Experimental investigation of PWR accident scenarios at the PKL test facility

PKL is the only test facility in Europe that replicates the entire primary side and the most important parts of the secondary side of western-type Pressurized Water Reactors (PWR) in the scale of 1:1 in heights. It is also worldwide the only test facility with 4 identical reactor coolant loops arranged symmetrically around the Reactor Pressure Vessel (RPV) for simulation of nonsymmetrical boundary conditions between the reactor loops. Thermal-hydraulic phenomena observed in PWRs are simulated in the PKL test facility for over 40 years. The analyses carried out in these years encompass a large spectrum of accident scenario simulations and corresponding cool-down procedures. The overall goal of the PKL experiments is to show that under accident conditions - even for extreme and highly unlikely assumptions as additional loss of safety systems - the core cooling can be maintained or re-established by automatic or operator- performed procedures and that a severe accident e.g. a core melt-down can be avoided under all circumstances. Another goal of the tests performed in the PKL facility is the provision of data for validation of thermal-hydraulic system codes. This paper presents recent modifications of the PKL facility, applied in order to adapt the facility to the latest western-type designs currently built in the world. The paper discusses also important results obtained in the last years

Experimental investigation of LWR passive safety systems performance at the INKA test facility

INKA is a test facility designed by Framatome and built in the technical center in Karlstein. The original objective for establishing this test rig was the investigation of the performance of the passive safety systems developed in a new Framatome Boiling Water Reactor (BWR) design – KERENA. INKA was constructed in the scale of 1:1 in heights while the total volume of the containment was replicated in 1:24. Since the geometries of particular safety systems are faithfully reflected, their actual performance in the original plant can be investigated at the full scale. Due to the unquestionable interest of the nuclear community in the inherent safety, not only new BWR and PWR designs are equipped with the passive systems, but also particular passive solutions are considered to be applied into the already existing Light Water Reactors (LWR). In this context and due to the fact that both, single component tests and integral tests can be conducted at INKA, the facility can be employed for a demonstration/qualification of a large range of passive safety systems foreseen for quite different types of LWRs. Hence, the goal of the EASY project was the experimental confirmation of the passive systems performance and the analysis of their interactions between each other in the integral tests. Besides, the overarching target of all tests performed at INKA is provision of data for codes validation. This paper presents major outcomes and conclusions drawn on the basis of EASY project results

Modelowanie zachowania systemu kondensatora awaryjnego w przypadku awarii utraty chłodziwa w reaktorze BWR generacji III+

Emergency Condenser (EC) is a heat exchanger composed of a large number of slightly inclined U-tubes arranged horizontally. The inlet header of the condenser is connected with the top part of the Reactor Pressure Vessel (RPV), which is occupied by steam during critical operation. The lower header in turn is linked with the RPV below the liquid water level during normal operation of the reactor. The tube bundle is filled with cold water and it is located in a vessel filled with water of the same temperature. Thus, the EC and RPV form together a system of communicating vessels. In case of an emergency and a decrease of the water level in the RPV, the water flows gravitationally from U-tubes to the RPV. At the same time the steam from the RPV enters to the EC and condenses due to its contact with cold walls of the EC. The condensate flows then back to the RPV due to the tubes inclination. Hence, the system removes heat from the RPV and serves as a high- and low-pressure injection system at the same time. In this paper a model of the EC system is presented. The model was developed with Modelica modeling language and OpenModelica environment which had not been used in this scope before. The model was verified against experimental data obtained during tests performed at INKA (Integral Test Facility Karlstein) ̶ a test facility dedicated for investigation of the passive safety systems performance of KERENA ̶ generation III+ BWR developed by Framatome.Kondensator awaryjny jest wymiennikiem ciepła złożonym z dużej ilości U-rurek lekko nachylonych względem pozycji horyzontalnej. Kolektor wlotowy kondensatora połączony jest pojedynczym przewodem z górną częścią zbiornika ciśnieniowego reaktora, w której w trakcie normalnej pracy reaktora znajduje się para wodna. Dolny kolektor połączony jest natomiast ze zbiornikiem ciśnieniowym poniżej lustra wody w stanie ciekłym. Wiązka rurek kondensatora, w trakcie krytycznej pracy reaktora, wypełniona jest zimną wodą i zanurzona jest w basenie z wodą o tej samej temperaturze. Wiązka rurek kondensatora oraz rur doprowadzających tworzą wraz ze zbiornikiem ciśnieniowym zespół naczyń połączonych. W razie sytuacji awaryjnej, w przypadku spadku poziomu wody w zbiorniku ciśnieniowym, woda z kondensatora spływa grawitacyjnie do zbiornika ciśnieniowego, a para, która dostaje się do U-rurek kondensuje na skutek wymiany ciepła z zimną wodą otaczającą kondensator od zewnątrz. W ten sposób kondensator działając pasywnie, zastępuje wysokociśnieniowy oraz niskociśnieniowy wtrysk wody chłodzącej do zbiornika ciśnieniowego. W artykule przedstawiono model systemu kondensatora awaryjnego wraz ze zbiornikiem ciśnieniowym. Model został wykonany przy użyciu niestosowanego wcześniej w tym zakresie języka Modelica oraz środowiska OpenModelica. Następnie opracowany kod został zweryfikowany poprzez porównanie wyników z pomiarami eksperymentalnymi przeprowadzonymi na obiekcie INKA (Integral Test Facility Karlstein) – obiekcie testowym dedykowanym badaniom nad pasywnymi systemami bezpieczeństwa reaktora KERENA – reaktora BWR generacji III+ opracowanego przez firmę Framatome

