L’étude des conséquences de la réaction sodium-eau (RSE) est un enjeu dans le cadre de la sûreté des futurs réacteurs à neutrons rapides à caloporteur sodium. Afin d'évaluer les conséquences de RSE dans des situations d'accident majeur, il est nécessaire de mieux comprendre la phénoménologie et notamment la quantité d'énergie libérée et la cinétique de libération. L'objectif est donc d’améliorer la compréhension de telles RSE pour prédire au mieux ses conséquences sur les équipements mécaniques alentours. Trois axes de travail ont été privilégiés, à savoir la recherche du déroulement des séquences accidentelles, un examen expérimental paramétrique, et une analyse de la phénoménologie avant le contact explosif. Dans un premier temps, une méthode arborescente d'analyse de risques a été croisée avec des méthodes de calcul d'effets. Cette analyse a permis d’imaginer comment le contact peut s'effectuer. Des études expérimentales démonstratives de l'influence du mode de mise en contact ont ensuite été effectuées afin d’approfondir certains aspects pratiques. L’analyse des nombreuses données recueillies conduit au développement d’un modèle d'interprétation phénoménologique, intégré dans une plateforme de simulation multi-physique. Bien que de nombreuses hypothèses simplificatrices soient réalisées, la prise en compte des transferts de chaleur transitoires permet de reproduire les observations expérimentales et notamment l'influence des conditions de mélange (masse de sodium et températures initiales) sur la phénoménologie. Ce travail d'étude de la phase de pré-mélange de l'explosion sodium-eau est pertinent au regard des méthodes de prédiction des chargements sur les structures. / Study of sodium-water reaction (SWR) consequences in open air represents a challenge in the frame of safety assessments of sodium fast reactors (SFR). In case of major accident and to predict consequences of SWR, it is necessary to better appreciate phenomena and especially quantity and rate of the energy releasement. The objective is thus to strengthen the understanding of such reactions in order to predict with lore accuracy its consequences on mechanical equipment in the surroundings. This work focuses on three areas : research of accidental sequences, experimental investigation, and phenomenological analysis before the explosive contact. At first, a tree structure risk analysis with calculations of dangerous phenomena permitted to suggest how the contact between reactants may happen. Then, demonstrative experimental studies were performed to deepen some practical aspects of the phenomenology, like the influence of the way the reactants get in contact. Data analysis conducted to the development of a phenomenological model, implemented into a software platform for numerical simulations. Although numerous hypothesis, transient heat transfer consideration enables to reproduce experimental observations, especially the influence of mixing conditions (sodium mass and initial temperatures) on the phenomenology. This study of the premixing step of sodium-water explosion is relevant in the frame of current prediction methods of mechanical loadings on structures.

<p>  </p> <p>A <em>digital twin</em> (DT), which keeps track of nuclear reactor history to provide real-time predictions, has been recently proposed for nuclear reactor monitoring. A digital twin can be implemented using either a differential equations-based physics model, or a data-driven machine learning model<strong>. </strong>The principal challenge in physics model-based DT consists of achieving sufficient model fidelity to represent a complex experimental system, while the main challenge in data-driven DT appears in the extensive training requirements and potential lack of predictive ability. </p> <p>In this thesis, we investigate the performance of a hybrid approach, which is based on physics-informed neural networks (PINNs) that encode fundamental physical laws into the loss function of the neural network. In this way, PINNs establish theoretical constraints and biases to supplement measurement data and provide solution to several limitations of purely data-driven machine learning (ML) models. We develop a PINN model to solve the point kinetic equations (PKEs), which are time dependent stiff nonlinear ordinary differential equations that constitute a nuclear reactor reduced-order model under the approximation of ignoring the spatial dependence of the neutron flux. PKEs portray the kinetic behavior of the system, and this kind of approach is the basis for most analyses of reactor systems, except in cases where flux shapes are known to vary with time. This system describes the nuclear parameters such as neutron density concentration, the delayed neutron precursor density concentration and reactivity. Both neutron density and delayed neutron precursor density concentrations are the vital parameters for safety and the transient behavior of the reactor power. </p> <p>The PINN model solution of PKEs is developed to monitor a start-up transient of the Purdue University Reactor Number One (PUR-1) using experimental parameters for the reactivity feedback schedule and the neutron source. The facility under modeling, PUR-1, is a pool type small research reactor located in West Lafayette Indiana. It is an all-digital light water reactor (LWR) submerged into a deep-water pool and has a power output of 10kW. The results demonstrate strong agreement between the PINN solution and finite difference numerical solution of PKEs. We investigate PINNs performance in both data interpolation and extrapolation. </p> <p>The findings of this thesis research indicate that the PINN model achieved highest performance and lowest errors in data interpolation. In the case of extrapolation data, three different test cases were considered, the first where the extrapolation is performed in a five-seconds interval, the second where the extrapolation is performed in a 10-seconds interval, and the third where the extrapolation is performed in a 15-seconds interval. The extrapolation errors are comparable to those of interpolation predictions. Extrapolation accuracy decreases with increasing time interval.</p>

