Chemical reactions at nuclear fuel-clad interfaces

This thesis presents a study of the chemical reactions that may occur at the fuel- clad interfaces of fuel elements used in advanced gas-coooled reactors (A.G.R.) The initial investigation involved a study of the inner surfaces of irradiated stainless steel clad and evidence was obtained to show that fission products, in particular tellerium, were associated with reaction products on these surfaces. An accelerated rate of oxidation was observed on the inner surfaces of a failed A.G.R. fuel pin. It is believed that fission product caesium was responsible for this enhancement. A fundamental study of the reaction between 20%Cr/25%Ni/niobium stabilised stainless steel and tellerium was then undertaken over the range 350 - 850 degrees C. Reaction occurred with increasing rapidity over this range and long term exposure at ≤ 750 degrees resulted in intergranular attack of the stainless steel and chromium depletion. The reaction on unoxidised steel surfaces involved the formation of an initial iron-nickel-tellerium layer which subsequently transformed to a chromium telluride product during continued exposure. The thermodynamic stabilities of the steel tellurides were determined to be chromium telluride > nickel telluride > iron telluride. Oxidation of the stainless steel surface prior to tellerium exposure inhibited the reaction. However reaction did occur in regions where the oxide layer had either cracked or spalled.

Experimental analysis of pressure drop and flow redistribution in axial flows in rod bundles

Fuel elements of PWR type nuclear reactors consist of rod bundles, arranged in a square array, and held by spacer grids. The coolant flows, mainly, axially along the rods. Although such elements are laterally open, experiments are performed in closed type test sections, originating the appearance of subchannels with different geometries. In the present work, utilizing a test section of two bundles of 4x4 pins each, experiments were performed to determine the friction and the grid drag coefficients for the different subchannels and to observe the effect of the grids in the crossflow, in cases of inlet flow maldistribution.

Neutronisäteilylle altistuminen käytetyn polttoaineen siirrossa Olkiluodon voimalaitoksella

Tässä työssä on tutkittu OL1/OL2-ydinvoimalaitosten käytetyn polttoaineen siirrossa aiheutuvaa altistusta neutronisäteilylle. Käytetty polttoaine siirretään vedellä täytetyssä käytetyn polttoaineen siirtosäiliössä Castor TVO:ssa OL1/OL2-laitoksilta käytetyn polttoaineen varastolle. Siirtotyön aikana useat eri ammattiryhmiin kuuluvat henkilöt työskentelevät siirtosäiliön välittömässä läheisyydessä, altistuen käytetystä polttoaineesta emittoituvalle fotoni- ja neutronisäteilylle. Aikaisemmista neutronisäteilyannosten mittauksista on todettu, ettei jatkuvalle altistuksen seurannalle ole ollut tarvetta. Tämän työn tarkoitus on selvittää teoreettisilla laskelmilla siirtotyöhön osallistuvan henkilön mahdollisuus saada kirjausrajan ylittävä annos neutronisäteilyä. Neutronisäteilyn annosnopeudet siirtosäiliötä ympäröivässä tilassa on laskettu yhdysvaltalaisella Monte Carlo-menetelmään perustuvalla MCNP-ohjelmalla. MCNP:llä mallinnettiin siirtosäiliö, siirtosäiliön sisältämä polttoaine ja ympäröivä tila kolmella jäähtymisajalla ja kolmella keskimääräisellä maksimipoistopalamalla. Polttoainenippujen isotooppikonsentraatiot ja säteilylähteiden voimakkuudet on laskettu Studsvik SNF-ohjelmalla. Simuloinnin perusteella voidaan todeta, ettei neutronisäteilyannosten jatkuvalle seurannalle ole tarvetta käytetyn polttoaineen siirrossa. Vaikka neutronisäteilyn annosnopeudet voivat nousta siirtosäiliön läheisyydessä suhteellisen suuriksi, ovat siirtosäiliön lähellä tehtävät työt niin lyhytaikaisia, että kirjausrajan ylitystä voidaan pitää hyvin epätodennäköisenä. Johtopäätökset varmistetaan työssä suunnitellulla mittausjärjestelyllä.

