Comparison of the behaviour of two core designs for ASTRID in case of severe accidents

Bertrand, Frédèric; Marie, N.; Prulhiere, G.; Lecerf, J.; Seiler, Jean-Marie

doi:10.1016/j.nucengdes.2015.04.020

Cited by 33 publications

(24 citation statements)

References 4 publications

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“…24 In recent years, several concepts of the ASTRID core were examined. 8 The main characteristics of the CFV-V0 core are summarized in Table 1 and shown in Figure 6, and the approximate plutonium vector of the ASTRID core is summarized in Table 2. 8 The main characteristics of the CFV-V0 core are summarized in Table 1 and shown in Figure 6, and the approximate plutonium vector of the ASTRID core is summarized in Table 2.…”

Section: Examined Configurationsmentioning

confidence: 99%

“…8 In the current study, several accidental scenarios are investigated in the CFV design context, in particular, separated degradation of the upper and lower fissile regions, followed by fuel crossing the fertile zone to form a large fissile molten pool. 8 In the current study, several accidental scenarios are investigated in the CFV design context, in particular, separated degradation of the upper and lower fissile regions, followed by fuel crossing the fertile zone to form a large fissile molten pool.…”

Section: Introductionmentioning

confidence: 99%

“…With the development of the French ASTRID SFR industrial demonstrator, the low void fraction (Coeur à Faible effet de Vide sodium or CFV) core design presents particularities that potentially deserve an experimental validation, in terms of core physics. 8 In the current study, several accidental scenarios are investigated in the CFV design context, in particular, separated degradation of the upper and lower fissile regions, followed by fuel crossing the fertile zone to form a large fissile molten pool. Large uncertainties on the behavior of the fuel could lead to local power excursions.…”

Section: Introductionmentioning

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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Section: Examined Configurationsmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Section: Introductionmentioning

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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“…The previous work focused on the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) reduced void fraction (CFV) core . Recent developments related to the ASTRID project experienced substantial modifications, with a significant reduction in power (from 600 MWe to 100‐200 MWe).…”

Section: Introductionmentioning

confidence: 99%

“…7 The previous work 7 focused on the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) reduced void fraction (CFV) core. 11 Recent developments related to the ASTRID project experienced substantial modifications, 12 with a significant reduction in power (from 600 MWe to 100-200 MWe). Nevertheless, the ASTRID CFV concept presents an interesting challenge in terms of severe accident studies, due to the high heterogeneity of the core.…”

Section: Introductionmentioning

confidence: 99%

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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Severe accident calculation tools used for severe accident and mitigation studies

Seiler,

Gubernatis,

Trotignon

et al. 2025

Severe Accidents in Sodium Fast Reactors

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Comparison of the behaviour of two core designs for ASTRID in case of severe accidents

Cited by 33 publications

References 4 publications

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

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

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

Severe accident calculation tools used for severe accident and mitigation studies

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