Luiz C Leal

Retroactive Generation of Covariance Matrix of Nuclear Model Parameters Using Marginalization Techniques

Abstract An uncertainty propagation methodology relying on marginalization techniques was recently developed to produce covariance matrices between existing model parameters involved in describing neutron-induced reactions. This work has been implemented in the nuclear data assimilation tool CONRAD. The performance of the code was demonstrated through simplified test cases based on a Reich-Moore description of the 155Gd(n,γ) reaction. Results are compared with those produced via Monte Carlo techniques.

R-matrix analysis of 235U neutron transmission and cross-section measurements in the 0- to 2.25-keV energy range

A new R-matrix analysis of the {sup 235}U cross-section data in the 0- to 2250-eV energy region is presented. The analysis was performed with the SAMMY computer code that has recently been updated to permit, for the first time, inclusion of both differential and integral data within the analysis process. Fourteen differential data sets and six integral quantities were used in this evaluation: two measurements of fission plus capture, one of fission plus absorption, six of fission alone, two of transmission, and one of eta, plus standard values of thermal cross sections for fission and capture, and of K1 and the Westcott g factors for both fission and absorption. An excellent representation was obtained for the high-resolution transmission, fission, and capture cross-section data as well for the integral quantities. The result is a single set of resonance parameters spanning the entire range up to 2250 eV, a decided improvement over the present ENDF/B-VI evaluation, in which 11 discrete resonance parameter sets are required to cover that same energy range. This new evaluation is expected to greatly improve predictability of the criticality safety margins for nuclear systems in which {sup 235}U is present.

Automatic rapid process for the generation of problem-dependent SAS2H/ORIGEN-S cross-section libraries

A methodology is described that serves as an alternative to the SAS2H path of the SCALE system to generate cross sections for point-depletion calculations with the ORIGEN-S code. Automatic Rapid Processing (ARP) is an algorithm that allows the generation of cross-section libraries suitable to the ORIGEN-S code by interpolation over pregenerated SAS2H libraries. The interpolations are carried out on the following variables. burnup, enrichment, and water density. The adequacy of the methodology is evaluated by comparing measured and computed spent-fuel isotopic compositions for pressurized water reactor and boiling water reactor systems.

Evaluation of 238U Resonance Parameters from 0 to 20 keV

The neutron resonance parameters of 238U were obtained in the energy range 0 to 20 keV from a sequential SAMMY analysis of the most recent high‐resolution neutron transmission and neutron capture cross‐section measurements. Special care was taken in the analysis of the lowest s‐wave resonances leading to resonance parameters slightly different from those of ENDF/B‐VI (Moxon‐Sowerby resonance parameters). The resolved‐resonance range was extended to 20 keV, taking advantage of the high‐resolution neutron transmission data of Harvey and neutron capture data of Macklin et al. Preliminary integral tests were performed with the new resonance parameters; thermal low‐enriched benchmark calculations show an improvement of the keff prediction, mainly due to a 1.5% decrease of the capture cross section at 0.0253 eV and about a 0.4% decrease of the effective shielded resonance capture integral.

CALCULATING PROBABILITY TABLES FOR THE UNRESOLVED-RESONANCE REGION USING MONTE CARLO METHODS

Abstract A new module, Probability tables for the Unresolved Region using Monte Carlo (PURM), has been developed for the AMPX-2000 cross-section–processing system. PURM uses a Monte Carlo approach to calculate probability tables on an evaluator-defined energy grid in the unresolved-resonance region. For each probability table, PURM samples a Wigner spacing distribution for pairs of resonances surrounding the reference energy (i.e., energy specified in the cross-section evaluation). The resonance distribution is sampled for each spin sequence (i.e., l-J pair), and PURM uses the Δ3-statistics test to determine the number of resonances to sample for each spin sequence. For each resonance, PURM samples the resonance widths from a chi-square distribution for a specified number of degrees of freedom. Once the resonance parameters are sampled, PURM calculates the total, capture, fission, and scatter cross sections at the reference energy using the single-level Breit-Wigner formalism with appropriate treatment for temperature effects. Probability tables have been calculated and compared with NJOY. The probability tables and cross-section values that are calculated by PURM and NJOY are in agreement, and the verification studies with NJOY establish the computational capability for generating probability tables using the new AMPX module PURM.

