Physical tool for Unprotected Loss Of Flow transient simulations in a Sodium Fast Reactor

Droin, Jean-Baptiste; Marie, N.; Bachrata, A.; Bertrand, Frédèric; Merle, E.; Seiler, Jean-Marie

doi:10.1016/j.anucene.2017.03.035

Cited by 21 publications

(4 citation statements)

References 11 publications

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“…Within the framework of 4th-generation sodium-cooled fast reactor ASTRID: Advanced Sodium Technological Reactor for Industrial Demonstration (see Figure 4), the CEA (French Commissariat à l'Énergie atomique et aux Énergies alternatives) provides numerical tools to model severe accident scenarios and assess the safety. Among them, a numerical tool called MACARENa (French: Modélisation de l'ACcident d'Arrêt des pompes d'un REacteur refroidi au sodium) developed by Droin et al 39 simulates a primary phase of an unprotected loss of flow (ULOF) accident. During this type of accident, the power loss of primary pumps and the dysfunction of shutdown systems cause a gradual decrease of the sodium flow in the primary circuit, which subsequently may increase the temperature of sodium until it boils.…”

Section: Nuclear Safety Applicationmentioning

confidence: 99%

Second‐level global sensitivity analysis of numerical simulators with application to an accident scenario in a sodium‐cooled fast reactor

Meynaoui

Marrel

Laurent

2022

Quality & Reliability Eng

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Numerical simulators are widely used to model physical phenomena and global sensitivity analysis (GSA) aims at studying the global impact of the input uncertainties on the simulator output. To perform GSA, statistical tools based on inputs/output dependence measures are commonly used. We focus here on the Hilbert–Schmidt independence criterion (HSIC). Sometimes, the probability distributions modeling the uncertainty of inputs may be themselves uncertain and it is important to quantify their impact on GSA results. We call it here the second‐level global sensitivity analysis (GSA2). However, GSA2, when performed with a Monte Carlo double‐loop, requires a large number of model evaluations, which is intractable with CPU time expensive simulators. To cope with this limitation, we propose a new statistical methodology based on a Monte Carlo single‐loop with a limited calculation budget. First, we build a unique sample of inputs and simulator outputs, from a well‐chosen probability distribution of inputs. From this sample, we perform GSA for various assumed probability distributions of inputs by using weighted HSIC measures estimators. Statistical properties of these weighted estimators are demonstrated. Subsequently, we define 2nd‐level HSIC‐based measures between the distributions of inputs and GSA results, which constitute GSA2 indices. The efficiency of our GSA2 methodology is illustrated on an analytical example, thereby comparing several technical options. Finally, an application to a test case simulating a severe accidental scenario on nuclear reactor is provided.

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Section: Nuclear Safety Applicationmentioning

confidence: 99%

Second‐level global sensitivity analysis of numerical simulators with application to an accident scenario in a sodium‐cooled fast reactor

Meynaoui

Marrel

Laurent

2022

Quality & Reliability Eng

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“…Based on a large set of calculations (typically several thousands), a statistical treatment of the calculation results will permit to estimate the probability to exceed a limit on a safety barrier. Finally, the combination of the difference between the best-estimate calculation and the criterion and the risk that the calculated parameter exceeds the criterion gives the design margin [35,36].…”

Section: Demonstration Tools For Safety Studiesmentioning

confidence: 99%

A Comparative Study on Severe Accident Phenomena Related to Melt Progression in Sodium Fast Reactors and Pressurized Water Reactors

