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.

Investigation of Fuel Burnup Impacts on Nuclear Reactor Safety Parameters in the Canadian Pressure Tube Supercritical Water-Cool Reactor

2017 ◽  
Vol 4 (1) ◽  
Author(s):  
Ahmad Moghrabi ◽  
David Raymond Novog
Keyword(s):  
Nuclear Reactor ◽  
High Efficiency ◽  
Fission Products ◽  
Neutron Cross Section ◽  
Fuel Mixture ◽  
Pressure Tube ◽  
Candu Reactors ◽  
Safety Parameters ◽  
Nuclear Reactor Safety ◽  
The Impact

The Canadian pressure-tube super critical water-cooled reactor (PT-SCWR) is an advanced generation IV reactor concept which is considered as an evolution of the conventional Canada Deuterium Uranium (CANDU) reactor that includes both pressure tubes and a low temperature and pressure heavy water moderator. The Canadian PT-SCWR fuel assembly utilizes a plutonium and thorium fuel mixture with supercritical light water coolant flowing through the high-efficiency re-entrance channel (HERC). In this work, the impact of fuel depletion on the evolution of lattice physics phenomena was investigated starting from fresh fuel to burnup conditions (25 MW d kg−1 [HM]) through sensitivity and uncertainty analyses using the lattice physics modules in standardized computer analysis for licensing evaluation (SCALE). Given the evolution of key phenomena such as void reactivity in traditional CANDU reactors with burnup, this study focuses on the impact of fission products, 233U breeding, and minor actinides on fuel performance. The work shows that the most significant change in fuel properties with burnup is the depletion of fission isotopes of Pu and the buildup of high-neutron cross section fission products, resulting in a decrease in cell k∞ with burnup as expected. Other impacts such as the presence of protactinium and uranium-233 are also discussed. When the feedback coefficients are assessed in terms of reactivity, there is considerable variation as a function of fuel depletion; however, when assessed as Δk (without normalization to the reference reactivity which changes with burnup), the net changes are almost invariant with depletion.

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.

scholarly journals Impact of the cross section library on 93mNb activity in VVER-1000 reactor dosimetry

2020 ◽  
Vol 225 ◽  
pp. 03009
Author(s):  
P. Haroková ◽  
M. Lovecký
Keyword(s):  
Cross Section ◽  
Neutron Transport ◽  
Nuclear Data ◽  
Nuclear Data Library ◽  
The Cross ◽  
Materials Used ◽  
The Impact ◽  
Data Library ◽  
Reactor Dosimetry

One of the objectives of reactor dosimetry is determination of activity of irradiated dosimeters, which are placed on reactor pressure vessel surface, and calculation of neutron flux in their position. The uncertainty of calculation depends mainly on the choice of nuclear data library, especially cross section used for neutron transport and cross section used as the response function for neutron activation. Nowadays, number of libraries already exists and can be still used in some applications. In addition, new nuclear data library was recently released. In this paper, we have investigated the impact of the cross section libraries on activity of niobium, one of the popular materials used as neutron fluence monitor. For this purpose, a MCNP6 model of VVER-1000 was made and we have compared the results between 14 commonly used cross section libraries. A possibility of using IRDFF library in activation calculations was also considered. The results show good agreement between the new libraries, with the exception of the most recent ENDF/B-VIII.0, which should be further validated.

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 Comparison of Deterministic and Stochastic Depletions in Graphite-Filled MOX Fuel Assembly Design

2020 ◽  
Vol 36 (4) ◽  
Author(s):  
Thanh Mai Vu ◽  
Donny Hartanto
Keyword(s):  
Fuel Assembly ◽  
Nuclear Data ◽  
Absolute Difference ◽  
Assembly Design ◽  
Mox Fuel ◽  
Nuclear Data Library ◽  
Calculation Results ◽  
The Impact ◽  
Data Library ◽  
Good Agreement

A comparison between stochastic and deterministic depletion calculations based on a graphite-filled MOX fuel assembly configuration is presented in this paper. The infinite multiplication factors and isotope inventory changes as a function of burnup obtained by Monte Carlo method module SCALE/KENO and deterministic method module SCALE/NEWT are compared with those obtained by deterministic code HELIOS. The impact in calculation results by using different nuclear data library is also investigated. The SCALE/KENO results show a good agreement with SCALE/NEWT results in the eigenvalue as a function of burnup (less than 0.1%). However, the absolute difference in the initial k¥ between SCALE/KENO and NEWT modules and HELIOS results is quite large (around 1.1%) and the isotope inventory changes show quite differently at the end of cycle. The uranium and plutonium depletion rates calculated by SCALE/KENO and SCALE/NEWT have quite good agreement. By using the same data library, the good agreement between stochastic and deterministic code’s results were confirmed.

scholarly journals Evolution of the importance of neutron-induced reactions along the cycle of an LFR

2020 ◽  
Vol 239 ◽  
pp. 22010
Author(s):  
Pablo Romojaro ◽  
Francisco Álvarez-Velarde
Keyword(s):  
Fast Reactor ◽  
Safety Assessment ◽  
Nuclear Energy ◽  
Nuclear Data ◽  
Point Of View ◽  
Sensitivity Analyses ◽  
Safety Level ◽  
Energy Needs ◽  
Nuclear Data Library ◽  
Data Library

