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.

The RELAP5 Simulation of 2″ LOCA of APEX Facility

2010 ◽  
Author(s):  
Heng Xie
Keyword(s):  
Test Facility ◽  
State University ◽  
Pressurized Water Reactor ◽  
Reduced Pressure ◽  
Pressurized Water ◽  
Sequence Of Events ◽  
Comparison Results ◽  
Integral Systems ◽  
Oregon State University ◽  
Good Agreement

The RELAP5/SCDAP Mod3.2(am5) code is employed to simulate the OSU-AP1000-05 test conducted in the Advanced Plant Experimental (APEX) test facility at Oregon State University (OSU). The APEX-1000 test facility is an one-fourth height, one-half time scale, and reduced pressure integral systems facility to simulate the Westinghouse Advanced Passive 1000 MW (AP1000) pressurized water reactor. OSU-AP1000-05 is a two-inch break at the bottom of cold leg #4 with 3 out of 4 ADS-4 valves of OSU-APEX-1000 facility. RELAP5 predictions are compared to the experimental data generated by the test. The comparison shows good agreement between the predicted and measured sequence of events of some key parameters during the transient. From the comparison results, it could be preliminary concluded that the RELAP5/SCDAP Mod3.2(am5) code are suitable to simulate the small LOCA of APEX.

Dynamic PRA Based on System Codes Coupling for Passive Safety System in Integral Pressurized Water Reactor

2020 ◽  
Author(s):  
Yi Mi ◽  
Akira Tokuhiro
Keyword(s):  
Natural Circulation ◽  
Passive Safety ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Active Systems ◽  
Water Reactor ◽  
System Parameters ◽  
Event Ordering ◽  
Passive Safety Systems ◽  
The Difference

Abstract An integral Pressurized Water Reactor (iPWR) type SMR was studied featuring Passive Safety Systems (PSSs). Different from active systems, PSSs are easily influenced by system parameters referred to as phenomenological factors such as heat loss, flow friction, oxidation, non-condensable gases and void fraction due to the low driving force of natural circulation. The system parameters also contribute to the uncertainty and dependency of PSS leading to the system unreliability. Thus, efforts are made to improve the reliability of PSS. A classical Probabilistic Risk Assessment (PRA) model describing active systems does not consider time evolution nor event ordering for PSS that dynamic PRA can accommodate. Here we developed and realized coupling between LabVIEW and CAFTA. Isolation Condenser System (ICS) is taken as the benchmark system due to the simple design in single phase without phase change phenomena in order to mainly remove decay heat and secondarily depressurize the reactor pressure vessel (RPV). A classical PRA model of ICS using CAFTA is coupled with real-time simulation of primary loop and ICS in LabVIEW, leading to a dynamic simulation result. The difference in failure probability using dynamic versus classical PRA revealed that for one there are more component demands with different event ordering, such that improved PSS reliability in the iPWR-type SMR designs is possible.

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.

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.

Simulation of JAERI Downcomer Effective Water Head Experiments With WCOBRA/TRAC-TF2

2018 ◽  
Author(s):  
Jeffrey R. Kobelak ◽  
Jun Liao ◽  
Katsuhiro Ohkawa
Keyword(s):  
Hydraulic System ◽  
Large Scale ◽  
Reactor Vessel ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
System Pressure ◽  
Water Head ◽  
Effective Water

During the reflood phase of a postulated large break loss-of-coolant accident (LBLOCA), the liquid head in the reactor vessel downcomer provides the driving force to reflood the core. Since the reflood rate is a function of the downcomer inventory, the calculation of the downcomer liquid inventory is critical in simulating the reflood phase of a postulated LBLOCA accident in a pressurized water reactor. Since the reactor coolant system pressure decreases rapidly after the onset of a LBLOCA transient, the walls surrounding the downcomer become superheated for the duration of the transient. The Japan Atomic Energy Research Institute (JAERI) downcomer effective water head test facility was designed to study boiling and steam-water interaction in the reactor vessel downcomer under prototypical reflood conditions. A number of tests were conducted at this facility with varying degrees of wall superheating (among other things) that cover the expected degree of superheating in a pressurized water reactor. The wall superheating achieved at the JAERI facility is greater than that of other large-scale facilities that are typically simulated to validate thermal-hydraulic system codes. WCOBRA/TRAC-TF2 is the thermal-hydraulic system code utilized in the FULL SPECTRUM™ LOCA (FSLOCA™) evaluation model (EM). The ability of the WCOBRA/TRAC-TF2 code to predict phenomena occurring in the reactor vessel downcomer during the reflood phase of a postulated LBLOCA has been previously validated. However, only limited wall superheating was present in the existing validation basis. As such, two experiments conducted at the JAERI downcomer effective water head test facility are simulated to provide additional information on the capability of WCOBRA/TRAC-TF2 to predict the liquid inventory in the reactor vessel downcomer during the reflood phase of a postulated LBLOCA. The code captured all the trends observed in the experimental data for both Run 115 and Run 121. The various collapsed liquid levels tended to be well-predicted or under-predicted by the code after the initial simulated accumulator injection period.

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.

Sign in / Sign up

Export Citation Format

Share Document