Experimental characterization of cement–bentonite interaction using core infiltration techniques and 4D computed tomography

Deep geological storage of radioactive waste foresees cementitious materials as reinforcement of tunnels and as backfill. Bentonite is proposed to enclose spent fuel drums, and as drift seals. The emplacement of cementitious material next to clay material generates an enormous chemical gradient in pore water composition that drives diffusive solute transport. Laboratory studies and reactive transport modeling predict significant mineral alteration at and near interfaces, mainly resulting in a decrease of porosity in bentonite. The goal of this project is to characterize and quantify the cement/bentonite skin effects spatially and temporally in laboratory experiments. A newly developed mobile X-ray transparent core infiltration device was used, which allows performing X-ray computed tomography (CT) periodically without interrupting a running experiment. A pre-saturated cylindrical MX-80 bentonite sample (1920 kg/m3 average wet density) is subjected to a confining pressure as a constant total pressure boundary condition. The infiltration of a hyperalkaline (pH 13.4), artificial OPC (ordinary Portland cement) pore water into the bentonite plug alters the mineral assemblage over time as an advancing reaction front. The related changes in X-ray attenuation values are related to changes in phase densities, porosity and local bulk density and are tracked over time periodically by non-destructive CT scans.

Experimental Characterization and Quantification of Cement-Bentonite Interaction using Core Infiltration Techniques Coupled with X-ray Tomography

Deep geological storage of radioactive waste foresees cementitious materials as reinforcement of tunnels and as backfill. Bentonite is proposed to enclose spent fuel canisters and as drift seals. Sand/bentonite (s/b) is foreseen as backfill material of access galleries or as drift seals. The emplacement of cementitious material next to clay material generates an enormous chemical gradient in pore-water composition that drives diffusive solute transport. Laboratory studies and reactive transport modeling predicted significant mineral alteration at and near interfaces, mainly resulting in a decrease of porosity in bentonite. The goal of this thesis was to characterize and quantify the cement/bentonite interactions both spatially and temporally in laboratory experiments. A newly developed mobile X-ray transparent core infiltration device was used to perform X-ray computed tomography (CT) scans without interruption of running experiments. CT scans allowed tracking the evolution of the reaction plume and changes in core volume/diameter/density during the experiments. In total 4 core infiltration experiments were carried out for this study with the compacted and saturated cores consisting of MX-80 bentonite and sand/MX-80 bentonite mixture (s/b; 65/35%). Two different high-pH cementitious pore-fluids were infiltrated: a young (early) ordinary Portland cement pore-fluid (APWOPC; K+–Na+–OH-; pH 13.4; ionic strength 0.28 mol/kg) and a young ‘low-pH’ ESDRED shotcrete pore-fluid (APWESDRED; Ca2+–Na+–K+–formate; pH 11.4; ionic strength 0.11 mol/kg). The experiments lasted between 1 and 2 years. In both bentonite experiments, the hydraulic conductivity was strongly reduced after switching to high-pH fluids, changing eventually from an advective to a diffusion-dominated transport regime. The reduction was mainly induced by mineral precipitation and possibly partly also by high ionic strength pore-fluids. Both bentonite cores showed a volume reduction and a resulting transient flow in which pore-water was squeezed out during high-pH infiltration. The outflow chemistry was characterized by a high ionic strength, while chloride in the initial pore water got replaced as main anionic charge carrier by sulfate, originating from gypsum dissolution. The chemistry of the high-pH fluids got strongly buffered by the bentonite, consuming hydroxide and in case of APWESDRED also formate. Hydroxide got consumed by mineral reactions (saponite and possibly talc and brucite precipitation), while formate being affected by bacterial degradation. Post-mortem analysis showed reaction zones near the inlet of the bentonite core, characterized by calcium and magnesium enrichment, consisting predominately of calcite and saponite, respectively. Silica got enriched in the outflow, indicating dissolution of silicate-minerals, identified as preferentially cristobalite. In s/b, infiltration of APWOPC reduced the hydraulic conductivity strongly, while APWESDRED infiltration had no effect. The reduction was mainly induced by mineral precipitation and probably partly also by high ionic strength pore-fluids. Not clear is why the observed mineral precipitates in the APWESDRED experiment had no effect on the fluid flow. Both s/b cores showed a volume expansion along with decreasing ionic strengths of the outflow, due to mineral reactions or in case of APWESDRED infiltration also mediated by microbiological activity, consuming hydroxide and formate, respectively. The chemistry of the high-pH fluids got strongly buffered by the s/b. In the case of APWESDRED infiltration, formate reached the outflow only for a short time, followed by enrichment in acetate, indicating most likely biological activity. This was in agreement to post-mortem analysis of the core, observing black spots on the inflow surface, while the sample had a rotten-egg smell indicative of some sulfate reduction. Post-mortem analysis showed further in both cores a Ca-enrichment in the first 10 mm of the core due to calcite precipitation. Mg-enrichment was only observed in the APWOPC experiment, originating from newly formed saponite. Silica got enriched in the outflow of both experiments, indicating dissolution of silicate-minerals, identified in the OPC experiment as cristobalite. The experiments attested an effective buffering capacity for bentonite and s/b, a progressing coupled hydraulic-chemical sealing process and also the preservation of the physical integrity of the interface region in this setup with a total pressure boundary condition on the core sample. No complete pore-clogging was observed but the hydraulic conductivity got rather strongly reduced in 3 experiments, explained by clogging of the intergranular porosity (macroporosity). Such a drop in hydraulic conductivity may impact the saturation time of the buffer in a nuclear waste repository, although the processes and geometry will be more complex in repository situation.

