New feature of DARWIN/PEPIN2 inventory code: Propagation of nuclear data uncertainties to decay heat and nuclide density

2021 ◽  
Vol 164 ◽  
pp. 108579
Author(s):  
A. Tsilanizara ◽  
T.D. Huynh
Keyword(s):  
Nuclear Data ◽  
Decay Heat ◽  
New Feature

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.

scholarly journals A COMPARISON OF OXIDE DECAY HEAT SIMULATIONS AND NUCLEAR DATA LIBRARIES WITH FUSION IRRADIATION EXPERIMENTS

2021 ◽  
Vol 247 ◽  
pp. 09011
Author(s):  
L W Packer ◽  
M Angelone ◽  
M R Gilbert ◽  
S Loreti ◽  
C R Nobs ◽  
...  
Keyword(s):  
Nuclear Data ◽  
Data Sets ◽  
Section Data ◽  
Fusion Neutron ◽  
Decay Heat ◽  
Decay Profile ◽  
Data Points ◽  
Fast Extraction ◽  
Data Libraries ◽  
Nuclear Data Libraries

We compare existing experimental decay heat data sets measured at the JAEA fusion neutron source (FNS), which employed a fast extraction rabbit system that in certain cases allowed the measurements to capture, at short cooling times, the decay profile of 16N in a range of oxides. Focussing on those data points and timescales that can be attributed to 16N, we compare measurements to simulations performed using the FISPACT-II inventory code together with evaluated nuclear data libraries. Making small corrections for other contributions at these short times, we derive integral cross section data estimates for 16O(n,p)16N from 12 oxide sample measurements and compare with previously obtained measurements in the IAEA EXFOR database and evaluations in the nuclear data libraries themselves.

scholarly journals Uncertainty quantification works relevant to fission yields and decay data

2018 ◽  
Vol 4 ◽  
pp. 43
Author(s):  
Go Chiba ◽  
Shunsuke Nihira
Keyword(s):  
Perturbation Theory ◽  
Uncertainty Quantification ◽  
Nuclear Data ◽  
Code System ◽  
Reactor Physics ◽  
Delayed Neutrons ◽  
Decay Data ◽  
Decay Heat ◽  
Numerical Tools ◽  
Fission Yields

In the present paper, firstly, we review our previous works on uncertainty quantification (UQ) of reactor physics parameters. This consists of (1) development of numerical tools based on the depletion perturbation theory (DPT), (2) linearity of reactor physics parameters to nuclear data, (3) UQ of decay heat and its reduction, and (4) correlation between decay heat and β-delayed neutrons emission. Secondly, we show results of extensive calculations about UQ on decay heat with several different numerical conditions by the DPT-based capability of a reactor physics code system CBZ.

scholarly journals Integral data assimilation of the MERCI-1 experiment for the nuclear data associated with the PWR decay heat computation

2019 ◽  
Vol 211 ◽  
pp. 07004
Author(s):  
J. Huyghe ◽  
C. De Saint-Jean ◽  
D. Lecarpentier ◽  
C. Reynard-Carette ◽  
C. Vaglio-Gaudard ◽  
...  
Keyword(s):  
Reactor Core ◽  
Nuclear Data ◽  
Nuclear Decay ◽  
Crucial Issue ◽  
Decay Heat ◽  
Uncertainty Calculation ◽  
Experimental Values ◽  
The Impact ◽  
Fission Yields ◽  
Integral Data

Nuclear decay heat is a crucial issue for PWR in-core safety after reactor shutdown and back-end cycle. It is a dimensioning parameter for safety injection systems (SIS) to avoid a dewatering of the reactor core. The decay heat uncertainty needs to be controlled over the largest range of applications. The assimilation of the MERCI-1 experiment was studied to provide feedbacks on nuclear data. This experiment consisted in the measurement of the decay heat of a PWR UOX fuel sample irradiated in the OSIRIS reactor, for cooling times between 45 minutes and 42 days. More specifically, the consideration of several experimental values of MERCI-1 at different cooling times was tested. This raised issues about correlations to consider between different measurements. Besides, the impact of considering correlations between independent fission yields in covariance matrices on the decay heat uncertainty calculation and on the feedbacks on nuclear data is discussed.

