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 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 Nuclear data sensitivity and uncertainty analyses on the first core criticality of the RSG GAS multipurpose research reactor

2020 ◽  
Vol 239 ◽  
pp. 22011 ◽  
Author(s):  
Peng Hong Liem ◽  
Zuhair ◽  
Donny Hartanto
Keyword(s):  
Nuclear Data ◽  
Worst Case ◽  
Control Rod ◽  
Nuclear Data Library ◽  
Uncertainty Analyses ◽  
Effective Neutron ◽  
Data Libraries ◽  
And Control ◽  
Analytical Tools ◽  
Nuclear Data Libraries

The results of criticality, sensitivity and uncertainty (S\U) analyses on the first core criticality of the Indonesian 30 MWth Multipurpose Reactor RSG GAS (MPR-30) using the recent nuclear data libraries (ENDF/B-VII.1 and JENDL-4.0) and analytical tools available at present (WHISPER-1.1) are presented. Two groups of criticality benchmark cases were carefully selected from the experiments conducted during the first criticality approach and control rod calibrations. The C/E values of effective neutron multiplication factor (k) for the worst case was found around 1.005. Large negative sensitivities were found in (n,e-mail:γ) reaction of H-1, U-235, Al-27, U-238 and Be-9 while large positive sensitivities were found in U-235 (total nu and fission), H-1 (elastic), Be-9 (free gas, elastic) and H-1 S(α,β) (lwtr.20t, inelastic). The S\U analysis results concluded that the uncertainties of k originated from the nuclear data were found around 0.6% which covered well the [C/E-1] values. Differences in the sensitivities amongst the two nuclear data libraries were also identified, and recommendation for improving the nuclear data library was given.

scholarly journals FAST REACTOR MULTIPHYSICS AND UNCERTAINTY PROPAGATION WITHIN WIMS

2021 ◽  
Vol 247 ◽  
pp. 06002
Author(s):  
Ben Lindley ◽  
Brendan Tollit ◽  
Peter Smith ◽  
Alan Charles ◽  
Robert Mason ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Uncertainty Propagation ◽  
Predictive Modelling ◽  
Nuclear Data ◽  
Reactor Physics ◽  
Fuel Performance ◽  
Horizon 2020 ◽  
Calculation Methodology ◽  
International Projects ◽  
Fast Reactor Fuel

For liquid metal-cooled fast reactors (LMFRs), improved predictive modelling is desirable to facilitate reactor licensing and operation and move towards a best estimate plus uncertainty (BEPU) approach. A key source of uncertainty in fast reactor calculations arises from the underlying nuclear data. Addressing the propagation of such uncertainties through multiphysics calculations schemes is therefore of importance, and is being addressed through international projects such as the Sodium-cooled Fast Reactor Uncertainty Analysis in Modelling (SFR-UAM) benchmark. In this paper, a methodology for propagation of nuclear data uncertainties within WIMS is presented. Uncertainties on key reactor physics parameters are calculated for selected SFR-UAM benchmark exercises, with good agreement with previous results. A methodology for coupled neutronic-thermal-hydraulic calculations within WIMS is developed, where thermal feedback is introduced to the neutronic solution through coupling with the ARTHUR subchannel code within WIMS and applied to steady-state analysis of the Horizon 2020 ESFR-SMART project reference core. Finally, integration of reactor physics and fuel performance calculations is demonstrated through linking of the WIMS reactor physics code to the TRAFIC fast reactor fuel performance code, through a Fortran-C-Python (FCP) interface. Given the 3D multiphysics calculation methodology, thermal-hydraulic and fuel performance uncertainties can ultimately be sampled alongside the nuclear data uncertainties. Together, these developments are therefore an important step towards enabling propagation of uncertainties through high fidelity, multiphysics SFR calculations and hence facilitate BEPU methodologies.

Assessment of WIMS-D5 Applicability to CANDU NPPs

2005 ◽  
Vol 277-279 ◽  
pp. 741-746
Author(s):  
W.Y. Kim ◽  
B.J. Min
Keyword(s):  
Large Deviation ◽  
Nuclear Data ◽  
Lattice Calculation ◽  
Reactor Physics ◽  
Candu Reactors ◽  
Energy Group ◽  
Code Comparison ◽  
Fuel Burn ◽  
Nuclear Data Library ◽  
Data Library

The purpose of this study is to develop a WIMS/CANDU code for a lattice calculation on the basis of WIMS-D5 code for the safety analysis of CANDU reactors. To assess WIMS-D5 applicability to a CANDU reactor, a lattice model was developed for CANDU-6 reactors at the Wolsong site. As for the benchmark of the code validation, a code-to-code comparison was performed between the WIMS-D5 code with both the 69- and 172-energy groups of ENDF/B-VI nuclear data library and the WIMS-AECL code with the 89-energy group. The comparison studies of the reactor physics parameters such as the void reactivity and the coolant/fuel/moderator temperature coefficients were conducted with a change of the internal isotopic composition due to the fuel burning-up using both WIMS-AECL and POWDERPUFS-V (PPV) codes. The results show that the present results from WIMS-D5 code and WIMS-AECL code agreed well with those of the PPV at the beginning of the fuel burn-up phase. As the burning-up progresses, the results of WIMS-D5 show a large deviation from those of PPV for the CANDU 6 reactors.

