scholarly journals Investigation of TASS/SMR Capability to Predict a Natural Circulation in the Test Facility for an Integral Reactor

2014 ◽  
Vol 2014 ◽  
pp. 1-6 ◽  
Author(s):  
Young-Jong Chung ◽  
Sung-Won Lim ◽  
Kyoo-Hwan Bae
Keyword(s):  
Natural Circulation ◽  
Thermal Power ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Integral Type ◽  
Low Velocity ◽  
Pressurized Water ◽  
Safety Improvement ◽  
Safety Features ◽  
Integral Reactor

System-integrated modular advanced reactor (SMART) is a small-sized advanced integral type pressurized water reactor (PWR) with a rated thermal power of 330 MW. It can produce 100 MW of electricity or 90 MW of electricity and 40,000 ton of desalinated water concurrently, which is sufficient for 100,000 residents. The design features contributing to safety enhancement are basically inherent safety improvement and passive safety features. TASS/SMR code was developed for an analysis of design based events and accidents in an integral type reactor reflecting the characteristics of the SMART design. The main purpose of the code is to analyze all relevant phenomena and processes. The code should be validated using experimental data in order to confirm prediction capability. TASS/SMR predicts well the overall thermal-hydraulic behavior under various natural circulation conditions at the experimental test facility for an integral reactor. A pressure loss should be provided a function of Reynolds number at low velocity conditions in order to simulate the mass flow rate well under natural circulations.

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.

scholarly journals Analyses of the OSU-MASLWR Experimental Test Facility

2012 ◽  
Vol 2012 ◽  
pp. 1-19 ◽  
Author(s):  
F. Mascari ◽  
G. Vella ◽  
B. G. Woods ◽  
F. D'Auria
Keyword(s):  
Natural Circulation ◽  
Developed Countries ◽  
System Level ◽  
Test Facility ◽  
State University ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Standard Problem ◽  
Oregon State University

Today, considering the sustainability of the nuclear technology in the energy mix policy of developing and developed countries, the international community starts the development of new advanced reactor designs. In this framework, Oregon State University (OSU) has constructed, a system level test facility to examine natural circulation phenomena of importance to multi-application small light water reactor (MASLWR) design, a small modular pressurized water reactor (PWR), relying on natural circulation during both steady-state and transient operation. The target of this paper is to give a review of the main characteristics of the experimental facility, to analyse the main phenomena characterizing the tests already performed, the potential transients that could be investigated in the facility, and to describe the current IAEA International Collaborative Standard Problem that is being hosted at OSU and the experimental data will be collected at the OSU-MASLWR test facility. A summary of the best estimate thermal hydraulic system code analyses, already performed, to analyze the codes capability in predicting the phenomena typical of the MASLWR prototype, thermal hydraulically characterized in the OSU-MASLWR facility, is presented as well.

PKL Tests on Heterogeneous Inherent Boron Dilution Following SB-LOCA: Applicability to Reactor Scale

2010 ◽  
Author(s):  
Klaus Umminger ◽  
Simon Philipp Schollenberger ◽  
Se´bastien Cornille ◽  
Claire Agnoux ◽  
Delphine Quintin ◽  
...  
Keyword(s):  
Cooling System ◽  
Natural Circulation ◽  
Limited Range ◽  
Test Facility ◽  
Test Results ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
The Core ◽  
Physical Phenomena ◽  
Boron Dilution

In the course of a small break LOCA in a Pressurized Water Reactor (PWR) the flow regime in the Reactor Cooling System (RCS) passes through a number of different phases and the filling level may decrease down to the point where the decay heat is transferred to the secondary side under Reflux-Condenser (RC) conditions. During RC, the steam formed in the core condensates in the Steam Generator (SG) U-tubes. For a limited range of break size and configuration, a continuous accumulation of condensate may cause the formation of boron-depleted slugs. If natural circulation reestablishes, as the RCS is refilled, boron-depleted slugs might be transported to the Reactor Pressure Vessel (RPV) and to the core. To draw conclusions on the risk of boron dilution processes in SB-LOCA transients, two important issues, the limitation of slug size and the onset of Natural Circulation (NC) have to be assessed on the basis of experimental data, as system Thermal-Hydraulic codes are limited in their capability to replicate the complex physical phenomena involved. The OECD PKL III tests were performed at AREVA’s PKL test facility in Erlangen, Germany, to evaluate important phases of the boron dilution transient in PWRs. Several integral and separate effect tests were conducted, addressing the inherent boron dilution issue. The PKL III integral transient test runs provide sufficient data to state major conclusions on the formation and maximum possible size of the boron-depleted slugs, their boron concentration and their transport into the RPV with the restart of NC. Some of these conclusions can be applied to reactor scale. It has to be mentioned, that even though this paper is based on PKL test results obtained within the OECD PKL project, the conclusions of this paper reflect the views of the authors and not necessarily of all the members of the OECD PKL project.

