scholarly journals Impact of Different Nuclear Data Library on Control Rod Reactivity Worth Calculation of Small Pebble Bed Reactor

2021 ◽  
Vol 2048 (1) ◽  
pp. 012029
Author(s):  
Suwoto ◽  
H Adrial ◽  
T Setiadipura ◽  
Zuhair ◽  
S Bakhri
Keyword(s):  
Nuclear Data ◽  
Pebble Bed ◽  
Integral Control ◽  
Pebble Bed Reactor ◽  
Control Rod ◽  
Nuclear Data Library ◽  
Data Libraries ◽  
And Control ◽  
Nuclear Data Libraries ◽  
Data Library

Abstract One of the main critical issues on a nuclear reactor is safety and control system. The control rod worth plays an important role in the safety and control of nuclear reactors. The control rods worth calculation is used to specify the safety margin of the reactor. The main objective of this work is to investigate impact of different nuclear data libraries on calculating the control rod reactivity worth on small pebble bed reactor. Calculation of the control rod reactivity worth in small high temperature gas cooled reactor has been conducted using the Monte Carlo N-Particle 6 (MCNP6) code coupled with a different nuclear data library. Famous evaluated nuclear data libraries such as JENDL-40u, ENDF/B-VII.1 and JEFF-3.2 continuous cross section-energy data libraries were used. The overall calculation results of integral control rod worth show that the ENDF/B-VII.1, JENDL-40u and JEFF-3.2 files give values of - 17.814%☐k/k, -18.0204 %☐k/k and -18.0267%☐k/k, respectively. Calculations using ENDF/B-VII.1 give a slightly lower value than the others, while the JENDL-4.0u file gives results that are close to JEFF-3.2 file. The different nuclear data libraries have a relatively small impact on the control rod worth of small pebble bed reactor. Accurate prediction by simulation of control rod worth is very important for the safety operation of all reactor types, especially for new reactor designs.

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 Benchmarking the new ENDF/B-VIII.0 nuclear data library for the first core of Indonesian multipurpose research reactor (RSG GAS)

2020 ◽  
Vol 239 ◽  
pp. 22007 ◽  
Author(s):  
Donny Hartanto ◽  
Victor Gillette ◽  
Tagor Malem Sembiring ◽  
Peng Hong Liem
Keyword(s):  
Research Reactor ◽  
Nuclear Data ◽  
Monte Carlo Code ◽  
Kinetics Parameters ◽  
Uranium Fuel ◽  
Nuclear Data Library ◽  
Enriched Uranium ◽  
Data Libraries ◽  
Nuclear Data Libraries ◽  
Data Library

The Indonesian Multipurpose Research Reactor namely Reaktor Serba Guna G.A. Siwabessy (RSG GAS) is a 30 MWth (max.) pool-type reactor loaded with plate-type low-enriched uranium fuel, using light water as coolant and moderator, and beryllium as reflector. The benchmark of the 1st criticality core of RSG GAS using different nuclear data libraries such as JENDL-4.0, JENDL-3.3, ENDF/B-VII.0, and JEFF-3.1 have been performed in the previous work and compared with the experiment result. In this work, the newly released ENDF/B-VIII.0 neutron reaction and thermal neutron scattering libraries will be used and the important neu-tronics parameters such as multiplication factor, kinetics parameters, and fission reaction rate will be calculated using Monte Carlo code MCNP6.2 and compared against the previous work and the experiment result.

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 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 Calculation and experimental analysis of benchmark experiments with a fast neutron spectrum and models of sodium and lead cooled fast reactors using different evaluated nuclear data libraries

2021 ◽  
Vol 7 (2) ◽  
pp. 103-109
Author(s):  
Olga N. Andrianova ◽  
Yury Ye. Golovko ◽  
Gleb B. Lomakov ◽  
Yevgeniya S. Teplukhina ◽  
Gennady M. Zherdev
Keyword(s):  
Nuclear Data ◽  
Monte Carlo Code ◽  
Fast Reactors ◽  
Neutron Data ◽  
Critical Configurations ◽  
Nuclear Data Library ◽  
Data Libraries ◽  
Nuclear Data Libraries ◽  
Data Library ◽  
Evaluated Nuclear Data

The paper presents the results of a comparative analysis of criticality calculations using a Monte-Carlo code with the BNAB-93 and BNAB-RF neutron group constants, as well as with evaluated neutron data files from the Russian ROSFOND evaluated nuclear data library and other evaluated nuclear data libraries (ENDF, JEFF, JENDL) from different years. A set of integral experiments on BFS critical assemblies carried out in different years at the Institute of Physics and Power Engineering (60 different critical configurations) was analyzed. The considered integral experiments are included in the database of evaluated experimental neutronic data used to justify the neutronic performance of sodium and lead cooled fast reactors, to verify codes and nuclear data as well as to estimate uncertainties in neutronic parameters due to the nuclear data uncertainties. It has been shown that the ROSFOND evaluated nuclear data library is a library that minimizes the calculation and experimental discrepancies for the considered set of integral experiments. The paper also presents the results of criticality calculations for models of sodium and lead cooled fast reactors based on different evaluated neutron data libraries and provides estimates for the uncertainty in criticality associated with nuclear data.

