steam generator (OTSG) is used widely in integrated Pressurized Water Reactor (PWR) The casing OTSG in concentric annuli tube is a new type of steam generator which applies double sides to transfer heat. The water of the secondary side goes through complicated phase change processes, and the flow pattern and heat transfer cases are much more complex than those of the natural circulation steam generator used in PWR. It is necessary to study their steady-state and dynamic characteristics. By means of THEATRe code which was based on the two-phase drift flux model and was modified by adding module calculating the effect of rolling motion, the casing OTSG simulation model in rolling motion was built. It can describe the parameters change in every section of OTSG accurately in rolling motion, and can embody dynamics characteristics from different aspects. By comparing the operational data in different rolling amplitude and rolling period, flow operational characteristics and principles were analyzed. The results can be used to analyze the thermal-hydraulics characteristics of the integrated PWR in rolling motion.

EVALUATION OF THE HOOP TENSILE PROPERTIES OF A STEAM GENERATOR TUBE

2006 ◽  
Vol 20 (25n27) ◽  
pp. 4129-4134
Author(s):  
SUNG-KEUN CHO ◽  
CHANG-SUNG SEOK ◽  
BONG-KOOK BAE ◽  
JAE-MEAN KOO
Keyword(s):  
Tensile Properties ◽  
Steam Generator ◽  
Tensile Tests ◽  
Tensile Specimen ◽  
Original Material ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Steam Generator Tube ◽  
Turbine Generators ◽  
Axial Tensile

The steam generators in a pressurized water reactor (PWR) are large heat exchangers that use the heat from the primary reactor coolant to make steam on the secondary-side to drive turbine generators. Hoop stress is known to be the main cause of fracture of inner pressurized tubes such as the steam generator tube. However, because the steam generator tube is too small to be manufactured to a standard tensile specimen in the hoop direction, the axial tensile properties of the steam generator tube (or original material properties) instead of hoop tensile properties have been used to estimate the fracture properties of a steam generator tube. In this study, we have conducted not only axial tensile tests but also ring tensile tests. From these test, both the axial and hoop tensile properties of steam generator tubes were obtained, and the reliability of the hoop tensile properties were confirmed by burst test of a real steam generator tube.

Natural Circulation Characteristics of an Integral Pressurized Water Reactor

2006 ◽  
Author(s):  
Junli Gou ◽  
Suizheng Qiu ◽  
Guanghui Su ◽  
Dounan Jia
Keyword(s):  
Steady State ◽  
Steam Generator ◽  
Natural Circulation ◽  
Heat Removal ◽  
Primary Circuit ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Secondary Side ◽  
Circulation Characteristics

Natural circulation potential is of great importance to the inherent safety of a nuclear reactor. This paper presents a theoretical investigation on the natural circulation characteristics of an integrated pressurized water reactor. Through numerically solved the one-dimensional model, the steady-state single phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the once-through steam generator, the natural circulation characteristics are studied. Based on the preliminary calculation analysis, it is found that natural circulation mass flow rate is proportional to the exponential function of the power, and the value of the exponent is related to working conditions of the steam generator secondary side. The higher height difference between the core center and the steam generator center is favorable to the heat removal capacity of the natural circulation.

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