Accident Management Actions In An Upper-Head Small-Break Loss-Of-Coolant Accident With High-Pressure Safety Injection Failed

Since the Three Mile Island accident, an important focus of pressurized water reactor (PWR) transient analyses has been a small-break loss-of-coolant accident (SBLOCA). In 2002, the discovery of thinning of the vessel head wall at the Davis Besse nuclear power plant reactor indicated the possibility of an SBLOCA in the upper head of the reactor vessel as a result of circumferential cracking of a control rod drive mechanism penetration nozzle - which has cast even greater importance on the study of SBLOCAs. Several experimental tests have been performed at the Large Scale Test Facility to simulate the behavior of a PWR during an upper-head SBLOCA. The last of these tests, Organisation for Economic Co-operation and Development Nuclear Energy Agency Rig of Safety Assessment (OECD/NEA ROSA) Test 6.1, was performed in 2005. This test was simulated with the TRACE 5.0 code, and good agreement with the experimental results was obtained. Additionally, a broad analysis of an upper-head SBLOCA with high-pressure safety injection failed in a Westinghouse PWR was performed taking into account different accident management actions and conditions in order to check their suitability. This issue has been analyzed also in the framework of the OECD/NEA ROSA project and the Code Applications and Maintenance Program (CAMP). The main conclusion is that the current emergency operating procedures for Westinghouse reactor design are adequate for these kinds of sequences, and they do not need to be modified.

Experiments for CFD-modelling of cooling water and Insulation debris two-phase flow phenomena during loss of coolant accidents

The knowledge of insulation debris generation and transport gains in importance regarding reactor safety research for PWR and BWR. The insulation debris released near the break consists of a mixture of very different fibres and particles concerning size, shape, consistence and other properties. Some fraction of the released insulation debris will be transported into the reactor sump where it may affect emergency core cooling. Experiments are performed to blast original samples of mineral wool insulation material by steam under original thermal-hydraulic break conditions of BWR. The gained fragments are used as initial specimen for further experiments at acrylic glass test facilities. The quasi ID-sinking behaviour of the insulation fragments are investigated in a water column by optical high speed video techniques and methods of image processing. Drag properties are derived from the measured sinking velocities of the fibres and observed geometric parameters for an adequate CFD modelling. In the test rig "Ring line-II" the influence of the insulation material on the head loss is investigated for debris loaded strainers. Correlations from the filter bed theory are adapted with experimental results and are used to model the flow resistance depending on particle load, filter bed porosity and parameters of the coolant flow. This concept also enables the simulation of a particular blocked strainer with CFDcodes. During the ongoing work further results of separate effect and integral experiments and the application and validation of the CFD-models for integral test facilities and original containment sump conditions are expected.

APROS code benchmarking inboiling water reactor accident analyses

Diplomityö käsittelee kiehutusvesilaitosten transienttien ja onnettomuuksien analysointia APROS-ohjelmiston avulla. Työ on tehty Teollisuuden Voima Oy:n (TVO) Olkiluoto 1 ja 2 laitosyksiköiden mallin pohjalta. Raportissa esitetään ohjelmiston käyttämiä yhtälöitäja laskentamalleja yleisellä tasolla. Työssä esitellään laitoksen yleispiirteet turvallisuustoimintoineen ja kuvataan ohjelmaan suureksi osaksi aiemmin luotua laskentamallia. Työssä on luetteloitu voimassa olevatlisensiointianalyysit, joiden joukosta on valittu laskentatapauksia ohjelmiston suorituskyvyn arviointia varten. Lisäksi työhön on valittu laskentatapauksia muilla kuin lisensointiin käytetyillä ohjelmilla lasketuista analyyseistä. Lisäksi on suoritettu vertailulaskuja konservatiivisen ja realistisen mallin erojen esille saamiseksi. Laskentatapauksia ovat mm. ylipainetransientti, jäähdytteen menetysonnettomuus ja oletettavissa oleva käyttöhäiriö, jossa pikasulku ei toimi (ATWS). Diplomityön edetessä laitosmallia on kehitetty edelleen lisäämällä joitakin järjestelmiä ja tarkentamalla joidenkin komponenttien kuvausta. Työssä ilmeni, että APROS soveltuu jäähdytteenmenetysonnettomuuden ja suojarakennuksen yhtäaikaiseen analyysiin. APROS.n vaste nopeisiin transientteihin jäi kuitenkin vertailutasosta. Tämän työn perusteella APROS-mallia kehitys jatkuu edelleen siten, että se soveltuisi entistä paremmin myös nopeiden transienttien ja ATWS-tilanteiden kuvaamiseen. Työssä olevaa lisensointianalyysien kuvausta tullaan käyttämään hyväksi selvitettäessä laitoksen turvallisuuden väliarviossa tarvittavien analyysien määrää ja laatua. Nyt saatuja kokemuksia voidaan hyödyntää myös mahdollisen kolmiulotteisen sydänmallin hankinnassa APROS-ohjelmistoon. Tässä diplomityössä esitettyjä parannuksia voidaan käyttää hyväksi SAFIRtutkimusohjelman hankkeiden suunnittelussa.