Thesis (PhD)--Stellenbosch University, 2012. / ENGLISH ABSTRACT: The objective of this research was to develop a neutron detection system that incorporates a mass spectrometer to measure high neutron flux in a nuclear reactor environment. This system consists of slow and fast neutron detector elements for measuring fluxes in those energy regions respectively. The detector should further be capable of withstanding the harsh conditions associated with a high temperature reactor. This novel detector which was initially intended for use in the PBMR reactor has possible applications as an in-core neutron and indirect temperature-monitoring device in any of the HTGR. Simulations of a generic HTGR core model were performed in order to obtain the neutron energy spectrum with emphasis on the behavior of three energy regions, slow, intermediate and fast neutrons within the core at different temperatures. The slow neutron flux which has the characteristic of a Maxwell- Boltzmann distribution were found to shift to larger values of neutron flux at higher energies as the fuel temperature increased, while fast neutron flux spectra remained relatively constant. In addition, the results of the fit of the slow neutron flux with a modified Maxwell-Boltzmann equation confirmed that in the presence of the neutron source, leakage and absorption, the effective neutron temperatures is above the medium temperatures. From these results, it was clear that the detection system will need to monitor both slow and fast neutron flux. Placing neutron detectors inside the reactor core, that are sensitive to a particular energy range of slow and fast neutrons, would thus provide information about the change of temperature in the fuel and hence act as an in-core temperature monitor. A detection mechanism was developed that employs the neutron-induced break-up reaction of 6Li and 12C into α-particles. These materials make excellent neutron converters without interference due to γ-rays, as the contributions from 6Li(γ,np)4He and 12C(γ,3α) reactions are negligible. The mass spectrometer measures the 4He partial pressure as a function of time under high vacuum with the help of pressure gradient provided by a high-vacuum turbomolecular pump and a positive-displacement fore-vacuum pump connected in series. A cryogenic trap, which contains a molecular sieve made of pellets 1.6 mm in diameter, was also designed and manufactured to remove impurities which cause a background in the lighter mass region of the spectrum. The development and testing of the high flux neutron detection system were performed at the iThemba Laboratory for Accelerator Based Sciences (LABS), South Africa. These tests were carried out with a high energy proton beam at the D-line neutron facility, and with a fast neutron beam at the neutron radiation therapy facility. To test the principle and capability of the detection system in measuring high fluxes, a high intensity 66 MeV proton beam was used to produce a large yield of α-particles. This was done because the proton inelastic scattering cross-section with 12C nuclei is similar to that of neutrons, with a threshold energy of about 8 MeV for both reactions. Secondly, the secondary fast neutrons produced from the 9Be(p,n)9B reaction were also measured with the fast neutron detector. The response of this detection system during irradiation was found to be relatively fast, with a rise time of a few seconds. This is seen as a sharp increase in the partial pressure of 4He gas as the proton or neutron beam bombards the 12C material. It was found that the production of 4He with the proton beam was directly proportional to the beam intensity. The number of 4He atoms produced per second was deduced from the partial pressure observed during the irradiation period. With a neutron beam of 1010 s−1 irradiating the detector, the deduced number of 4He atoms was 109 s−1. When irradiation stops, the partial pressure drops exponentially. This response is attributed to a small quantity of 4He trapped in the present design. Overall, the measurements of 4He partial pressure produced during the tests with proton and fast neutron beams were successful and demonstrated proof of principle of the new detection technique. It was also found that this system has no upper neutron flux detection limit; it can be even higher than 1014 n·cm−2·s−1. The lifetime of this detection system in nuclear reactor environment is practically unlimited, as determined by the known ability of stainless steel to keeps its integrity under the high radiation levels. Hence, it is concluded that this high flux