Dissertations / Theses on the topic 'Nuclear reactor vessel'

Thesis (S.M. and S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2009.
Cataloged from PDF version of thesis.
Includes bibliographical references (p. 92).
Many desirable locations for Pebble Bed Modular Reactors are currently out of consideration as construction sites for current designs due to limitations on the mode of transportation for large RPVs. In particular, the PBMR-400 design developed by PBMR Pty of South Africa uses an RPV with an outer diameter of 6.4 meters. Since current SCHNABEL railcars can only haul components up to 4.3 meters wide, the only other possibility for transport is by barge, which limits construction to sites accessible by river, lake or coast. Designing a PBMR with a core able to fit within an RPV able to be transported by railcar would be extremely valuable, especially for potential inland sites only accessible by railway, such as those in the Canadian Oil Sands at which the PBMR would be utilized for oil extraction processes. Therefore, a study was conducted to determine the feasibility of a Pebble Bed Modular Reactor design operating at 250 MWth with a core restricted to fitting inside an RPV with an outer diameter of 4.3 meters. After reviewing the performance of various core configurations satisfying this constraint, an optimized PBMR design operating at this power was found. This new design uses the same fuel management scheme as the PBMR 400, as well as similar inlet and outlet coolant temperatures. This MPBR-250 design includes a pebble bed with an outer diameter of 2.7 meters, an outer reflector 50 cm thick and 12.5% enriched fuel. A mixture of graphite pebbles of 11.7% is also included in the pebble bed to produce an equilibrium core with minimal excess reactivity.
(cont.) This thesis shows that the MPBR-250 can perform up to the standards of the PBMR-400 design with respect to power peaking factors, peak temperatures and RPV fast fluences and can also increase fuel burnup to nearly 110 GWd/T. In addition, the MPBR-250 is a much more agile design, able to be deployed at a wider variety of locations because its RPV can be transported by railcar.
by Matthew S. Everson.
S.M.and S.B.

This master thesis has been performed by request of Dekra Industrial AB. Dekra Industrial AB is a Swedish subsidiary company of the German company Dekra and works for example with safety inspections within the nuclear power industry. The inspections performed by the company are often non-destructive testing (NDT) such as visual inspections of nuclear reactor pressure vessels. The inspection methods used today are considered to be further developed and there is a strong demand of improving the visual inspection. 3D-Imaging techniques are starting to be used as a measuring tool within the industry and could be a potential aid tool for the visual inspection. The purpose with the master thesis is to gain an understanding of 3D-Imaging technique to propose a suitable implementation so that it may be used as an aid tool for visual inspection. The main goal with the master thesis work is to gain knowledge about 3D-Imaging techniques and propose an implementation which may be used in the nuclear power industry. However there are different types of techniques and all of them have advantages and disadvantages. The method began with a comprehensive study about 3D-Imaging techniques, optics of 3D-Imaging and behaviour of electronics in radioactive environment. Information that could not be acquired by literature alone is acquired by interviews and meetings. The chosen 3D-Imaging technique that was considered to be the most suitable was structural light. Structural light is built on a triangulation principle that uses a projector and a camera for acquiring 3D coordinates. By using patterns displayed by the projector onto the object the camera may detect the reflected patterns and thus creating 3D coordinates. A structural light system was built and tested. The main test consisted about a two-level factorial design. The tested factors were triangulation angle, brightness and measurement distance. The test run that had the largest triangulation angle, highest brightness and shortest measurement distance gave the best accuracy. The accuracy was determined by measuring the flatness of the object. The best accuracy was measured to 91.5 μm. Besides the accuracy the technique has proven its potential by being able to scan weld tests and reconstruct well defined point clouds of the weld profiles. In conclusion the goal of the master thesis was reached and the demanded accuracy was reached. The accuracy is comparable with some industrial systems available today. This was possible due to use of a high resolution still camera. Since the camera and projector are commercially available products the tests proves that there is room for further improvements in order to reach better and a more robust accuracy. Keywords: Dekra Industrial AB, Visual testing, Imaging technique, Structural light
Detta examensarbete har utförts på uppdrag av Dekra Industrial AB. Dekra Industrial AB är ett dotterbolag till Dekra. Dekra Industrial AB arbetar främst med kontroller och provningar inom industrin. Kärnkraftindustrin är en industrigren där DEKRA arbetar med sådan kontroll Inspektionerna som utförs består huvudsakligen av oförstörande provning såsom visuell provning. Metoderna som används idag behöver vidareutvecklas och det finns en stark efterfrågan att förbättra den visuella inspektionen. 3D-avbildningsteknik är allt vanligare inom industrin idag och skulle kunna användas som ett mäthjälpmedel för att komplettera den visuella inspektionen. Syftet med examensarbetet är att få en förståelse för hur väl tekniken fungerar samt att föreslå en tillämpning där den kan komma att användas som ett komplement till den visuella inspektionen. Målet med arbetet är att ta fram underlag och föreslå en tillämpning för provning i högstrålande miljö. 3D-avbildningsteknik är ett generellt namn för många olika typer av tekniker som har sina fördelar respektive nackdelar. Arbetet inleds med en litteraturstudie kring 3D-avbildningstekniker, fysik med avseende på avbildningsteknik, den visuella proceduren idag samt hur elektronik påverkas av högstrålande miljö. Information som inte kan fås via studier inhämtas via intervjuer och möten. Tekniken som valdes att analyseras var strukturerat ljus. Tekniken bygger på en trianguleringsprincip som använder en projektor och kamera för att tillförskaffa 3D-koordinater. Genom att projicera mönster på ett objekt kan kameran detektera det reflekterade mönstret och på så vis skapa 3D koordinater. Ett strukturerat ljus system ställdes upp och testades. Testet bestod huvudsakligen av en försöksplanering där de testade faktorerna var trianguleringsvinkel, ljusstyrka och mätavstånd. Testuppställningen som gav bäst resultat var med störst trianguleringsvinkel, högsta ljusstyrka samt kortast mätavstånd. Noggrannheten bestämdes genom att mäta planheten på objektet. Den bästa noggrannheten som uppnåddes med testet var 91.5 μm. Förutom den goda noggrannheten har tekniken visat sin potential genom att avbilda ett svetsprov som genererade ett väldefinierat punktmoln av svetsprofilen. Sammanfattningsvis uppfylldes målen och det uppställda systemet gav en noggrannhet som är jämförbar med en del system ute på marknaden. Detta var möjligt på grund av att en högupplöst stillbildskamera användes. Det finns potential för förbättringar då komponenterna som används i systemet är kommersiella produkter. Nyckelord: Dekra Industrial AB, Visuell inspektion, Avbildningsteknik, Strukturerat ljus

