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Uncertainty quantification in decay heat calculation of spent nuclear fuel by STREAM/RAST-K. NUCLEAR ENGINEERING AND TECHNOLOGY 2021. [DOI: 10.1016/j.net.2021.03.010] [Citation(s) in RCA: 2] [Impact Index Per Article: 0.7] [Reference Citation Analysis] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/23/2022]
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Mercatali L, Beydogan N, Sanchez-Espinoza V. Simulation of low-enriched uranium burnup in Russian VVER-1000 reactors with the Serpent Monte-Carlo code. NUCLEAR ENGINEERING AND TECHNOLOGY 2021. [DOI: 10.1016/j.net.2021.03.014] [Citation(s) in RCA: 2] [Impact Index Per Article: 0.7] [Reference Citation Analysis] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/17/2022]
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Shama A, Rochman D, Pudollek S, Caruso S, Pautz A. Uncertainty analyses of spent nuclear fuel decay heat calculations using SCALE modules. NUCLEAR ENGINEERING AND TECHNOLOGY 2021. [DOI: 10.1016/j.net.2021.03.013] [Citation(s) in RCA: 1] [Impact Index Per Article: 0.3] [Reference Citation Analysis] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 10/21/2022]
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Jang J, Kong C, Ebiwonjumi B, Jo Y, Lee D. Uncertainties of PWR spent nuclear fuel isotope inventory for back-end cycle analysis with STREAM/RAST-K. ANN NUCL ENERGY 2021. [DOI: 10.1016/j.anucene.2021.108267] [Citation(s) in RCA: 2] [Impact Index Per Article: 0.7] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 10/21/2022]
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Nuclear Data Uncertainty Quantification in Criticality Safety Evaluations for Spent Nuclear Fuel Geological Disposal. APPLIED SCIENCES-BASEL 2021. [DOI: 10.3390/app11146499] [Citation(s) in RCA: 4] [Impact Index Per Article: 1.3] [Reference Citation Analysis] [Abstract] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/17/2022]
Abstract
Presently, a criticality safety evaluation methodology for the final geological disposal of Swiss spent nuclear fuel is under development at the Paul Scherrer Institute in collaboration with the Swiss National Technical Competence Centre in the field of deep geological disposal of radioactive waste. This method in essence pursues a best estimate plus uncertainty approach and includes burnup credit. Burnup credit is applied by means of a computational scheme called BUCSS-R (Burnup Credit System for the Swiss Reactors–Repository case) which is complemented by the quantification of uncertainties from various sources. BUCSS-R consists in depletion, decay and criticality calculations with CASMO5, SERPENT2 and MCNP6, respectively, determining the keff eigenvalues of the disposal canister loaded with the Swiss spent nuclear fuel assemblies. However, the depletion calculation in the first and the criticality calculation in the third step, in particular, are subject to uncertainties in the nuclear data input. In previous studies, the effects of these nuclear data-related uncertainties on obtained keff values, stemming from each of the two steps, have been quantified independently. Both contributions to the overall uncertainty in the calculated keff values have, therefore, been considered as fully correlated leading to an overly conservative estimation of total uncertainties. This study presents a consistent approach eliminating the need to assume and take into account unrealistically strong correlations in the keff results. The nuclear data uncertainty quantification for both depletion and criticality calculation is now performed at once using one and the same set of perturbation factors for uncertainty propagation through the corresponding calculation steps of the evaluation method. The present results reveal the overestimation of nuclear data-related uncertainties by the previous approach, in particular for spent nuclear fuel with a high burn-up, and underline the importance of consistent nuclear data uncertainty quantification methods. However, only canister loadings with UO2 fuel assemblies are considered, not offering insights into potentially different trends in nuclear data-related uncertainties for mixed oxide fuel assemblies.