neutron detection system is excellent for neutron detection in the presence of high γ-radiation level and provides real-time flux measurements. / AFRIKAANSE OPSOMMING: Die doel van hierdie navorsing was om ’n neutrondetektorstelsel te ontwikkel wat hoë neutronvloed binne in ’n kernreaktor kan meet. Die stelsel bevat twee aparte detektorelemente sodat die termiese sowel as snelneutronvloed gemeet kan word. Die detektor moet verder in staat wees om die strawwe toestande, kenmerkend aan ’n hoë temperatuur reaktor, te kan weerstaan. Die innoverende detektorstelsel, oorspronklik geoormerk vir gebruik in die PBMR reaktor, het toepassingsmoontlikhede as in-kern neutron- sowel as indirekte temperatuurmonitor. Simulasies van ’n generiese model van ’n HTGR reaktorkern is uitgevoer ten einde die neutronenergiespektrum in die kern by verskillende temperature te bekom met klem op die gedrag van neutrone in drie energiegroepe: stadig (termies), intermediêr en snel (vinnig). Daar is bevind dat die stadige neutrone, wat ’n Maxwell-Boltzman verdeling toon, in intensiteit toeneem en dat die piek na hoër energie verskuif met toename in temperatuur, terwyl die vinnige neutronspektrum relatief onveranderd bly. ’n Passing van die stadige spektrum op ’n gemodifiseerde Maxwell-Boltzmann verdeling het bevestig dat die effektiewe neutrontemperatuur weens die teenwoordigheid van bronterme, verliese en absorpsie, hoër as die temperatuur van die medium is. Hierdie resultate maak dit duidelik dat die detektorstelsel beide die stadige sowel as die vinnige neutronvloed moet kan waarneem. Deur detektorelemente wat sensitief is vir die onderskeie spekrale gebiede in die reaktorhart te plaas, kan informasie bekom word wat tot in-kern temperatuur herleibaar is sodat die stelsel inderdaad as indirekte temperatuurmonitor kan dien. Die feit dat alfa-deeltjies geproduseer word in neutron-geïnduseerde opbreekreaksies van 6Li en 12C is as die basis van die nuwe opsporingsmeganisme aangewend. Hierdie materiale funksioneer uitstekend as neutron-selektiewe omsetters in die teenwoordigheid van gamma-strale aangesien laasgenoemde se bydraes tot helium produksie via die 6Li(γ,np)4He en 12C(γ,3α) reaksies, weglaatbaar is. Die massaspektrometer meet die tydgedrag van die 4He parsiële druk binne ’n hoogvakuum wat met behulp van ’n seriegeskakelde kombinasie van ’n turbomolekulêre en positiewe-verplasingsvoorpomp verkry word. ’n Koueval met ’n molekulêre sif, bestaande uit 1.6 mm diameter korrels, is ontwerp en vervaardig om onsuiwerhede te verwyder wat andersins as agtergrond by die ligter gedeelte van die massaspektrum sou wys. Die ontwikkeling en toetsing van die hoëvloed detektorstelsel is te iThembaLABS (iThemba Laboratories for Accelerator Based Sciences) gedoen. Dit is uitgevoer deur gebruik te maak van die hoë energie protonbundel van die D-lyn neutronfasiliteit asook van die bundel vinnige neutrone by die neutronterapiefasiliteit. Om die beginsel en vermoë te toets om by ’n hoë neutronvloed te kan meet, is van die intense 66 MeV protonbudel gebruik gemaak om ’n hoë opbrengs alfa-deeltjies te verkry. Dit is gedoen omdat die reaksiedeursnit vir onelastiese verstrooiing van protone vanaf 12C kerne soortgelyk is aan die van neutrone, met ’n drumpelenergie van 8 MeV vir beide reaksies. Tweedens is die sekondêre vinnige neutrone afkomstig van die 9Be(p,n)9B reaksie ook met die neutrondetektor gemeet. Daar is bevind dat die reaksietyd van die deteksiestelsel tydens bestraling relatief vinnig is, soos gekenmerk deur ’n stygtyd van etlike sekondes. Laasgenoemde manifesteer as ’n toename in die parsiële druk van die 4He sodra die proton- of neutronbundel op die 12C teiken inval. Daar is verder bevind dat die 4He produksie direk eweredig aan die bundelintensiteit is. Vir ’n neutronbundel van nagenoeg 1010 s−1, invallend op die neutrondetektor, is vanaf die gemete parsiële druk afgelei dat die produksie van 4He atome sowat 109 s−1 beloop. In die geheel beoordeel, was die meting van die 4He parsiële druk tydens die toetse met vinnige protone en neutrone suksesvol en het dit die nuwe meetbeginsel bevestig. Dit is verder bevind dat die meetstelsel nie ’n beperking op die boonste neutronvloed plaas nie, maar dat dit vloede van selfs hoër as 1014 s−1 kan hanteer. Die leeftyd van die detektorstelsel in die reaktor is prakties onbeperk en onderhewig aan die bevestigde integriteit van vlekvrystaal onder hoë bestraling. Die gevolgtrekking is dus dat die nuwe detektorstelsel uitstekend geskik is vir die in-tyd meting van ’n baie hoë vloed van neutrone ook in die teenwoordigheid van intense gammabestraling.

