scholarly journals Study on the Sensitivity and Uncertainty of Nuclear Data to the Sodium-Cooled Linear Breed-and-Burn Fast Reactor Using SCALE6.2 Code

2021 ◽  
Vol 2021 ◽  
pp. 1-13
Author(s):  
Thanh Mai Vu ◽  
Donny Hartanto
Keyword(s):  
Elastic Scattering ◽  
Energy Range ◽  
Fast Reactor ◽  
Spent Nuclear Fuel ◽  
Nuclear Data ◽  
High Energy ◽  
Data Uncertainty ◽  
Breed And Burn ◽  
Calculation Results ◽  
The Impact

Previously, the neutronics design of a small and compact linear breed-and-burn fast reactor (B&BR) was completed. The reactor produces 400 MWth power, and it can operate with excess reactivity of less than 1$ for more than 50 years without refuelling. As the blanket fuel, the spent nuclear fuel (SNF) from existing light water reactors (LWRs) is used to reduce the burden from the problematic long-lived isotopes in SNF. However, by loading massive nuclides at the initial core, the impact of nuclear data uncertainty on the reactivity calculation results of SNF-fuelled B&BR at the beginning of life (BOL) is expected to be significant because these nuclides have different credentials in evaluated nuclear data libraries. In this study, the impact of nuclear library uncertainty from ENDF/B-VII.0 and ENDF/B-VII.1 on reactivity calculation of B&BR is evaluated using the continuous-energy TSUNAMI-3D module in the SCALE6.2 code package. The uncertainty of reactivity calculation results of B&BR caused by the inaccuracy of two libraries is significant (more than 2000 pcm), mainly from the uncertainty of 235,238U and 56Fe cross section. The energy-dependent sensitivity profiles show that they are significant at the fast energy range. The uncertainty of coolant void reactivity (CVR) is about 18%, and that of fuel temperature coefficient (FTC) is about 15% of the reactivity effect. The top five contributions for CVR accounted for elastic scattering of 238U, capture of 235,238U, and elastic scattering of 23Na and 56Fe. Meanwhile, the top contributors for FTC were accounted for elastic scattering of 238U and 56Fe, capture of 235U, and elastic scattering of 94Zr and 57Fe. It is highly recommended to improve the accuracy of those isotopes’ cross sections at the high energy range to provide a more reliable reactivity calculation for the fast system.

scholarly journals Nuclear Data Uncertainty Quantification in Criticality Safety Evaluations for Spent Nuclear Fuel Geological Disposal

2021 ◽  
Vol 11 (14) ◽  
pp. 6499
Author(s):  
Matthias Frankl ◽  
Mathieu Hursin ◽  
Dimitri Rochman ◽  
Alexander Vasiliev ◽  
Hakim Ferroukhi
Keyword(s):  
Uncertainty Quantification ◽  
Nuclear Fuel ◽  
Evaluation Method ◽  
Spent Nuclear Fuel ◽  
Nuclear Data ◽  
Data Uncertainty ◽  
Evaluation Methodology ◽  
Geological Disposal ◽  
Fuel Assemblies ◽  
Criticality Safety

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.

Nuclear Data Uncertainty Propagation for a Sodium Fast Reactor

2011 ◽  
Vol 59 (2(3)) ◽  
pp. 1191-1194 ◽  
Author(s):  
D. Rochman ◽  
A. J. Koning ◽  
D. F. Dacruz ◽  
S. C. van der Marck
Keyword(s):  
Fast Reactor ◽  
Uncertainty Propagation ◽  
Nuclear Data ◽  
Data Uncertainty ◽  
Sodium Fast Reactor

scholarly journals Effects of different nuclear evaluation data on the RMC keff calculation

2020 ◽  
Vol 239 ◽  
pp. 19005
Author(s):  
Zhang Wenxin ◽  
Qiang shenglong ◽  
Yin qiang ◽  
Cui Xiantao
Keyword(s):  
Cross Section ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Neutron Cross Section ◽  
Nuclear Data ◽  
Section Data ◽  
Calculation Results ◽  
The Impact ◽  
Physical Calculation

Neutron cross section data is the basis of nuclear reactor physical calculation and has a decisive influence on the accuracy of calculation results. AFA3Gassemble is widely used in nuclear power plants. CENACE is an ACE format multiple-temperature continuous energy cross section library that developed by China Nuclear Data Centre. In this paper, we calculated the AFA3G assemble by RMC.We respectively used ENDF6.8/, ENDF/7 and CENACE data for calculation. The impact of nuclear data on RMC calculation is studied by comparing the results of different nuclear data.

