Machine learning of LWR spent nuclear fuel assembly decay heat measurements

Ebiwonjumi, Bamidele,Cherezov, Alexey,Dzianisau, Siarhei,Lee, Deokjung Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.11

Measured decay heat data of light water reactor (LWR) spent nuclear fuel (SNF) assemblies are adopted to train machine learning (ML) models. The measured data is available for fuel assemblies irradiated in commercial reactors operated in the United States and Sweden. The data comes from calorimetric measurements of discharged pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies. 91 and 171 measurements of PWR and BWR assembly decay heat data are used, respectively. Due to the small size of the measurement dataset, we propose: (i) to use the method of multiple runs (ii) to generate and use synthetic data, as large dataset which has similar statistical characteristics as the original dataset. Three ML models are developed based on Gaussian process (GP), support vector machines (SVM) and neural networks (NN), with four inputs including the fuel assembly averaged enrichment, assembly averaged burnup, initial heavy metal mass, and cooling time after discharge. The outcomes of this work are (i) development of ML models which predict LWR fuel assembly decay heat from the four inputs (ii) generation and application of synthetic data which improves the performance of the ML models (iii) uncertainty analysis of the ML models and their predictions.

Uncertainty quantification of PWR spent fuel due to nuclear data and modeling parameters

Ebiwonjumi, Bamidele,Kong, Chidong,Zhang, Peng,Cherezov, Alexey,Lee, Deokjung Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.3

Uncertainties are calculated for pressurized water reactor (PWR) spent nuclear fuel (SNF) characteristics. The deterministic code STREAM is currently being used as an SNF analysis tool to obtain isotopic inventory, radioactivity, decay heat, neutron and gamma source strengths. The SNF analysis capability of STREAM was recently validated. However, the uncertainty analysis is yet to be conducted. To estimate the uncertainty due to nuclear data, STREAM is used to perturb nuclear cross section (XS) and resonance integral (RI) libraries produced by NJOY99. The perturbation of XS and RI involves the stochastic sampling of ENDF/B-VII.1 covariance data. To estimate the uncertainty due to modeling parameters (fuel design and irradiation history), surrogate models are built based on polynomial chaos expansion (PCE) and variance-based sensitivity indices (i.e., Sobol' indices) are employed to perform global sensitivity analysis (GSA). The calculation results indicate that uncertainty of SNF due to modeling parameters are also very important and as a result can contribute significantly to the difference of uncertainties due to nuclear data and modeling parameters. In addition, the surrogate model offers a computationally efficient approach with significantly reduced computation time, to accurately evaluate uncertainties of SNF integral characteristics.

Propagation of radiation source uncertainties in spent fuel cask shielding calculations

Ebiwonjumi Bamidele Francis,Trong Mai Nhan Nguyen,이현철,이덕중 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.8

The propagation of radiation source uncertainties in spent nuclear fuel (SNF) cask shielding calculations is presented in this paper. The uncertainty propagation employs the depletion and source term outputs of the deterministic code STREAM as input to the transport simulation of the Monte Carlo (MC) codes MCS and MCNP6. The uncertainties of dose rate coming from two sources: nuclear data and modeling parameters, are quantified. The nuclear data uncertainties are obtained from the stochastic sampling of the cross-section covariance and perturbed fission product yields. Uncertainties induced by perturbed modeling parameters consider the design parameters and operating conditions. Uncertainties coming from the two sources result in perturbed depleted nuclide inventories and radiation source terms which are then propagated to the dose rate on the cask surface. The uncertainty analysis results show that the neutron and secondary photon dose have uncertainties which are dominated by the cross section and modeling parameters, while the fission yields have relatively insignificant effect. Besides, the primary photon dose is mostly influenced by the fission yield and modeling parameters, while the cross-section data have a relatively negligible effect. Moreover, the neutron, secondary photon, and primary photon dose can have uncertainties up to about 13%, 14%, and 6%, respectively.