scholarly journals Validation of Advanced Computer Codes for VVER Technology: LB-LOCA Transient in PSB-VVER Facility

2012 ◽  
Vol 2012 ◽  
pp. 1-15 ◽  
Author(s):  
A. Del Nevo ◽  
M. Adorni ◽  
F. D'Auria ◽  
O. I. Melikhov ◽  
I. V. Elkin ◽  
...  
Keyword(s):  
Natural Circulation ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Mass Inventory ◽  
Computer Codes ◽  
Set Up ◽  
Advanced Computer

The OECD/NEA PSB-VVER project provided unique and useful experimental data for code validation from PSB-VVER test facility. This facility represents the scaled-down layout of the Russian-designed pressurized water reactor, namely, VVER-1000. Five experiments were executed, dealing with loss of coolant scenarios (small, intermediate, and large break loss of coolant accidents), a primary-to-secondary leak, and a parametric study (natural circulation test) aimed at characterizing the VVER system at reduced mass inventory conditions. The comparative analysis, presented in the paper, regards the large break loss of coolant accident experiment. Four participants from three different institutions were involved in the benchmark and applied their own models and set up for four different thermal-hydraulic system codes. The benchmark demonstrated the performances of such codes in predicting phenomena relevant for safety on the basis of fixed criteria.

Feasibility Study on the Natural Circulation Cooling Procedure at the Loop Unbalanced Condition

2017 ◽  
Author(s):  
Wataru Sakuma ◽  
Shinya Miyata ◽  
Manabu Maruyama ◽  
Junto Ogawa
Keyword(s):  
Large Scale ◽  
Natural Circulation ◽  
Test Facility ◽  
Coolant Temperature ◽  
Pressurized Water Reactor ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Scale Test ◽  
Safety Enhancement

In typical pressurized water reactor (PWR) plant, in case that one steam generator (SG) is dried out and cannot be credited for the primary cooldown, at least one reactor coolant pump (RCP) has to be operated in order to homogenize the primary coolant temperature distribution among loops when the plant is cooled down to the cold shutdown state. For example, an accident such as steam line break (SLB) and feedwater line break (FLB) leads to this situation. If the natural circulation condition is established due to unavailability of all the RCPs, the natural circulation in the primary loop connected to the affected SG would be interrupted in the plant cooldown phase. In this situation, the continuous cooldown disturbs the smooth depressurization because it leads to void generation at the top of the affected SG tube where the high temperature coolant is left. In addition, there is a possibility that all RCPs cannot be operated in case of the earthquake or the fire if the RCPs are not earthquake-proof and fire-resistant. Therefore the establishment of the cooldown procedure without RCPs operation under the temperature unbalanced condition among the primary loops can contribute to the safety enhancement for typical PWR plants. The several experiments have been already performed to observe the natural circulation phenomena under the temperature unbalanced condition. It has been reported that the plant can be continuously cooled down with smooth depressurization by stepwise cooling manner using MSRVs of the intact SGs. In this study, Mitsubishi Heavy Industries, Ltd. (MHI) performed the transient analyses to simulate the natural circulation cooldown test under the temperature unbalanced condition among loops performed by Large Scale Test Facility (JAEA ROSA/LSTF) using M-RELAP5, which was a modified plant system transient code by MHI based on RELAP5-3D. Based on the analysis results, the thermal hydraulic phenomena of natural circulation cooldown under the temperature unbalanced condition were investigated. As a result, the mechanism of natural circulation interruption was clarified, and this paper shows the outline of the cooldown procedure under the temperature unbalanced condition which could be applied to the PWR plants.

Post Small-Break LOCA Long Term Core Cooling Performance in ACME Test Facility

2016 ◽  
Author(s):  
Zhanfei Qi ◽  
Sheng Zhu
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Safety Margin ◽  
Test Facility ◽  
Passive Safety ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Safety Features ◽  
Core Cooling

CAP1400 Pressurized Water Reactor is developed by China’s State Nuclear Power Technology Corporation (SNPTC) based on the passive safety concept and advanced system design. The Advanced Core-cooling Mechanism Experiment (ACME) integral effect test facility, which was constructed at Tsinghua University, represents a 1/3-scale height of CAP1400 RCS and passive safety features. It is designed to simulate the performance of CAP1400 passive core cooling system in the small break loss of coolant accidents (SBLOCA) for design certification, safety review and safety analysis code development. The Long Term Core Cooling (LTCC) post-LOCA could be simulated by ACME as well. A series of test cases with various break sizes and locations with post-LOCA LTCC period were conducted in ACME facility. This paper describes the post-LOCA LTCC test conducted in ACME test facility. The LTCC phenomena in different cases are very similar. It’s found that the interval that switching from IRWST injection to sump recirculation has the least safety margin. However, it’s shown that the post-LOCA LTCC in ACME could be well maintained by passive core cooling system according to the test results even though the recirculation water level in ACME IRWST-2 is lower than the containment recircualtion level in CAP1400 conservatively.