Thesis (Ph.D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2001.
Includes bibliographical references (p. 357-366).
The feasibility of a lead-bismuth (Pb-Bi) cooled fast reactor that eliminates the need for steam generators and coolant pumps was explored. The working steam is generated by direct contact vaporization of water and liquid metal in the chimney above the core and then is sent to the turbine. The presence of a lighter fluid in the chimney drives the natural circulation of the Pb- Bi within the reactor pool. Three key technical issues were addressed: 1) the maximum thermal power removable by direct contact heat transfer without violating the fuel, clad and vessel temperature limits, 2) the consequences of Pb-Bi aerosol transport on the design and operation of the turbine and 3) the release of radioactive polonium (a product of coolant activation) to the steam. Modeling of the multi-phase phenomena occurring in the chimney confirmed the effectiveness of the direct contact heat transfer mode within a well-defined design envelope for the reactor power, chimney height and steam superheat. A 1260MWth power is found possible for 10m chimney height and 25°C superheat. The temperature of the low-nickel steel clad is maintained below 600°C, which results in limited corrosion if tight control of the coolant oxygen concentration is adopted.
Generation, transport and deposition of Pb-Bi aerosols were also modeled. It was found that the design of a chevron steam separator reduces the heavy liquid metal in the steam lines by about three orders of magnitude. Nevertheless, the residual Pb-Bi is predicted to cause embrittlement of the turbine blades. Four solutions to this problem were assessed: blade coating, employment of alternative materials, electrostatic precipitation and oxidation of the Pb-Bi droplets. An experimental campaign was conducted to investigate the polonium release from a hot Pb- Bi bath to a gas-streamn. Th thermodynamics of the polonium hydride formation reaction (free- energy vs. temperature). as welQ as the vapor pressure of the lead-polonide were measured and then utilized to model the polonium transport in the reactor. It was found that the polonium concentration in the steam and on the surface of the power cycle components is significantly above the acceptable limits, which makes the very concept of a direct contact reactor open to question.
by Jacopo Buongiorno.
Ph.D.