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Jang J, Ebiwonjumi B, Kim W, Cherezov A, Park J, Lee D. Verification and validation of isotope inventory prediction for back-end cycle management using two-step method. NUCLEAR ENGINEERING AND TECHNOLOGY 2021. [DOI: 10.1016/j.net.2021.01.009] [Citation(s) in RCA: 3] [Impact Index Per Article: 1.0] [Reference Citation Analysis] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 10/22/2022]
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Ebiwonjumi B, Kong C, Zhang P, Cherezov A, Lee D. Uncertainty quantification of PWR spent fuel due to nuclear data and modeling parameters. NUCLEAR ENGINEERING AND TECHNOLOGY 2021. [DOI: 10.1016/j.net.2020.07.012] [Citation(s) in RCA: 10] [Impact Index Per Article: 3.3] [Reference Citation Analysis] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/15/2022]
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A new methodology to estimate stochastic uncertainty of MCNP-predicted isotope concentrations in nuclear fuel burnup simulations. ANN NUCL ENERGY 2021. [DOI: 10.1016/j.anucene.2020.107911] [Citation(s) in RCA: 3] [Impact Index Per Article: 1.0] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/20/2022]
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Rochman D, Vasiliev A, Ferroukhi H, Hursin M. Analysis for the ARIANE BM1 and BM3 samples: nuclide inventory and decay heat. EPJ NUCLEAR SCIENCES & TECHNOLOGIES 2021. [DOI: 10.1051/epjn/2021017] [Citation(s) in RCA: 1] [Impact Index Per Article: 0.3] [Reference Citation Analysis] [Abstract] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/14/2022] Open
Abstract
The Mixed Oxide samples (MOX) ARIANE Post Irradiation Examination samples BM1 and BM3 have been analyzed in this work, based on various two- and three-dimensional models. Calculated and measured nuclide inventories are compared based on CASMO5, SIMULATE and SNF simulations, and calculated values for the decay heat of the assembly containing the samples are also provided. For uncertainty propagation, the covariance information from three different nuclear data libraries are used. Uncertainties from manufacturing tolerances and operating conditions are also considered. The results from these two samples are compared with the ones from two UO2 samples, namely GU1 and GU3, also from the ARIANE program, applying the same calculation scheme and uncertainty assumptions. It is shown that a two-dimensional assembly model provides better agreement with the measurements than a two-dimensional single pin model, and that the full core three-dimensional model provides similar results compared to the assembly model, although no 148Nd normalization is applied for the full core model. For the MOX assembly decay heat, as expected, heavy actinides have a higher contribution compared to the cases with the UO2 samples; additionally, decay heat uncertainties are moderately smaller in the case of the MOX assembly.
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Rochman D, Vasiliev A, Ferroukhi H, Hursin M, Ichou R, Taforeau J, Simeonov T. Analysis for the ARIANE GU3 sample: nuclide inventory and decay heat. EPJ NUCLEAR SCIENCES & TECHNOLOGIES 2021. [DOI: 10.1051/epjn/2021013] [Citation(s) in RCA: 1] [Impact Index Per Article: 0.3] [Reference Citation Analysis] [Abstract] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/14/2022] Open
Abstract
This study presents an analysis of the ARIANE GU3 sample, in terms of nuclide inventory, as well as sample rod and assembly decay heat. The validation of a number of CASMO5 and library versions are performed with regards to the measured nuclide inventory, taking into account two dimensional lattice simulations. Uncertainties due to various sources (nuclear data, operating conditions and manufacturing tolerances) are also provided, and are combined with biases into expanded uncertainties. This study is similar to a previous one on the GU1 sample and fit in the framework of code validation, as well as in the estimation of code predictive power for spent fuel characterization.
