Benchmark of Advanced Burner Test Reactor model using MCNPX 2.6.0 and ERANOS 2.1 (original) (raw)

MCNPX Monte Carlo burnup simulations of the isotope correlation experiments in the NPP Obrigheim

Annals of Nuclear Energy, 2010

This paper describes the simulation work of the Isotope Correlation Experiment (ICE) using the MCNPX Monte Carlo computer code package. The Monte Carlo simulation results are compared with the ICE-Experimental measurements for burnup up to 30 GWD/t. The comparison shows the good capabilities of the MCNPX computer code package for predicting the depletion of the uranium fuel and the buildup of the plutonium isotopes in a PWR thermal reactor. The Monte Carlo simulation results show also good agreements with the experimental data for calculating several long-lived and stable fission products. However, for the americium and curium actinides, it is difficult to judge the predication capabilities for these actinides due to the large uncertainties in the ICE-Experimental data. In the MCNPX numerical simulations, a pin cell model is utilized to simulate the fuel lattice of the nuclear power reactor. Temperature dependent libraries based on JEFF3.1 nuclear data files are utilized for the calculations. In addition, temperature dependent libraries based ENDF/B-VII nuclear data files are utilized and the obtained results are very close to the JEFF3.1 results, except for $10% differences in the prediction of the minor actinide isotopes buildup.

Preparation of a multi-isotope plutonium AMS standard and preliminary results of a first inter-lab comparison

Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, 2015

The motivation of this work is to establish a new multi-isotope plutonium standard for isotopic ratio measurements with accelerator mass spectrometry (AMS), since stocks of existing solutions are declining. To this end, certified reference materials (CRMs) of each of the individual isotopes 239 Pu, 240 Pu, 242 Pu and 244 Pu were obtained from JRC IRMM (Joint Research Center Institute for Reference Materials and Measurements). These certified reference materials (IRMM-081a, IRMM-083, IRMM-043 and IRMM-042a) were diluted with nitric acid and mixed to obtain a stock standard solution with an isotopic ratio of approximately 1.

VALMOX: validation of nuclear data for high burn-up MOX fuels

Nuclear Engineering and Design, 2005

as part of the research on Nuclear Fission Safety in the 5 th R&D Framework Programme of the European Commission (1999)(2000)(2001)(2002)(2003)(2004) under contract number FIKS-CT-00191. VALMOX was one of the projects of the cluster EVOL (Evolutionary Fuel Concepts: High Burnup and MOX Fuels).

Nuclear data adjustment based on the interpretation of post-irradiation experiments with the DARWIN2.3 package

EPJ Nuclear Sciences & Technologies, 2018

DARWIN2.3 is the French reference package dedicated to fuel cycle applications, computing fuel inventory as well as decay heat, neutron emissions, α, β and γ spectra. The DARWIN2.3 package fuel inventory calculation was experimentally validated with Post-Irradiation Experiments (PIEs), mainly consisting in irradiated fuel pellets analysis. This paper presents a method to assimilate these integral trends for improving nuclear data. In this study, the method is applied to 137Cs/238U concentration ratio. Results suggest an increase of the JEFF-3.1.1 235U cumulated thermal fission yield in 137Cs by (+3.8 ± 2.1)%, from 6.221E-02 to 6.460E-02 ± 2.1%.

I.C. Gauld, J. M. Giaquinto, J. S. Delashmitt, J. Hu, G. Ilas, T.J. Haverlock, C. Romano, " Re-evaluation of Spent Nuclear Fuel Assay Data for the Three Mile Island Unit 1 Reactor and Application to Code Validation ", Annals of Nuclear Energy vol.87, pp. 267-281 (January 2016)

Annals of Nuclear Energy

Destructive radiochemical assay measurements of spent nuclear fuel rod segments from an assembly irradiated in the Three Mile Island unit 1 (TMI-1) pressurized water reactor have been performed at Oak Ridge National Laboratory (ORNL). Assay data are reported for five samples from two fuel rods of the same assembly. The TMI-1 assembly was a 15 × 15 design with an initial enrichment of 4.013 wt % 235 U, and the measured samples achieved burnups between 45.5 and 54.5 gigawatt days per metric ton of initial uranium (GWd/t). Measurements were performed mainly using inductively coupled plasma mass spectrometry after elemental separation via high performance liquid chromatography. High precision measurements were achieved using isotope dilution techniques for many of the lanthanides, uranium, and plutonium isotopes. Measurements are reported for more than 50 different isotopes and 16 elements. One of the two TMI-1 fuel rods measured in this work had been measured previously by Argonne National Laboratory (ANL), and these data have been widely used to support code and nuclear data validation. The recent measurements performed by ORNL provided an important opportunity to independently cross check results against previous measurements performed at ANL. These measurements serve to improve confidence in the data, to verify reported uncertainties, and to investigate previous anomalies noted in the plutonium measurements. The measured nuclide concentrations are used to validate burnup calculations using the SCALE nuclear systems modeling and simulation code suite. These results show that the new measurements provide reliable benchmark data for computer code validation.

Analyzing Nuclear Fuel Cycles from Isotopic Ratios of Waste Products Applicable to Measurement by Accelerator Mass Spectrometry

Nuclear Science and Engineering, 2007

An extensive study was conducted to determine isotopic ratios of nuclides in spent fuel that may be utilized to reveal historical characteristics of a nuclear reactor cycle. This forensic information is important to determine the origin of unknown nuclear waste. The distribution of isotopes in waste products provides information about a nuclear fuel cycle, even when the isotopes of uranium and plutonium are removed through chemical processing. Several different reactor cycles of the PWR, BWR, CANDU, and LMFBR were simulated for this work with the ORIGEN-ARP and ORIGEN 2.2 codes. The spent fuel nuclide concentrations of these reactors were analyzed to find the most informative isotopic ratios indicative of irradiation cycle length and reactor design. Special focus was given to long-lived and stable fission products that would be present many years after their creation. For such nuclides, mass spectrometry analysis methods often have better detection limits than classic gamma-ray spectroscopy. The isotopic ratios Sm Sm 146 151 , Sm Sm 146 149 , and Cm Cm 246 244 were found to be good indicators of fuel cycle length and are well suited for analysis by accelerator mass spectroscopy.