The investigation of reactor pressure vessel (RPV) material from the decommissioned Greifswald nuclear power plant representing the first generation of Russian-type WWER-440/V-230 reactors offers the opportunity to evaluate the real toughness response. The Greifswald RPVs of 4 units represent different material conditions as follows: • Irradiated (Unit 4), • irradiated and recovery annealed (Units 2 and 3), and • irradiated, recovery annealed and re-irradiated (Unit1). The recovery annealing of the RPV was performed at a temperature of 475° for about 152 hours and included a region covering ±0.70 m above and below the core beltline welding seam. Material samples of a diameter of 119 mm called trepans were extracted from the RPV walls. The research program is focused on the characterisation of the RPV steels (base and weld metal) across the thickness of the RPV wall. This report presents test results measured on the trepans from the beltline welding seam No. SN0.1.4. and forged base metal ring No. 0.3.1. of the Units 1 2 and 4 RPVs. The key part of the testing is focussed on the determination of the reference temperature T0 of the Master Curve (MC) approach following the ASTM standard E1921 to determine the facture toughness, and how it degrades under neutron irradiation and is recovered by thermal annealing. Other than that the mentioned test results include Charpy-V and tensile test results. Following results have been determined: • The mitigation of the neutron embrittlement of the weld and base metal by recovery annealing could be confirmed. • KJc values of the weld metals generally followed the course of the MC though with a large scatter. • There was a large variation in the T0 values evaluated across the thickness of the multilayered welding seams. • The T0 measured on T-S oriented SE(B) specimens from different thickness locations of the welding seams strongly depended on the intrinsic structure along the crack front. • The reference temperature RT0 determined according to the “Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs - VERLIFE” and the fracture toughness lower bound curve based thereon are applicable on the investigated weld metals. • A strong scatter of the fracture toughness KJc values of the recovery annealed and re-irradiated and the irradiated base metal of Unit 1 and 4, respectively is observed with clearly more than 2% of the values below the MC for 2% fracture probability. The application of the multimodal MC-based approach was more suitable and described the temperature dependence of the KJc values in a satisfactory manner. • It was demonstrated that T0 evaluated according to the SINTAP MC extension represented the brittle fraction of the data sets and is therefore suitable for the nonhomogeneous base metal. • The efficiency of the large-scale thermal annealing of the Greifswald WWER 440/V230 Unit 1 and 2 RPVs could be confirmed.

Formation of a high density of Mn-Ni-Si nanoscale precipitates in irradiated reactor pressure vessel steels could lead to severe, unexpected embrittlement, which may limit the lifetimes of our nation’s light water reactors. While the existence of these precipitates was hypothesized over 20 years ago, they are currently not included in embrittlement prediction models used by the Nuclear Regulatory Commission. This work aims to investigate the mechanisms and variables that control Mn-Ni-Si precipitate (MNSP) formation as well as correlate their formation with hardening and embrittlement.

A series of RPV model steels with systematic variations in Cu and Ni contents, two variables that have been shown to have a dominant effect on hardening, were irradiated in a series of test reactor and power reactor surveillance irradiations. Atom probe tomography (APT) measurements show that large volume fractions (f_v) of MNSPs form in all the steels irradiated at high fluence, even those containing no added Cu, which were previously believed to have low sensitivity to embrittlement. It is demonstrated that while Cu enhances the rate of MNSP formation, it does not appear to significantly alter their saturation f_v or composition. The high fluence MNSPs have compositions consistent with known intermetallic phases in the Mn-Ni-Si system and have f_v very near those predicted by equilibrium thermodynamic models. In addition, X-ray diffraction experiments by collaborators shows that these precipitates also have the expected crystal structure of the predicted Mn-Ni-Si phases.

Post irradiation annealing experiments are used to measure the hardness recovery at various temperatures as well as to determine if the large f _v of MNSPs that form under high fluence neutron irradiation are thermodynamically stable phases or non-equilibrium solute clusters, enhanced or induced by irradiation, respectively. Notably, while post irradiation annealing of a Cu-free, high Ni steel at 425°C results in dissolution of most precipitates, a few larger MNSPs appear to remain stable and may begin to coarsen after long times. A cluster dynamics model rationalizes the dissolution and reduction in precipitate number density, since most are less than the critical radius at the annealing temperature and decomposed matrix composition. The stability of larger precipitates suggests that they are an equilibrium phase, consistent with thermodynamic models.

