The path for innovative severe accident neutronics studies in ZPRs. Part I.1 - Analysis of SNEAK-12A experiments for core disruption in LMFBRs

Margulis, Marat; Blaise, Patrick; Gabrielli, F.; Gruel, A.; Mellier, F.; Gilad, Erez

doi:10.1016/j.pnucene.2016.11.001

Cited by 10 publications

(5 citation statements)

References 18 publications

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“…The results indicate that there are six major isotopes that contribute to the total propagated uncertainties, ie, 56 Fe, 206,207 Pb, 238 U, and 239,240 Pu. The propagation of uncertainties, performed with the ENDF covariance data, follows the same trend obtained with the JEFF covariance data evaluation, 32,33 similarly to previous analysis of experimental programs related to SCA studies (the SNEAK-12A/B experimental program 34,35 ). The uncertainty related to 238 U and 239 Pu remains high in the two different evaluations.…”

Section: Uncertainty Propagation In the Elsy Designmentioning

confidence: 73%

Modeling representative Gen-IV molten fuel reactivity effects in the ZEPHYR ZPR - LFR analysis

2018

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Summary Studies related to severe core accidents constitute a crucial element in the safety design of Gen‐IV systems. A new experimental program, related to severe core accidents studies, is proposed for the zero‐power experimental physics reactor (ZEPHYR) future reactor. The innovative program aims at studying reactivity effects at high temperature during degradation of Gen‐IV cores by using critical facilities and surrogate models. The current study introduces the European lead‐cooled system (ELSY) as an additional Gen‐IV system into the representativity arsenal of the ZEPHYR, in addition to the sodium‐cooled fast reactors. Furthermore, this study constitutes yet another step towards the ultimate goal of studying severe core accidents on a full core scale. The representation of the various systems is enabled by optimizing the content of plutonium oxide in the ZEPHYR fuel assembly. The study focuses on representing reactivity variation from 900°C at nominal state to 3000°C at a degraded state in both ELSY and Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) cores. The study utilizes the previously developed calculation scheme, which is based on the coupling of stochastic optimization process and Serpent 2 code for sensitivity analysis. Two covariance data are used: the ENDF 175 groups for ELSY and the Covariance Matrix Cadarache (COMAC) 33 groups for ASTRID. The effect of the energy group structure of the covariance data on the representativity process is found to be significant. The results for single degraded ELSY fuel assembly demonstrate high representativity factor (>0.95) for reactivity variation and for the criticality level. Also, it is shown that the finer energy group structure of the covariance matrices results in dramatic improvement in the representation level of reactivity variations.

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Section: Uncertainty Propagation In the Elsy Designmentioning

confidence: 73%

Modeling representative Gen-IV molten fuel reactivity effects in the ZEPHYR ZPR - LFR analysis

2018

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“…15,16,22,23,31 UCOM-V01AB was created for the sole purpose of this project to studying the sensitivity of the optimization process to the changes in the ND uncertainties. The COMAC-V01 contains rather high uncertainties on several key isotopes, as presented in the paper, while UCOM-V01AB is updated by using older experimental data available from SNEAK-12A/B severe accident studies experimental program.…”

Section: Discussionmentioning

confidence: 99%

“…The COMAC-V01 contains rather high uncertainties on several key isotopes, as presented in the paper, while UCOM-V01AB is updated by using older experimental data available from SNEAK-12A/B severe accident studies experimental program. 15,16,22,23,31 UCOM-V01AB was created for the sole purpose of this project to studying the sensitivity of the optimization process to the changes in the ND uncertainties. It should be noted that COMAC considering some data assimilation from SFR representative configurations, thus consequently, these data are also available in UCOM.…”

Section: Discussionmentioning

confidence: 99%

“…The UCOM-V01AB is an adjusted version of the COMAC-V01 version using an experimental data available from the recently published SNEAK-12A/B benchmark, 15,16,22,23 which focused on severe accident in liquid metal-cooled fast reactors. The results of the benchmark calculation based on the JEFF-3.1.1 library with the Serpent 2 code show excellent agreement.…”

Section: Variance-covariance Datamentioning

confidence: 99%

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Modeling representative Gen-IV molten fuel reactivity effects in the ZEPHYR fast/thermal coupled ZPRs. Part I-Assembly level