Mathematical Modeling of High Energy Ball Milling (HEBM) Process

2014 ◽  
Vol 353 ◽  
pp. 126-130 ◽  
Author(s):  
Maya Radune ◽  
A. Radune ◽  
Svetlana Lugovskoy ◽  
M. Zinigrad ◽  
David Fuks ◽  
...  
Keyword(s):  
Reasonable Agreement ◽  
Rotation Speed ◽  
Weight Ratio ◽  
High Energy ◽  
Impact Angle ◽  
Velocity Impact ◽  
Calculation Results ◽  
Angle Rotation ◽  
Energy Ball ◽  
The Impact

In the present study the modeling of the HEBM process is presented. The impact velocity, impact angle, rotation speed, mass of balls, ball-to-powder weight ratio and time of milling have been taken into account in order to calculate the energy transferred from the balls to the powder. Two different systems, namely, TiN-AlN and polysalicylic acid were experimentally investigated in order to confirm the validity of the model. The calculation results are in a reasonable agreement with the results of experimental research.

Pseudopotential calculations of elastic scattering of low-energy electrons (< 20 eV) from neon atoms

1992 ◽  
Vol 70 (5) ◽  
pp. 305-310 ◽  
Author(s):  
Y. Frongillo ◽  
B. Plenkiewicz ◽  
P. Plenkiewicz ◽  
J.-P. Jay-Gerin
Keyword(s):  
Elastic Scattering ◽  
Energy Range ◽  
Cross Sections ◽  
Impact Energy ◽  
Theoretical Data ◽  
Low Energy ◽  
Pseudopotential Calculations ◽  
Low Energy Electrons ◽  
The Impact ◽  
Good Agreement

Pseudopotential calculations of phase shifts, differential, total, and momentum-transfer cross sections for electrons elastically scattered from neon atoms are reported in the impact energy range 0–20 eV. The results are found to be in very good agreement with existing experimental and other theoretical data.

scholarly journals Nuclear Data Uncertainty Propagation to Reactivity Coefficients of a Sodium Fast Reactor

2014 ◽  
Vol 118 ◽  
pp. 535-537
Author(s):  
J.J. Herrero ◽  
R. Ochoa ◽  
J.S. Martínez ◽  
C.J. Díez ◽  
N. García-Herranz ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Uncertainty Propagation ◽  
Nuclear Data ◽  
Data Uncertainty ◽  
Sodium Fast Reactor ◽  
Reactivity Coefficients

scholarly journals Impact of nuclear data evaluations on data assimilation for an LFR

2020 ◽  
Vol 239 ◽  
pp. 13007
Author(s):  
Pablo Romojaro ◽  
Francisco Álvarez-Velarde
Keyword(s):  
Data Assimilation ◽  
Fast Reactor ◽  
Nuclear Reactor ◽  
Critical Mass ◽  
Nuclear Data ◽  
Operational Experience ◽  
Operation Conditions ◽  
Nuclear Data Library ◽  
Safety Parameters ◽  
The Impact