Charged particle irradiations using Fe³⁺ ions are also used to investigate the precipitates which form under irradiation. Two steels irradiated to a dose of 0.2 dpa using both neutrons and ions show precipitates with very similar compositions. The ion irradiation shows a smaller f _v, likely due to the much higher dose rate, which has been previously shown to delay precipitation to higher fluences. While the precipitates in the ion irradiated condition are slightly deficient in Mn and enriched in Ni and Si compared to neutron irradiated condition, the overall similarities between the two conditions suggest that ion irradiations can be a very useful tool to study the susceptibility of a given steel to irradiation embrittlement.

Finally, the large f_v of MNSPs that are shown to form in all steels, including those low in Cu, at high fluence, even those without added Cu, result in large amounts of hardening and embrittlement. A preliminary embrittlement prediction model, which incorporates MNSPs at high fluence, is presented, along with results from a recent test reactor irradiation to fluences representative of extended lifetimes. This model shows very good agreement with the data.

In this work a cooling system connected to a reactor pressure vessel has been studied using the CFD method for the purpose of investigating the strengths and shortcomings of using CFD as a tool in similar fluid flow problems within nuclear power plants. The cooling system is used to transport water of 288K (15°C) into a nuclear reactor vessel filled with water of about 555K (282°C) during certain operating scenarios. After the system has been used, the warm water inside the vessel will be carried into the cooling system by buoyancy forces. It was of interest to investigate how quickly the warm water moves into the cooling system and how the temperature field of the water changes over time. Using the open source CFD code OpenFOAM 2.3.x and the LES turbulence modelling method, a certain operating scenario of the cooling system was simulated. A simplified computational domain was created to represent the geometries of the downcomer region within the reactor pressure vessel and the pipe structure of the cooling system. Boundary conditions and other domain properties were chosen and motivated to represent the real scenario as good as possible. For the geometry, four computational grids of different sizes and design were generated. Three of these were generated using the ANSA pre-processing tool, and they all have the same general structure only with different cell sizes. The fourth grid was made by the OpenFOAM application snappyHexMesh, which automatically creates the volume mesh with little user input. It was found that for the case at hand, the different computational grids produced roughly the same results despite the number of cells ranging from 0,14M to 3,2M. A major difference between the simulations was the maximum size of the time steps which ranged from 0,3ms for the finest ANSA mesh to 2ms for the snappy mesh, a difference which has a large impact on the total time consumption of the simulations. Furthermore, a comparison of the CFD results was made with those of a simpler 1D thermal hydraulic code, Relap5. The difference in time consumption between the two analyses were of course large and it was found that although the CFD analysis provided more detailed information about the flow field, the cheaper 1D analysis managed to capture the important phenomena for this particular case. However, it cannot be guaranteed that the 1D analysis is sufficient for all similar flow scenarios as it may not always be able to sufficiently capture phenomena such as thermal shocks and sharp temperature gradients in the fluid. Regardless of whether the CFD method or a simpler analysis is used, conservativeness in the flow simulation results needs to be ensured. If the simplifications introduced in the computational models cannot be proved to always give conservative results, the final simulation results need to be modified to ensure conservativeness although no such modifications were made in this work.

Maintaining the safe operation of nuclear power plants (NPPs) is crucial. This requires fully understanding the mechanism of long term irradiation and thermal ageing, as well as their effects, on components including the reactor pressure vessel (RPV). The research community is collecting data that will be required to support the case for extending the operation of western-type NPPs beyond that of 60 years. One of the current dilemmas faced by the long-term operation of RPVs is the formation of nanometre scale precipitates. These precipitates are known to cause embrittlement where it increases the ductile-to-brittle transition temperature of the RPV steels. The chemistry of these precipitates is strongly dependent on the chemistry of the RPV steels. In general, these precipitates can be categorised into two types, copper-rich precipitates (CRPs) and manganese-nickel (-enriched) precipitates (MNPs) [1, 2]. The concentration of copper in the precipitates depends on the bulk content of the steel [3]. The formation mechanism of the precipitates under neutron irradiation and thermal ageing, and their influence on material degradation at high neutron fluence (Φt), is still unclear. To understand the long term precipitation under irradiation and thermal ageing, high nickel and copper containing RPV steels with a similar microstructure an chemical composition as those currently in service were subjected to either neutron irradiation (to high neutron fluences, Φt ≥ 5 x 10²³ neutrons.m^-2) or thermal ageing (for as long as ≈ 50,000 hours). CRPs and MNPs were both detected. The co-precipitation of the CRPs and MNPs were observed in thermally aged steels. The development of crystal structures in the CRPs is believed to be dependent on the size of the precipitates and the ambient temperature. When the CRPs reached a critical size, they underwent the martensitic transformation from BCC→9R→3R→FCC or FCT. The CRPs preferentially nucleate heterogeneously at the dislocation lines. Chemical analysis suggests that most of the CRPs are iron free. Under thermal ageing, the MNPs were found to precipitate at the interface of the CRPs and the matrix. These MNPs are found to be iron free too. Larger MNPs were often found to be at CPRs that were associated with dislocation lines. Also, based on the volume fraction observed, it is possible to suggest that the kinetics of nucleation and growth of the MNPs are relatively slow compared to the CRPs. This is in good agreement with the simulations reported in Refs. [4, 5]. It is the first time the MNPs are directly imaged from neutron irradiation low copper steels using electron microscopy. These irradiation-induced MNPs are densely populated in the neutron irradiated samples. It was found that the irradiation-induced MNPs are more sensitive to electron beams. It was thought that this was due to a relatively large amount of point defects present in the irradiation-induced MNPs.