Impact of nuclear data uncertainties in the Phase-1B Benchmark

Accurate control over the spent nuclear fuel content is essential for its safe and optimized transportation, storage and management. Consequently, the reactivity of spent fuel and its isotopic content must be accurately determined. Nowadays, to predict isotopic evolution throughout irradiation and decay periods is not a problem thanks to the development of powerful codes and methodologies. In order to have a realistic confidence level in the prediction of spent fuel isotopic content, it is desirable to determine how uncertainties in the basic nuclear data affect isotopic prediction calculations by quantifying their associated uncertainties

Burnup Uncertainty Analysis for PWR-HFP Exercise I-1

Burn-up credit analyses are based on depletion calculations that provide an accurate prediction of spent fuel isotopic contents, followed by criticality calculations to assess keff

Impact of Nuclear Data Uncertainties in the Phase-1B Benchmark

Accurate control over the spent nuclear fuel content is essential for its safe and optimized transportation, storage and management. Consequently, the reactivity of spent fuel and its isotopic content must be accurately determined.

Isotopic prediction calculation methodologies: application to Vandellos-II Reactor cycles 7-11

Determining as accurate as possible spent nuclear fuel isotopic content is gaining importance due to its safety and economic implications. Since nowadays higher burn ups are achievable through increasing initial enrichments, more efficient burn up strategies within the reactor cores and the extension of the irradiation periods, establishing and improving computation methodologies is mandatory in order to carry out reliable criticality and isotopic prediction calculations. Several codes (WIMSD5, SERPENT 1.1.7, SCALE 6.0, MONTEBURNS 2.0 and MCNP-ACAB) and methodologies are tested here and compared to consolidated benchmarks (OECD/NEA pin cell moderated with light water) with the purpose of validating them and reviewing the state of the isotopic prediction capabilities. These preliminary comparisons will suggest what can be generally expected of these codes when applied to real problems. In the present paper, SCALE 6.0 and MONTEBURNS 2.0 are used to model the same reported geometries, material compositions and burn up history of the Spanish Van de llós II reactor cycles 7-11 and to reproduce measured isotopies after irradiation and decay times. We analyze comparisons between measurements and each code results for several grades of geometrical modelization detail, using different libraries and cross-section treatment methodologies. The power and flux normalization method implemented in MONTEBURNS 2.0 is discussed and a new normalization strategy is developed to deal with the selected and similar problems, further options are included to reproduce temperature distributions of the materials within the fuel assemblies and it is introduced a new code to automate series of simulations and manage material information between them. In order to have a realistic confidence level in the prediction of spent fuel isotopic content, we have estimated uncertainties using our MCNP-ACAB system. This depletion code, which combines the neutron transport code MCNP and the inventory code ACAB, propagates the uncertainties in the nuclide inventory assessing the potential impact of uncertainties in the basic nuclear data: cross-section, decay data and fission yields

