Design and automated operation of a condensation-induced depressurization system

Typical examples of the use of the vacuum environments in industry include vacuum coating, vacuum drying, vacuum packing, vacuum casting, vacuum heat treatment, vacuum cooling for food storage and leakage detection. Most of these applications require a comparatively small volume of vacuum environment. However, there are also many applications where enormous vacuum chambers are needed. For example, a large vacuum chamber is used to simulate the conditions of space. This study investigates the design and the automated operation of a vacuum generation system based on the idea of the condensation-induced depressurization with prefilled condensable medium in a confined, adiabatic chamber. The operation process of this new system includes mainly four phases, steam filling phase, cooling phase, usage phase and transition phase, operating sequentially. An automated control system based on these phases is designed and implemented on a laboratory scaled experimental system. This experimental system serves as a vacuum environment application to demonstrate the automatic and continuous operation of the condensation-induced depressurization system. In order to obtain a better understanding on the selection of parameters for performance improvement, models of the first three process phases are developed and analyzed. These models provide a reference for the design of systems for other industry applications as well. Methods of improving the system design and operation are investigated. The analysis shows that high pressure and fast flow steam source will accelerate the steam filling process. Removal of the transition phase improves the steam filling phase and speeds up the vacuum preparation. Improvement can also be achieved on the coordination between vacuum generation and usage through control elements, as well as the proper selection of volume ratio between usage chamber and preparation chambers. The ultimate objective of this study is that the results can be used to develop guidelines for the design and operation of vacuum generation systems according to specific usage patterns of the vacuum environment applications

Depressurization characteristics of steam-based reciprocating vacuum pump

This dissertation introduces a novel vacuum technology that leverages low-pressure saturated steam and cooling-controlled condensation, offering an efficient way to utilize low-grade thermal energy sources like waste heat, steam, or solar energy. At the heart of this technology is a unique duo-chamber vacuum pump system, featuring a reciprocating piston and a heat-conductive wall, designed to generate a vacuum through steam-condensation and cooling processes. The core of this research lies in developing and validating mechanistic models for the steam-condensation depressurization process, a complex phenomenon involving phase change and transport mechanisms. Prior to this work, these mechanisms were not sufficiently modeled or understood, limiting the potential for quantitative assessment and optimization of the technology. This study addresses this gap by establishing comprehensive models that simulate the dynamic depressurization process, incorporating factors like transient cooling and non-uniform steam condensation. Dynamic depressurization characteristics, such as chamber pressure, condensation rate, and vapor temperature, have been investigated, along with parametric effects of key operation and system parameters, including initial vapor pressure/temperature, coolant flowrate and temperature, and system geometric dimensions and material selections. Specifically, a parametric model of depressurization process is developed, which is based on a modified formulation of film condensation within an enclosed cylinder. This model, based on simplified lumped-heat capacity approximation of system components, can reasonably predict various parametric effects on the depressurization process, including the parametric effects of initial steam vapor pressure or temperature, coolant flow rate, and inlet temperature of coolant. These parametric analyses are vital to the optimized system design selections and operations. To account for the transient and nonuniformity of heat and mass transfer with coolant-flow-influenced vapor condensation in a three-dimensional system, a more complicated numerical model and associated computational fluid dynamics (CFD) simulation is conducted. The depressurization-process CFD model of the condensing vapor-liquid two-phase flow is based on a Eulerian approach with the volume of fraction (VOF) method and space condensation modeling approximation. The CFD simulation reveals some strong three-dimensional non-uniform temperature distributions of both steam vapor and coolant flow. The steam condensation is also strongly non-uniformly distributed near the cooling wall of inner cylinders. However, the transient vapor pressure is nearly uniformly distributed within the chamber at any moment. The deviation between the non-uniform vapor temperature distribution and the uniform pressure distribution clearly suggests that there exists strong thermodynamic non-equilibrium in the vapor phase during the vapor condensation process, and the depressurization is mainly due to the vapor condensation rather than the cooling of vapor. A lab-scale prototype of the vacuum pump system was constructed to provide the quantitative proof-of-concept assessment of the innovative vacuum generation technology. The experimental system is also used to provide important measurements on vacuum generation processes for the validation of the proposed parametric and CFD models. In addition, a preliminary design of the automatic operation of the dual-chamber steam-based reciprocating vacuum pump is proposed. All these results and preliminary studies not only demonstrate the practical viability of the proposed vacuum technology but also provide critical insights for its optimized design and automated operation, which also lay down some foundation for the follow-up research