scholarly journals Summation Calculations for Reactor Antineutrino Spectra, Decay Heat and Delayed Neutron Fractions Involving New TAGS Data and Evaluated Databases

2019 ◽  
Vol 211 ◽  
pp. 01001
Author(s):  
M. Estienne ◽  
M. Fallot ◽  
L. Giot ◽  
V. Guadilla-Gomez ◽  
L. Le Meur ◽  
...  
Keyword(s):  
Fission Product ◽  
Neutron Emission ◽  
Fission Products ◽  
Nuclear Data ◽  
Delayed Neutron ◽  
Data Bases ◽  
Decay Data ◽  
Decay Heat ◽  
Reactor Antineutrino ◽  
Decay Properties

Three observables of interest for present and future reactors depend on the β decay properties of the fission products: antineutrinos from reactors, the reactor decay heat and delayed neutron emission. In these proceedings, we present new results from summation calculations of the first two quantities quoted above, performed with evolved independent yields coupled with fission product decay data, from various nuclear data bases or models. New TAGS results from the latest experiment of the TAGS collaboration at the JYFL facility of Jyväskylä will be displayed as well as their impact on the antineutrino spectra and the decay heat associated to fission pulses of the main actinides.

scholarly journals Testing JEFF-3.1.1 and ENDF/B-VII.1 Decay and Fission Yield Nuclear Data Libraries with Fission Pulse Neutron Emission and Decay Heat Experiments

2014 ◽  
Vol 118 ◽  
pp. 472-475
Author(s):  
O. Cabellos ◽  
V. de Fusco ◽  
C.J. Diez de la Obra ◽  
J.S. Martinez ◽  
E. Gonzalez ◽  
...  
Keyword(s):  
Neutron Emission ◽  
Nuclear Data ◽  
Pulse Neutron ◽  
Fission Yield ◽  
Decay Heat ◽  
Data Libraries ◽  
Nuclear Data Libraries

scholarly journals Uncertainties for Swiss LWR spent nuclear fuels due to nuclear data

2018 ◽  
Vol 4 ◽  
pp. 6 ◽  
Author(s):  
Dimitri A. Rochman ◽  
Alexander Vasiliev ◽  
Abdelhamid Dokhane ◽  
Hakim Ferroukhi
Keyword(s):  
Cross Sections ◽  
Power Plants ◽  
Neutron Emission ◽  
Nuclear Data ◽  
Monte Carlo Sampling ◽  
Boiling Water Reactors ◽  
Nuclear Fuels ◽  
Decay Heat ◽  
Water Reactors ◽  
The Impact

This paper presents a study of the impact of the nuclear data (cross sections, neutron emission and spectra) on different quantities for spent nuclear fuels (SNF) from Swiss power plants: activities, decay heat, neutron and gamma sources and isotopic vectors. Realistic irradiation histories are considered using validated core follow-up models based on CASMO and SIMULATE. Two Pressurized and one Boiling Water Reactors (PWR and BWR) are considered over a large number of operated cycles. All the assemblies at the end of the cycles are studied, being reloaded or finally discharged, allowing spanning over a large range of exposure (from 4 to 60 MWd/kgU for ≃9200 assembly-cycles). Both UO2 and MOX fuels were used during the reactor cycles, with enrichments from 1.9 to 4.7% for the UO2 and 2.2 to 5.8% Pu for the MOX. The SNF characteristics presented in this paper are calculated with the SNF code. The calculated uncertainties, based on the ENDF/B-VII.1 library are obtained using a simple Monte Carlo sampling method. It is demonstrated that the impact of nuclear data is relatively important (e.g. up to 17% for the decay heat), showing the necessity to consider them for safety analysis of the SNF handling and disposal.

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