Radiation damage occurs in reactor pressure vessel (RPV) steel, causing microstructural changes such as point defect clusters, interstitial loops, vacancy-solute clusters, and precipitates, that cause material embrittlement. Radiation damage is a crucial concern in the nuclear industry since many nuclear plants throughout the US are entering the first period of life extension and older plants are currently undergoing assessment of technical basis to operate beyond 60 years. The result of extended operation is that the RPV and other components will be exposed to higher levels of neutron radiation than they were originally designed to withstand. There is currently no nondestructive evaluation technique that can unambiguously assess the amount of radiation damage in RPV steels. Nonlinear ultrasound (NLU) is a nondestructive evaluation technique that is sensitive to microstructural features such as dislocations, precipitates, and their interactions in metallic materials. The physical effect monitored by NLU is the generation of higher harmonic frequencies in an initially monochromatic ultrasonic wave, arising from the interaction of the ultrasonic wave with microstructural features. This effect is quantified with the measurable acoustic nonlinearity parameter, beta. In this work, nonlinear ultrasound is used to characterize radiation damage in reactor pressure vessel steels over a range of fluence levels, irradiation temperatures, and material composition. Experimental results are presented and interpreted with newly developed analytical models that combine different irradiation-induced microstructural contributions to the acoustic nonlinearity parameter.

Les cuves des reacteurs a eau pressurisee se fragilisent sous irradiation neutronique. L'etude des mecanismes responsables donne des informations sur la duree de vie de ces cuves, ainsi que le choix des aciers constitutifs. De nombreuses interpretations font intervenir la precipitation sous irradiation, ou la formation de grosses cavites. Pour notre part, nous avons utilise la technique consistant a mesurer la duree de vie de positions injectes dans les materiaux soumis a l'etude. Cette technique est interessante pour traiter ce probleme, car les positions sont attires et pieges aux espaces vides de l'echantillon. Dans les cas favorables, elle permet l'evaluation de la taille et la concentration des espaces vides, dans une gamme bien inferieure a la limite de detection des meilleurs microscopes. Ainsi, la technique des positions excelle dans une gamme de concentration de 0. 1 a 100 lacunes par million d'atomes, et pour des tailles de cavites comprises entre 0. 5 et 50 volumes atomiques. Les irradiations etaient simulees par une dose standard de 10#1#9 electrons par cm#-#2, a 27, 150 et 288c, obtenus avec l'accelerateur d'electrons de 3 mev du ceng. Nos resultats different notablement des publications des autres ecoles, neanmoins nous precisons que nos resultats sont valables sous irradiation electronique. Nous avons mis en evidence que le phenomene essentiel n'est pas la precipitation sous irradiation, mais le stockage transitoire d'une forte concentration lacunaire par piegeage aux impuretes cu, c ou n. La liberation de celle-ci formerait de petites cavites de 6 a 10 lacunes, qui par elles memes ne semblent pas responsables de la fragilisation. Mais ces cavites grossissent dangereusement (taille superieures a 50 volumes atomiques), si des dislocations non decorees sont presentes. De telles dislocations peuvent etre introduites fortuitement par un ecrouissage accidentel et une temperature d'irradiation legerement inferieure a la temperature de fonctionnement, a savoir 288c. (les dislocations resultant des traitements thermiques et presentes en grand nombre dans les alliages industriels n'ont pas d'influence car elles sont fortement decorees par du carbone. ) l'azote exerce une action importante. Son etude est difficile car son action est souvent confondue avec celle du carbone. Nous avons dispose d'echantillons de fen, avec une teneur en carbone inferieure a 10#-#5 c/at. , ou l'azote avait ete introduit en phase liquide. Dans ce cas, l'azote stabilise considerablement le reseau de dislocations et favorise sous irradiation la germination de cavites de grandes tailles

