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

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.

Appraisal of Core Analysis Methods against Zero-Power Experiments on Full-Scale BWR Fuel Assemblies

Flavio Dante Giust

The present doctoral research aims at the appraisal of nodal core simulators used for the calculation of commercial boiling water reactor (BWR) cores, against measurements carried out at the Paul Scherrer Institute under the LWR-PROTEUS experimental programme. The research focuses mainly on the prediction of radial and axial total-fission rate (i.e. power) distributions in the vicinity of core heterogeneities, caused by features such as the presence of control blades, enrichment boundaries or partial length rods. As such, this thesis complements previous validation work performed for LWR calculational methods on the basis of integral experiments in zero-power research reactors, the latter corresponding largely to the validation of two-dimensional, reflected fuel-assembly calculations. Two commercial code systems have been investigated currently: HELIOS/PRESTO-2, presently used for core monitoring at the Leibstadt Nuclear Power Plant (Switzerland), and CASMO-5/SIMULATE-5, which represents the most recent generation of the widely applied CASMO/SIMULATE system. Six different LWR-PROTEUS configurations have been modelled, for which the nodal reconstructed total-fission rate distributions have been compared against experimental results, as well as against reference, three-dimensional whole-reactor Monte Carlo calculations using the MCNPX code. To start with, a methodology has been developed and tested for appropriate representation of the multi-zone experimental configurations, as employed in the LWR-PROTEUS programme, by means of reduced-geometry models set up using the investigated nodal code systems. The approach adopted has been to apply, in each case, case-dependent three-dimensional boundary conditions to the nodal code's modelling of the 3x3 array of BWR assemblies constituting the LWR-PROTEUS test zone. This has been done in terms of so-called Partial Current Ratios (PCRs), which describe the relation between the incoming and outgoing neutron currents across the test-zone boundary. These PCRs have been derived from the fore-mentioned MCNPX calculation of the multi-zone PROTEUS reactor, thus allowing for the adequate description of the three-dimensional effects associated with the interaction between the test zone and its surroundings. The results obtained for the reference LWR-PROTEUS configuration, with a test zone consisting of an unperturbed regular array of identical, axially uniform BWR fuel assemblies (of type SVEA-96+), showed that both the investigated nodal methodologies reproduce the experimental total-fission rate distribution with very good accuracy, equivalent to that achievable with high-order transport calculations. This has provided adequate indication that the developed methodology of using three-dimensional PCRs, calculated by means of an appropriate whole-reactor MCNPX model, offers a reliable platform for the desired validation. The full insertion of a L-shaped hafnium control blade, especially fabricated for the LWR-PROTEUS programme, has permitted the study of the behaviour of calculated total-fission rate distributions in the presence of strong radial and azimuthal flux gradients. In this case, the use of reflective boundary conditions in the lattice calculations, combined with the azimuthal non-uniformity of the currents that couple the test zone with the outer zones of the reactor, produce systematic deviations which, together, lead to some loss of accuracy in comparison with the regular, uncontrolled case. The three-dimensional effects of the changes in enrichment and gadolinium content between two different axial sections of the fuel assembly have also been studied. In this case, the calculations of the axial total-fission rate distributions are seen to generally reproduce the experimental data with good accuracy, although certain systematic effects are observed. In particular, at local level, and very near the strong axial heterogeneity caused by the enrichment boundary, relatively large deviations have been found to occur. However, due to their very local character, these deviations are not considered to represent a relevant issue in connection with power reactor monitoring and design. Also for the demanding case of the LWR-PROTEUS configuration with a partially inserted control blade, the results obtained in this thesis show a very good behaviour. As in the boundary enrichment case, however, some very localized deviations occur in the vicinity of the strong axial flux gradients. In a separate phase of the LWR-PROTEUS programme, BWR assemblies with partial length rods (PLRs), viz. of type SVEA-96 Optima2, were studied in the central test zone. The results of the comparisons made currently for the corresponding LWR-PROTEUS configurations have shown that, also for fuel assemblies with axial heterogeneities of this type, the performance of both the investigated nodal methodologies is very satisfactory for predicting the global distribution of the total-fission rate. The local three-dimensional effects caused by the PLRs are also predicted with good accuracy, similar, in fact, to that obtained with high-order three-dimensional transport calculations. The present research shows that the two nodal code systems investigated – HELIOS/PRESTO-2 and CASMO-5/SIMULATE-5 – reproduce the LWR-PROTEUS experimental results with high accuracy, both for the radial and the three-dimensional (i.e. at pellet level) comparisons. Significant deviations occur almost exclusively within a short distance from the nodal interface, in cases featuring strong gradients across the core midplane. The axial and radial shapes of the total-fission rate distributions are well predicted in most cases, which represents an important observation concerning the monitoring of operational limits in power reactor cores. Thus, overall – within the range of applicability of the experimental conditions studied in the LWR-PROTEUS programme – the results of the comparisons performed in this thesis confirm that nodal methodologies with pin-power reconstruction have a very high level of performance. In fact, the accuracy that they can achieve for the assembly-internal total-fission rate distribution is, in general, higher than that practically achievable in an operating nuclear power plant. This is due to unavoidable, additional uncertainties in the characterisation of a BWR core under power reactor conditions. Finally, the present research has provided quantitative insights into the applicability of integral data, produced in critical facilities such as PROTEUS, for validating the calculation of power-reactor core heterogeneities. One has thus been able to demonstrate that the experimental evidence from the LWR-PROTEUS programme indeed constitutes a very valuable basis for the validation and appraisal of three-dimensional nodal methodologies with pin-power reconstruction.

EPFL2012

A Hybrid Approach to Neutron Transport with Thermal-hydraulic Feedback for Reactor Transient Analysis

Alexander Aures

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

Appraisal of Core Analysis Methods against Zero-Power Experiments on Full-Scale BWR Fuel Assemblies

Flavio Dante Giust

EPFL2012

A Hybrid Approach to Neutron Transport with Thermal-hydraulic Feedback for Reactor Transient Analysis

Alexander Aures

EPFL2021