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Hursin M, Rochman D, Vasiliev A, Ferroukhi H, Pautz A. IMPACT OF VARIOUS SOURCE OF COVARIANCE INFORMATION ON INTEGRAL PARAMETERS UNCERTAINTY DURING DEPLETION CALCULATIONS WITH CASMO-5. EPJ WEB OF CONFERENCES 2021. [DOI: 10.1051/epjconf/202124709005] [Citation(s) in RCA: 1] [Impact Index Per Article: 0.3] [Reference Citation Analysis] [Abstract] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/14/2022] Open
Abstract
This paper describes the effect of input uncertainties on a set of integral parameters (kinf, nuclide compositions) associated with the validation of CASMO-5 against PIE data. The nuclear data under consideration are the cross-sections, fission spectrum and neutron multiplicities and fission yields. Various sources of covariance information are considered, novel ones (ENDFB-VIII.0, JEFF-3.3) as well as more widely distributed ones (JENDL-4.0, ENDF/B-VII.1, Scale 6.1 and Scale 6.2). All possible nuclide reaction pairs (cross sections, fission spectrum and averaged number of neutron per fission) have been perturbed, e.g. all isotopes available in both the respective covariance libraries and the CASMO-5 library. The evolution of the uncertainty estimates with exposure is complemented with sensitivity analysis to determine the main contributors to the uncertainty. The Pearson coefficient defined between the model output and a given input is used in this work to assess the part of the variance in the model output coming from the considered input uncertainty. It is a very promising measure of sensitivity as it is computationally cheap even though it assumes linearity of the output with respect to input perturbations. The evolution of the uncertainty with exposure, both in terms of trends and magnitude are however very different. Sensitivity analysis allows determining why the trends and magnitudes are different.
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Perret G, Rochman D, Vasiliev A, Ferroukhi H. NEUTRON EMISSION MEASUREMENTS OF PWR SPENT FUEL SEGMENTS AND PRELIMINARY VALIDATION OF DEPLETION CALCULATIONS. EPJ WEB OF CONFERENCES 2021. [DOI: 10.1051/epjconf/202124710004] [Citation(s) in RCA: 0] [Impact Index Per Article: 0] [Reference Citation Analysis] [Abstract] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/15/2022] Open
Abstract
Assessing neutron emission of LWR spent fuel is necessary for the back-end of the fuel cycle, such as the dimensioning of transport and storage casks of spent fuel. Although core and depletion codes can calculate the isotopic composition of the discharged fuel and therefore infer its neutron source, accurate measured neutron emission values remain rare mainly because of the difficulty to prepare, handle and characterize spent fuel. Measured neutron emission values are, however, extremely relevant to code validation, as neutrons emitted by LWR spent fuel mainly originates from spontaneous fissions of minor actinides (e.g., 242Cm, 244Cm and 252Cf) that are produced only after a large number of neutron captures in the reactor core. This paper reports on neutron emission measurements of selected LWR-PROTEUS spent fuel samples and their comparisons with a core and depletion calculation chains based on CASMO-5, SIMULATE-3 and the SNF codes. The measured LWR-PROTEUS samples are comprised of 11 samples irradiated in a Swiss PWR. The samples are UO2 or MOX and have discharge burn-ups ranging from 20 to 120 GWd/t. We measured the 40-cm long samples in a hot-cell of the Paul Scherrer Institut using a measurement station made of polyethylene and a BF3 detector. We repeated the measurements several times and in different conditions to ensure the accuracy and reproducibility of the results. We derived ratios of neutron rates emitted by the different samples and absolute neutron emission rates by comparison with a reference 252Cf source, which we re-calibrated for this exercise. The experimental uncertainty (1σ) on the absolute neutron emission varies from 3% to 4%. We compared a subset of the measured values to the calculation predictions and showed an agreement within less than 7% for all but one sample.
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Rochman DA, Bauge E. Fission yields and cross sections: correlated or not? EPJ NUCLEAR SCIENCES & TECHNOLOGIES 2021. [DOI: 10.1051/epjn/2021005] [Citation(s) in RCA: 1] [Impact Index Per Article: 0.3] [Reference Citation Analysis] [Abstract] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/14/2022] Open
Abstract
Cross sections and fission yields can be correlated, depending on the selection of integral experimental data. To support this statement, this work presents the use of experimental isotopic compositions (both for actinides and fission products) from a sample irradiated in a reactor, to construct correlations between various cross sections and fission yields. This study is therefore complementing previous analysis demonstrating that different types of nuclear data can be correlated, based on experimental integral data.