Evaluation of Bubble Formation and Break Up in Suppression Pools by Using Pattern Recognition Methods

During a possible loss of coolant accident in BWRs, a large amount of steam will be released from the reactor pressure vessel to the suppression pool. Steam will be condensed into the suppression pool causing dynamic and structural loads to the pool. The formation and break up of bubbles can be measured by visual observation using a suitable pattern recognition algorithm. The aim of this study was to improve the preliminary pattern recognition algorithm, developed by Vesa Tanskanen in his doctoral dissertation, by using MATLAB. Video material from the PPOOLEX test facility, recorded during thermal stratification and mixing experiments, was used as a reference in the development of the algorithm. The developed algorithm consists of two parts: the pattern recognition of the bubbles and the analysis of recognized bubble images. The bubble recognition works well, but some errors will appear due to the complex structure of the pool. The results of the image analysis were reasonable. The volume and the surface area of the bubbles were not evaluated. Chugging frequencies calculated by using FFT fitted well into the results of oscillation frequencies measured in the experiments. The pattern recognition algorithm works in the conditions it is designed for. If the measurement configuration will be changed, some modifications have to be done. Numerous improvements are proposed for the future 3D equipment.

Analysis of PACTEL small break LOCA experiment using TRACE

A small break loss-of-coolant accident (SBLOCA) is one of problems investigated in an NPP operation. Such accident can be analyzed using an experiment facility and TRACE thermal-hydraulic system code. A series of SBLOCA experiments was carried out on Parallel Channel Test Loop (PACTEL) facility, exploited together with Technical Research Centre of Finland VTT Energy and Lappeenranta University of Technology (LUT), in order to investigate two-phase phenomena related to a VVER-type reactor. The experiments and a TRACE model of the PACTEL facility are described in the paper. In addition, there is the TRACE code description with main field equations. At the work, calculations of a SBLOCA series are implemented and after the calculations, the thesis discusses the validation of TRACE and concludes with an assessment of the usefulness and accuracy of the code in calculating small breaks.

Analysis of PWR-PACTEL Small Break LOCA experiment using TRACE

Currently, the power generation is one of the most significant life aspects for the whole man-kind. Barely one can imagine our life without electricity and thermal energy. Thus, different technologies for producing those types of energy need to be used. Each of those technologies will always have their own advantages and disadvantages. Nevertheless, every technology must satisfy such requirements as efficiency, ecology safety and reliability. In the matter of the power generation with nuclear energy utilization these requirements needs to be highly main-tained, especially since accidents on nuclear power plants may cause very long term deadly consequences. In order to prevent possible disasters related to the accident on a nuclear power plant strong and powerful algorithms were invented in last decades. Such algorithms are able to manage calculations of different physical processes and phenomena of real facilities. How-ever, the results acquired by the computing must be verified with experimental data.

Manufacturing of self-passivating tungsten based alloys by different powder metallurgical routes

Self-passivating tungsten based alloys will provide a major safety advantage compared to pure tungsten when used as first wall armor of future fusion reactors, due to the formation of a protective oxide layer which prevents the formation of volatile and radioactive WO3 in case of a loss of coolant accident with simultaneous air ingress. Bulk WCr10Ti2 alloys were manufactured by two different powder metallurgical routes: (1) mechanical alloying (MA) followed by hot isostatic pressing (HIP) of metallic capsules, and (2) MA, compaction, pressureless sintering in H2 and subsequent HIPing without encapsulation. Both routes resulted in fully dense materials with homogeneous microstructure and grain sizes of 300 nm and 1 μm, respectively. The content of impurities remained unchanged after HIP, but it increased after sintering due to binder residue. It was not possible to produce large samples by route (2) due to difficulties in the uniaxial compaction stage. Flexural strength and fracture toughness measured on samples produced by route (1) revealed a ductile-to-brittle-transition temperature (DBTT) of about 950 °C. The strength increased from room temperature to 800 °C, decreasing significantly in the plastic region. An increase of fracture toughness is observed around the DBTT.