This master‘s thesis deals with radiation damage of reactor pressure vessels, specifically of NPP Dukovany Unit No. 3. In general damage mechanisms of reactor steels are described and possibilities of monitoring of material degradation and its recovery used at NPP’s all over the world are mentioned as well. A practical part of the thesis is focused on interpretation of analyses carried out with the assistance of MOBY DICK code. The ground of these analyses is a neutron fluence value development on different locations of RPV for the whole life of operation up to 24th cycle. The analyses results are put into context with performed in-service inspections. The thesis follows up with neutron fluence computation for the future cycles containing new types of nuclear fuel up to 34th cycle. The outcome of practical part of the master‘s thesis is a comparison between new types of nuclear fuel with respect to radiation damage of RPV’s.

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Tese(doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering; and, (S.M.)--Massachusetts Institute of Technology, Engineering Systems Division, Technology and Policy Program, 2007.
Includes bibliographical references (p. 126-130).
Nanofluids at very low concentrations experimentally exhibit a substantial increase in Critical Heat Flux (CHF) compared to water. The use of a nanofluid in the In-Vessel Retention (IVR) severe accident management strategy, employed in Advanced Light Water Reactors, was investigated. A model simulating the two-phase flow and heat transfer on the reactor vessel outer surface quantified the increase in decay power that can be removed using a nanofluid, predicting that the use of a nanofluid will allow a stable operating power ~40% greater than the power allowable using water to be achieved, while holding the Minimum Departure from Nucleate Boiling Ratio (MDNBR) constant. A nanofluid injection system that would take advantage of the enhanced CHF properties of the nanofluid in order to provide a higher safety margin than the current IVR strategy or, for given margin, enable IVR at higher core power, is proposed. A risk-informed analysis has revealed that this injection system has a reasonably high success probability of 0.99, comparable to the success probability without the injection system. Potential regulatory, environmental, and health risk issues were analyzed, and it was concluded that the current regulatory regimes are adequate for ensuring that the implementation of nanofluids in IVR will not endanger public health and safety. However, experimental verification of nanofluid CHF enhancement at prototypical IVR conditions and periodic nanofluid property testing as a surveillance requirement are needed to reduce the key uncertainties related to nanofluid performance. Finally, a periodic review of the health and environmental risks of nanofluids and, if necessary,follow-up research are ecommended to ensure the health of the public and environment.
by Ryan Christopher Hannink.
S.M.

Dans les structures de grandes dimensions réalisées en acier 16mnd5 et revêtues par soudage àa l'arc avec deux couches d'acier austénitique, des décohésions intergranulaires peuvent être présentes dans le métal de base. Pour pouvoir garantir l'intégrité de la structure, p. Ex. De la cuve dans les centrales nucléaires, on cherche à savoir pourquoi ces défauts se forment, et dans quelles conditions accidentelles, une rupture devient possible. Lors d'une rupture déclenchée à partir d'une DIDR, deux mécanismes se superposent. D'une part, la petite taille de ces défauts diminue la température de la transition de la ténacité, et d'autre part, la fragilité intergranulaire est caracterisée par des valeurs de ténacité plus faibles qu'en clivage. Des essais sur des éprouvettes contenant des DIDR confirment avec 36mpam à -90 C la faible ténacité associée à la rupture intergranulaire. En revanche, pour des températures plus élevées, c'est l'effet de taille qui abaisse la température de transition. Pour la rupture en clivage, on observe une diminution de la température de transition de 60 C pour une réduction du rapport A/W de 0. 5 à 0. 1. Dus à la perte de confinement de la plasticité à la pointe de la fissure, le volume plastifié et la ténacité sont augmentes. L'approche locale avec m=29. 8, =0. 5, #u=2460mpa permet de prédire la transition pour des éprouvettes renfermant des petites ou des grandes fissures. Les facies des décohésions intergranulaires provoquées lors du réchauffage (DIDR) sont caracterisés par un taux élevé de soufre qui diminue en s'éloignant de la ligne de fusion. Fragilisant les joints de grains, les contraintes résiduelles sont partiellement relaxées par fissuration. Il en résulte un facies intergranulaire lisse sans aucune trace de déformation. En revanche, la forme finale du défaut est due à un mécanisme combiné de micro-fluage et de fragilisation par du soufre ségrége qui conduit à la formation de facettes intergranulaires cavitées