Review of current Severe Accident Management (SAM) approaches for Nuclear Power Plants in Europe

The Fukushima accidents highlighted that both the in-depth understanding of such sequences and the development or improvement of adequate Severe Accident Management (SAM) measures are essential in order to further increase the safety of the nuclear power plants operated in Europe. To support this effort, the CESAM (Code for European Severe Accident Management) R&D project, coordinated by GRS, started in April 2013 for 4 years in the 7th EC Framework Programme of research and development of the European Commission. It gathers 18 partners from 12 countries: IRSN, AREVA NP SAS and EDF (France), GRS, KIT, USTUTT and RUB (Germany), CIEMAT (Spain), ENEA (Italy), VUJE and IVS (Slovakia), LEI (Lithuania), NUBIKI (Hungary), INRNE (Bulgaria), JSI (Slovenia), VTT (Finland), PSI (Switzerland), BARC (India) plus the European Commission Joint Research Center (JRC). The CESAM project focuses on the improvement of the ASTEC (Accident Source Term Evaluation Code) computer code. ASTEC,, jointly developed by IRSN and GRS, is considered as the European reference code since it capitalizes knowledge from the European R&D on the domain. The project aims at its enhancement and extension for use in severe accident management (SAM) analysis of the nuclear power plants (NPP) of Generation II-III presently under operation or foreseen in near future in Europe, spent fuel pools included. In the frame of the CESAM project one of the tasks consisted in the preparation of a report providing an overview of the Severe Accident Management (SAM) approaches in European Nuclear Power Plants to serve as a basis for further ASTEC improvements. This report draws on the experience in several countries from introducing SAMGs and on substantial information that has become available within the EU “stress test”. To disseminate this information to a broader audience, the initial CESAM report has been revised to include only public available information. This work has been done with the agreement and in collaboration with all the CESAM project partners. The result of this work is presented here.JRC.F.5-Nuclear Reactor Safety Assessmen

NuScalen pienen modulaarisen reaktorin simulointi ja turvallisuustoiminnot

Small modular reactor (SMR) is a relatively recent concept in the nuclear power industry. Whereas the traditional large scale reactors are facing challenges due to their size, the small modular reactors intend to bypass the problems by utilizing aspects that their small size and modularity enables. However, the transition to SMRs includes many open questions. The energy industry and nuclear safety authorities are accustomed to large scale power plants, whereas experience with SMRs are shallow. This effectively creates a need for SMR studies, some of which this thesis aims to answer to. NuScale SMR was chosen as a focus of this thesis due to the concept’s reasonable level of maturity and its interesting utilization of passive safety systems. In the framework of this thesis, a NuScale SMR simulation model is built using the Apros program. The model is used to validate Apros calculation required for SMR simulation.The second objective of this thesis is to present the specific safety features of NuScale SMR that differ from current nuclear power plants. They are presented from a technical point of view and are briefly projected on Finnish regulatory guides. Feature projection reveals that many of the design features are fundamentally compatible with current guides with a few exceptions. In this thesis we perceive that among others the modular reactor mass production and passive functions could face challenges. The simulation results show that Apros code is capable of SMR and passive safety system modelling. However, the results also show that precise simulation of passive safety systems would benefit from further code development on those fields. The thesis also presents modelling guidelines that are beneficial when Apros is used for SMR modelling.Pienet modulaariset ydinreaktorit (SMR) ovat verrattain uusi konsepti ydinenergiateollisuudessa. Siinä missä tavalliset reaktorit kohtaavat huomattavia haasteita suuren kokonsa vuoksi, pienet modulaariset reaktorit pyrkivät kiertämään ne hyödyntämällä pienen kokonsa ja modulaarisuutensa mahdollistamia ominaisuuksia. Siirtyminen SMR:iin sisältää tosin myös avoimia kysymyksiä. Energiateollisuus ja ydinturvallisuusviranomaiset ovat tottuneet käsittelemään suuren kokoluokan laitoksia ja niihin liittyviä ilmiöitä, kun taas kokemukset SMR:istä ja niiden ilmiöistä ovat vähäisiä. Käytännössä tämä luo tarpeen SMR:iin kohdistuvalle tutkimukselle, jota tämäkin opinnäytetyö pyrkii tukemaan. NuScale SMR valittiin tämän työn tutkimuksen kohteeksi kyseisen konseptin kohtuullisen korkean valmiusasteen ja siinä hyödynnettävien mielenkiintoisten passiivisten ilmiöiden takia. Työssä rakennetaan Apros-ohjelmalla simulointimalli NuScalen konseptin mukaisesta SMR -koelaitteistosta. Mallin avulla validoidaan Aprosin SMR:ien simuloimiseen tarvittavaa laskentaa. Työn toinen tavoite on esittää NuScalen konseptille suunniteltuja erityisesti nykyisistä ydinvoimalaitoksista eriäviä ominaisuuksia teknisestä näkökulmasta, ja verrata niitä Suomen ydinvoimalain ja ydinturvallisuusohjeiden (YVL-ohjeet) turvallisuusvaatimuksiin. Ominaisuuksien projisointi paljastaa, että osa suunnitteluominaisuuksista sopii perustavanlaatuisella tasolla nykyisiin määräyksiin muutamin poikkeuksien. Työssä huomataan, että haasteita on muun muassa modulaaristen reaktoreiden massatuotannon ja passiivisten ominaisuuksien osalta. Simulointitulokset osoittavat Aprosin nykytilassaan kykenevän SMR:ien ja passiivisten turvallisuustoimintojen mallinnukseen. Tulokset kuitenkin osoittavat, että passiivisten systeemien tarkka simulointi hyötyisi kyseisien alueiden koodien jatkokehityksestä.Työssä myös esitetään mallinnusperiaatteita, joita olisi hyvä noudattaa, kun Aprosilla mallinnetaan SMR:iä