R-Matrix Analysis of 232Th Neutron Transmissions and Capture Cross Sections in the Energy Range Thermal to 4 keV

Abstract Neutron resonance parameters of 232Th were obtained from the Reich-Moore SAMMY analysis of high-resolution neutron transmission measurements performed at the Oak Ridge Electron Linear Accelerator (ORELA) by Olsen in 1981, along with the high-resolution neutron capture measurements performed in 2005 at the Geel Linear Accelerator (GELINA, Belgium) by Schillebeeckx and at the n-TOF facility (CERN, Switzerland) by Aerts. The ORELA data were analyzed previously by Olsen with the Breit-Wigner multilevel code SIOB, and the results were used in the ENDF/B-VI evaluation. In the new analysis of the Olsen neutron transmissions by the modern computer code SAMMY, better accuracy is obtained for the resonance parameters by including in the experimental database the recent experimental neutron capture data. The experimental database and the method of analysis are described in the report. The neutron transmissions and the capture cross sections calculated with the resonance parameters are compared to the experimental values. A description is given of the statistical properties of the resonance parameters. The new evaluation results in a decrease in the capture resonance integral.

Reevaluation and validation of the 241Pu resonance parameters in the energy range thermal to 20 eV

Abstract A new SAMMY analysis of the 241Pu resonance parameters from thermal to 20 eV is presented. This evaluation takes into account the trends given by integral experiments [post-irradiation experiments performed in French pressurized water reactors (PWRs)]. Compared to the previous evaluations performed by Derrien and de Saussure, the capture cross section increases especially in the 0.26-eV resonance. It is shown that the new resonance parameters proposed in this work improve the prediction of the 242Pu buildup in a PWR, which was significantly underestimated with the previous evaluations.

New Neutron Cross‐Section Measurements at ORELA for Improved Nuclear Data Calculations

The Oak Ridge Electron Linear Accelerator (ORELA) was used to measure neutron total and capture cross sections of aluminum, silicon, chlorine, fluorine, and potassium in the energy range from 100 eV to ~600 keV. These measurements were carried out to support the Nuclear Criticality Safety Program (NCSP). Concerns about the use of existing cross section data in nuclear criticality calculations have been a prime motivator for the new cross section measurements. Our results are substantially different from the evaluated nuclear data files of ENDF/B-VI and JENDL-3.2.

High-Resolution Transmission Measurements of 233U Using a Cooled Sample at the Temperature T=11 K

Abstract For the first time, high-resolution transmission data of 233U have been obtained using a cooled sample. The samples were cooled to T = 11 K using a cryogenic device, which reduced the Doppler broadening of resonances by 50% compared to room-temperature measurements. The measurements were carried out at the Oak Ridge Electron Linear Accelerator over the energy range from 0.6 eV to 300 keV at the 80-m flight path station. Corrections were made for experimental effects, and the average total cross section in this energy range was determined. Results are compared to previous measurements.

ORIGEN-ARP, A Fast and Easy-to-Use Source Term Generation Tool

ORIGEN-ARP is a new SCALE analytical sequence for spent fuel characterization and source term generation that serves as a faster alternative to the SAS2H sequence by using the Automatic Rapid Processing (ARP) methodology for generating problem-dependent ORIGEN-S cross-section libraries. ORIGEN-ARP provides an easy-to-use menu-driven input processor. This new sequence is two orders of magnitude faster than SAS2H while conserving the rigor and accuracy of the SAS2H methodology. ORIGEN-ARP has been validated against pressurized water reactor (PWR) and boiling water reactor (BWR) spent fuel chemical assay data.