Extension of the nodal code DYN3D to SFR applications

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.

Development of the control assembly pattern and dynamic analysis of the generation IV large gas-cooled fast reactor (GFR)

Gaëtan Girardin

During the past ten years, different independent factors, such as the rapidly increasing worldwide demand in energy, societal concerns about greenhouse gas emissions, and the high and volatile prices for fossil fuels, have contributed to the renewed interest in nuclear technology. It is in this context that the Generation IV international forum (GIF) launched the initiative, in 2000, to collaborate on the research and development (R&D) efforts needed for the next generation, i.e. Generation IV, of nuclear reactors. These advanced systems will be ideally deployed beyond the year 2030, following the Generation III or III+ nuclear power plants, which are mainly based on light water technology and are currently entering deployment. A particular goal set for Generation IV systems is closure of the nuclear fuel cycle. Thus, apart from improvements in safety, they are expected to offer a better utilization of natural resources, as also a minimization of long-lived radioactive wastes. Among the systems selected by the GIF, the Gas-cooled Fast Reactor (GFR) is a highly innovative system with advanced fuel geometry and materials (fuel pellets of mixed uranium-plutonium carbide within a plate-type, honeycomb structure made of SiC). It is in the context of the large, 2400 MWth reference GFR design that the present doctoral research has been conducted, the principal aim having been to develop and qualify the control assembly (CA) pattern and corresponding CA implementation scheme for this system. The work has been carried out in three successive and complementary phases: (1) validation of the neutronics tools, (2) the CA pattern development and related static analysis, and (3) dynamic core behavior studies for hypothetical CA driven transients. The deterministic code system ERANOS and its associated nuclear data libraries for fast reactors were developed and validated for the previous generation of sodium-cooled reactors. The validation of ERANOS for GFR applications was, therefore, the first task to be realized in the present research. This has entailed a systematic reanalysis of the GFR-relevant, integral data generated at PSI during the GCFR-PROTEUS experimental program of the 1970's. Thus, during the first phase of the thesis, the reference PROTEUS test lattice from these experiments has been analyzed with ERANOS-2.0 and its associated, adjusted nuclear data library ERALIB1, in order to derive a reference computational scheme to be used later for the GFR analysis. Additionally, benchmark calculations were performed with the Monte Carlo code MCNPX, allowing one to both check the deterministic results and to analyze the sensitivity to different modern data libraries. It has been found that, for the main reaction rate ratios, the new analysis of the GCFR-PROTEUS reference lattice generally yields good agreement – within 1σ measurement uncertainty – with experimental values and with the Monte Carlo simulations. As shown by the analysis, the predictions were in somewhat better agreement in the case of the adjusted ERALIB1 library. The applicability of ERANOS-2.0/ERALIB1 as the reference neutronics tool for the GFR analysis could thus be demonstrated. Furthermore, neutronics aspects related to the novel features of the GFR, for which new experimental investigations are needed, were highlighted. In the second phase of the research, the CA pattern was developed for the GFR, based on iterative neutronics and thermal-hydraulics calculations, 2D and 3D neutronics models for the reactor core having first been set up using the reference ERANOS-2.0/ERALIB1 computational scheme. For the thermal-hydraulics analysis, the CEA code COPERNIC was used. This design work was followed by the study of an appropriate CA implementation scheme (number of CAs and corresponding positions within the core). Detailed neutronics studies revealed the existence of large CA interaction effects, so-called shadowing/anti-shadowing effects leading to an amplification/reduction of the CA worth. The interactions between the absorber pins within a CA, and between the CAs themselves, were investigated in detail, with the goal to optimize the CA efficiency, in terms of the absorber fraction and minimization of the associated heterogeneity effects. The proposed CA pattern consists of 54 absorber pins placed in a triangular lattice. Each absorber pin is a stainless-steel tube filled with highly enriched 10B boron carbide pellets. As a result of the detailed investigations, the absorber pin diameter could be chosen such as to minimize the pin-to-pin influence within the assembly. In particular, a central part of the CA was designed without any absorber pins (zone filled with stagnant helium). A final reduction of the heterogeneity effect (difference between homogeneous and heterogeneous treatments) to 13% was achieved through this feature. Of special importance, the neutronics investigations performed for the reference GFR core ("2004-Core"), especially those related to the CA interactions, have directly contributed to a new core design ("2007-Core"), with the height-to-diameter ratio having been increased to 0.6, compared to 0.3 for the reference core. During the third phase, detailed coupled, 3D neutron-kinetics (NK) and 1D thermal-hydraulics (TH) models were developed for the GFR core, the aim being to arrive at an in-depth understanding of the 3D core behavior during CA driven transients, especially from the viewpoint of spatial effects. The coupled models were developed using the PARCS code for the 3D NK and the TRACE code for the 1D TH modeling. Particular attention was paid to have each individual fuel sub-assembly and CA represented, in order to allow the analysis of local deformations of the 3D distributions of power and safety related parameters, such as the coolant, cladding and fuel temperatures. The validation of the coupled full-core models was performed against reference ERANOS-VARIANT calculations. In particular, the CA worth and reactivity feedbacks were benchmarked, the discrepancies being shown to be relatively low (

EPFL2009

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

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

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

Flavio Dante Giust

EPFL2012

Neutronics Experiments and Analysis Related to Strong Moderation Heterogeneity in LWRs

Dominik Rätz

EPFL2012

Development of the control assembly pattern and dynamic analysis of the generation IV large gas-cooled fast reactor (GFR)

Gaëtan Girardin

EPFL2009