Hydrodynamic modeling of mineral wool fiber suspensions in a two-dimensional flow

A consequence of a loss of coolant accident is that the local insulation material is damaged and maybe transported to the containment sump where it can penetrate and/or block the sump strainers. An experimental and theoretical study, which examines the transport of mineral wool fibers via single and multi-effect experiments is being performed. This paper focuses on the experiments and simulations performed for validation of numerical models of sedimentation and resuspension of mineral wool fiber agglomerates in a racetrack type channel. Three velocity conditions are used to test the response of two dispersed phase fiber agglomerates to two drag correlations and to two turbulent dispersion coefficients. The Eulerian multiphase flow model is applied with either one or two dispersed phases.

Simulations of agglomerate sedimentation and suspension

Mineral wool insulation material applied to the primary cooling circuit of a nuclear reactor maybe damaged in the course of a loss of coolant accident (LOCA). The insulation material released by the leak may compromise the operation of the emergency core cooling system (ECCS), as it maybe transported together with the coolant in the form of mineral wool fiber agglomerates (MWFA) suspensions to the containment sump strainers, which are mounted at the inlet of the ECCS to keep any debris away from the emergency cooling pumps. In the further course of the LOCA, the MWFA may block or penetrate the strainers. In addition to the impact of MWFA on the pressure drop across the strainers, corrosion products formed over time may also accumulate in the fiber cakes on the strainers, which can lead to a significant increase in the strainer pressure drop and result in cavitation in the ECCS. Therefore, it is essential to understand the transport characteristics of the insulation materials in order to determine the long-term operability of nuclear reactors, which undergo LOCA. An experimental and theoretical study performed by the Helmholtz-Zentrum Dresden-Rossendorf and the Hochschule Zittau/Görlitz is investigating the phenomena that maybe observed in the containment vessel during a primary circuit coolant leak. The study entails the generation of fiber agglomerates, the determination of their transport properties in single and multi-effect experiments and the long-term effects that particles formed due to corrosion of metallic containment internals by the coolant medium have on the strainer pressure drop. The focus of this presentation is on the numerical models that are used to predict the transport of MWFA by CFD simulations. A number of pseudo-continuous dispersed phases of spherical wetted agglomerates can represent the MWFA. The size, density, the relative viscosity of the fluid-fiber agglomerate mixture and the turbulent dispersion all affect how the fiber agglomerates are transported. In the cases described here, the size is kept constant while the density is modified. This definition affects both the terminal velocity and volume fraction of the dispersed phases. Application of such a model to sedimentation in a quiescent column and a horizontal flow are examined. The scenario also presents the suspension and horizontal transport of a single fiber agglomerate phase in a racetrack type channel.

Hydrodynamic modeling of mineral wool fiber suspensions in a two-dimensional channel flow

A consequence of a loss of coolant accident is the damage of adjacent insulation materials (IM). IM may then be transported to the containment sump strainers where water is drawn into the ECCS (emergency core cooling system). Blockage of the strainers by IM lead to an increased pressure drop acting on the operating ECCS pumps. IM can also penetrate the strainers, enter the reactor coolant system and then accumulate in the reactor pressure vessel. An experimental and theoretical study that concentrates on mineral wool fiber transport in the containment sump and the ECCS is being performed. The study entails fiber generation and the assessment of fiber transport in single and multi-effect experiments. The experiments include measurement of the terminal settling velocity, the strainer pressure drop, fiber sedimentation and resuspension in a channel flow and jet flow in a rectangular tank. An integrated test facility is also operated to assess the compounded effects. Each experimental facility is used to provide data for the validation of equivalent computational fluid dynamic models. The channel flow facility allows the determination of the steady state distribution of the fibers at different flow velocities. The fibers are modeled in the Eulerian-Eulerian reference frame as spherical wetted agglomerates. The fiber agglomerate size, density, the relative viscosity of the fluid-fiber mixture and the turbulent dispersion of the fibers all affect the steady state accumulation of fibers at the channel base. In the current simulations, two fiber phases are separately considered. The particle size is kept constant while the density is modified, which affects both the terminal velocity and volume fraction. The relative viscosity is only significant at higher concentrations. The numerical model finds that the fibers accumulate at the channel base even at high velocities; therefore, modifications to the drag and turbulent dispersion forces can be made to reduce fiber accumulation.