Traceability of Thermal Power Measurments: Modified Venturi Tubes

2005 ◽  
Author(s):  
Herb Estrada ◽  
Don Augenstein ◽  
Ernie Hauser
Keyword(s):  
Thermal Power ◽  
Pressurized Water Reactor ◽  
Steam Generators ◽  
Pressurized Water ◽  
Flow Measurements ◽  
Flow Meters ◽  
Tube Type ◽  
Traceable Calibration ◽  
Flow Element

This is the second of two papers describing the traceability of nuclear feedwater flow measurements. The first considered the challenges and methodology for establishing the traceability of chordal ultrasonic flow meters. This paper considers the challenges of establishing the traceability in a measurement using a flow element of the modified venturi tube type. It specifically considers the assumptions and uncertainties associated with the extrapolation, for use in the field, of tube calibration factors measured in the laboratory. To quantify these uncertainties, the in-situ performance of four modified venturi tubes is compared with the performance of four 8-path chordal ultrasonic flowmeters. The data analyzed were collected in the feeds of four steam generators in a large pressurized water reactor plant, each feed containing one meter of each type. The meters were initially calibrated in this series arrangement in a NIST traceable calibration lab and then operated in the same arrangement in the field.

Thermal-Hydraulic Studies of the Steam Separation in Horizontal Steam Generator at PGV Test Facility

2009 ◽  
Author(s):  
Yuriy V. Parfenov ◽  
Oleg I. Melikhov ◽  
Vladimir I. Melikhov ◽  
Ilya V. Elkin
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Test Facility ◽  
Evolutionary Development ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Tube Bundles ◽  
Horizontal Steam Generator ◽  
Calculation Results ◽  
Steam Separation

A new design of nuclear power plant (NPP) with pressurized water reactor “NPP-2006” was developed in Russia. It represents the evolutionary development of the designs of NPPs with VVER-1000 reactors. Horizontal steam generator PGV-1000 MKP with in-line arrangement of the tube bundles will be used in “NPP-2006”. PGV test facility was constructed at the Electrogorsk Research and Engineering Center on NPP Safety (EREC) to investigate the process of the steam separation in steam generator. The description of the PGV test facility and tests, which will be carried out at the facility in 2009, are presented in this paper. The experimental results will be used for verification of the 3D thermal-hydraulic code STEG, which is developed in EREC. STEG pretest calculation results are presented in the paper.

Validation of Trio_U: Numerical Simulations of a ROCOM Buoyancy Driven Test Case

2004 ◽  
Author(s):  
T. Ho¨hne ◽  
U. Bieder ◽  
S. Kliem ◽  
H.-M. Prasser
Keyword(s):  
Turbulent Flows ◽  
Natural Circulation ◽  
Density Difference ◽  
Test Facility ◽  
Test Case ◽  
Pressurized Water ◽  
Constant Flow Rate ◽  
Sensor Position ◽  
Nuclear Systems ◽  
Water Reactors

A generic investigation of the influence of density differences between the primary loop inventory and the ECC water on the mixing in the downcomer was made at the ROCOM Mixing Test Facility at Forschungszentrum Rossendorf (FZR)/Germany. ROCOM is designed for experimental coolant mixing studies over a wide variety of possible scenarios. It is equipped with advanced instrumentation, which delivers high-resolution information characterizing either temperature or boron concentration fields in the investigated pressurized water reactor. For the validation of the Trio_U code an experiment with 5% constant flow rate in one loop (magnitude of natural circulation) and 10% density difference between ECC and loop water was taken. Trio_U is a CFD code developed by the CEA France, aimed to supply an efficient computational tool to simulate transient thermal-hydraulic single-phase turbulent flows encountered in the nuclear systems as well as in the industrial processes. For this study a LES approach was used for mesh sizes according to between 300000–2 million control volumes. The results of the experiment as well as of the numerical calculations show, that a streak formation of the water with higher density is observed. At the upper sensor, the ECC water covers a small azimuthal sector. The density difference partly suppresses the propagation of the ECC water in circumferential direction. The ECC water falls down in an almost straight streamline and reaches the lower downcomer sensor position directly below the affected inlet nozzle. Only later, coolant containing ECC water appears at the opposite side of the downcomer. The study showed, that density effects play an important role during natural convection with ECC injection in pressurized water reactors. Furthermore it was important to point out, that Trio_U is able to cope the main flow and mixing phenomena.

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