Severe Accident Phenomenology Analyses and Fission Gas Release in Advanced Nuclear Reactors

The aim of this work is to contribute to qualify a model in order to simulate the progression of a severe accident (SA), evaluating the Source Term during SA scenarios for a III Generation nuclear power plant (NPP), as AP1000 and EPR. The present Ph.D. thesis is articulated in 3 different parts. The first part is a status of the art on the SA phenomenology and on the Lumped Parameter Codes involved in the analyses. The aim of second part of the work is to give an overview on the phenomena and on the ability of code models to follow the progression of the test scenarios. The third part of the thesis concerns SA sequence analyses of AP1000 plant

Further development of Severe Accident Management Strategies for a German PWR Konvoi Plant based on the European Severe Accident Code ASTEC

Der Fukushima-Unfall hat gezeigt, dass weitere Verbesserungen des Handbuchs für Mitigative Notfallschutzmaßnahmen (SAMGs) notwendig sind. Dafür ist es erforderlich, eine umfangreiche Datenbank von Risiko-relevanten Szenarien unter Einsatz deterministischer Analysen basierend auf dem Stand von Wissenschaft und Technik zu generieren.In Rahmen dieser Doktorarbeit wird das Störfallcode ASTEC verifiziert und zur Optimierung und Entwicklung verschiedener Notfallschutzmaßnahmen (SAM) für eine Deutsche Reaktoranlage-Anlage unter Berücksichtigung der aus dem Fukushima-Unfall abgeleiteten Lehren eingesetzt. Zu diesem Zweck werden die physikalischen Modelle der Frühphase eines schweren Unfalls von ASTECV2.0 anhand vom QUENCH-08 Versuch validiert, wobei einen beheizten Kern mit gesättigtem Wasserdampfgeflutet wird. Der Vergleich zwischen den gerechneten und experimentellen Ergebnissenhat zeigt, dass ASTECV2.0 alle wichtigen Phänomenewie z.B. Oxidation, Konvektions-, und Strahlungswärme vom QUENCH-08 Versuch mit guten Genauigkeit beschreiben kann. Daher liegen die berechneten Oxidationsprofile und diefreigesetzte Wasserstoffs-Menge nah an den Messdaten. Allerdings wird die Steigerung der Temperaturen während der Flutphase vom ASTEC unterschätzt. Ursachen für diese Unterschätzung sind u.a. die niedrigen, radialen Temperaturgradientensowie die Wärmeübertragung über den aktivenTeil des Bündels. Mit dem validierten ASTECV2.0 Code wurde die Wirksamkeit verschiedenerSAM-Maßnahmenfür unterschiedliche SA-Sequenzen der Konvoi DWR-Anlage wie z.B. primär- und sekundärseitige Druckentlastung und/oder Wassereinspeisung in den Sekundär-, oder Primärkreislauf umfassend untersucht. Zu den ausgewählten SA-Störfallszenarien gehören u.a. den mittleren und kleinen Bruch im Primärkreislauf (MBLOCA und SBLOCA) sowie der Ausfall der Drehstromversorgung (SBO). Unter Berücksichtigung der zahlreichen durchgeführten ASTEC-Analysen der genannten Szenarien haben sich folgende Notfallschutzmaßnahmen als „vielversprechenden und sehr wirksam“ zur Verzögerung oder Verhinderung des Reaktordruckbehälter (RPV)-Versagens herauskristallisiert: 1) Sekundärseitige Druckentlastung und Dampferzeugerbespeisung mit mehr als 15 kg/s durchgeführt bevor die Überschreitung der Kernaustrittstemperatur (CET) von 400 °C erreicht wurde um Kernschmelze zu vermeiden (gemäß SBO). 2) Primärseitige Druckentlastung durchgeführt beim Erreichen der Kernaustrittstemperatur (CET) von 400 °C oder mit einer maximalen Verspätung von20-30 min, um Kernschmelze und RPV-Versagen zu verzögern (gemäß SBLOCA und SBO). 