Numerical models used for the modelling of the transport of fibrous insulation debris

Mineral wool insulation material applied to the primary cooling circuit of a nuclear reactor maybe damaged in the course of a loss of coolant accident (LOCA). The insulation material released by the leak may compromise the operation of the emergency core cooling system (ECCS), as it maybe transported together with the coolant in the form of mineral wool fiber agglomerates (MWFA) suspensions to the containment sump strainers, which are mounted at the inlet of the ECCS to keep any debris away from the emergency cooling pumps. In the further course of the LOCA, the MWFA may block or penetrate the strainers. In addition to the impact of MWFA on the pressure drop across the strainers, corrosion products formed over time may also accumulate in the fiber cakes on the strainers, which can lead to a significant increase in the strainer pressure drop and result in cavitation in the ECCS. Therefore, it is essential to understand the transport characteristics of the insulation materials in order to determine the long-term operability of nuclear reactors, which undergo LOCA. An experimental and theoretical study performed by the Helmholtz-Zentrum Dresden-Rossendorf and the Hochschule Zittau/Görlitz1 is investigating the phenomena that maybe observed in the containment vessel during a primary circuit coolant leak. The study entails the generation of fiber agglomerates, the determination of their transport properties in single and multi-effect experiments and the long-term effects that particles formed due to corrosion of metallic containment internals by the coolant medium have on the strainer pressure drop. The focus of this presentation is on the numerical models that are used to predict the transport of MWFA by CFD simulations. A number of pseudo-continuous dispersed phases of spherical wetted agglomerates can represent the MWFA. The size, density, the relative viscosity of the fluid-fiber agglomerate mixture and the turbulent dispersion all affect how the fiber agglomerates are transported. In the cases described here, the size is kept constant while the density is modified. This definition affects both the terminal velocity and volume fraction of the dispersed phases. Only one of the single effect experimental scenarios is described here that are used in validation of the numerical models. The scenario examines the suspension and horizontal transport of the fiber agglomerates in a racetrack type channel. The corresponding experiments will be described in an accompanying presentation (see abstract of Seeliger et al.).

Laskentamalli PKL-koelaitteiston pienen vuodon kokeen simuloimiseksi Apros-ohjelmistolla

Työn tavoitteena oli tehdä Apros-laskentamalli PKL-koelaitteistosta ja testata kuinka hyvin Apros pystyy laskemaan PKL-koelaitteistolla suoritetun E2.2 pienen vuodon kokeen. Tavoitteena oli myös tarkastella boorittoman veden tulpan etenemistä pienen vuodon kokeen aikana. PKL-koelaitteisto vastaa saksalaista sähköteholtaan 1300 MW olevaa Philippsburg 2 painevesilaitosta. Koelaitteiston tilavuudet ja teho on skaalattu kertoimella 145. Työssä tehdyllä laskentamallilla tarkasteltiin boorittoman veden tulpan liikkumista pienen vuodon kokeen aikana. Kun malli oli valmis, laskenta suoritettiin Apros 5.05 versiolla. Boorittoman veden tulpan etenemisen laskennassa käytettiin toisen kertaluvun diskretointia, jolla booripitoisuuden muutokset säilyvät teräväreunaisina. Laskentamalli pystyi kuvaamaan koelaitteistolla suoritetussa pienen vuodon kokeessa tapahtuneet ilmiöt varsin hyvin. Eroa koetuloksiintuli pääkiertopiirien luonnonkiertojen alkamishetkistä ja primääripaineen käyttäytymisessä. Kokeen alkutilanne ei ollut stationääritila, joten alkutilanteen asettamisessa oli hankaluuksia. Varsinkin pääkiertopiirien veden pinnankorkeuksienasettamisessa oli vaikeuksia, koska veden pinnankorkeuksien erot pyrkivät tasoittumaan nopeasti kokeen aikana. Apros pystyi laskemaan PKL-koelaitteistolla suoritetun pienen vuodon kokeen hyvin. Mallilla tulisi kuitenkin laskea vielä toisentyyppisiäkin kokeita, ennen kuin voidaan varmuudella tietää mallin toimivuus. PKL-koelaitteisto vastaa pääpiirteiltään Suomeen rakennettavaa Olkiluoto 3 ydinvoimalaitosta. Tehty työ antaa lisävarmuutta, kun Olkiluoto 3 laitoksen turvallisuustarkasteluita tehdään.

