Sandro Pelloni

COMPARISONS OF CELL CALCULATIONS FOR URANIUM-FREE LIGHT WATER REACTOR FUELS

An effective way to reduce the large quantities of Pu currently accumulated worldwide would be to use uranium-free fuel in light water reactors (LWRs) so that no new Pu is produced. Such a possibility could be provided by an LWR fuel consisting of Pu in a neutronically inert matrix. It may be necessary to add a burnable absorber or thorium to reduce the reactivity swing during burnup. The methods and data currently used for LWR analyses have not been tested in conjunction with such exotic fuel materials. An international exercise has accordingly been launched to compare the relative performance of different code systems and the accuracy of the basic data. Comparison of the results of cell calculations done with fixed isotopic densities against reference Monte Carlo results shows fairly small but systematic differences in the multiplication factors. A sensitivity analysis done with different basic cross section libraries and the same code system allows one to distinguish between the effects of the codes and those of the databases. The results of the burnup calculations indicate a fair agreement in k∞ both at beginning of life (BOL) and after 1200 days of irradiation [end of life (EOL)] under conditions representative of a present-day pressurized water reactor. At BOL, the fuel temperature coefficients agree fairly well among the different contributions, but unacceptably large differences are observed at EOL. The void coefficients agree well for low voidage, but for void fractions >90%, there are significant effects mostly due to the databases used. The agreement in the calculated boron worths is good.

FAST Code System: Review of Recent Developments and Near-Future Plans

The FAST code system is currently being developed and used at the Paul Scherrer Institut for static and transient analysis of the main Generation 4 fast-spectrum reactor concepts: sodium-, helium-, and gas-cooled fast reactors. The code system includes the ERANOS code system for static neutronics calculations, as well as coupled TRACE/PARCS/FRED for neutron kinetics, thermal hydraulic, and fuel transient analysis. The paper presents the status of the recent developments in neutronics (new 3D procedure for equilibrium cycle simulation and new transient cross section generation procedure), in thermal hydraulics and chemistry (equations-of-state for new coolants, two-phase flow models for sodium, and new model for oxide layer buildup in heavy-metal flow), and in fuel behavior (new model for the dispersed gas-cooled fast reactor fuel). Near-future plans for the further development of FAST are outlined. 2010 by ASME.

Sensitivities of the proton beam current resulting from variations in the source term for a Pb–Bi cooled accelerator driven system with a Pb–Bi Target

Abstract Numerical simulations of a subcritical reactor coupled to a neutron spallation source, 600 MeV protons impinging on a Pb–Bi target, have been performed using the deterministic code system ERANOS. The investigations are focussed on the sensitivity of the subcriticality parameters resulting from variations in the external source term, the same neutron yield per source proton being assumed in each calculation. The analysis is performed using a detailed three-dimensional model representative of the experimental accelerator driven system being designed by Ansaldo. The reactor is based on a pool-type liquid metal fast critical reactor with a thermal output of about 80 MW and has been currently considered with the option of using existing Superphenix MOX fuel (Ansaldo Nucleare, 2001. XADS Pb–Bi cooled eperimental accelerator driven system reference configuration, Summary Report by Division of Ansaldo Energia SpA, ADS 1 SIFX 0500, Rev. 0, June 2001.). In the paper, it is highlighted that for such Pb–Bi cooled systems the proton beam current required in order to obtain the foreseen operational power level is not significantly sensitive to the source characteristics. Clearly, this result is hoped to facilitate the further overall optimisation of the Ansaldo design.

A Time-Dependent Solver for Coupled Neutron-Transport Thermal-Mechanics Calculations and Simulation of a Godiva Prompt-Critical Burst

The present paper describes a first step taken at the Paul Scherrer Institut in the development of a new multi-physics platform for reactor analysis. Such platform is based on the finite-volume software OpenFOAM and aims at a tightly coupled description of neutron transport, thermal mechanics and fluid dynamics. For this purpose, a steady-state 3-D discrete ordinates/thermal-mechanics solver was first developed in collaboration with the Politecnico di Milano. The present work briefly discusses such solver and its preliminary validation, which will be described in detail in parallel publications. It then focuses on its extension to time-dependent simulations. The solver is first tested by simulating different step-wise reactivity insertions in a critical configuration constituted by an infinite slab of highly enriched uranium. Subsequently, a super-prompt-critical power burst in the Godiva reactor has been simulated. Godiva was a spherical assembly of highly enriched uranium built and operated at the Los Alamos National Laboratory (US) during the Fifties. A prompt-critical transient in such system configures as a quick power excursion (up to ∼10 GW), which causes a temperature rise, and a subsequent reactivity reduction via expansion of the sphere. The overall transient lasts for few fractions of a millisecond. Results obtained with the newly developed model have been compared to experimental results, showing a relatively good agreement.Copyright

Validation of Light Water Reactor Calculation Methods and JEF-1—Based Data Libraries by TRX and BAPL Critical Experiments

The capability of various code systems and JEF-1—based nuclear data libraries to compute light water reactor lattices is analyzed by comparing calculations with results from thermal reactor benchma...

