S. C. van der Marck

Efficient use of Monte Carlo : Uncertainty Propagation

Abstract A new and faster Total Monte Carlo (TMC) method for the propagation of nuclear data uncertainties in Monte Carlo nuclear simulations is presented (the fast TMC method). It addresses the main drawback of the original TMC method, namely, the necessary large time multiplication factor compared to a single calculation. With this new method, Monte Carlo simulations can now be accompanied with an uncertainty propagation (other than statistical), with small additional calculation time. The fast TMC method is presented and compared with the TMC and fast GRS methods for criticality and shielding benchmarks and burnup calculations. Finally, to demonstrate the efficiency of the method, uncertainties due to uncertainties in 235,238U, 239Pu, and thermal scattering nuclear data, for the local deposited power in 12.7 million cells, are calculated for a full-size reactor core.

New nuclear data libraries for lead and bismuth and their impact on accelerator-driven systems design

Abstract New ENDF-6 formatted nuclear data libraries are presented for 204,206,207,208Pb and 209Bi, for incident neutrons and protons. Apart from the resonance range, which we have adopted from the best available source in existing libraries, the nuclear data evaluations are completely revised in the 0 to 20 MeV energy range and moreover extend up to 200 MeV. This collection of isotopic evaluations is created by using the nuclear model code TALYS with a consistent set of input parameters for all isotopes. The most important nuclear reaction models and parameters needed for our data files are described. We have intended to make these evaluations complete in their description of reaction channels, and use a consistent method to store the data in ENDF-6 format, which includes cross sections, angular distributions, double-differential spectra, discrete and continuum photon production cross sections, and residual production (activation) cross sections including isomers. It is shown that the data present in our libraries give an improved agreement with existing basic experimental data. Moreover, we have validated the new libraries with criticality and shielding benchmarks, where available. We present the results of neutronics calculations on subcritical accelerator-driven systems to show the impact of our new nuclear data on critical reactor parameters, such as keff, when compared with the existing ENDF/B-VI, JENDL, and JEFF libraries.

Benchmark Results for Delayed Neutron Data

We have calculated the effective delayed neutron fraction βeff for 32 benchmark configurations for which measurements have been reported. We use these results to test the delayed neutron data of JEFF‐3.0, ENDF/B‐VI.8, and JENDL‐3.3.

New Nuclear Data Evaluations for Ca, Sc, Fe, Ge, Pb, and Bi Isotopes

New ENDF‐6 formatted nuclear data evaluations are presented for all natural isotopes of Ca, Sc, Fe, Ge, Pb, and Bi. Apart from the resonance range and low‐energy total cross sections, which we have adopted from the best available library, the nuclear data evaluations are completely revised up to 20 MeV, and moreover are extended up to 200 MeV. This collection of isotopic evaluations, called NRG‐2003, is created with the nuclear‐model code TALYS using input parameters that deviate only slightly, or not at all, from the default values. These model‐input parameters have been tuned to the available experimental data. All evaluations presented here are thus of comparable quality: For each isotope, we use the same set of nuclear models and, equally important, the same set of ENDF‐6 formatting procedures. We have made these evaluations complete in their description of reaction channels. The data, which are stored in a compact way, include cross sections, angular distributions, double‐differential spectra, discre...

Uncertainties for criticality‐safety benchmarks and consequences for nuclear data measurements

We have developed a new method to propagate the uncertainties of fundamental nuclear physics models and parameters used in nuclear data evaluation to the design and performance of future nuclear energy systems. Using Monte Carlo simulation, it is for the first time possible to couple these two fields at the extremes of nuclear science without any loss of information in between. With the help of a large database of nuclear reaction measurements, we have determined the uncertainties of theoretical nuclear reaction models such as the optical, compound nucleus, pre‐equilibrium and fission models. A similar assessment is done for the parameters that describe the resolved resonance range. We are now able to quantify the required quality of theoretical nuclear reaction models and measurements directly from the reactor design requirements. Examples will be presented for actinides using criticality‐safety benchmarks with feedback on experimental requirements.

Neutron dosimetry and 3D neutron transport calculations in the HFR Petten — a powerful tool in detailed neutron monitoring and damage analysis

Dosimetry at the High Flux Reactor in Petten, the Netherlands, serves a variety of purposes ranging from demonstrating compliance with the license for operating the reactor to validating the predictions of DPA in material irradiations. It is also used in the validation of the 3D neutron transport calculations during post-irradiation analysis to normalize the spectrum obtained from these calculations. In this paper, a thorough examination of irradiated steel specimens is carried out using the measurements performed with the neutron activation technique as well as the 3D neutron transport calculations to check the consistency in the resulting DPA values.