scholarly journals Feasibility study of decay heat uncertainty reduction using nuclear data adjustment method with experimental data

2016 ◽  
Vol 54 (2) ◽  
pp. 213-222 ◽  
Author(s):  
Yosuke Kawamoto ◽  
Go Chiba
Keyword(s):  
Experimental Data ◽  
Feasibility Study ◽  
Nuclear Data ◽  
Uncertainty Reduction ◽  
Adjustment Method ◽  
Decay Heat ◽  
Data Adjustment

scholarly journals Nuclear data adjustment based on the interpretation of post-irradiation experiments with the DARWIN2.3 package

2018 ◽  
Vol 4 ◽  
pp. 47 ◽  
Author(s):  
Axel Rizzo ◽  
Claire Vaglio-Gaudard ◽  
Gilles Noguere ◽  
Julie-Fiona Martin ◽  
Vanessa Vallet ◽  
...  
Keyword(s):  
Concentration Ratio ◽  
Fuel Cycle ◽  
Nuclear Data ◽  
Fission Yield ◽  
Post Irradiation ◽  
Decay Heat ◽  
Fuel Pellets ◽  
Irradiated Fuel ◽  
Data Adjustment

DARWIN2.3 is the French reference package dedicated to fuel cycle applications, computing fuel inventory as well as decay heat, neutron emissions, α, β and γ spectra. The DARWIN2.3 package fuel inventory calculation was experimentally validated with Post-Irradiation Experiments (PIEs), mainly consisting in irradiated fuel pellets analysis. This paper presents a method to assimilate these integral trends for improving nuclear data. In this study, the method is applied to 137Cs/238U concentration ratio. Results suggest an increase of the JEFF-3.1.1 235U cumulated thermal fission yield in 137Cs by (+3.8 ± 2.1)%, from 6.221E-02 to 6.460E-02 ± 2.1%.

Validation of CATHARE V2.5 Thermal-Hydraulic Code Against Full-Scale PERSEO Tests for Decay Heat Removal in LWRs

2010 ◽  
Author(s):  
Giacomino Bandini ◽  
Paride Meloni ◽  
Massimiliano Polidori ◽  
Calogera Lombardo
Keyword(s):  
Experimental Data ◽  
Heat Exchangers ◽  
Heat Removal ◽  
Full Scale ◽  
Experimental Program ◽  
Decay Heat ◽  
The Past ◽  
Reserve Pool ◽  
Large Water ◽  
Safety Systems

The PERSEO experimental program was performed in the framework of a domestic research program on innovative safety systems with the purpose to increase the reliability of passive decay heat removal systems implementing in-pool heat exchangers. The conceived system was tested at SIET laboratories by modifying the existing PANTHERS IC-PCC facility utilized in the past for testing a full scale module of the GE-SBWR in-pool heat exchanger. Integral tests and stability tests were conducted to verify the operating principles, the steadiness and the effectiveness of the system. Two of the more representative tests have been analyzed with CATHARE V2.5 for code validation purposes. The paper deals with the comparison of code results against experimental data. The capabilities and the limits of the code in simulating such kind of tests are highlighted. An improvement in the modeling of the large water reserve pool is suggested trying to reduce the discrepancies observed between code results and test measurements.

Decay Heat Characterization for the European Sodium Fast Reactor

2021 ◽  
Author(s):  
Antonio Jiménez-Carrascosa ◽  
Nuria Garcia Herranz ◽  
Jiri Krepel ◽  
Marat Margulis ◽  
Una Baker ◽  
...  
Keyword(s):  
Fast Reactor ◽  
State Of The Art ◽  
Nuclear Data ◽  
Summation Method ◽  
Heat Power ◽  
Coupled Transport ◽  
Decay Heat ◽  
Sodium Fast Reactor ◽  
Detailed Assessment ◽  
Commercial Size

Abstract In this work a detailed assessment of the decay heat power for the commercial-size European Sodium-cooled Fast Reactor (ESFR) at the end of its equilibrium cycle has been performed. The summation method has been used to compute very accurate spatial- and time-dependent decay heat by employing state-of-the-art coupled transport-depletion computational codes and nuclear data. This detailed map provides basic information for subsequent transient calculations of the ESFR. A comprehensive analysis of the decay heat has been carried out and interdependencies among decay heat and different parameters characterizing the core state prior to shutdown, such as discharge burnup or type of fuel material, have been identified. That analysis has served as a basis to develop analytic functions to reconstruct the spatial-dependent decay heat power for the ESFR for cooling times within the first day after shutdown.