Validation of nuclide depletion capabilities in Monte Carlo code MCS

Ebiwonjumi, Bamidele,Lee, Hyunsuk,Kim, Wonkyeong,Lee, Deokjung Korean Nuclear Society 2020 Nuclear Engineering and Technology Vol.52 No.9

In this work, the depletion capability implemented in Monte Carlo code MCS is investigated to predict the isotopic compositions of spent nuclear fuel (SNF). By comparison of MCS calculation results to post irradiation examination (PIE) data obtained from one pressurized water reactor (PWR), the validation of this capability is conducted. The depletion analysis is performed with the ENDF/B-VII.1 library and a fuel assembly model. The transmutation equation is solved by the Chebyshev Rational Approximation Method (CRAM) with a depletion chain of 3820 isotopes. 18 actinides and 19 fission products are analyzed in 14 SNF samples. The effect of statistical uncertainties on the calculated number densities is discussed. On average, most of the actinides and fission products analyzed are predicted within ±6% of the experiment. MCS depletion results are also compared to other depletion codes based on publicly reported information in literature. The code-to-code analysis shows comparable accuracy. Overall, it is demonstrated that the depletion capability in MCS can be reliably applied in the prediction of SNF isotopic inventory.

Verification and validation of isotope inventory prediction for back-end cycle management using two-step method

Jang, Jaerim,Ebiwonjumi, Bamidele,Kim, Wonkyeong,Cherezov, Alexey,Park, Jinsu,Lee, Deokjung Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.7

This paper presents the verification and validation (V&V) of a calculation module for isotope inventory prediction to control the back-end cycle of spent nuclear fuel (SNF). The calculation method presented herein was implemented in a two-step code system of a lattice code STREAM and a nodal diffusion code RAST-K. STREAM generates a cross section and provides the number density information using branch/history depletion branch calculations, whereas RAST-K supplies the power history and three history indices (boron concentration, moderator temperature, and fuel temperature). As its primary feature, this method can directly consider three-dimensional core simulation conditions using history indices of the operating conditions. Therefore, this method reduces the computation time by avoiding a recalculation of the fuel depletion. The module for isotope inventory calculates the number densities using the Lagrange interpolation method and power history correction factors, which are applied to correct the effects of the decay and fission products generated at different power levels. To assess the reliability of the developed code system for back-end cycle analysis, validation study was performed with 58 measured samples of pressurized water reactor (PWR) SNF, and code-to-code comparison was conducted with STREAM-SNF, HELIOS-1.6 and SCALE 5.1. The V&V results presented that the developed code system can provide reasonable results with comparable confidence intervals. As a result, this paper successfully demonstrates that the isotope inventory prediction code system can be used for spent nuclear fuel analysis.

Validation of spent nuclear fuel decay heat calculation by a two-step method

Jang, Jaerim,Ebiwonjumi, Bamidele,Kim, Wonkyeong,Park, Jinsu,Choe, Jiwon,Lee, Deokjung Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.1

In this paper, we validate the decay heat calculation capability via a two-step method to analyze spent nuclear fuel (SNF) discharged from pressurized water reactors (PWRs). The calculation method is implemented with a lattice code STREAM and a nodal diffusion code RAST-K. One of the features of this method is the direct consideration of three-dimensional (3D) core simulation conditions with the advantage of a short simulation time. Other features include the prediction of the isotope inventory by Lagrange non-linear interpolation and the use of power history correction factors. The validation is performed with 58 decay heat measurements of 48 fuel assemblies (FAs) discharged from five PWRs operated in Sweden and the United States. These realistic benchmarks cover the discharge burnup range up to 51 GWd/MTU, 23.2 years of cooling time, and spanning an initial uranium enrichment range of 2.100-4.005 wt percent. The SNF analysis capability of STREAM is also employed in the code-to-code comparison. Compared to the measurements, the validation results of the FA calculation with RAST-K are within ±4%, and the pin-wise results are within ±4.3%. This paper successfully demonstrates that the developed decay heat calculation method can perform SNF back-end cycle analyses.