Approved for public release; distribution is unlimited / Twenty-three years ago, Israeli fighter pilots destroyed the Osiraq nuclear reactor and made a profound statement about global nuclear proliferation. In light of the recent preventive regime change in Iraq, a review of this strike reveals timely lessons for future counterproliferation actions. Using old, new, and primary source evidence, this thesis examines Osiraq for lessons from a preventive attack on a non-conventional target. Before attacking Osiraq, Israeli policymakers attempted diplomatic coercion to delay Iraq's nuclear development. Concurrent with diplomatic actions, Israeli planners developed a state of the art military plan to destroy Osiraq. Finally, Israeli leaders weathered the international storm after the strike. The thesis examines Israeli decisionmaking for each of these phases. The thesis draws two conclusions. First, preventive strikes are valuable primarily for two purposes: buying time and gaining international attention. Second, the strike provided a one-time benefit for Israel. Subsequent strikes will be less effective due to dispersed/hardened nuclear targets and limited intelligence. / Major, United States Air Force

The SuperCritical Water-cooled nuclear Reactor (SCWR) is one of six Generation-IV nuclear-reactor concepts currently being designed. It will operate at pressures of 25 MPa and temperatures up to 625°C. These operating conditions make a SuperCritical Water (SCW) Nuclear Power Plant (NPP) suitable to support thermochemical-based hydrogen production via co-generation. The Copper-Chlorine (Cu‒Cl) cycle is a prospective thermochemical cycle with a maximum temperature requirement of ~530°C and could be linked to an SCW NPP through a piping network. An intermediate Heat eXchanger (HX) is considered as a medium for heat transfer with operating fluids selected to be SCW and SuperHeated Steam (SHS). Thermalhydraulic calculations based on an iterative energy balance procedure are performed for counter-flow double-pipe design concept HXs integrated at several locations on an SCW NPP coolant loop. Using various test cases, design and operating parameters are recommended for detailed future research. In addition, predicted effects of heat transfer enhancement on HX parameters are evaluated considering theoretical improvements from helically-corrugated HX piping. The effects of operating fluid pressure drop are briefly discussed for applicability in future studies. / UOIT

Experimental data on SuperCritical-Water (SCW) cooled bundles are very limited. Major problems with performing such experiments are: 1) small number of operating SCW experimental setups and 2) difficulties in testing and experimental costs at very high pressures, temperatures and heat fluxes. However, SuperCritical Water-cooled nuclear Reactor (SCWRs) designs cannot be finalized without such data. Therefore, as a preliminary approach experiments in SCW-cooled bare tubes and in bundles cooled with SC modeling fluids can be used. One of the SC modeling fluids typically used is Freon-12 (R-12) where the critical pressure is 4.136 MPa and the critical temperature is 111.97ºC. These conditions correspond to a critical pressure of 22.064 MPa and critical temperature of 373.95ºC in water. A set of experimental data obtained in a Freon-12 cooled vertical bare bundle at the Institute of Physics and Power Engineering (IPPE, Obninsk, Russia) was analyzed. This set consisted of 20 cases of a vertically oriented 7-element bundle installed in a hexagonal flow channel. To secure the bundle in the flow channel 3 thin spacers were used. The dataset was obtained at equivalent parameters of the proposed SCWR concepts. Data was collected at pressures of about 4.65 MPa for several different combinations of wall and bulk-fluid temperatures that were below, at, or above the pseudocritical temperature. Heat fluxes ranged from 9 kW/m2 to 120 kW/m2 and mass fluxes ranged from 440 kg/m2s to 1320 kg/m2s. Also inlet temperatures ranged from 70ºC – 120ºC. The test section consisted of fuel elements that were 9.5 mm in diameter with the total heated length of 1 m. Bulk-fluid and wall temperature profiles were recorded using a combination of 8 different thermocouples.The data was analyzed with respect to its temperature profile and heat transfer coefficient along the heated length of the test section. In a previous study it was confirmed that there is the existence of three distinct regimes for forced convention with supercritical fluids. (1) Normal heat transfer; (2) Deteriorated heat transfer, characterized by higher than expected temperatures; and (3) Improved heat transfer, characterized by lower than expected temperatures. All three regions were observed for the 7 rod bundle experiments. This work compares the experimental data to predictions based upon current 1-D correlations for heat transfer in supercritical fluids. Results show that no current 1-D correlation was able to accurately predict heat transfer coefficients within ±50%. A parametric analysis of the data was also completed to determine if continuity in the experiment was present. Results of this study show that two distinct regions are present in the data. For cases with a mass flux below 1200 kg/m2s wall temperature profiles appear to be normal while in cases with mass flux above 1200 kg/m2s temperature given by the wall thermocouples were higher than normal. This phenomenon occurred regardless of heat flux-to-mass flux ratios. / UOIT

