Supercritical water reactor | EPFL Graph Search

The supercritical water reactor (SCWR) is a concept Generation IV reactor, designed as a light water reactor (LWR) that operates at supercritical pressure (i.e. greater than 22.1 MPa). The term critical in this context refers to the critical point of water, and must not be confused with the concept of criticality of the nuclear reactor. The water heated in the reactor core becomes a supercritical fluid above the critical temperature of 374 °C, transitioning from a fluid more resembling liquid water to a fluid more resembling saturated steam (which can be used in a steam turbine), without going through the distinct phase transition of boiling. In contrast, the well-established pressurized water reactors (PWR) have a primary cooling loop of liquid water at a subcritical pressure, transporting heat from the reactor core to a secondary cooling loop, where the steam for driving the turbines is produced in a boiler (called the steam generator). Boiling water reactors (BWR) oper

Neutronics Experiments and Analysis Related to Strong Moderation Heterogeneity in LWRs

Dominik Rätz

The Supercritical-Water-Cooled Reactor (SCWR) is the Generation IV reactor concept most closely related to current light water reactors (LWRs). The SCWR builds on the vast experience with today's LWRs and supercritical coal-fired power plants. Water at supercritical state is used as moderator and coolant, and reaches a temperature of about 500 °C at the core outlet, which enables a much higher thermal efficiency (∼ 44%) than possible with current-day LWRs. In the operating range of an SCWR, the supercrical water density can become as low as one seventh of the density of water at room temperature. In order to ensure a thermal neutron spectrum, large moderator regions are introduced into the fuel assembly designs, resulting in lattices with strong moderation heterogeneity. As such, the neutronics of these assemblies differs significantly from that of standard LWRs, and lies outside the validated domain of reactor physics codes. Previous comparisons between the deterministic code CASMO-4 and reference calculations carried out with the Monte Carlo code MCNP4C showed large differences in calculated pin-wise reaction rate distributions in perturbed SCWR lattices. The experimental validation of standard reactor physics codes for SCWR-representative neutronics conditions is thus clearly of key importance for the further development of this technology. The goal of this thesis is to provide an experimental database for such validation and to assess the level of performance of current-day codes for SCWR analysis. An SCWR-like fuel lattice, based on a Japanese assembly design proposal from 2001, has been investigated in the PROTEUS zero-power research reactor at the Paul Scherrer Institute. Measurements have been carried out on the unperturbed test lattice, as also on six other PROTEUS configurations corresponding to different types of perturbations of the reference lattice. The investigated perturbations include control rod related effects, moderator density changes, and the replacement of a fuel pin with gadolinium-poisoned fuel. For each experimental configuration, pin-wise distributions of the total fission rate (Ftot) and the 238U capture rate (C8), as well as of their ratio (C8/Ftot), have been obtained across the assembly and compared to computed values. The neutronics codes used for the calculations are the LWR assembly code CASMO-4E and the Monte Carlo code MCNPX. Additionally, the reactivity effects of removing individual pins from the unperturbed SCWR-like lattice have been measured and compared to predictions obtained using the two codes. The pin-wise reaction rate distributions predicted with MCNPX have been found to agree within two standard deviations with the measured values for the unperturbed, as well as all the perturbed, configurations. The 1σ uncertainty was in the order of 0.4%, 0.8%, and 2.2% for Ftot, C8, and C8/Ftot, respectively. MCNPX could thus be validated for the various SCWR-like conditions investigated and has, in turn, been used to validate CASMO-4E on simplified geometries. For the latter purpose, the unperturbed and perturbed SCWR-like lattices were modeled as reflected assemblies, and reaction rate distributions predicted with the two codes were compared. For the unperturbed lattice and for the configurations with local perturbations of the neutron absorption, the agreement between the codes was within ∼ 1% for all reaction rates. However, for lattices with locally perturbed moderation conditions, CASMO-4E was initially found to overestimate the reaction rates in the vicinity of the perturbations by up to 5%. The discrepancies were identified as resulting from the default leakage treatment in CASMO-4E, which applies a DB2 correction in a homogenized sense across the lattice. In this way, cases with a global leakage gradient were not being treated properly. Usage of the optional input card BZ2, which allows a region-wise leakage treatment, resolved this problem, and the codes then agreed mostly within 1% for all the configurations. The pin removal worth measurements in the unperturbed SCWR-like lattice have provided integral data complementary to the reaction rate distributions. MCNPX results were found to agree with experiment within the statistical uncertainty of typically ∼ 10% (1σ). The comparison of reflected-assembly results from MCNPX and CASMO-4E yielded significant differences in the calculated pin removal worths (up to 4.3σ). A decomposition analysis of these differences has indicated that the discrepancies, especially for the better moderated pin positions, are linked to the calculation of the additional moderator effect caused by the pin removal. Finally, the transferability of the code validation, carried out under the PROTEUS experimental conditions, to the most recently proposed SCWR assembly designs has been assessed. Since, under power-reactor conditions, the moderator and coolant water densities are considerably lower in the proposed SCWR assemblies than that in the PROTEUS test lattices, their neutron spectra are much harder. Assembly-averaged values of the integral parameters C8/Ftot and F8/Ftot have been found to be as much as 65% and 80% higher, respectively, than for the PROTEUS reference lattice. However, the agreement between CASMO-4E and MCNPX predictions for the proposed SCWR assembly designs is still very good, indicating that the accuracy of the deterministic calculations does not deteriorate markedly when considering these new designs under power-reactor conditions.

EPFL2012

Extension of the nodal code DYN3D to SFR applications

Evgeny Nikitin

DYN3D is a well-established Light Water Reactor (LWR) simulation tool and is being extended for safety analyses of Sodium cooled Fast Reactors (SFRs) at the Helmholtz-Zentrum Dresden-Rossendorf. This thesis focuses on the first stage of the development process, that is, the extension and application of DYN3D for steady-state and transient SFR calculations on reactor core level. In contrast to LWRs, the SFR behavior is especially sensitive to thermal expansions of the reactor components. Therefore, a new thermal-mechanical module accounting for thermal expansions is implemented into DYN3D. At first step, this module is capable of treating two important thermal expansion effects occurring within the core, namely axial expansion of fuel rods and radial expansion of diagrid. In order to perform nodal calculations with DYN3D, pre-generated homogenized few-group cross sections (XS) are necessarily needed. Therefore, prior to the development of thermal expansion models, a general methodology for XS generation is established for SFR nodal calculations based on the use of the Monte Carlo code Serpent. The new methodological developments presented in this thesis are verified against the Monte Carlo solutions of Serpent. Two SFR cores are used for testing: the large oxide core of the OECD/NEA benchmark and a smaller core from the Phenix end-of-life tests. Finally, the extended DYN3D is validated against selected IAEA benchmark tests on the Phenix end-of-life experiments that contain both steady-state and transient calculations. The contribution to the SFR-related developments at the HZDR, as presented in this thesis, makes it possible of performing steady-state and transient calculations for SFRs on reactor core level by using DYN3D. With this study, the basis of the next stage of DYN3D developments is established, that is, the up-scale of SFR analysis to system level can continue by coupling with a sodium capable thermal-hydraulic system code.

EPFL2019

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