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.

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 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 Monte Carlo with fuel burnup method for the ENHS benchmark calculations

2003 ◽  
Vol 18 (2) ◽  
pp. 3-11
Author(s):  
Miodrag Milosevic ◽  
Ehud Greenspan ◽  
Jasmina Vujic
Keyword(s):  
Neutron Transport ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Nuclear Data ◽  
Zone Model ◽  
Geometry Model ◽  
Nuclear Heat ◽  
Nuclear Data Library ◽  
Full Power ◽  
Data Libraries

Estimates of the uncertainties arising from approximations in the methods used in different nuclear data processing and neutron transport codes are usually obtained by inter-comparing calculations made using different code systems. This paper gives details of an investigation of differences between results obtained by using different codes for a single zone model of the Encapsulated Nuclear Heat Source (ENHS) benchmark core fuelled with metallic alloy of Pu, U, and Zr. The ENHS is a new lead-bismuth or lead cooled novel reactor concept for 20 effective full power years without refuelling and with very small reactivity swing. The computational tools benchmarked include MOCUP, a coupled MCNP-4C and ORIGEN2.1 utility codes with MCNP data libraries based on ENDF/B-VI evaluation; KENO-V.a/ORIGEN2.1 code system, recently developed by authors of this paper, with the ENDFB-V based 238 group library; the design-oriented procedure based on the simplified one-dimensional (1D) geometry model and SAS2H control module; and the well-established fast reactor neutronics design tools in use at Argonne National Laboratory. Calculations made for the ENHS benchmark have shown that the differences between the results obtained when using different code schemes are quite significant and should be taken into account in assessing the quality of the nuclear data library.

scholarly journals ANALYSIS OF THE BEAVRS BENCHMARK USING CASMO5/SIMULATE5 WITH JENDL-4.0 AND ENDF/B-VII.1

2021 ◽  
Vol 247 ◽  
pp. 02014
Author(s):  
Fujita Tatsuya ◽  
Sakai Tomohiro
Keyword(s):  
Measurement Data ◽  
Nuclear Data ◽  
Isothermal Temperature ◽  
Control Rod ◽  
Significant Difference ◽  
Nuclear Data Library ◽  
Calculation Results ◽  
The Difference ◽  
Data Library ◽  
Good Agreement

The BEAVRS benchmark was analyzed using the CASMO5/SIMULATE5 in order to compare the measurement data and the calculation results based on the JENDL-4.0 and ENDF/B-VII.1 and investigate the difference between those calculation results. For the hot zero power (HZP) physics test, the calculation results showed good agreement with the measurement data for both of cycles 1 and 2. For cycle 1, the calculation results of the isothermal temperature coefficient (ITC) differed from the measurement data by approximately 1 pcm/℉, and the same tendency has been reported in previous studies. For the cycle operation, the calculation results of the boron letdown agreed well with the measurement data. On the other hand, some calculation results of the axial detector signals had a large difference from the measurement data, which is supposedly attributed to the discrepancy of the axial offset (AO) caused by the authors’ approximation for the control rod and shutdown bank positions. In terms of the comparison between the JENDL-4.0 and ENDF/B-VII.1, although approximately 15 ppm difference of the boron letdown in the cycle operation was observed, no significant difference was seen for other core parameters, thus, the influence of the two nuclear data library was small on the present results.

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 Analysis of the BEAVRS PWR benchmark using SCALE and PARCS

Nukleonika ◽  
2019 ◽  
Vol 64 (3) ◽  
pp. 87-96 ◽  
Author(s):  
Piotr Darnowski ◽  
Michał Pawluczyk
Keyword(s):  
Fuel Cycle ◽  
Neutron Transport ◽  
Measurement Data ◽  
Nuclear Data ◽  
Data File ◽  
Power Measurement ◽  
Future Research ◽  
Pressurized Water ◽  
Control Rod ◽  
Nodal Diffusion

Abstract This paper presents an analysis of the Benchmark for Evaluation And Validation of Reactor Simulations (BEAVRS) performed using SCALE 6.1.2 and PARCS 3.2 computer codes. The benchmark specification contains a detailed design, operational data and measurements for a real 4-loop Westinghouse pressurized water reactor (PWR). The lattice physics simulations were prepared using TRITON depletion sequence and NEWT neutron transport solver (SCALE package). The 238-neutron group library based on evaluated nuclear data file – ENDF/B-VII nuclear data libraries was applied. A set of branch and burnup calculations was prepared, and group constants in the form of PMAXS files were generated with GenPMAXS. The full-core models were prepared using the PARCS nodal-diffusion core simulator. The PMAXS libraries were used with PARCS to investigate the core operation. The hot zero power measurement data, including control rod worths and critical boron concentrations, were compared using simulations, and satisfactory results were achieved. The first fuel cycle was simulated, and acceptable agreement with boron letdown curve and measurements were obtained. Finally, conclusions and recommendations for future research were presented.

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 Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL

2015 ◽  
Vol 2015 ◽  
pp. 1-9 ◽  
Author(s):  
A. Rais ◽  
D. Siefman ◽  
G. Girardin ◽  
M. Hursin ◽  
A. Pautz
Keyword(s):  
Power Distribution ◽  
Reactor Core ◽  
Transient Behavior ◽  
Nuclear Data ◽  
Safety Analysis ◽  
Monte Carlo Code ◽  
Thermal Hydraulics ◽  
Control Rod ◽  
Nuclear Data Library ◽  
And Control

In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor.

scholarly journals Reaction Rate Benchmark Experiments with Miniature Fission Chambers at the Slovenian TRIGA Mark II Reactor

2018 ◽  
Vol 170 ◽  
pp. 04023 ◽  
Author(s):  
Žiga Štancar ◽  
Tanja Kaiba ◽  
Luka Snoj ◽  
Loïc Barbot ◽  
Christophe Destouches ◽  
...  
Keyword(s):  
Reaction Rate ◽  
Neutron Transport ◽  
Experimental Parameter ◽  
Sensitivity Analyses ◽  
Fission Rate ◽  
Control Rod ◽  
Local Maxima ◽  
Detector Position ◽  
Rate Profile ◽  
Fission Chambers

A series of fission rate profile measurements with miniature fission chambers, developed by the Commisariat á l’énergie atomique et auxénergies alternatives, were performed at the Jožef Stefan Institute’s TRIGA research reactor. Two types of fission chambers with different fissionable coating (235U and 238U) were used to perform axial fission rate profile measurements at various radial positions and several control rod configurations. The experimental campaign was supported by an extensive set of computations, based on a validated Monte Carlo computational model of the TRIGA reactor. The computing effort included neutron transport calculations to support the planning and design of the experiments as well as calculations to aid the evaluation of experimental and computational uncertainties and major biases. The evaluation of uncertainties was performed by employing various types of sensitivity analyses such as experimental parameter perturbation and core reaction rate gradient calculations. It has been found that the experimental uncertainty of the measurements is sufficiently low, i.e. the total relative fission rate uncertainty being approximately 5 %, in order for the experiments to serve as benchmark experiments for validation of fission rate profiles. The effect of the neutron flux redistribution due to the control rod movement was studied by performing measurements and calculations of fission rates and fission chamber responses in different axial and radial positions at different control rod configurations. It was confirmed that the control rod movement affects the position of the maximum in the axial fission rate distribution, as well as the height of the local maxima. The optimal detector position, in which the redistributions would have minimum effect on its signal, was determined.

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