Thesis (M.A. in Security Studies (Defense Decision-Making and Planning))--Naval Postgraduate School, Sept. 2004. / Thesis advisor(s): Peter R. Lavoy. Includes bibliographical references (p. 61-62). Also available online.

In this thesis a numerical study has been undertaken to investigate turbulent flow and heat transfer in a number of flow problems, representing the gas-cooled reactor core flows. The first part of the research consisted of a meticulous assessment of various advanced RANS models of fluid turbulence against experimental and numerical data for buoyancy-modified mixed convection flows, such flows being representative of low-flow-rate flows in the cores of nuclear reactors, both presently-operating Advanced Gas-cooled Reactors (AGRs) and proposed ‘Generation IV’ designs. For this part of the project, an in-house code (‘CONVERT’), a commercial CFD package (‘STAR-CD’) and an industrial code (‘Code_Saturne’) were used to generate results. Wide variations in turbulence model performance were identified. Comparison with the DNS data showed that the Launder-Sharma model best captures the phenomenon of heat transfer impairment that occurs in the ascending flow case; v^2-f formulations also performed well. The k-omega-SST model was found to be in the poorest agreement with the data. Cross-code comparison was also carried out and satisfactory agreement was found between the results.The research described above concerned flow in smooth passages; a second distinct contribution made in this thesis concerned the thermal-hydraulic performance of rib-roughened surfaces, these being representative of the fuel elements employed in the UK fleet of AGRs. All computations in this part of the study were undertaken using STAR-CD. This part of the research took four continuous and four discrete design factors into consideration including the effects of rib profile, rib height-to-channel height ratio, rib width-to-height ratio, rib pitch-to-height ratio, and Reynolds number. For each design factor, the optimum configuration was identified using the ‘efficiency index’. Through comparison with experimental data, the performance of different RANS turbulence models was also assessed. Of the four models, the v^2-f was found to be in the best agreement with the experimental data as, to a somewhat lesser degree were the results of the k-omega-SST model. The k-epsilon and Suga models, however, performed poorly. Structured and unstructured meshes were also compared, where some discrepancies were found, especially in the heat transfer results. The final stage of the study involved a simulation of a simplified 3-dimensional representation of an AGR fuel element using a 30 degree sector configuration. The v^2-f model was employed and comparison was made against the results of a 2D rib-roughened channel in order to assess the validity and relevance of the precursor 2D simulations of rib-roughened channels. It was shown that although a 2D approach is extremely useful and economical for ‘parametric studies’, it does not provide an accurate representation of a 3D fuel element configuration, especially for the velocity and pressure coefficient distributions, where large discrepancies were found between the results of the 2D channel and azimuthal planes of the 3D configuration.

The objective of this diploma thesis is to assess the suitability of a small nuclear reactor for the purpose of heat supply in small towns and cities. First, all the heat demands of these towns and cities are analysed. Then, a small nuclear reactor to serve these needs is designed and a computational model for burn-up fuel is created. The thesis aims to propose a design of a new heat exchanger. Economic and environmental criteria are taken into account as well and including safety measures. The outcome of this thesis is an evaluation of the whole concept with respect to potential future implementation.

No description available.