scholarly journals Validation of JEFF-3.3 and ENDF/B-VIII.0 nuclear data libraries in ANSWERS codes

2020 ◽  
Vol 239 ◽  
pp. 19001
Author(s):  
Tim Ware ◽  
David Hanlon ◽  
Glynn Hosking ◽  
Ray Perry ◽  
Simon Richards
Keyword(s):  
Cross Sections ◽  
Nuclear Data ◽  
Scattering Data ◽  
Reactor Physics ◽  
Group Data ◽  
Nuclear Data Library ◽  
Data Libraries ◽  
Neutron Cross Sections ◽  
Nuclear Data Libraries ◽  
Data Library

The JEFF-3.3 and ENDF/B-VIII.0 evaluated nuclear data libraries were released in December 2017 and February 2018 respectively. Both evaluations represent a comprehensive update to their predecessor evaluations. The ANSWERS Software Service produces the MONK® and MCBEND Monte Carlo codes, and the WIMS deterministic code for nuclear criticality, shielding and reactor physics applications. MONK and MCBEND can utilise continuous energy nuclear data provided by the BINGO nuclear data library and MONK and WIMS can utilise broad energy group data (172 group XMAS scheme) via the WIMS nuclear data library. To produce the BINGO library, the BINGO Pre-Processor code is used to process ENDF-6 format evaluations. This utilises the RECONR-BROADR-PURR sequence of NJOY2016 to reconstruct and Doppler broaden the free gas neutron cross sections together with bespoke routines to generate cumulative distributions for the S(α,β) tabulations and equi-probable bins or probability functions for the secondary angle and energy data. To produce the WIMS library, NJOY2016 is again used to reconstruct and Doppler broaden the cross sections. The THERMR module is used to process the thermal scattering data. Preparation of data for system-dependent resonance shielding of some nuclides is performed. GROUPR is then used to produce the group averaged data before all the data are transformed into the specific WIMS library format. The MONK validation includes analyses based on around 800 configurations for a range of fuel and moderator types. The WIMS validation includes analyses of zero-energy critical and sub-critical, commissioning, operational and post-irradiation experiments for a range of fuel and moderator types. This paper presents and discusses the results of MONK and WIMS validation benchmark calculations using the JEFF-3.3 and ENDF/B-VIII.0 based BINGO and WIMS nuclear data libraries.

scholarly journals Verification and application of SuperMC3.3 to lead-bismuth-cooled fast reactor

2019 ◽  
Vol 34 (2) ◽  
pp. 122-128
Author(s):  
Hui Ding ◽  
Quan Gan ◽  
Lijuan Hao ◽  
Jing Song ◽  
Yican Wu
Keyword(s):  
Fast Reactor ◽  
Large Scale ◽  
Bench Mark ◽  
Fast Reactors ◽  
Reactor Physics ◽  
Kinetics Parameters ◽  
Power Profile ◽  
Integrated Software ◽  
Relationship Of ◽  
The Relationship

Lead-cooled fast reactors have multilayered designs and large internal temperature differences, which cause challenges in simulating reactor physics. SuperMC, a large-scale integrated software system for neutronics design, is inherently able to address complex geometries and multi-temperature problems. The purpose of this study is to verify the applicability of SuperMC to the lead-bismuth-cooled fast reactor RBEC-M. The multi-temperature cross-section generation function of SuperMC was employed and showed good performance. Based on the ENDF/B-VII.1 library, the effective multiplication factor keff obtained by SuperMC showed good agreement with those from previous works. The relationship of keff and 15N enrichment applied to the fuel material was also studied, with the results showing that in creased 15N could significantly improve keff. The axial power profile and kinetics parameters for the bench mark were then calculated and analyzed. This work thus verified the applicability of SuperMC for comprehensive neutronics simulations for lead-bismuth-cooled fast reactors.

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 Neutronics performances of gas-cooled fast reactor for 300-600 MWt Output Power with Modified CANDLE burn-up scheme in radial direction

2021 ◽  
Vol 2072 (1) ◽  
pp. 012013
Author(s):  
F H Irka ◽  
Z Suud ◽  
D Irwanto ◽  
S N Khotimah ◽  
H Sekimoto
Keyword(s):  
Output Power ◽  
Fast Reactor ◽  
Energy Production ◽  
Radial Direction ◽  
Fuel Cycle ◽  
Nuclear Data ◽  
Closed Fuel Cycle ◽  
Nuclear Data Library ◽  
Reactor Type ◽  
Data Library

Abstract Gas-Cooled Fast Reactor-GFR is a Generation IV reactor that is helium-cooled and has a closed fuel cycle. Due to the target operation on 2022-2030, this reactor type still needs further research and development technologies. We investigated the neutronics performances of a GFR balance type core with some modification of CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy production) burn-up scheme in the radial direction. The output power varied from 300 to 600 MWt. The neutronics calculation was performed using SRAC 2002 with JENDL 4.0 nuclear data library. The analysis indicate the reactor could operate critically for ten years without refueling with burn-up level 20% HM.

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