Pressurized water reactor

Summary

A pressurized water reactor (PWR) is a type of light-water nuclear reactor. PWRs constitute the large majority of the world's nuclear power plants (with notable exceptions being the UK, Japan and Canada). In a PWR, the primary coolant (water) is pumped under high pressure to the reactor core where it is heated by the energy released by the fission of atoms. The heated, high pressure water then flows to a steam generator, where it transfers its thermal energy to lower pressure water of a secondary system where steam is generated. The steam then drives turbines, which spin an electric generator. In contrast to a boiling water reactor (BWR), pressure in the primary coolant loop prevents the water from boiling within the reactor. All light-water reactors use ordinary water as both coolant and neutron moderator. Most use anywhere from two to four vertically mounted steam generators; VVER reactors use horizontal steam generators. PWRs were originally designed to serve as nuclear marine pro

Official source

https://en.wikipedia.org/wiki/Pressurized_water_reactor

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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

Understanding the time-dependent behaviour of a nuclear reactor following an intended or unintended change of the reactor conditions is of crucial importance to the safe operation of nuclear reactors. Safety evaluations of nuclear reactors involve the analysis of normal and abnormal reactor operation states with adequate and validated simulation tools. There is a growing interest in supplementing the results of best-estimate reactor calculations with uncertainties. A major source of uncertainty in reactor calculations is the nuclear data. The propagation of uncertainties allows to study their impact on response values of reactor calculations and to establish safety margins in nuclear reactor safety evaluations. For the simulation of reactor transients, the two-step approach consisting of the generation of group constants in infinite lattice geometry and the full-core calculation using a neutron-kinetic/thermal-hydraulic code system has been state-of-the-art. Nowadays, Monte Carlo codes allow for the determination of group constants based on a three-dimensional full-core model. Thus, effects from neighbouring fuel assemblies can be considered. In this thesis, the applicability of the neutron-kinetic code DYN3D coupled with the thermal-hydraulic system code ATHLET for the simulation of transients in small reactor cores was assessed. Experimental reactivity accident tests carried out at the SPERT III research reactor were simulated. Appropriate group constants were generated with the Monte Carlo code Serpent. The overall good agreement between the calculated and the experimental results prove the applicability of the applied codes for the analysis of such reactor transients. Nuclear data uncertainties were propagated through the simulations of two control rod withdrawal transients in a mini-core model of a pressurised water reactor. The random sampling-based method XSUSA in combination with the SCALE code package was used to determine varied two-group constants. Based on these two-group constants, 1,000 DYN3D-ATHLET calculations were performed and statistically analysed to obtain uncertainty information for the reactor power and the reactivity. Since the assumption of normality could not be maintained for the reactor power, distribution-free Wilks tolerance limits were applied. By sensitivity analyses, the number of delayed neutrons per fission of U-235 and the scattering cross sections of U-238 were identified as the major contributors to the reactor power uncertainty. For a potential improvement of transient simulations, an approach was developed that was comprised of the determination of group constants using a Serpent full-core model at selected times during a transient simulation and the updating of the original infinite lattice group constants with these full-core group constants. Various challenges and possible solutions were identified. The applicability of the group constants obtained by a full-core model was assessed by their application in full-core diffusion calculations.

EPFL2021

Modelling of fuel fragmentation, relocation and dispersal during Loss-of-Coolant Accident in Light Water Reactor

Vladimir Vladimirov Brankov

Recent LOCA tests with high burnup fuel at the OECD Halden Reactor Project and at Studsvik demonstrated the susceptibility of the fuel to fragment to small pieces, to relocate and possibly cause a hot-spot effect and to be dispersed in the event of cladding rupture. However, the LOCA safety criteria defined by the US NRC are still based on fuel tests with fresh and low burnup fuel and therefore require revision for high burnup fuel. In this context a PhD project with the goal of developing new models for high burnup fuel fragmentation, relocation and dispersal during Loss of Coolant Accident in Light Water Reactors was launched at Paul Scherrer Institute in June 2013 with a financial support from swissnuclear. The work continued with base irradiation simulation with a closer look at the fission gas release measurements of high burnup BWR fuel rods.The data showed significant scatter for fuel rods at the same average burnup, originating from the same fuel assembly and having almost the same enrichment. The scatter was explained with a BWR-specific mechanism for fission gas trapping. It was motivated with Scanning Electron Microscopy images at the pellet-cladding interface that showed very strong fuel-cladding bonding layer on one side of the fuel rod. Base irradiation with the EPRI's FALCON code coupled with an in-house advanced fission gas release and gaseous swelling model GRSW-A was done for selected high burnup BWR fuel rods. The calculated fission gas release was overestimated in all cases. This was expected, because at the time the modelling did not account for fission gas trapping. From the available modelling and experimental data, a model for fission gas trapping was proposed. After calibration, the agreement between calculated and measured fission gas release was significantly improved. Fuel fragmentation modelling was focused on fuel pulverization. This is a high burnup fuel-specific phenomenon, in which the fuel pellet periphery may fragment to sizes less than 100 ÎŒm. Such fragments are very mobile and can be easily relocated and dispersed in the event of cladding rupture regardless of the rupture opening size. Therefore, the fuel fragmentation model addresses the potentially most important mode of fragmentation - fuel pulverization. Fuel relocation inside fuel rod is simulated during cladding ballooning and until cladding failure. The model takes as input the time-dependent cladding deformation supplied by FALCON and the fragment size distribution either provided directly by experimental data or by the fuel fragmentation model. Fuel relocation was modelled under the specific assumption that outermost fragments (e.g. pulverized fuel) relocate first, resulting in a large packing factor and local cladding temperature increase at the balloon (i.e. hot-spot effect) and, as a consequence, in enhanced cladding oxidation at the balloon. Fuel dispersal is modelled by solving the mass, energy and momentum conservation equations for two phases - the gas inside the fuel rod and the fraction of fuel which is considered "moveable". Although the model uses simplified geometrical representation of the fuel rod and some other simplifying assumptions, the underlying reason for fuel dispersal (namely the interfacial friction between the gas outflow and the solid) is explicitly simulated. The model for fuel dispersal is calibrated using Halden and Studsvik LOCA tests.

EPFL2017