Review and outlook of the integral test facility PKL III corresponding studies

Kerntechnik ◽  
2021 ◽  
Vol 86 (6) ◽  
pp. 391-399
Author(s):  
H. Xu
Keyword(s):  
Initial Conditions ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Model Combination ◽  
Integral Test ◽  
Pressurized Water ◽  
Safety Research ◽  
Core Section ◽  
Accident Scenarios ◽  
Complex Phenomena

Abstract This paper reviews an important integral test facility (ITF) named PKL (primary loop in German), which is designed based on a 4-loop pressurized water reactor (PWR) with the power 1 300 MWe, and especially concentrates on two aspects: (1) the tests at each developmental period of the facility until 2020, which is a typical microcosm of nuclear safety research; (2) the simulation of the PKL facility tests by using system thermal-hydraulic (STH) codes, especially RELAP5, TRACE and ATHLET. The results from the literature showed that all of these codes could reproduce the accident scenarios on the PKL facility to some extent, and simulate the complex phenomena both in the reactor pressurized vessel (RPV) and in the loops well, except some local phenomena (e. g., peak cladding temperature (PCT)). Furthermore, this paper presents some suggestions on PKL further tests. Especially, the sensitivity studies of initial conditions (ICs) and boundary conditions (BCs), test studies related to Extensive damage mitigation guidelines (EDMGs) and FLEX strategies, anticipated transients without scram (ATWS), detailed core section model, combination with other ITF or separate effects test (SET) facilities, and tests on advanced conception reactors are emphasized.

scholarly journals Integral Test Facility PKL: Experimental PWR Accident Investigation

2012 ◽  
Vol 2012 ◽  
pp. 1-16 ◽  
Author(s):  
Klaus Umminger ◽  
Lars Dennhardt ◽  
Simon Schollenberger ◽  
Bernhard Schoen
Keyword(s):  
Systematic Investigation ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Integral Test ◽  
Pressurized Water ◽  
Height Scale ◽  
Water Reactors ◽  
Accident Conditions ◽  
Secondary Side ◽  
Reactor Pressure

Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR) at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circulation pumps and steam generators (SGs) arranged symmetrically around the reactor pressure vessel (RPV). The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermal-hydraulic phenomena. This paper presents a survey of test objectives and programs carried out to date. It also describes the test facility in its present state. Some important results obtained over the years with focus on investigations carried out since the beginning of the international cooperation are exemplarily discussed.

scholarly journals The SPES3 Experimental Facility Design for the IRIS Reactor Simulation

2009 ◽  
Vol 2009 ◽  
pp. 1-12 ◽  
Author(s):  
Mario Carelli ◽  
Lawrence Conway ◽  
Milorad Dzodzo ◽  
Andrea Maioli ◽  
Luca Oriani ◽  
...  
Keyword(s):  
Nuclear Fission ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Integral Test ◽  
Pressurized Water ◽  
Hydraulic Conditions ◽  
Final Design ◽  
System Verification ◽  
Data Extrapolation ◽  
Relap5 Code

IRIS is an advanced integral pressurized water reactor, developed by an international consortium led by Westinghouse. The licensing process requires the execution of integral and separate effect tests on a properly scaled reactor simulator for reactor concept, safety system verification, and code assessment. Within the framework of an Italian R&D program on Nuclear Fission, managed by ENEA and supported by the Ministry of Economic Development, the SPES3 facility is under design and will be built and operated at SIET laboratories. SPES3 simulates the primary, secondary, and containment systems of IRIS with 1 : 100 volume scale, full elevation, and prototypical thermal-hydraulic conditions. The simulation of the facility with the RELAP5 code and the execution of the tests will provide a reliable tool for data extrapolation and safety analyses of the final IRIS design. This paper summarises the main design steps of the SPES3 integral test facility, underlying choices and phases that lead to the final design.

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.

Experimental Research on Non-Condensable Gases Effects in Passive Decay Heat Removal System

2016 ◽  
Author(s):  
Yang Liu ◽  
Haijun Jia ◽  
Li Weihua
Keyword(s):  
High Pressure ◽  
Heat Removal ◽  
Volume Ratio ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Decay Heat ◽  
Optimizing Design ◽  
Heat Removal System ◽  
Removal System

Passive decay heat removal (PDHR) system is important to the safety of integral pressurized water reactor (IPWR). In small break LOCA sequence, the depressurization of the reactor pressure vessel (RPV) is achieved by the PDHR that remove the decay heat by condensing steam directly through the SGs inside the RPV at high pressure. The non-condensable gases in the RPV significantly weaken the heat transfer capability of PDHR. This paper focus on the non-condensable gas effects in passive decay heat removal system at high pressure. A series of experiments are conducted in the Institute of Nuclear and New Energy Technology test facility with various heating power and non-condensable gas volume ratio. The results are significant to the optimizing design of the PDHR and the safety operation of the IPWR.

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.

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.

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.

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.

Numerical Investigation on the Gas Filling Process of Hydrogen-Air-Steam Mixtures in a Semi-Confined Experimental Pipe

2017 ◽  
Author(s):  
Min Li ◽  
Youyou Xu ◽  
Xiaojian Wen ◽  
Songlin Liu ◽  
Guangnan Luo
Keyword(s):  
Severe Accident ◽  
Blockage Ratio ◽  
Gas Composition ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Safety Research ◽  
Filling Time ◽  
Time Required ◽  
Hydrogen Safety

Hydrogen gathering in the containment may occur followed by a severe accident in a nuclear power plant. A flammable mixture can be formed when hydrogen is mixed with air. The ignition of the gas mixture could threaten the integrity of the containment. In order to provide technology base and experiment data for optimization of hydrogen safety technology of Chinese advanced pressurized water reactor CAP1400, a major project regarding hydrogen safety research of pressurized water reactor containment is underway. As an important part of the project, an experimental facility (A4Q-DH) for the study of hydrogen combustion will be built. Gas displacement method is used to filling the premixed hydrogen-air-steam mixtures into the experimental pipe. Flow behaviors of the gases in the pipe are complicated because fluid flow can be disturbed by the built-in obstacles and gas density can be changed with variation of gas composition concentration. Therefore, it is necessary to evaluate the effectiveness of the gas filling method. In this paper, gas filling processes for the experimental pipe with different obstacles and gas composition concentrations were simulated using computational fluid dynamics software ANSYS Fluent. The results indicated that hydrogen-air-steam mixtures can be uniformly distributed in the experimental pipe within tens of seconds. The obstacles with modest blockage ratio in the pipe are conducive to shorten the required gas filling time. The hindering effect of annular obstacles is greater than the one of circular and square obstacles. The time required for air to achieve uniformly distribution increases with the increase of the inlet concentration of steam and hydrogen. However, the time required for hydrogen and steam to be evenly distributed in the pipe are relatively close regardless of the shape and blockage ratio of obstacle and the inlet gas concentration.

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