Bachrata

Bertrand

Marie

et al. 2020

Journal of Nuclear Engineering and Radiation Science

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The nuclear safety approach has to cover accident sequences involving core degradation in order to develop reliable mitigation strategies for both existing and future reactors. In particular, the long-term stabilization of the degraded core materials and their coolability has to be ensured after a severe accident. This paper focuses on severe accident phenomena in Pressurized Water Reactors (PWR) compared to those potentially occurring in future GenIV-type Sodium Fast Reactors (SFR). Firstly, the two considered reactor concepts are introduced by focusing on safety aspects. The severe accident scenarios leading to core melting are presented and the initiating events are highlighted. The paper focuses on in-vessel severe accident phenomena, including the chronology of core damage, major changes in the core configuration and molten core progression. Regarding the mitigation means, the in-vessel retention phenomena and the core catcher characteristics are reviewed for these different nuclear generation concepts (II, III and IV). A comparison between the PWR and SFR severe accident evolution is provided as well as the relation between governing physical parameters and the adopted mitigation provisions for each reactor concept. Finally, it is highlighted how the robustness of the safety demonstration is established by means of a combined probabilistic and deterministic approach.

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“…This reference bounding scenario, which was an Unprotected Loss of Flow accident (ULOF), was simulated by mechanistic codes such as SAS4A [4] and SIMMER-III [5]. Currently the safety approach as regards mitigation of severe accident requires to simulate different possible accident scenarios classified into three main families: USAF (Unprotected Sub-Assembly Fault) [6], ULOF (Unprotected Loss Of Flow) [7], UTOP (Unprotected Transient OverPower [2]).…”

Section: Introductionmentioning

confidence: 99%

Advanced Studies and Statistical Treatment for Sodium-Cooled Fast Reactor Pin Failures During Unprotected Transient Overpower Accident

Marie

Herbreteau

Marrel

et al. 2020

Nuclear Science and Engineering

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Usually, simulation tools are validated on experimental data considering a Best Estimate simulation case and there is no quantification of this validation, which remains based on a rough expert judgment. This paper presents an advanced validation treatment of the simulation tool OCARINa, devoted to Unprotected Transient OverPower (UTOP) accidents, on two CABRI tests, considering this time, a Best Estimate Plus Uncertainties (BEPU) approach. The output results of interest are both scalar physical data such as the time and location of the pin failure and associated molten mass and vector data such as temperature axial distribution or temperature evolution versus time. This approach is a first step in quantifying the degree of agreement between the calculation results and the experimental results. It is of great interest for the VV&UQ (Verification, Validation and Uncertainty Quantification) approach, which leads to the qualification of scientific calculation tools.Within the framework of the Generation IV SFR R&D project in which the CEA is involved, OCARINa is a physical tool, relevant for performing pre-conceptual design studies, devoted to simulation of UTOP accidents on heterogeneous cores. Such accidents could not be simulated with mechanistic calculation tools such as SAS4A or SIMMER with their current capabilities; the thermomechanical models are not finalized in SIMMER tool and the SAS4A tool is only validated for homogeneous cores. The final objective aims at deriving the variability of the main results of interest to quantify the safety margins.The final use of the OCARINa tool being to perform sensitivity studies on the various possible sodium fast nuclear pre-conceptual core designs, the validation of this tool is first discussed at the pin scale (where separate effect test measurements are available) based on statistical treatment. This enables to determine the lacks and uncertainties of this tool. The modeling is then extended from local pin behavior to global core behavior adding a point kinetic neutronic model. Final simulations of UTOP accidents caused by a uniform space reactivity ramp on an SFR (Sodium-cooled Fast Reactor) core are realized taking into account the specificities of the pins of the various assemblies. The orders of magnitude of mechanical energy released are derived.

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Physical tool for Unprotected Loss Of Flow transient simulations in a Sodium Fast Reactor

Cited by 21 publications

References 11 publications

Second‐level global sensitivity analysis of numerical simulators with application to an accident scenario in a sodium‐cooled fast reactor

Second‐level global sensitivity analysis of numerical simulators with application to an accident scenario in a sodium‐cooled fast reactor

A Comparative Study on Severe Accident Phenomena Related to Melt Progression in Sodium Fast Reactors and Pressurized Water Reactors

Advanced Studies and Statistical Treatment for Sodium-Cooled Fast Reactor Pin Failures During Unprotected Transient Overpower Accident

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