Modelowanie procesów zachodzących w zbiorniku ciśnieniowym reaktora wodnego wrzącego podczas spadku ciśnienia w warunkach pracy awaryjnej

Pressure Vessel resulting from an accident scenario is an important aspect in the reactor safety analysis. This paper discusses results of simulations of the water dynamics and heat transfer during the process of an abrupt depressurization of a vessel filled up to a certain level with saturated liquid water and with the rest of the vessel occupied by steam under saturation conditions. During the pressure decrease e.g. due to a break in the steam pipeline, the liquid water evaporates abruptly leading to strong transients in the vessel. These transients and the sudden emergence of void in the area occupied by liquid at the beginning, result in the elevation of the two-phase mixture. This work presents several approaches for modelling of the void fraction, the level swell and the collapse level. The first approach was based on the churn turbulent drift-flux correlation and an explicit analytic equation for the averge void fraction as a function of dimendsionless superficial vapor velocity. The second and the third aproaches were based on dimensionless analysis and purely empirical corelations. The models were verified against independent experimental data. The models represent the Reactor Pressure Vessel of the Integral Test Facility Karlstein (INKA) – a dedicated test facility for experimental investigation of KERENA – a new medium size Boiling Water Reactor design of Framatome. The comparison of the simulations results against the reference data shows a good agreement.Kontrola poziomu mieszaniny dwufazowej wody w warunkach nagłego obniżenia ciśnienia w zbiorniku ciśnieniowym reaktora, wynikających z pracy awaryjnej jest ważnym aspektem analizy bezpieczeństwa reaktora jądrowego. Artykuł opisuje i weryfikuje wyniki symulacji zjawisk mechaniki płynów i wymiany ciepła w zbiorniku ciśnieniowym podczas gwałtownego spadku ciśnienia. W trakcie normalnej pracy zbiornik wypełniony jest do pewnego poziomu wodą w stanie nasycenia. Powyżej tego poziomu znajduje się para wodna będąca również w stanie nasycenia. W przypadku szybkiego spadku ciśnienia w zbiorniku np. w wyniku uszkodzenia rurociągu pary, woda w stanie ciekłym gwałtownie odparowuje, prowadząc do stanu nieustalonego w zbiorniku. Stan nieustalony oraz pojawienie się pary w rejonie zajmowanym wcześniej przez ciecz prowadzą do podwyższenia poziomu mieszaniny dwufazowej w zbiorniku. Artykuł prezentuje i porównuje kilka sposobów modelowania udziału fazy parowej oraz zależnych od tego udziału poziomu mieszaniny dwufazowej i wysokości słupa cieczy. Pierwszy z modeli został oparty o równanie analityczne przedstawiające średnią porowatość przepływu jako funkcję bezwymiarowej prędkości pary. Drugi i trzeci model jest oparty o analizę bezwymiarową i równania otrzymane na drodze empirycznej. Modele zostały zweryfikowane z niezależnymi danymi eksperymentalnymi. Modele reprezentują zbiornik ciśnieniowy reaktora obiektu testowego INKA (Integral Test Facility Karlstein) – obiektu dedykowanego do analizy eksperymentalnej reaktora KERENA – średniej mocy reaktora wodnego wrzącego, zaprojektowanego przez firmę Framatome. Porównanie wyników symulacji z danymi referencyjnymi wskazuje na zadowalającą zgodność obliczeń

The modelling of condensation in horizontal tubes and the comparison with experimental data