S-CO2 Turbine Design for Decay Heat Removal System of Sodium Cooled Fast Reactor

2016 ◽  
Author(s):  
Seong Kuk Cho ◽  
Jekyoung Lee ◽  
Jeong Ik Lee ◽  
Jae Eun Cha
Keyword(s):  
Experimental Data ◽  
Fast Reactor ◽  
Design Methodology ◽  
Nuclear Reactors ◽  
Heat Removal ◽  
Design Code ◽  
Decay Heat ◽  
Turbine Design ◽  
Heat Removal System ◽  
Removal System

A Sodium-cooled Fast Reactor (SFR) has receiving attention as one of the promising next generation nuclear reactors because it can recycle the spent nuclear fuel produced from the current commercial nuclear reactors and accomplish higher thermal efficiency than the current commercial nuclear reactors. However, after shutdown of the nuclear reactor core, the accumulated fission products of the SFR also decay and release heat via radiation within the reactor. To remove this residual heat, a decay heat removal system (DHRS) with supercritical CO2 (S-CO2) as the working fluid is suggested with a turbocharger system which achieves passive operational capability. However, for designing this system an improved S-CO2 turbine design methodology should be suggested because the existing methodology for designing the S-CO2 Brayton cycle has focused only on the compressor design near the critical point. To develop a S-CO2 turbine design methodology, the non-dimensional number based design and the 1D mean line design method were modified and suggested. The design methodology was implemented into the developed code and the code results were compared with existing turbine experimental data. The data were collected under air and S-CO2 environment. The developed code in this research showed a reasonable agreement with the experimental data. Finally using the design code, the turbocharger design for the suggested DHRS and prediction of the off design performance were carried out. As further works, more effort will be put it to expand the S-CO2 turbine test data for validating the design code and methodology.

scholarly journals Observations of poorly-known features of the 239Pu and 235U prompt fission neutron spectra

2020 ◽  
Vol 239 ◽  
pp. 05010
Author(s):  
Keegan J. Kelly ◽  
Jaime A. Gomez ◽  
John M. O'Donnell ◽  
Matthew Devlin ◽  
Robert C. Haight ◽  
...  
Keyword(s):  
Experimental Data ◽  
Neutron Spectrum ◽  
Neutron Emission ◽  
National Laboratory ◽  
Nuclear Data ◽  
Fission Neutron ◽  
Fission Neutron Spectrum ◽  
Wide Range ◽  
Prompt Fission ◽  
Prompt Fission Neutron

Prompt fission neutron spectrum (PFNS) evaluations use provide nuclear data for the PFNS across a wide range of incident and outgoing neutron energies. However, experimental data underlying the evaluation are sparse, inconsistent, and incomplete with respect to the desired incident and outgoing energy coverage. As such, evaluations sometimes predict features of the PFNS, such those relating to multi-chance fission and pre-equilibrium pre-fission neutron emission, without any experimental validation. The Chi-Nu experiment at Los Alamos National Laboratory has recently obtained high-precision results for the 239Pu and 235U PFNS which, for the first time in both cases, have shed light on multi-chance fission and pre-equilibrium contributions to the observed fission neutron spectrum. In addition to providing the first experimental data on some of these fission properties, the angular coverage of the Chi-Nu experiment allows for the extraction of angular distributions of pre-equilibrium pre-fission neutrons. PFNS results of multi-chance fission and pre-equilibrium pre-fission neutron emission are discussed in this proceedings in terms of the observed neutron spectrum and the average PFNS energies.

scholarly journals Impact of delayed neutron constants on reactivity effects measured in a fast reactor

2019 ◽  
Vol 211 ◽  
pp. 03004
Author(s):  
Antonín Krása ◽  
Anatoly Kochetkov ◽  
Nadia Messaoudi ◽  
Alexey Stankovskiy ◽  
Guido Vittiglio ◽  
...  
Keyword(s):  
Experimental Data ◽  
Reactor Core ◽  
Nuclear Data ◽  
Delayed Neutron ◽  
Neutron Data ◽  
Monte Carlo Calculations ◽  
Data Libraries ◽  
Experimental Data Analysis ◽  
The Impact ◽  
Nuclear Data Libraries

Delayed neutron parameters of fast VENUS-F reactor core configurations are determined with Monte Carlo calculations using various nuclear data libraries. Differences in the calculated effective delayed neutron fraction and the impact of the delayed neutron data (6- or 8-group precursors) that are applied in the experimental data analysis on the measured reactivity effects are studied. Considerable differences are found due to application of 235U and 238U delayed neutron data from JEFF, JENDL and ENDF evaluations.

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