Chernobyl Nuclear Power Plant | EPFL Graph Search

The Chernobyl Nuclear Power Plant (ChNPP; Чорнобильська атомна електростанція; Чернобыльская атомная электростанция) is a nuclear power plant undergoing decommissioning. ChNPP is located near the abandoned city of Pripyat in northern Ukraine northwest of the city of Chernobyl, from the Belarus–Ukraine border, and about north of Kyiv. The plant was cooled by an engineered pond, fed by the Pripyat River about northwest from its juncture with the Dnieper. Originally named for Vladimir Lenin, the plant was commissioned in phases with the four reactors entering commercial operation between 1978 and 1984. In 1986, in what became known as the Chernobyl disaster, reactor No. 4 suffered a catastrophic meltdown and explosion; as a result of this, the power plant is now within a large restricted area known as the Chernobyl Exclusion Zone. Both the zone and the power plant are administered by the State Agency of Ukraine on Exclusion Zone Management. The three other reactors remained operati

Helium effects on irradiation assisted stress corrosion cracking susceptibility of 316L austenitic stainless steel

Ignasi Villacampa Roses

The formation and growth of cracks by irradiation assisted stress corrosion cracking (IASCC) in light water reactor internals is a critical issue for a safe long-term operation of nuclear power plants. The IASCC susceptibility at relatively low dose is dominated by conventional mechanisms such as radiation-induced segregation and radiation hardening. However, the ageing of the nuclear fleet combined with the increase of their life-span reveals other mechanisms that could play an important role on IASCC susceptibility. Recent studies show that a huge amount of helium (He) can be accumulated in reactor internal components of pressurized water reactors (PWR) after long-term operation. This occurrence could significantly increase the IASCC susceptibility at high doses. The main objective was to investigate the He effects on IASCC susceptibility of the solution-annealed (SA) and 20% cold-worked (CW) austenitic steel 316L. SA and CW miniaturized tensile samples and a SA plate were homogeneously implanted up to 1000 appm He. He was implanted at ~45 MeV and 300°C at the cyclotron in CEMHTI/CNRS (France). Slow strain rate tests (SSRT) in air and in high-temperature water were then carried out on SA, CW, post-implantation annealed (PIA) and as-implanted samples; instrumented nanoindentation tests were performed at room temperature on non-implanted and implanted specimens; and scanning electron microscope and transmission electron microscope were employed for fractography and microstructural characterization. The results of SSRTs in high-temperature air and in hydrogenated high-temperature water showed that homogenized implanted He up to 1000 appm corresponding to a displacement damage of about 0.16 dpa (displacement per atom) does not produce intergranular cracking and IASCC. However, the deformation microstructure of non-implanted SA/CW is characterized by high dislocation density arranged in cell walls separated by relatively dislocation-free regions, while the as-implanted SA samples (> 0.05 dpa) exhibit a more planar deformation microstructure. Characterization of as-implanted CW and PIA samples did not show any significant implantation effect on the deformation microstructure. PIA of the He implanted plate was carried out from 650 to 1000°C for 1h to increase the helium grain boundary coverage, and to reproduce the observed microstructure of the replaced reactor internal components that showed IASCC and use the results for the PIA of tensile samples. The transmission electron microscope investigations showed that an increase of the annealing temperature causes an average bubble size increase and density decrease. The average He bubble size and distribution in the grain interior and on the grain boundary were similar in the range of temperatures studied. In both cases, bubbles grew by the Ostwald ripening mechanism. No preferential He build-up took place on the grain boundaries. The increase of yield stress produced by He bubbles was calculated with three distinct dislocation-localized obstacle and compared to the tensile test results. The bubble strength estimated for these models was used to assess their validity and to compare the results to published data. Finally, nano-indentation tests in SA, CW, as-implanted SA and PIA (650 to 850°C for 1h) samples did not show He effects on grain boundaries strength. However, the average hardness of the grain interior increased with the He implantation and decreased increasing the PIA temperature.

