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

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.

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NUCLEAR DATA UNCERTAINTY QUANTIFICATION FOR THE DECAY HEAT OF PWR MOX FUELS USING DATA ASSIMILATION OF ELEMENTARY FISSION BURSTS

EPJ Web of Conferences ◽

10.1051/epjconf/202124710002 ◽

2021 ◽

Vol 247 ◽

pp. 10002

Author(s):

V. Vallet ◽

J. Huyghe ◽

C. Vaglio-Gaudard ◽

D. Lecarpentier ◽

C. Reynard-Carette

Keyword(s):

Data Assimilation ◽

Uncertainty Quantification ◽

Uncertainty Propagation ◽

Fission Products ◽

Nuclear Data ◽

Data Uncertainty ◽

Decay Heat ◽

Using Data ◽

The Impact ◽

Fission Yields

Currently there is no integral experimental data for code validation regarding the decay heat of MOX fuels, excepted fission burst experiments (for fission products contributions at short cooling times) or post-irradiated experiments on nuclide inventories (restricted number of nuclide of interest for decay heat). The uncertainty quantification mainly relies on uncertainty propagation of nuclear data covariances. In the recent years, the transposition method, based on the data assimilation theory, was used in order to transpose the experiment-to-calculation discrepancies at a given set of parameters (cooling time, fuel burnup) to another set of parameters. As an example, this method was used on the CLAB experiments and the experiment-to-calculation discrepancies at 13 years were transposed to an UOX fuel between 5 and 27 years and for burnups from 10 to 50 GWd/t. The purpose of this paper is to study to what extent the transposition method could be used for MOX fuels. In particular, the Dickens fission burst experiment of 239Pu was considered for MOX fuels at short cooling times (< 1h30) and low burnup (< 10 GWd/t). The impact of fission yields (FY) correlations was also discussed. As a conclusion, the efficiency of the transposition process is limited by the experimental uncertainties larger than nuclear data uncertainties, and by the fact that fission burst experiments would only be representative to the FY contribution of the decay heat uncertainty of an irradiated reactor fuel. Nevertheless, this method strengthens the decay heat uncertainties at very short cooling times, previously based only on nuclear data covariance propagation through computation.

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

EPJ Web of Conferences ◽

10.1051/epjconf/201921103004 ◽

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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Uncertainty quantification works relevant to fission yields and decay data

EPJ Nuclear Sciences & Technologies ◽

10.1051/epjn/2018022 ◽

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.

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Extension of Bayesian inference for multi-experimental and coupled problem in neutronics − a revisit of the theoretical approach

EPJ Nuclear Sciences & Technologies ◽

10.1051/epjn/2018046 ◽

2018 ◽

Vol 4 ◽

pp. 19 ◽

Cited By ~ 2

Author(s):

Thomas Frosio ◽

Thomas Bonaccorsi ◽

Patrick Blaise

Keyword(s):

Bayesian Inference ◽

Bayesian Methods ◽

Theoretical Approach ◽

Nuclear Data ◽

Matrix Form ◽

Multidimensional System ◽

Coupled Problem ◽

New Methods ◽

Experimental Values ◽

The Impact

Bayesian methods are known for treating the so-called data re-assimilation. The Bayesian inference applied to core physics allows to get a new adjustment of nuclear data using the results of integral experiments. This theory leading to reassimliation encompasses a broader approach. In previous papers, new methods have been developed to calculate the impact of nuclear and manufacturing data uncertainties on neutronics parameters. Usually, adjustment is performed step by step with one parameter and one experiment by batch. In this document, we rewrite Orlov theory to extend to multiple experimental values and parameters adjustment. We found that the multidimensional system expression looks like can be written as the monodimensional system in a matrix form. In this extension, correlation terms appears between experimental processes (manufacturing and measurements) and we discuss how to fix them. Then formula are applied to the extension to the Boltzmann/Bateman coupled problem, where each term could be evaluated by computing depletion uncertainties, studied in previous papers.

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Fission yields and cross sections: correlated or not?

EPJ Nuclear Sciences & Technologies ◽

10.1051/epjn/2021005 ◽

2021 ◽

Vol 7 ◽

pp. 5

Author(s):

Dimitri Alexandre Rochman ◽

Eric Bauge

Keyword(s):

Experimental Data ◽

Cross Sections ◽

Previous Analysis ◽

Fission Products ◽

Nuclear Data ◽

Isotopic Compositions ◽

Different Types ◽

Fission Yields ◽

Integral Data ◽

Selection Of

Cross sections and fission yields can be correlated, depending on the selection of integral experimental data. To support this statement, this work presents the use of experimental isotopic compositions (both for actinides and fission products) from a sample irradiated in a reactor, to construct correlations between various cross sections and fission yields. This study is therefore complementing previous analysis demonstrating that different types of nuclear data can be correlated, based on experimental integral data.