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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

In a nuclear power plant, the last barrier under normal and accident operations is the containment building. This is normally constructed from concrete reinforced with steel bars, which are prestressed to enhance the overall capability to withstand thermodynamic stresses like over-pressurisation and high temperatures. The failure of this final barrier will lead to the release of radioactivity to the surrounding environment. To examine the effects of thermo-hydraulic stresses on PWR containment following a LOCA, a model is proposed with simulated scenarios performed at the Koeberg Nuclear Power Station as a case study. The accidents were simulated using the Koeberg engineering simulator to obtain the output data. The scenario for the proposed model correlates the critical mass flow from a double-ended guillotine break to the containment pressure and temperature increase. Different containment filtered venting systems (CFVS) are also investigated in this study as severe accident management systems. CFVS have historically been included in boiling water reactor (BWR) designs, but following the Fukushima Daiichi nuclear accident, they are being introduced as severe accident management systems to manage the threat of containment over-pressurisation in pressurised water reactors (PWR). Finally, the rate of change in containment pressure and temperature is analysed and compared to literature, with the incorporation of a simulated filtered venting system to the PWR containment building.

Durante um acidente em uma usina nuclear, a integridade do vaso pressurizado deve ser assegurada. Em resposta a um possível derretimento do combustível nuclear, a atual geração de usinas possui um sistema para a injeção de água potável na cavidade do vaso pressurizado com intuito de resfriar sua parede, prevenindo danos a sua estrutura e evitando o vazamento de material radioativo. Esse estudo considerou o uso de água marinha como refrigerante para inundar a cavidade do vaso pressurizado combinado com a fixação de um estrutura porosa em forma de grade em sua parede externa como meio de aprimorar a margem de segurança durante a contenção de acidentes. Experimentos de longa duração para a ebulição em piscina de água marinha artificial foram conduzidos em uma superfície circular de cobre plana com 30 mm de diâmetro. Foi encontrado um fluxo de calor crítico de 1; 6 MW/m2 sob pressão atmosférica. Esse valor é significantemente maior que aquele obtido (1; 0 MW/m2) nas mesmas condições experimentais. Foi verificado que os depósitos de sais marinhos podem aumentar a molhabilidade e a capilaridade da superfície de teste, aprimorando assim o fluxo crítico. Combinando a água marinha e a fixação da estrutura porosa sobre a superfície de teste, verificou-se um melhora no coeficiente de transmissão de calor e no fluxo de calor crítico de até 110 % (2; 1 MW/m2), quando comparado a água destilada na superfície limpa, sem a instalação da estrutura. Após os experimentos, foi identificado que muitos dos poros presentes nas superfícies da estrutura porosa encontravam-se bloqueados devido ao aglutinamento de sais marinhos. Isso levou a conclusão que o aumento no valor do fluxo crítico observado para a água marinha artificial ocorreu devido, principalmente, a separação das fases líquida e gasosa do fluido na região próxima a superfície de teste, efeito proporcionado pela forma de grade da estrutura porosa, e ao aumento da molhabilidade e capilaridade da superfície devido a formação dos depósitos marinhos.
During a severe nuclear power plant accident, the integrity of the reactor pressure vessel must be assured. In response to a possible fuel meltdown, operators of the current generation of nuclear power plants are likely to inject water into the reactor pressure vessel to cool down the reactor vessel wall, preserving its integrity and avoiding leakage of radioactive material. This study considers the use of seawater to flood a reactor pressure vessel combined with the attachment of a honeycomb porous plate (HPP) on the vessel outer wall as a way to improve the safety margins for in-vessel retention of fuel. In long-duration experiments, saturated pool boiling of artificial seawater was performed with an upward-facing plain copper heated surface 30 mm in diameter. The resulting value for critical heat flux (CHF) was 1; 6 MW/m2 at atmospheric pressure, a value significantly higher than the CHF obtained when the working fluid was distilled water (1; 0 MW/m2). It was verified that sea-salt deposits could greatly improve surface wettability and capillarity, enhancing the CHF. The combination of artificial seawater and an HPP attached to the heated surface improved the boiling heat transfer coefficient and increased the CHF up to 110% (2; 1 MW/m2) as compared to distilled water on a bare surface. After the artificial seawater experiments, most of the wall micropores of the HPP were clogged because of sea-salt aggregation on the HPP top and bottom surfaces. Thus, the CHF enhancement observed in this case was attributed mainly to the separation of liquid and vapor phases provided by the HPP channel structure and improvement of surface wettability and capillarity by sea-salt deposition.