2018

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Summary The comprehension of severe criticality accident is a key issue in Gen‐IV neutronics and safety. Within the future zero‐power experimental physics reactor (ZEPHYR), to be built in Cadarache in the next decade, innovative approaches to reproduce high temperature partially degraded Gen‐IV cores into a critical facility is being investigated. This work presents the first attempt to represent a fuel assembly of sodium‐cooled fast reactor severe criticality accident based on surrogate models. One identified way to construct such representative configuration is to use MASURCA plates stockpile (MOX, UOx, Na, U, and Pu metal) in a fast/thermal coupled core to model a stratified molten assembly. The present study is the first step in a more global approach to full core analysis. The approach is based on a nature‐inspired metaheuristic algorithm, the particle swarm optimization algorithm, to find relevant ZEPHYR configuration at 20°C that exhibits characteristics of (2000‐3000°C) molten MOX assembly in a stratified metal arrangement in a reference sodium‐cooled fast reactor core. Thus, the underlying research question of this study is whether it is possible to represent temperature‐related reactivity effects occurring at fuel meltdown temperatures in a power reactor as density‐related reactivity effects at the operation temperature of a zero‐power reactor, and if so, how should it be done? The calculations are based on a Serpent‐2 Monte Carlo sensitivity and representativity analysis using the Cadarache's cross sections covariance data (COMAC). The single fuel assembly studies show that it is possible to represent the multiplication factor with a representativity factor greater than 0.98. As for reactivity variations, it is possible to achieve a satisfactory representativity factor of above 0.85 in all the presented cases. The representativity process demonstrates that temperature effects could be translated into density effects with good confidence. A complementary analysis on modified nuclear data covariance matrix demonstrates the importance of selecting consistent and robust uncertainties in the particle swarm optimization algorithm. This work provides insights on the behavior of the representativity scheme in different core states and shades some light on the problem in hand.

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“…For example, different types of LWR (e.g., PWR, BWR, and MTR) FAs contain water or void channels for improved or reduced neutron moderation [21]; research reactors often assemble a core with a water gap at the center (thermal flux trap) to irradiate samples [22,23], and the neutron reflector in many reactors contains water channels for cooling and moderation [24,25]. Fuel deformation may also lead to water gaps due to either mechanical bowing [26][27][28][29][30] or meltdown during severe core accidents [31,32].…”

Section: Introductionmentioning

confidence: 99%

High-Fidelity Monte Carlo Analysis of Highly-Localized Fission Power Peaking near a Flux Trap in an MTR Core

Abuzlf,

Castagna,

Makmal

et al. 2023

International Journal of Energy Research

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In this study, a high-fidelity Monte Carlo model was utilized to investigate the localized behavior of the power peaking factor in a reactor core. The model considered the effect of water gaps, also known as flux traps, on the local power peaking. The model, developed using the Serpent code, employs a fine spatial mesh to accurately describe the fission power and burnup distribution. The results obtained from the model were validated through comparison with experimental measurements of the power distribution obtained from the IRR1 facility. The study found that the local accumulation of thermal flux caused by enhanced neutron moderation in the flux trap leads to a highly localized increase in power density, affecting the power peaking factor. The quantification of this effect was a key finding of the study. The results obtained from the study were highly dependent on the model’s fidelity, with significant differences being observed between the power peaking factor calculated on a fine mesh as opposed to that calculated at the plate or assembly scale. Furthermore, the experimental validation of the model enabled the prediction of the power peaking factor in other regions within the fuel assembly, which are not accessible to measurement instrumentation, with a high degree of confidence. The study’s findings may be useful for optimizing reactor design and operation and assessing the safety margins in reactors with similar characteristics.

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The path for innovative severe accident neutronics studies in ZPRs. Part I.1 - Analysis of SNEAK-12A experiments for core disruption in LMFBRs

Cited by 10 publications

References 18 publications

Modeling representative Gen-IV molten fuel reactivity effects in the ZEPHYR ZPR - LFR analysis

Modeling representative Gen-IV molten fuel reactivity effects in the ZEPHYR ZPR - LFR analysis

Modeling representative Gen-IV molten fuel reactivity effects in the ZEPHYR fast/thermal coupled ZPRs. Part I-Assembly level

High-Fidelity Monte Carlo Analysis of Highly-Localized Fission Power Peaking near a Flux Trap in an MTR Core

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