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Investigations About Released Debris and Corrosion Products and Their Impact to Head Loss Build Up and Decay Heat Removal Under Water-Chemical PWR-LOCA Conditions

Volume 3: Thermal-Hydraulics ◽

10.1115/icone24-60261 ◽

2016 ◽

Author(s):

Stefan Renger ◽

Sören Alt ◽

Wolfgang Kästner ◽

André Seeliger

Keyword(s):

Heat Transfer ◽

Boric Acid ◽

Reactor Core ◽

Head Loss ◽

Corrosion Products ◽

Zinc Borate ◽

Insulation Material ◽

Decay Heat ◽

The Impact

Investigations about the release, transportation and deposit of fibrous insulation material (FIM), corrosion products as well as resulting compounds and debris mixtures become more important to reactor safety research, when considering long-term behavior of emergency core cooling systems (ECCS) during loss of coolant accidents (LOCA). Debris released by a leakage jet leads to head loss buildups at the sump strainers, the debris filters and the spacers of fuel assemblies. However, these complex processes may influence the decay heat transfer out of the reactor core. A similar but newer scenario implies that the boric acid coolant in pressurized water reactors (PWR) can support corrosion processes at hot-dip galvanized installations, leading to a significant release of ionic zinc into the coolant during the sump recirculation phase. A long-term change of chemical properties of the coolant (e.g. pH, zinc ion concentration) has to be considered in safety analyses. Chemical analyses showed that the solubility of zinc in boric acid coolant is inversely proportional to the coolant temperature. Consequently, zinc ions can be dissolved at lower temperatures in the containment sump. Precipitations of zinc borate (ZBP) are possible at hot spots in the reactor core. The ZBP can be formed as a flocculent disperse phase in the coolant or as solidified layers at hot fuel rod surfaces. Layer spalling could lead to the release of further solid particles into coolant flow inside the reactor core. In several joint research projects between the Zittau/Görlitz University of Applied Sciences (HSZG) and the Helmholtz-Zentrum Dresden-Rossendorf (HZDR) investigations on the impact of these processes to the head loss buildup and the heat transfer from core were done at laboratory and semi-technical scaled test facilities. Generic experiments showed the formation of ZBP in heating rod configurations. The ZBP may remain in the core structures or can be transported on debris filter cakes in upstream and downstream components of ECCS and influence the head loss. After this, research addressed the systematical clarification of physico-chemical mechanisms and their influence on thermal-hydraulic-dynamic processes occurring as a consequence of flow induced corrosion effects during LOCA. This paper includes a description of the most important involved test facilities, applied measuring techniques, an overview of boundary conditions considered experimentally and selected results.

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Uncertainties for Swiss LWR spent nuclear fuels due to nuclear data

EPJ Nuclear Sciences & Technologies ◽

10.1051/epjn/2018005 ◽

2018 ◽

Vol 4 ◽

pp. 6 ◽

Cited By ~ 2

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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Feedback from experimental isotopic compositions of used nuclear fuels on neutron cross sections and cumulative fission yields of the JEFF-3.1.1 library by using integral data assimilation

EPJ Nuclear Sciences & Technologies ◽

10.1051/epjn/2019056 ◽

2019 ◽

Vol 5 ◽

pp. 24

Author(s):

Axel Rizzo ◽

Claire Vaglio-Gaudard ◽

Gilles Noguere ◽

Romain Eschbach ◽

Gabriele Grassi ◽

...

Keyword(s):

Data Assimilation ◽

Cross Sections ◽

Fission Products ◽

Nuclear Data ◽

Capture Cross ◽

Nuclear Fuels ◽

Fission Yields ◽

Integral Data ◽

Capture Cross Sections

Comparisons of calculated and experimental isotopic compositions of used nuclear fuels can provide valuable information on the quality of nuclear data involved in neutronic calculations. The experimental database used in the present study − containing more than a thousand isotopic ratio measurements for UOX and MOX fuels with burnup ranging from 10 GWd/t up to 85 GWd/t − allowed to investigate 45 isotopic ratios covering a large number of actinides (U, Np, Pu, Am and Cm) and fission products (Nd, Cs, Sm, Eu, Gd, Ru, Ce, Tc, Mo, Ag and Rh). The Integral Data Assimilation procedure implemented in the CONRAD code was used to provide nuclear data trends with realistic uncertainties for Pressurized Water Reactors (PWRs) applications. Results confirm the quality of the 235U, 239Pu and 241Pu neutron capture cross sections available in the JEFF-3.1.1 library; slight increases of +1.2 ± 2.4%, +0.5 ± 2.2% and +1.2 ± 4.2% are respectively suggested, these all being within the limits of the quoted uncertainties. Additional trends on the capture cross sections were also obtained for other actinides (236U, 238Pu, 240Pu, 242Pu, 241Am, 243Am, 245Cm) and fission products (103Rh, 153Eu, 154Eu) as well as for the 238U(n,2n) and 237Np(n,2n) reactions. Meaningful trends for the cumulative fission yields of 144Ce, 133Cs, 137Cs and 106Ru for the 235U(nth,f) and 239Pu(nth,f) reactions are also reported.

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