The internationally circulated Pool Critical Assembly (PCA) Pressure Vessel Benchmark was analyzed using the PENTRAN Parallel SN code system for the geometry, material, and source specifications as described in the PCA Benchmark documentation. Improvements to the benchmark are proposed through the application of more representative flux and volume weighted homogenized cross sections for the PCA reactor core, which were obtained from a rigorous heterogeneous modeling of all fuel assembly types in the core. A new source term definition is also proposed based on calculated relative power in each core fuel assembly with a spectrum based on the Uranium-235 fission spectra. This research focused on utilizing the BUGLE-96 cross section library and accompanying reaction rates, while examining both adaptive differencing on a coarse mesh basis, as well as the sole use of Directional Theta-Weighted (DTW) SN differencing scheme in order to compare the calculated PENTRAN results to measured data. The results show good comparison with the measured data, which suggests PENTRAN is a viable and reliable code system for calculation of light water reactor neutron shielding and dosimetry calculations. Furthermore, the improvements to the benchmark methodology resulting from this work provide a 6 percent increase in accuracy of the calculation (based on the average of all calculation points), when compared with experimentally measured results at the same spatial location in the PCA pressure vessel simulator.

Oregon State University has conducted research in collaboration with TerraPower, LLC, to perform a top-down scaling analysis of an integrated test facility. The goal of this facility is to simulate transient and quasi-steady phenomena at a reduced scale, including steady-state operation, pump coastdown, natural circulation, reactor head heat transfer, and coolant stratification. To support this goal, this thesis presents the methodology and analysis by which approximate facility dimensions were generated. This analysis includes implementation of the hierarchical two-tiered scaling methodology, as outlined by the Nuclear Regulatory Commission and optimization through the general reduced gradient methodology.
Graduation date: 2013

碩士
國立臺北科技大學
工業工程與管理研究所
97
The RPV(Reactor Pressure Vessel) water level is one of the most critical monitoring parameters in nuclear plant operation, however, the two RPV water level displays in the Lungmen Nuclear Power Plant: Wide Display Panel (WDP) and Video Display Unit (VDU) SDPS C92、C93 displays both are unable to effectively provide operators the RPV water level information; it may also cause the misread of the water level. Therefore, a new RPV water level displays which can assist operators to effectively monitor the water level is essential. In the concerns of feasibility and contribution, this research aim to redesign the VDU SDPS C92、C93 displays on the prerequisite of maintaining the data content. In addition, based on an interface display design theory of EID (Ecological Interface Design), this research combines document research and interview approach in order to elicit the operators’ actual needs of the RPV water level information. The newly designed RPV water level display was evaluated by 17 operators in the Lungmen Nuclear Power Plant by two kinds of questionnaires. Use revised SA (SWORD) questionnaire to evaluate workloads and situation awareness and revised SUS questionnaires to evaluate usability. The research has shown that new RPV water level displays are able to reduce human errors, enhance operator’s performance of and further improve the safety of nuclear plant.

碩士
國立臺灣大學
機械工程學研究所
99
Nowadays, we are facing problems of global energy shortage as well as the need of environmental protection. The advantage of low cost and small amount of CO2 discharge makes nuclear power an important choice for energy. However, the safety of structures, systems and mechanical components employed in a nuclear power plant has to be assured before a plant can be constructed. One of the most important pressure boundary components in the steam supply system of a nuclear power plant is the reactor pressure vessel (RPV). It is welded together by several steel plates. Cracks occur more frequently in welds rather than in base plates of a RPV. When a predominant crack grows along with operating time to a certain size, it may result in brittle fracture in the weld of a RPV. It has been pointed out that clad thickness and crack size affect the embrittlement and fracture of the weld. The present study employs a probabilistic fracture mechanics approach by taking into account radiation embrittlement to find fracture-failure probabilities of RPV welds. The result shows that, when the clad is thicker than 0.35 inch, the failure probability at axial weld should be paid more attention to. As for the effect of inspection and repair, it is found that adopting a more advanced inspection instrument reduces failure probability more than increasing inspection cycles or covering more inspection areas. It is also found the probability of failure at circumferential welds is smaller than that at axial welds. The finding reassures the proposition made by the United States Nuclear Regulatory Commission (USNRC) that inspection of circumferential welds of a RPV can be exempted.