EPFL2018

Pin-by-Pin Treatment of LWR Cores with Cross Section Equivalence Procedures and Higher-Order Transport Methods

Petra Mala

This thesis presents a systematic study on the merits and limitations on using pin-by-pin resolution and transport theory based approaches for nuclear core design calculations. Starting from the lattice codes and an optimal cross section generation scheme, it compares different methods, transport approximations, and spatial discretizations used in pin-by-pin homogenized codes. It is in the interest of nuclear power plant operators to employ more heterogeneous core loadings in order to improve the fuel utilization and decrease the amount of spent fuel. This necessarily increases the requirements on the accuracy of the computation tools used for the core design and safety analysis. One possibility is employing 3D core solvers with higher spatial resolution, e.g. pin-cell wise. The comparison of several lattice codes indicates that already the proper generation of diffusion coefficients and higher-order scattering moments for pin-cell geometry is not straightforward. Out of three available lattice codes, two generated unphysical diffusion coefficient when using the inscatter approximation, while the last code was not able to provide the higher scattering moments. Several few-assembly test cases with either high neutron leakage effects or MOX/uranium interfaces showed that the quality of the results of diffusion and SP3 solvers depends critically on the choice of the diffusion coefficient: in the outscatter approximation, it can cause major deviations, while the inscatter seems overall more adequate. Regarding the transport solvers, results obtained with the SN solver DORT showed very good performance for MOX/uranium interfaces, but major deviations occurred for problems with large power gradients. On the other hand, the MOC solver nTRACER showed in all cases small average error, but 2 - 3 % error around the interface of different assemblies. A study on spatial discretization indicated that the finite difference method applied on pin-cells does not properly capture the big flux changes between MOX and uranium fuel, while the nodal expansion method is more accurate but too slow. It was suggested to use the finite difference method with finer mesh in the outer assembly pin-cells, which increases the required computation time by only 50 % and decreases the pin power errors below 1 % with respect to lattice code results. Due to some problems which were observed with the available diffusion/SP3 solvers, a new SP3 solver was implemented in the DORT-TD platform. Several core tests showed that the SP3 pin-by-pin solver can significantly outperform the state-of-the-art nodal solver SIMULATE-5, in particular for reactor cores with inserted control rods. Finally, the pin-by-pin solvers were coupled to a depletion solver. For that, an accurate and fast interpolation routine had to be implemented. The obtained results of full-cycle depletion with complex core loading showed very good performance in comparison to a heterogeneous transport-based fine-group calculations.

EPFL2018

Modélisation de la tribocorrosion d'aciers inoxydables dans l'eau à haute pression et haute température

Jonathan Perret

Several metallic components of pressurized water reactors (PWR) used in nuclear power plants are exposed to the combined degradation by friction and corrosion (tribocorrosion). In PWR, corrosion arises from the water pressurized at 150 bars and maintained at 300°C and results in the build up of a micrometer thick surface oxide film. Tribocorrosion is of complex nature as it involves multi-scale interactions between mechanical, chemical and material factors. Nevertheless, PWR component manufacturers are interested in understanding the deterioration mechanisms in order to select appropriate materials as well as predicting the component lifetime. This thesis is a contribution towards the elucidation of the tribocorrosion mechanisms of a specific PWR component (austenitic stainless steel control rods). The research is focused on two inter-related aspects related to the subsurface deformation induced by friction and to the high temperature and high-pressure oxidation kinetics of austenitic stainless steel under tribocorrosion conditions. The subsurface deformation is investigated by using a model tribocorrosion system consisting in an alumina ball sliding against a 304L austenitic stainless steel in sulphuric acid at ambient temperature under electrochemical control. This model system allows studying subsurface deformation in absence of a thick high temperature oxide films that can complicate the interpretation of the results. Subsurface deformation was characterized using SEM, EBDS and FIB cross-sections. Good correlation was found with the deformation found in pressurized water. Subsurface deformation was found to result in the formation of a recrystallised layer of nano-grains close to the surface and an underlying layer exhibiting a deformation gradient. The thickness of the recrystallized zone is highly affected by the presence or the absence of a few nm thick oxide in the contact. However, no obvious correlation between the thickness of the recrystallized zone, or any other measured deformation appearance, and wear could be found. The effect of oxidation was studied using a simulator (Aurore), reproducing the chemical and mechanical PWR conditions of control rods. It includes an electrochemical set-up allowing measuring in-situ and in real time the oxidation kinetic. The oxidation kinetic was found to follow a parabolic law (diffusion controlled oxide growth). Rubbing resulted in repeated scratching of the surface leading to a thinning of the oxide film and thus to an enhanced anodic oxidation. A quantitative model of the electrochemical response of stainless steel in PWR conditions was developed. Based on these results and on the third body theory of wear, a mechanistic model was proposed for describing the material behaviour in contacts operating in pressurized water. This model explains a certain number of practical observations obtained in nuclear plants or in simulators concerning chemical and materials effects on component degradation.

EPFL2010