The condensation in horizontal tubes plays an important role in determining the operation mode of passive safety systems of modern nuclear power plants. In this paper, two different approaches for modelling of this phenomenon are compared and verified against experimental data. The first approach is based on the flow regime map developed by Tandon. Depending on the regime, the heat transfer coefficient is calculated according to corresponding semi-empirical correlation. The second approach uses a general, fully empirical correlation proposed by Shah. Both models are developed with utilization of the object-oriented, equation-based Modelica language and the open-source Open-Modelica environment. The results are compared with data obtained during a large scale integral test, simulating a Loss of Coolant Accident scenario performed at the dedicated Integral Test Facility Karlstein (INKA) which was built at the Components Testing Department of AREVA in Karlstein, Germany. The INKA facility was designed to test the performance of the passive safety systems of KERENA, the new AREVA boiling water reactor design. INKA represents the KERENA containment with a volume scaling of 1:24. Components heights and levels over the ground are in the full scale. The comparison of simulations results shows a good agreement

The modelling of condensation in horizontal tubes and the comparison with experimental data

The condensation in horizontal tubes plays an important role in determining the operation mode of passive safety systems of modern nuclear power plants. In this paper, two different approaches for modelling of this phenomenon are compared and verified against experimental data. The first approach is based on the flow regime map developed by Tandon. Depending on the regime, the heat transfer coefficient is calculated according to corresponding semi-empirical correlation. The second approach uses a general, fully empirical correlation proposed by Shah. Both models are developed with utilization of the object-oriented, equation-based Modelica language and the open-source Open-Modelica environment. The results are compared with data obtained during a large scale integral test, simulating a Loss of Coolant Accident scenario performed at the dedicated Integral Test Facility Karlstein (INKA) which was built at the Components Testing Department of AREVA in Karlstein, Germany. The INKA facility was designed to test the performance of the passive safety systems of KERENA, the new AREVA boiling water reactor design. INKA represents the KERENA containment with a volume scaling of 1:24. Components heights and levels over the ground are in the full scale. The comparison of simulations results shows a good agreement

Krakowskie Studia Międzynarodowe nr 2, 2010 (Miscellanea Americana)

The present volume of Krakowskie Studia Międzynarodowe [Krakow International Studies] is as diverse as America is. Many of the problems discussed here seem from the European perspective – or at least the Western European one – exotic, even parochial, but this is a misunderstanding of what the United States is. In Ame­rica they are real since America is a baroque, extremely pluralistic country, with the citizens devoid of an apologizing attitude towards the democratic process and debating fiercely in public. The first essay, by Marta Dębska, “A Brief History of Americanization”, is a general, concise historical-comparative study which explains the meaning of this term, crucial for America. Andrzej Bryk takes up an issue which Dębska touches on in the conclusion of her essay. Marta du Vall analyzes the very interesting phenomenon of American com­passionate conservatism as a new version of the welfare state, an issue which has been in the air for a long time. Maciej Brachowicz discusses the topic of abortion, which in the American context is especially contested. The subject of Tocqueville and slavery has always fascinated students of America, and Wojciech Kaczor is no exception. He analyzes the problem from the point of view of a French aristocrat. In turn Piotr Musie­wicz analyzes the question of the 19th-century movement reforming the doctrine of the Anglican Church and the repercussions of this reform for the American Episco­pal Church. Rafał Marek takes up another topic connected with this religious side of American life, the issue of the Orthodox Church in the United States in the context of American church-state relations. Marta Majorek takes up the work of one of the best-known scholars and thinkers of anarchism, Robert Paul Wolff, living proof of the robust presence of the anarchist streak in the American psyche full of mistrust of state power. Beata Szyjka addresses the topic of the visa lottery in the United States, pla­cing it within the historical, legal and social context of American immigration law. The last article in the volume is an exception to the entirely Polish group of mainly young students of America publishing in this volume. It is written by one of the most distinguished American scholars of political philosophy, Catherine H. Zuckert of the University of Notre Dame. It is devoted to the work of Ralph Elli­son. As usual the American volume of Krakowskie Studia Międzynarodowe con­tains its Archive section. This time we publish an excerpt from a work by Richard John Neuhaus

Fluid solid interactions – a novelty in industrial applications

The article deals with a current state-of-art of fluid solid interaction (FSI) – the new branch of continuum physics. Fluid-solid interaction is a new quality of modeling physical processes of continuum mechanics, it can be described as the interaction of various (so far treated separately from the point of view of mathematical modeling) physical phenomena occurring in continuous media systems. The most correct is the simultaneous application of the laws of the given physical disciplines, which implies that fluid solid interaction is a subset of multi-physical applications where the interactions between these subsets are exchanged on the surface in interconnected systems. Our purpose is to extend the fluid solid interaction aplications into new phenomena what follow from the industrial needs and inovative thechnologies. Selecting the various approaches, we prefer the arbitraty lagrangean-eulerian description within the bulk of fluid/solid domain and a new sort of advanced boundary condition on a surface of common contact