Influence of the Thermal Cutoff Energy on the Calculation of Neutronic Parameters for Light Water Reactor Lattices

Abstract Efforts are currently being pursued to validate present-day analysis methods for light water reactors (LWRs) in conjunction with advanced fuels having a high plutonium content. One particular problem, encountered in the framework of cell calculations performed at the Paul Scherrer Institute for infinite arrays of uranium-free plutonium fuel rods considered earlier in the framework of an international benchmark, is that significant k∞ decreases can be observed by just raising the upper energy boundary of the thermal range from 1.3 to 2.4 eV (some other LWR lattice codes utilize even smaller values for this cutoff energy). The present study indicates that the sensitivity of the computed k∞ with respect to the upper boundary of the thermal range results primarily from the different scattering matrices for hydrogen. The thermal motion of the scattering nucleus is taken into account in the thermal energy range, while a zero velocity of the scattering nucleus is assumed in the elastic scattering matrices employed in the epithermal energy range. Therefore, if the thermal cutoff energy is increased, the energy loss of the neutrons scattered down by hydrogen from energies between the original and the larger cutoff value is reduced, resulting in a shift of the neutron spectrum toward higher energies. The sensitivity of the computed k∞ with respect to the upper boundary of the thermal range, however, does not depend on upscattering effects, which were estimated to be negligibly small for energies greater than ~1 eV. Instead of further raising the upper energy boundary of the thermal range (the choice of an appropriate, sufficiently large value being strongly problem dependent), the analytical recalculation of the epithermal scattering matrices for the moderator nuclides, based on the free gas model (instead of the elastic scattering model, which assumes a zero velocity for the scattering nucleus), is proposed. Several beginning-of-life LWR systems have been analyzed with the new method. In particular, for a representative cell with uranium-free plutonium fuel consisting of a mixture of oxides of reactor-grade plutonium, zirconium, and the burnable poison erbium, the use of the new epithermal free gas scattering matrices results in a k∞ decrease of ~700 pcm (pcm = 1.0 × 10-5). For a mixed-oxide (MOX) fuel assembly with 4.8 wt% fissile plutonium in depleted uranium, typical of a present-day pressurized water reactor, the computed k∞ decreases by as much as ~400 pcm. For fresh assemblies with UO2 fuel, the decrease is less significant (<100 pcm). The effect is thus particularly important for cases with strong resonance absorptions in the lower-electron-volt range. Bearing in mind that present-day LWRs are being loaded with a larger number of assemblies with MOX fuel, a general recommendation is to compute the epithermal scattering matrices of hydrogen by accounting for the thermal motion of the nucleus below a sufficiently high energy limit. Upscattering effects being negligibly small for energies greater than ~1 eV, the specific choice of a thermal cutoff energy larger than ~1 eV is, of course, not important as far as upscattering is concerned. However, the choice of a sufficiently high thermal cutoff energy could help in reducing the inaccuracies produced by an inadequate model in the epithermal energy range.

Comparative transient analysis of critical and subcritical 80-MW Pb-Bi eutectic-cooled reactor systems

A consistent analytical comparison has been made of the transient behavior of critical and subcritical fast-spectrum reactor systems, the basic core design assumed in each case being that of the 80-MW(thermal) mixed-oxide-fueled, Pb-Bi-cooled, Experimental Accelerator Driven System (XADS). The transient calculations were performed using the FAST code system developed at the Paul Scherrer Institute. The present study demonstrates a high level of self-protection of both the critical and subcritical systems over a wide range of postulated events, including transient overpower due to reactivity insertion, loss of flow, station blackout, loss of coolant, and core overcooling accidents. The relative advantages and shortcomings of the two system types, from the viewpoint of transient behavior, are discussed on the basis of the corresponding simulation results obtained.

Data Sensitivity of Design Calculations for High Conversion D2O Reactors

Light water high conversion reactor (LWHCR) experiments carried out during the late 1980’s at the PROTEUS facility in Switzerland included the investigation of a 11wt% PuO2/UO2 lattice with a volumetric moderator-to-fuel ratio of 0.95, in which the H2O was replaced by D2O. The current paper presents a systematic investigation of the effects of different data libraries on calculation/experiment (C/E) comparisons of k∞ and reaction rate ratios for this particular test lattice. Effects of library changes for the corresponding dry (fully voided) lattice have also been studied, permitting conclusions to be drawn on the data sensitivity of the void coefficient and its components. The results obtained serve to quantify, in the rather stringent testing situation provided, the data sensitivity of reactivity and reaction rate ratio predictions for a high conversion D2O-reactor under both normal and fully voided conditions. Two parallel self-consistent calculational routes have been employed. These are the Monte Carlo code, MCNP, in conjunction with pointwise cross sections based on JEF-2, ENDF/B-VI and ENDF/B-V data, and a deterministic route, MICROX-2/ONEDANT, using pointwise and 193-group libraries generated via NJOY processing of JENDL-3 and JEFF-3T, in addition to JEF-2, ENDF/B-VI and ENDF/B-V data.