Heat Removal Tests for Pressurized Water Reactor Containment Spray by Large-Scale Facility

This paper presents the results of a scaled aero acoustic test that modeled a side branch resonance observed in the residual heat removal suction line of a large pressurized water reactor. Resolution of the acoustic resonance was sought by detuning the eddy shedding frequency from the fundamental side branch acoustic mode. The specific physical modifications and their ability to detune the coupled system are presented.

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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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Analysis of PRHRS for an Integral Pressurized Water Reactor Using RELAP5/MOD3.4

18th International Conference on Nuclear Engineering: Volume 4, Parts A and B ◽

10.1115/icone18-29987 ◽

2010 ◽

Author(s):

Xuhua Ye ◽

Minjun Peng ◽

Jiange Liu

Keyword(s):

Natural Circulation ◽

Heat Removal ◽

Parameter Study ◽

Pressurized Water Reactor ◽

Steam Generators ◽

Primary Coolant ◽

Pressurized Water ◽

Water Reactor ◽

Pressure Peak ◽

Residual Heat

An investigation on the thermal hydraulic characteristics of the passive residual heat removal system (PRHRS) which is used in an integral pressurized water reactor (INSURE-100) is presented in this paper. The main components of primary coolant system are enclosed in reactor vessel. Primary fluid flow circle is natural circulation. The PRHRS can remove the energy from the primary side as long as the residual heat exchanger (RHE) is submerged in the emergency cooldown tank (ECT). The parameter study is performed by considering the effects of an effective height between the steam generators and the RHE and a valve actuation time, which are useful for the design of the PRHRS. The mass flow in the PRHRS has been affected by the height difference between the steam generators and the RHE. The pressure peak of the primary side and PRHRS has been affected by the valve action time.

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Loss of Feed Water Accident in an Integral Pressurized Water Reactor (IPWR)

Advanced Materials Research ◽

10.4028/www.scientific.net/amr.171-172.379 ◽

2010 ◽

Vol 171-172 ◽

pp. 379-384

Author(s):

Khan Salah Ud Din ◽

Min Jun Peng ◽

Muhammad Zubair

Keyword(s):

Heat Removal ◽

Feed Water ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor ◽

Heat Removal System ◽

Water Accident ◽

Relap5 Code ◽

Removal System ◽

Residual Heat

In this paper research has been carried out on Loss of Feed Water Accident (LOFW) scenario of the Integral Pressurized Water Reactor ( IPWR) under two circumstances by the use of thermal hydraulic system code i.e Relap5/Mod3.4. In the first one, Passive Residual Heat Removal System (PRHRS) which is designed to absorb core residual heat in case of transient conditions is included which has the function of operating under the accident vulnerabilities. Concerning with the second case i.e without the use of PRHRS rather a tank of water which has the capacity of about 8% of the total feed water supply and is operated under accident scenario is considered. Taken into account these conditions,first the nodalization diagram of the two cases have been figured out then according to the LOFW accident time event scenario use the Relap5 code to simulate the accident. Finally the graphical explanation (separately) of the two cases with graphical approach as well as the conclusion is given at the end.

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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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Natural Circulation Characteristics of an Integral Pressurized Water Reactor

Volume 2: Thermal Hydraulics ◽

10.1115/icone14-89390 ◽

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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Evaluation of a SGTR Accident in the Multi-Application Integrated Pressurized Water Reactor

18th International Conference on Nuclear Engineering: Volume 4, Parts A and B ◽

10.1115/icone18-29953 ◽

2010 ◽

Author(s):

Liguo Jiang ◽

Minjun Peng ◽

Jiange Liu

Keyword(s):

Steam Generator ◽

Flow Model ◽

Heat Removal ◽

Injection System ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor ◽

Heat Removal System ◽

Removal System ◽

Residual Heat

One of more frequent events in the Pressurized Water Reactor (PWR) is Steam Generator Tube Rupture (SGTR) accident, which is among the main accidents in the field of nuclear safety. This paper studies the SGTR event in the Multi-application Integrated Pressurized Water Reactor (IPWR) using the best-estimate thermal-hydraulic code RELAP5/MOD3.4. In the reactor of IPWR, several Once-Through Steam Generator (OTSG) cassettes are used and located between the core support and the pressure vessel. The tube rupture location is on the top of the tube sheet of a steam generator. Three different tube rupture modeling methods and several different subcooled discharge coefficients in the critical flow model are considered and compared. In the safety analysis, high pressure safety injection system, core makeup system and Passive Residual Heat Removal System (PRHRS) that would affect the accident consequences are considered.

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