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Nuclear data uncertainty propagation and modeling uncertainty impact evaluation in neutronics core simulation. PROGRESS IN NUCLEAR ENERGY 2020. [DOI: 10.1016/j.pnucene.2020.103443] [Citation(s) in RCA: 4] [Impact Index Per Article: 1.0] [Reference Citation Analysis] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/22/2022]
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Siefman D, Hursin M, Perret G, Pautz A. Applying SHARK-X to perform data assimilation with the LWR-PROTEUS Phase II integral experiments. PROGRESS IN NUCLEAR ENERGY 2020. [DOI: 10.1016/j.pnucene.2020.103245] [Citation(s) in RCA: 0] [Impact Index Per Article: 0] [Reference Citation Analysis] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 10/25/2022]
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Kim W, Hursin M, Pautz A, Vincent L, Pavel F, Lee D. Determination of the activity inventory and associated uncertainty quantification for the CROCUS zero power research reactor. ANN NUCL ENERGY 2020. [DOI: 10.1016/j.anucene.2019.107034] [Citation(s) in RCA: 0] [Impact Index Per Article: 0] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 10/26/2022]
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Siefman D, Hursin M, Sjostrand H, Schnabel G, Rochman D, Pautz A. Data assimilation of post-irradiation examination data for fission yields from GEF. EPJ NUCLEAR SCIENCES & TECHNOLOGIES 2020. [DOI: 10.1051/epjn/2020015] [Citation(s) in RCA: 4] [Impact Index Per Article: 1.0] [Reference Citation Analysis] [Abstract] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/14/2022] Open
Abstract
Nuclear data, especially fission yields, create uncertainties in the predicted concentrations of fission products in spent fuel which can exceed engineering target accuracies. Herein, we present a new framework that extends data assimilation methods to burnup simulations by using post-irradiation examination experiments. The adjusted fission yields lowered the bias and reduced the uncertainty of the simulations. Our approach adjusts the model parameters of the code GEF. We compare the BFMC and MOCABA approaches to data assimilation, focusing especially on the effects of the non-normality of GEF’s fission yields. In the application that we present, the best data assimilation framework decreased the average bias of the simulations from 26% to 14%. The average relative standard deviation decreased from 21% to 14%. The GEF fission yields after data assimilation agreed better with those in JEFF3.3. For Pu-239 thermal fission, the average relative difference from JEFF3.3 was 16% before data assimilation and after it was 12%. For the standard deviations of the fission yields, GEF’s were 100% larger than JEFF3.3’s before data assimilation and after were only 4% larger. The inconsistency of the integral data had an important effect on MOCABA, as shown with the Marginal Likelihood Optimization method. When the method was not applied, MOCABA’s adjusted fission yields worsened the bias of the simulations by 30%. BFMC showed that it inherently accounted for this inconsistency. Applying Marginal Likelihood Optimization with BFMC gave a 2% lower bias compared to not applying it, but the results were more poorly converged.
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Siefman D, Hursin M, Pautz A. Data assimilation of post irradiation examination experiments to adjust fission yields. EPJ WEB OF CONFERENCES 2020. [DOI: 10.1051/epjconf/202023913004] [Citation(s) in RCA: 0] [Impact Index Per Article: 0] [Reference Citation Analysis] [Abstract] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/14/2022] Open
Abstract
Nuclear data, especially fission yields, create uncertainties in the predicted concentrations of fission products in spent fuel. Herein, we present a new framework that extends data assimilation methods to burnup simulations by using data from post-irradiation examination experiments. The adjusted fission yields improve the bias and reduce the uncertainty of predicted fission product concentrations in spent fuel. Our approach modifies fission yields by adjusting the model parameters of the code GEF with post-irradiation examination experiments. We used the BFMC data assimilation method to account for the non-normality of GEF's fission yields. In the application that we present, the assimilation decreased the average bias of the predicted fission product concentrations from 26% to 15%. The average relative standard deviation decreased from 21% to 14%. The GEF fission yields after data assimilation agreed better with those in ENDF/B-VIII.O. For Pu-239 thermal fission, the average relative difference from ENDF/B-VIII.O was 16% before data assimilation and 11% after. For the standard deviations of the fission yields, GEF's were, on average, 16% larger than those from ENDF/B-VIII.O before data assimilation and 15% smaller after.