scholarly journals Sensitivity and Uncertainty of Criticality

2021 ◽  
pp. 215-243
Author(s):  
Masao Yamanaka
Keyword(s):  
Reactor Core ◽  
Nuclear Data ◽  
Japan Atomic Energy Agency ◽  
Reactor Physics ◽  
Energy Agency ◽  
Control Rod ◽  
The Core ◽  
Data Libraries ◽  
And Control ◽  
Nuclear Data Libraries

AbstractExcess reactivity and control rod worth are generally considered important reactor physics parameters for experimentally examining the neutron characteristics of criticality in a core, and for maintaining safe operation of the reactor core in terms of neutron multiplication in the core. For excess reactivity and control rod worth at KUCA, as well as at the Fast Critical Assembly in the Japan Atomic Energy Agency, special attention is given to analyzing the uncertainty induced by nuclear data libraries based on experimental data of criticality in representative cores (EE1 and E3 cores). Also, the effect of decreasing uncertainty on the accuracy of criticality is discussed in this study. At KUCA, experimental results are accumulated by measurements of excess reactivity and control rod worth. To evaluate the accuracy of experiments for benchmarks, the uncertainty originated from modeling of the core configuration should be discussed in addition to uncertainty induced by nuclear data, since the uncertainty from modeling has a potential to cover the eigenvalue bias more than uncertainty by nuclear data. Here, to investigate the uncertainty of criticality depending on the neutron spectrum of cores, it is very useful to analyze the reactivity of a large number of measurements in typical hard (EE1) and soft (E3) spectrum cores at KUCA.

Neutronic Analyses of the FREYA Experiments in Support of the ALFRED Lead-Cooled Fast Reactor Core Design and Licensing

2019 ◽  
Vol 6 (1) ◽  
Author(s):  
Massimo Sarotto ◽  
Gabriele Firpo ◽  
Anatoly Kochetkov ◽  
Antonin Krása ◽  
Emil Fridman ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Nuclear Data ◽  
Control Rod ◽  
The Core ◽  
Major Interest ◽  
Reactor Spectrum ◽  
Data Libraries ◽  
Nuclear Data Libraries ◽  
Sensitivity And Uncertainty Analysis

Abstract During the EURATOM FP7 project FREYA, a number of experiments were performed in a critical core assembled in the VENUS-F zero-power reactor able to reproduce the ALFRED lead-cooled fast reactor spectrum in a dedicated island. The experiments dealt with the measurements of integral and local neutronic parameters, such as the core criticality, the control rod and the lead void reactivity worth, the axial distributions of fission rates for the nuclides of major interest in a fast spectrum, the spectral indices of important actinides (238U, 239Pu, 237 Np) with respect to 235U. With the main aim to validate the neutronic codes adopted for the ALFRED core design, the VENUS-F core and its characterization measurements were simulated with both deterministic (ERANOS) and stochastic (MCNP, SERPENT) codes, by adopting different nuclear data libraries (JEFF, ENDF/B, JENDL, TENDL). This paper summarizes the main results obtained by highlighting a general agreement between measurements and simulations, with few discrepancies for some parameters that are discussed here. Additionally, a sensitivity and uncertainty analysis was performed with deterministic methods for the core reactivity: it clearly indicates that the small over-criticality estimated by the different codes/libraries resulted to be lower than the uncertainties due to nuclear data.

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.

Nuclear Data Library Effects on Fast to Thermal Flux Shapes Around PWR Control Rod Tips

2014 ◽  
Vol 118 ◽  
pp. 575-578 ◽  
Author(s):  
A. Vasiliev ◽  
H. Ferroukhi ◽  
T. Zhu ◽  
A. Pautz
Keyword(s):  
Thermal Flux ◽  
Nuclear Data ◽  
Control Rod ◽  
Nuclear Data Library ◽  
Data Library

scholarly journals THE SOURCE JERK INTEGRALMETHOD FOR SUB-CRITICALITY MEASUREMENTS IN ADS

2021 ◽  
Vol 247 ◽  
pp. 08002
Author(s):  
A. Kochetkov ◽  
A. Billebaud ◽  
S. Chabod ◽  
A. Krása ◽  
F.-R. Lecolley ◽  
...  
Keyword(s):  
Dead Time ◽  
Power Reactor ◽  
Nuclear Data ◽  
Start Up ◽  
Nuclear Data Library ◽  
And Control ◽  
Control Rods ◽  
Fission Chambers ◽  
Data Library ◽  
Evaluated Nuclear Data

Three sub-critical (SC) core configurations were investigated in the VENUS-F zero power reactor coupled with the GENEPI-3C accelerator. The SC10 and SC12 were a mock-up of a MYRRHA start-up core and SC11 represented a more complex MYRRHA core loaded with various types of in-pile-sections. The sub-criticality of 11 variants of these VENUS-F cores was changed in several steps from -6$ down to -30$ using the safety and control rods. Their sub-criticalities were determined with the Source Jerk Integral (SJI) method using 11 fission chambers located all over the reactor. For the data analysis, the 8-group delayed neutron parameters from the JEFF-3.1.2 evaluated nuclear data library were used. Reliability and reproducibility of the experimental results were tested by repeating the measurements, swapping the detectors and varying the accelerator beam intensity, thus changing the detector count rates and verifying the validity of the dead time corrections. The obtained results are compared with MCNP calculations.

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