Passive Safety Injection Experiments with a Large-Scale Pressurized Water Reactor Simulator

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.

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

Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation ◽

10.1115/icone26-82610 ◽

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.

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Large Scale Pressurized Water Reactor Passive Containment Cooling System Wind Tunnel Test

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4042495 ◽

2019 ◽

Vol 5 (3) ◽

Author(s):

Huang Jingyu ◽

Pan Xinxin ◽

Song Chunjing

Keyword(s):

Wind Tunnel ◽

Large Scale ◽

Cooling System ◽

Air Flow ◽

Wind Tunnel Test ◽

Pressurized Water Reactor ◽

Scaled Model ◽

Pressurized Water ◽

Water Reactor ◽

Wind Speeds

The objective of the current work is to shed light on studying the air flow features of the air path which is part of the passive containment cooling system (PCS) in a pressurized water reactor design. A wind tunnel test using a 1:100 scaled model is established to study the characteristic called “wind-neutrality” of the air flow in the air path, which indicates that the environmental wind should not be beneficial or detrimental to the air flow for containment cooling. Test results show that the pressure distribution in the air path is uniform, and wind speeds, wind angles, and surroundings have little effect on air flow uniformity. These investigations show that it is possible to understand air flows in the air path of PCS with a scale wind tunnel test.

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Thermodynamic consequences of hydrogen combustion within a containment of pressurized water reactor

Archives of Thermodynamics ◽

10.2478/v10173-011-0032-2 ◽

2011 ◽

Vol 32 (4) ◽

pp. 67-79

Author(s):

Tomasz Bury

Keyword(s):

Nuclear Power ◽

Large Scale ◽

Nuclear Reactor ◽

Computer Code ◽

Hydrogen Combustion ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor ◽

Lumped Parameter ◽

Loss Of Coolant

Thermodynamic consequences of hydrogen combustion within a containment of pressurized water reactor Gaseous hydrogen may be generated in a nuclear reactor system as an effect of the core overheating. This creates a risk of its uncontrolled combustion which may have a destructive consequences, as it could be observed during the Fukushima nuclear power plant accident. Favorable conditions for hydrogen production occur during heavy loss-of-coolant accidents. The author used an own computer code, called HEPCAL, of the lumped parameter type to realize a set of simulations of a large scale loss-of-coolant accidents scenarios within containment of second generation pressurized water reactor. Some simulations resulted in high pressure peaks, seemed to be irrational. A more detailed analysis and comparison with Three Mile Island and Fukushima accidents consequences allowed for withdrawing interesting conclusions.

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Massively Parallelized Simulation of Deflagration-to-Detonation Transition in a Konvoi-Type Pressurized Water Reactor

Volume 5: Student Paper Competition ◽

10.1115/icone24-60093 ◽

2016 ◽

Cited By ~ 2

Author(s):

Josef Hasslberger ◽

Peter Katzy ◽

Thomas Sattelmayer ◽

Lorenz R. Boeck

Keyword(s):

Large Scale ◽

Initial Conditions ◽

Combustion Process ◽

Three Dimensional ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Flow Solver ◽

Water Reactor ◽

Highly Sensitive ◽

Transition Criteria

For the purpose of nuclear safety analysis, a reactive flow solver has been developed to determine the hazard potential of large-scale hydrogen explosions. Without using empirical transition criteria, the whole combustion process (including DDT) is computed within a single solver framework. In this paper, we present massively parallelized three-dimensional explosion simulations in a full-scale pressurized water reactor of the Konvoi type. Several generic DDT scenarios in globally lean hydrogen-air mixtures are examined to assess the importance of different input parameters. It is demonstrated that the explosion process is highly sensitive to mixture composition, ignition location and thermodynamic initial conditions. Pressure loads on the confining structure show a profoundly dynamic behavior depending on the position in the containment.

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