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Park J, Kim W, Hursin M, Perret G, Vasiliev A, Rochman D, Pautz A, Ferroukhi H, Lee D. Uncertainty quantification of LWR-PROTEUS Phase II experiments using CASMO-5. ANN NUCL ENERGY 2019. [DOI: 10.1016/j.anucene.2019.03.023] [Citation(s) in RCA: 3] [Impact Index Per Article: 0.6] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 10/27/2022]
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Price D, Radaideh MI, O'Grady D, Kozlowski T. Advanced BWR criticality safety part II: Cask criticality, burnup credit, sensitivity, and uncertainty analyses. PROGRESS IN NUCLEAR ENERGY 2019. [DOI: 10.1016/j.pnucene.2019.03.039] [Citation(s) in RCA: 7] [Impact Index Per Article: 1.4] [Reference Citation Analysis] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 10/27/2022]
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Preliminary Assessment of Criticality Safety Constraints for Swiss Spent Nuclear Fuel Loading in Disposal Canisters. MATERIALS 2019; 12:ma12030494. [PMID: 30764572 PMCID: PMC6384812 DOI: 10.3390/ma12030494] [Citation(s) in RCA: 8] [Impact Index Per Article: 1.6] [Reference Citation Analysis] [Abstract] [Key Words] [Track Full Text] [Download PDF] [Figures] [Subscribe] [Scholar Register] [Received: 12/18/2018] [Revised: 01/24/2019] [Accepted: 01/28/2019] [Indexed: 11/19/2022]
Abstract
This paper presents preliminary criticality safety assessments performed by the Paul Scherrer Institute (PSI) in cooperation with the Swiss National Cooperative for the Disposal of Radioactive Waste (Nagra) for spent nuclear fuel disposal canisters loaded with Swiss Pressurized Water Reactor (PWR) UO2 spent fuel assemblies. The burnup credit application is examined with respect to both existing concepts: taking into account actinides only and taking into account actinides plus fission products. The criticality safety calculations are integrated with uncertainty quantifications that are as detailed as possible, accounting for the uncertainties in the nuclear data used, fuel assembly and disposal canister design parameters and operating conditions, as well as the radiation-induced changes in the fuel assembly geometry. Furthermore, the most penalising axial and radial burnup profiles and the most reactive fuel loading configuration for the canisters were taken into account accordingly. The results of the study are presented with the help of loading curves showing what minimum average fuel assembly burnup is required for the given initial fuel enrichment of fresh fuel assemblies to ensure that the effective neutron multiplication factor, keff, of the canister would comply with the imposed criticality safety criterion.
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On the options for incorporating nuclear data uncertainties in criticality safety assessments for LWR fuel. ANN NUCL ENERGY 2018. [DOI: 10.1016/j.anucene.2018.01.046] [Citation(s) in RCA: 10] [Impact Index Per Article: 1.7] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/19/2022]
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Rochman DA, Vasiliev A, Dokhane A, Ferroukhi H. Uncertainties for Swiss LWR spent nuclear fuels due to nuclear data. EPJ NUCLEAR SCIENCES & TECHNOLOGIES 2018. [DOI: 10.1051/epjn/2018005] [Citation(s) in RCA: 11] [Impact Index Per Article: 1.8] [Reference Citation Analysis] [Abstract] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/14/2022] Open
Abstract
This paper presents a study of the impact of the nuclear data (cross sections, neutron emission and spectra) on different quantities for spent nuclear fuels (SNF) from Swiss power plants: activities, decay heat, neutron and gamma sources and isotopic vectors. Realistic irradiation histories are considered using validated core follow-up models based on CASMO and SIMULATE. Two Pressurized and one Boiling Water Reactors (PWR and BWR) are considered over a large number of operated cycles. All the assemblies at the end of the cycles are studied, being reloaded or finally discharged, allowing spanning over a large range of exposure (from 4 to 60 MWd/kgU for ≃9200 assembly-cycles). Both UO2 and MOX fuels were used during the reactor cycles, with enrichments from 1.9 to 4.7% for the UO2 and 2.2 to 5.8% Pu for the MOX. The SNF characteristics presented in this paper are calculated with the SNF code. The calculated uncertainties, based on the ENDF/B-VII.1 library are obtained using a simple Monte Carlo sampling method. It is demonstrated that the impact of nuclear data is relatively important (e.g. up to 17% for the decay heat), showing the necessity to consider them for safety analysis of the SNF handling and disposal.