Isotopic uncertainty assessment due to nuclear data uncertainties in high-burnup samples.

The accurate prediction of the spent nuclear fuel content is essential for its safe and optimized transportation, storage and management. This isotopic evolution can be predicted using powerful codes and methodologies throughout irradiation as well as cooling time periods. However, in order to have a realistic confidence level in the prediction of spent fuel isotopic content, it is desirable to determine how uncertainties affect isotopic prediction calculations by quantifying their associated uncertainties.

Performance of a transmutation advanced device for sustainable energy application

Preliminary studies have been performed to design a device for nuclear waste transmutation and hydrogen generation based on a gas-cooled pebble bed accelerator driven system, TADSEA (Transmutation Advanced Device for Sustainable Energy Application). In previous studies we have addressed the viability of an ADS Transmutation device that uses as fuel wastes from the existing LWR power plants, encapsulated in graphite in the form of pebble beds, cooled by helium which enables high temperatures (in the order of 1200 K), to generate hydrogen from water either by high temperature electrolysis or by thermochemical cycles. For designing this device several configurations were studied, including several reflectors thickness, to achieve the desired parameters, the transmutation of nuclear waste and the production of 100 MW of thermal power. In this paper new studies performed on deep burn in-core fuel management strategy for LWR waste are presented. The fuel cycle on TADSEA device has been analyzed based on both: driven and transmutation fuel that had been proposed by the General Atomic design of a gas turbine-modular helium reactor. The transmutation results of the three fuel management strategies, using driven, transmutation and standard LWR spent fuel were compared, and several parameters describing the neutron performance of TADSEA nuclear core as the fuel and moderator temperature reactivity coefficients and transmutation chain, are also presented

Surveillance and diagnostics of the beam mode vibrations of the ringhals pwrs

Surveillance of core barrel vibrations has been performed in the Swedish Ringhals PWRs for several years. This surveillance is focused mainly on the pendular motion of the core barrel, which is known as the beam mode. The monitoring of the beam mode has suggested that its amplitude increases along the cycle and decreases after refuelling. In the last 5 years several measurements have been taken in order to understand this behaviour. Besides, a non-linear fitting procedure has been implemented in order to better distinguish the different components of vibration. By using this fitting procedure, two modes of vibration have been identified in the frequency range of the beam mode. Several results coming from the trend analysis performed during these years indicate that one of the modes is due to the core barrel motion itself and the other is due to the individual flow induced vibrations of the fuel elements. In this work, the latest results of this monitoring are presented.

Surveillance and diagnostics of the beam mode vibrations of the ringhlas PWRs

Surveillance of core barrel vibrations has been performed in the Swedish Ringhals PWRs for several years. This surveillance is focused mainly on the pendular motion of the core barrel, which is known as the beam mode. The monitoring of the beam mode has suggested that its amplitude increases along the cycle and decreases after refuelling. In the last 5 years several measurements have been taken in order to understand this behaviour. Besides, a non-linear fitting procedure has been implemented in order to better distinguish the different components of vibration. By using this fitting procedure, two modes of vibration have been identified in the frequency range of the beam mode. Several results coming from the trend analysis performed during these years indicate that one of the modes is due to the core barrel motion itself and the other is due to the individual flow induced vibrations of the fuel elements. In this work, the latest results of this monitoring are presented.