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. Several LFR concepts are now in design phase, such as MYRRHA and ALFRED, and accurate nuclear data are required for the neutronic and safety assessment of the fast reactor designs. In this work, an assessment of the evolution of the importance of neutron-induced reactions along the cycle of a reference LFR design (i.e., ALFRED) with the state-of-the-art JEFF-3.3 nuclear data library is performed. Sensitivity analyses have been carried out with MCNP6 code in order to identify the most relevant isotopes and reactions from the neutronic point of view at BoL, BoC and EoC. Furthermore, an uncertainty quantification has been performed with the SUMMON system to study the evolution of uncertainties in the keff along the reactor cycle. The results from this work provide an exhaustive picture on the influence of nuclear data on core criticality performance, identifying key quantities and nuclear data needs relevant to achieve an improved safety level for LFR.

scholarly journals Benchmarking the new ENDF/B-VIII.0 nuclear data library for OECD/NEA medium 1000 MWth sodium-cooled fast reactor

2020 ◽  
Vol 239 ◽  
pp. 22006
Author(s):  
Donny Hartanto ◽  
Bassam Khuwaileh ◽  
Peng Hong Liem
Keyword(s):  
Monte Carlo ◽  
Fast Reactor ◽  
Nuclear Data ◽  
Control Rod ◽  
Continuous Energy ◽  
Nuclear Data Library ◽  
And Control ◽  
Benchmark Evaluation ◽  
Data Library

This paper presents the benchmark evaluation of the new ENDF/B-VIII.0 nuclear library for the OECD/NEA Medium 1000 MWth Sodium-cooled Fast Reactor (SFR). There are 2 SFR cores: metallic fueled (MET-1000) and oxide fueled (MOX-1000). The continuous-energy Monte Carlo Serpent2 code was used as the calculation tool. Various nuclear libraries such as ENDF/B-VII.1 and JENDL-4.0 were included to be compared with the newest ENDF/B-VIII.0. The evaluated parameters are k,βeff, sodium void reactivity (∆ρNa), Doppler constant (∆ρDoppler), and control rod worth (∆ρCR).

scholarly journals Nuclear Data Sensitivity Study for the EBR-II Fast Reactor Benchmark Using SCALE with ENDF/B-VII.1 and ENDF/B-VIII.0

2021 ◽  
Vol 2 (4) ◽  
pp. 345-367
Author(s):  
Friederike Bostelmann ◽  
Germina Ilas ◽  
William A. Wieselquist
Keyword(s):  
Fast Reactor ◽  
Nuclear Data ◽  
Sensitivity Study ◽  
Scale Model ◽  
Reactor Physics ◽  
Power Profile ◽  
Nuclear Data Library ◽  
Uncertainty Analyses ◽  
235U Fission ◽  
Quantities Of Interest

The EBR-II benchmark, which was recently included in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, served as a basis for assessing the performance of the SCALE code system for fast reactor analyses. A reference SCALE model was developed based on the benchmark specifications. Great agreement was observed between the eigenvalue calculated with this SCALE model and the benchmark eigenvalue. To identify potential gaps and uncertainties of nuclear data for the simulation of various quantities of interest in fast spectrum systems, sensitivity and uncertainty analyses were performed for the eigenvalue, reactivity effects, and the radial power profile of EBR-II using the two most recent ENDF/B nuclear data library releases. While the nominal results are consistent between the calculations with the different libraries, the uncertainties due to nuclear data vary significantly. The major driver of observed uncertainties is the uncertainty of the 235U (n,γ) reaction. Since the uncertainty of this reaction is significantly reduced in the ENDF/B-VIII.0 library compared to ENDF/B-VII.1, the obtained output uncertainties tend to be smaller in ENDF/B-VIII.0 calculations, although the decrease is partially compensated by increased uncertainties in 235U fission and ν¯.

scholarly journals BEAVRS BENCHMARK EVALUATION WITH CASMO5 AND SIMULATE5

2021 ◽  
Vol 247 ◽  
pp. 06022 ◽  
Author(s):  
Tamer Bahadir
Keyword(s):  
Reaction Rate ◽  
Boron Concentration ◽  
Neutron Transport ◽  
Nuclear Data ◽  
Control Rod ◽  
Significant Difference ◽  
Nuclear Data Library ◽  
Full Power ◽  
The Impact ◽  
Benchmark Evaluation

The MIT BEAVRS benchmark problem, which was primarily setup for the verification and validation of high-fidelity tools that have coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion models, has also found extensive usage in the reactor physics community for validating core analysis tools. The primary purpose of this paper is to provide an accurate, comprehensive evaluation of the BEAVRS benchmark with CASMO5 and SIMULATE5 codes. The CMS5 calculated results for low-power physics tests (hot zero power critical boron, control rod worth and isothermal temperature coefficients) and full power operation (boron let-down and flux map reaction rate distributions) are compared to plant measured data provided in the benchmark specification. The CMS5 model, using ENDF/BVII.1 nuclear data library, predicts HZP critical boron concentration for all-rods-out conditions within 10 ppm for Cycle-1, and 25 ppm in Cycle-2; the control rod worth is predicted with a difference of 0.7% ± 3.8%, where the maximum difference is less than 10%. For the core follow calculations at the hot full power condition, the average difference in predicting the critical boron concentration is less than 20 ppm. In addition, the radial and nodal reaction rate distributions are predicted with a mean difference of about 1.6% and 3.8%, respectively. The CMS5 calculations are repeated using the most recent ENDF/B-VIII.0 library. No significant difference is observed in predicting measured plant parameters with different nuclear data libraries. Additionally, the impact of various modeling options, which are typically employed with nodal diffusion codes, on the predictions of important core parameters are presented as part of the benchmark evaluation.

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