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Jaffke P. Identifying Inconsistencies in Fission Product Yield Evaluations with Prompt Neutron Emission. NUCL SCI ENG 2018. [DOI: 10.1080/00295639.2018.1429173] [Citation(s) in RCA: 4] [Impact Index Per Article: 0.7] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 10/17/2022]
Affiliation(s)
- Patrick Jaffke
- Los Alamos National Laboratory, Theoretical Division, Los Alamos, New Mexico 87545
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Leray O, Ferroukhi H, Hursin M, Vasiliev A, Rochman D. Methodology for core analyses with nuclear data uncertainty quantification and application to Swiss PWR operated cycles. ANN NUCL ENERGY 2017. [DOI: 10.1016/j.anucene.2017.07.006] [Citation(s) in RCA: 28] [Impact Index Per Article: 4.0] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/29/2022]
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Terranova N, Serot O, Archier P, De Saint Jean C, Sumini M. Fission yield covariance matrices for the main neutron-induced fissioning systems contained in the JEFF-3.1.1 library. ANN NUCL ENERGY 2017. [DOI: 10.1016/j.anucene.2017.05.052] [Citation(s) in RCA: 6] [Impact Index Per Article: 0.9] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 10/19/2022]
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Grimm P, Hursin M, Perret G, Siefman D, Ferroukhi H. Analysis of reactivity worths of burnt PWR fuel samples measured in LWR-PROTEUS Phase II using a CASMO-5 reflected-assembly model. PROGRESS IN NUCLEAR ENERGY 2017. [DOI: 10.1016/j.pnucene.2017.03.018] [Citation(s) in RCA: 6] [Impact Index Per Article: 0.9] [Reference Citation Analysis] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/30/2022]
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Herrero J, Rochman D, Leray O, Vasiliev A, Pecchia M, Ferroukhi H, Caruso S. Impact of nuclear data uncertainty on safety calculations for spent nuclear fuel geological disposal. EPJ WEB OF CONFERENCES 2017. [DOI: 10.1051/epjconf/201714609028] [Citation(s) in RCA: 4] [Impact Index Per Article: 0.6] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/14/2022] Open
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Ilas G, Liljenfeldt H. Decay heat uncertainty for BWR used fuel due to modeling and nuclear data uncertainties. NUCLEAR ENGINEERING AND DESIGN 2017. [DOI: 10.1016/j.nucengdes.2017.05.009] [Citation(s) in RCA: 24] [Impact Index Per Article: 3.4] [Reference Citation Analysis] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/30/2022]
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Exploring Stochastic Sampling in Nuclear Data Uncertainties Assessment for Reactor Physics Applications and Validation Studies. ENERGIES 2016. [DOI: 10.3390/en9121039] [Citation(s) in RCA: 6] [Impact Index Per Article: 0.8] [Reference Citation Analysis] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/16/2022]
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Rochman D, Leray O, Vasiliev A, Ferroukhi H, Koning A, Fleming M, Sublet J. A Bayesian Monte Carlo method for fission yield covariance information. ANN NUCL ENERGY 2016. [DOI: 10.1016/j.anucene.2016.05.005] [Citation(s) in RCA: 26] [Impact Index Per Article: 3.3] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/29/2022]
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