Ydinvoimalaitoksen varalla olevien turvallisuusjärjestelmien määräaikaistestausten riittävyys ja kattavuus

Ydinvoimalaitoksen varalla olevien turvallisuusjärjestelmien tehtävänä on ehkäistä häiriö- ja onnettomuustilanteiden syntyminen sekä lieventää mahdollisen onnettomuuden seurauksia. Jotta saadaan tietoa näiden tärkeiden järjestelmien käyttökunnosta, on suoritettava riittäviä ja kattavia määräaikaistestauksia. Tutkimuksen pääkohteena ovat Olkiluodon voimalaitoksen matala- ja korkeapaineisten hätäjäähdytysjärjestelmien määräaikaistestaukset ja niiden ohjeet. Määräaikaistestauksista arvioidaan niiden kykyä havainnoida vikoja, mahdollisia vikaantumisia testauksissa, testausten taajuutta sekä vastaavuutta järjestelmien suunnitteluperusteena olevaan jäähdytteenmenetysonnettomuuteen (LOCA). Lisäksi selvitetään, mitä hyötyä testausten hajautuksilla ja diversifioinnilla on saavutettu, ja miten niitä tulisi jatkossa soveltaa. Testauksiin liittyviä ohjeita ja menettelyjä arvioidaan tarkastelemalla, täyttävätkö ne viranomaisen asettamat vaatimukset. Tulokseksi syntyi arvio järjestelmien testausten nykytilasta, joka on yleisesti ottaen hyvä. Tähän ovat vaikuttaneet testauksissa esiintyneiden puutteiden korjaaminen ja määräaikaistestausten määräajoin tapahtuvan arvioinnin kehittäminen. Vertailut LO-CA:an tuottivat tyydyttävän tuloksen, koska testausten todettiin olevan riittävän laajat ja vastaavan vuodessa kertyvien rasitusten osalta noin vuorokauden aikaista onnettomuutta lähes kaikilla laitteilla. Suositeltavaa olisi suorittaa pitkäaikaisempaa testausta apusyöttövesijärjestelmän pumpulle. Optimitestausvälin mukaisesti testausvälit ovat tällä hetkellä riittävän tiheät, ja muutamia testauksia pitäisi jopa harventaa. Hajautuksilla on saavutettu huomattava riskin väheneminen, ja nykyisin hajautusta sovelletaan hätäjäähdytysjärjestelmissä laajasti. Joistakin mittalaitteiden testauksista hajautus vielä puuttuu, joten näihin se olisi suositeltavaa lisätä. Järjestelmien testausten diversifiointi on nykyisellään riittävää.

OL1/OL2-suojarakennuksen jäähdytteenmenetysonnettomuusanalyysit APROS 5.09 -ohjelmistolla

Tässä työssä on tutkittu OL1/OL2-suojarakennuksen käyttäytymistä jäähdytteenmenetysonnettomuuden eli LOCA:n aikana. Onnettomuuden simulointiin on kehitetty suojarakennusmalli suomalaiseen APROS 5.09 - ohjelmistoon sisältyvällä LP-koodilla (Lumped Parameter Code). Työssä on keskitytty suojarakennuksen kannalta oleellisimpien suureiden: kaasutilavuuksien paineen sekä lämpötilan ja lauhdutusaltaan lämpötilan ja pinnankorkeuden ajalliseen käyttäytymiseen. Mallinnetut LOCA:t ovat päähöyrylinjan ja sammutetun reaktorin jäähdytysjärjestelmän putkikatkoksia. Simulointeja on tehty laitoksen täyden tehon ja kuumavalmiuden lähtötiloissa ja tarkasteltavien suureiden käyttäytymistä on tutkittu erikseen valituilla konservatiivisilla oletuksilla. Laskentatapaukset rajoittuvat lyhyeen aikaväliin 27.78 tuntia kuvitellusta putkirikosta eteenpäin. Tuloksia on verrattu OL1/OL2- laitostoimittajan, Westinghouse Electric Sweden AB:n, tekemiin lisensiointianalyyseihin. Työssä on myös kuvattu APROS-laskennan epävarmuustekijöitä. Suojarakennuksen on todettu käyttäytyvän fysikaalisesti yhtenevästi lisensiointianalyyseissä ja APROS:lla tehdyissä vertailulaskuissa.