Verification of photon transport capability of UNIST Monte Carlo code MCS

Lemaire, Matthieu,Lee, Hyunsuk,Ebiwonjumi, Bamidele,Kong, Chidong,Kim, Wonkyeong,Jo, Yunki,Park, Jinsu,Lee, Deokjung Elsevier 2018 Computer physics communications Vol.231 No.-

<P><B>Abstract</B></P> <P>A photon transport capability has been implemented and verified in the Monte Carlo code MCS of Ulsan National Institute of Science and Technology for the purpose of radiation shielding studies. The MCS photon fixed-source mode simulates the transport of photons between 1 keV and 20 MeV for all elements from hydrogen to fermium. The specific physics for the main four photo-atomic reactions (Rayleigh scattering, Compton scattering, photoelectric effect and pair production) and three secondary processes of photon production (positron–electron annihilation, atomic relaxation and electron/positron bremsstrahlung) are reviewed. Verification results against Monte Carlo codes MCNP6.1 and SERPENT2.1.29 are presented. The verification cases include the comparison of energy distributions of photon flux in an infinite medium, of spatial distributions of photon flux in a cylinder, of the spatial distribution of photon body-equivalent dose in a spent nuclear fuel transport cask, and of photon KERMA (Kinetic Energy Released per MAss) in photon detector calibration geometries. Good calculation/calculation agreement is observed overall, with some marked differences in the detailed photon flux comparison at given energy ranges traced back to differences in photon physics implementation. MCS can from now on be applied for the purpose of advanced photon studies and corresponding validation against experimental shielding benchmarks will follow in the future.</P>

Uncertainty quantification in decay heat calculation of spent nuclear fuel by STREAM/RAST-K

Jang, Jaerim,Kong, Chidong,Ebiwonjumi, Bamidele,Cherezov, Alexey,Jo, Yunki,Lee, Deokjung Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.9

This paper addresses the uncertainty quantification and sensitivity analysis of a depleted light-water fuel assembly of the Turkey Point-3 benchmark. The uncertainty of the fuel assembly decay heat and isotopic densities is quantified with respect to three different groups of diverse parameters: nuclear data, assembly design, and reactor core operation. The uncertainty propagation is conducted using a two-step analysis code system comprising the lattice code STREAM, nodal code RAST-K, and spent nuclear fuel module SNF through the random sampling of microscopic cross-sections, fuel rod sizes, number densities, reactor core total power, and temperature distributions. Overall, the statistical analysis of the calculated samples demonstrates that the decay heat uncertainty decreases with the cooling time. The nuclear data and assembly design parameters are proven to be the largest contributors to the decay heat uncertainty, whereas the reactor core power and inlet coolant temperature have a minor effect. The majority of the decay heat uncertainties are delivered by a small number of isotopes such as ²⁴¹Am, ¹³⁷Ba, ²⁴⁴Cm, ²³⁸Pu, and ⁹⁰Y.

On-the-fly energy release per fission model in STREAM with explicit neutron and photon heating

Mai Nhan Nguyen Trong,이웅희,김경원,Ebiwonjumi Bamidele Francis,김원경,이덕중 한국원자력학회 2023 Nuclear Engineering and Technology Vol.55 No.3

The on-the-fly energy release per fission (OTFK) model is implemented in STREAM to continuously update the Kappa values during the depletion calculation. The explicit neutron and photon energy distribution, which has not been considered in previous STREAM versions, is incorporated into the existing on-the-fly model. The impacts of the modified OTFK model with explicit neutron and photon heating in STREAM on the power distribution, fuel temperature, and other core parameters during depletion with feedback calculations are studied using several problems from the VERA benchmark suit. Overall, the explicit heating calculation provides a better power map for the feedback calculations particularly when strong gamma emitters are present. Generally, the fuel temperature decreases when neutron and photon heating is employed because fission neutrons and gamma rays are transported away from their points of generation. This energy release model in STREAM indicates that gamma energy accounts for approximately 9.5%e10% of the total energy released, and approximately 2.4%e2.6% of the total energy released will be deposited in the coolant for the VERA 5, NuScale, and Yonggwang Unit 3 2D cores.