3) Kernfluten beim Überschreitung der CET>650 °Cmit mehr als 20 kg/s Einspeiserate, um erhebliche Kernschmelze zu vermeiden (gemäß MBLOCA, SBLOCA und SBO). 4) Falls eine externe Bespeisung in den Primärkreislauf z.B. mit mobilen Pumpen berücksichtig wird, muss der Einsatz einerHochdruckmobilpumpe (>50 bar) in der darauf folgenden Stunde nach dem Verlust des ACserfolgen, damit das Fluten kurz nach dem Erreichen der CET=650 °C erfolgen kann (gemäß SBO). 5) Unabhängig vom Einspeiseraten kann RPV Versagen nicht vermieden werden, wenn sich mehr als 20 Tonnen Coriumlänger als 20 min im unteren Plenum angesammelt hat (gemäß MBLOCA, SBLOCA und SBO). Die aufgelisteten Notfallmaßnahmen sind von großer Bedeutung für den Betreiber und sollten mit der neusten ASTEC Version 2.1, welche verbesserte Zweiphasenströmungsmodelle für CESAR-Modul zur Beschreibung des Kernverhaltens aufweist, verifiziert werden. Des Weiteren ist es notwendig, eine Quantifizierung der Codeunsicherheiten bezüglich wesentliche physikalischen Phänomene wie z.B. Oxidation, Schmelzeverlagerung und Stratifizierung im unterm Plenum, RPV-Versagen, etc. durchzuführen. Die durchgeführten Analysenan dem generischen Deutschen Konvoi PWR demonstrieren die Leistungsfähigkeit vom ASTECV2.0 wesentlichen Phänomenen der frühen Kernschmelzphase sowie die Eignung zur Erarbeitung und Optimierung von Notfallschutzmaßnahmen. Abschließend kann herausgestellt werden, dass diese Doktorarbeit wichtige Beiträge zur Erweiterung der technischen Basis für die Entwicklung und Optimierung von SAMGs geleistet hat, welche zur Stärkung der Robustheit der Sicherheitseigenschaften einer PWR-Anlage gegenüber schweren Kernunfällen genutzt werden kann

RELAP5-3D modeling of ADS blowdown of MASLWR facility

Oregon State University has hosted an International Atomic Energy Agency (IAEA) International Collaborative Standard Problem (ICSP) through testing conducted on the Multi-Application Small Light Water (MASLWR) facility. The MASLWR facility features a full-time natural circulation loop in the primary vessel and a unique pressure suppression device for accident scenarios. Automatic depressurization system (ADS) lines connect the primary vessel to a high pressure containment (HPC) which dissipates steam heat through a heat transfer plate thermally connected to another vessel with a large cool water inventory. This feature drew the interest of the IAEA and an ICSP was developed where a loss of feedwater to the steam generators prompted a depressurization of the primary vessel via a blowdown through the ADS lines. The purpose of the ICSP is to evaluate the applicability of thermal-hydraulic computer codes to unique experiments usually outside of the validation matrix of the code itself. RELAP5-3D 2:4:2 was chosen to model the ICSP. RELAP5-3D is a best-estimate code designed to simulate transient fluid and thermal behavior in light water reactors. Modeling was conducted in RELAP5-3D to identify the strengths and weaknesses of the code in predicting the experimental trends of the IAEA ICSP. This extended to nodalization sensitivity studies, an investigation of built-in models and heat transfer boundary conditions. Besides a qualitative analysis, a quantitative analysis method was also performed

Fukushima Daiichi accident study : status as of April 2012.

Abstract Not Provide