The Lead-cooled Fast Reactor is one of the three technologies selected by the Sustainable Nuclear Energy Technology Platform that can meet future European energy needs. The main drawbacks for the industrial deployment of LFR are the lack of operational experience and the impact of uncertainties. In nuclear reactor design the uncertainties mainly come from material properties, fabrication tolerances, operation conditions, simulation tools and nuclear data. The uncertainty in nuclear data is one of the most important sources of uncertainty in reactor physics simulations. Furthermore, it is known that the uncertainties in reactor criti-cality safety parameters are severely dependent on the nuclear data library used to estimate them. However, the impact of using different evaluations while performing data assimilation to constraint the uncertainties in the criticality parameters has not been properly assessed yet. In this work, a data assimilation for the main isotopes contributing to the uncertainty in keff of the ALFRED lead-cooled fast reactor has been performed with the SUMMON system using JEFF-3.3, ENDF/B-VIII.0 and JENDL-4.0u2 state-of-the-art nuclear data libraries, together with critical mass experiments from the International Criticality Safety Benchmark Evaluation Project that are representative of ALFRED, in order to assess the impact of using different evaluations for data assimilation.

scholarly journals Trends on major actinides from an integral data assimilation

2020 ◽  
Vol 239 ◽  
pp. 13002
Author(s):  
Gerald Rimpault ◽  
Gilles Noguère ◽  
Cyrille de Saint Jean
Keyword(s):  
Data Assimilation ◽  
Energy Range ◽  
Cross Section ◽  
Capture Cross Section ◽  
Critical Mass ◽  
Nuclear Data ◽  
Data Uncertainty ◽  
Capture Cross ◽  
Plateau Region ◽  
Integral Data

The objective of this work is to revisit integral data assimilation for a better prediction of the characteristics of SFR cores. ICSBEP, IRPhE and MASURCA critical masses, PROFIL irradiation experiments and the FCA-IX experimental programme (critical masses and spectral indices) with well-mastered experimental technique have been used. As calculations are performed without modelling errors (with as-built geometries) and without approximations with the TRIPOLI4 MC code, highly reliable C/E are achieved. Assimilation results suggest a 2.5% decrease for 238U capture from 3 keV to 60 keV, and a 4-5% decrease for 238U inelastic in the plateau region. For this energy range, uncertainties are respectively reduced to 1-2% and to 2-2.5% for 238U capture and 238U inelastic respectively. The increase trends on 239Pu capture cross section of around 3% in the [2 keV-100 keV] energy range come from a low PROFIL 240Pu/239Pu ratio C/E. For 240Pu capture cross section, the increase trend of around 4% in the [3 keV-100 keV] energy range goes in the same direction as the recent ENDF/B.VIII evaluation though at a much lower level. The nuclear data uncertainty associated to SFR ASTRID critical mass is reduced to 470 pcm.

scholarly journals Impact of nuclear data uncertainty on safety calculations for spent nuclear fuel geological disposal

2017 ◽  
Vol 146 ◽  
pp. 09028 ◽  
Author(s):  
J.J. Herrero ◽  
D. Rochman ◽  
O. Leray ◽  
A. Vasiliev ◽  
M. Pecchia ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Nuclear Data ◽  
Data Uncertainty ◽  
Geological Disposal

scholarly journals Examples of Monte Carlo techniques applied for nuclear data uncertainty propagation

2019 ◽  
Vol 211 ◽  
pp. 07008 ◽  
Author(s):  
Oscar Cabellos ◽  
Luca Fiorito
Keyword(s):  
Monte Carlo ◽  
Large Scale ◽  
Uncertainty Propagation ◽  
Nuclear Data ◽  
Monte Carlo Technique ◽  
Data Uncertainty ◽  
Design Parameters ◽  
Monte Carlo Techniques ◽  
Bayesian Monte Carlo ◽  
The Impact

The aim of this work is to review different Monte Carlo techniques used to propagate nuclear data uncertainties. Firstly, we introduced Monte Carlo technique applied for Uncertainty Quantification studies in safety calculations of large scale systems. As an example, the impact of nuclear data uncertainty of JEFF-3.3 235U, 238U and 239Pu is demonstrated for the main design parameters of a typical 3-loop PWR Westinghouse unit. Secondly, the Bayesian Monte Carlo technique for data adjustment is presented. An example for 235U adjustment using criticality and shielding integral benchmarks shows the importance of performing joint adjustment based on different set of integral benchmarks.

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