| 1. |
- Chen, Yangli, et al.
(författare)
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A sensitivity study of MELCOR nodalization for simulation of in-vessel severe accident progression in a boiling water reactor
- 2019
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Ingår i: Nuclear Engineering and Design. - : ELSEVIER SCIENCE SA. - 0029-5493 .- 1872-759X. ; 343, s. 22-37
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Tidskriftsartikel (refereegranskat)abstract
- This paper presents a sensitivity study of MELCOR nodalization for simulation of postulated severe accidents in a Nordic boiling water reactor, with the objective to address the nodal effect on in-vessel accident progression, including thermal-hydraulic response, core degradation and relocation, hydrogen generation, source term release, melt behavior and heat transfer in the lower head, etc. For this purpose, three meshing schemes (coarse, medium and fine) of the COR package of MELCOR are chosen to analyze two severe accident scenarios: station blackout (SBO) accident and large break loss-of-coolant accident (LOCA) combined with station blackout. The comparative results of the MELCOR simulations show that the meshing schemes mainly affect the core degradation and relocation to the lower head of the reactor pressure vessel: the fine mesh leads to a delayed leveling process of a heap-like debris bed in the lower head, and a later breach of the vessel. The simulations with fine mesh also provide more detailed distributions of corium mass and temperature, as well as heat flux which is an important parameter in qualification assessment of the In-Vessel Melt Retention (IVR) strategy.
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| 2. |
- Chen, Yangli, et al.
(författare)
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Coupled MELCOR/COCOMO analysis on quench of ex-vessel debris beds
- 2022
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 165
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Tidskriftsartikel (refereegranskat)abstract
- The cornerstone of severe accident strategy of Nordic BWRs is to flood the reactor cavity for the long-termcoolability of an ex-vessel debris bed. As a prerequisite of the long-term coolability, the hot debris bedformed from fuel coolant interactions (FCI) should be quenched. In the present study, coupling of theMELCOR and COCOMO codes is realized with the aim to analyze the quench process of an ex-vessel debrisbed under prototypical condition of a Nordic BWR. In this coupled simulation, MELCOR performs an integralanalysis of accident progression, and COCOMO performs the thermal–hydraulic analysis of the debrisbed in the flooded cavity. The effective diameter of the particles is investigated. The discussion on thebed’s shape shows a significant effect on the propagation of the quench front, due to different flow patterns.Compared with MELCOR standalone simulation, the coupled simulation predicts earlier cavity poolsaturation and containment venting.
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| 3. |
- Chen, Yangli, et al.
(författare)
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Development and application of a surrogate model for quick estimation of ex-vessel debris bed coolability
- 2020
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Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 370
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Tidskriftsartikel (refereegranskat)abstract
- During a hypothetical severe accident of a Nordic boiling water reactor (BWR), an ex-vessel particulate debris bed is expected to form in the flooded lower drywell due to melt-coolant interactions after vessel failure. The key parameter to evaluate debris bed coolability is the dryout heat flux (DHF) or dryout power density, representing the limit of heat removal capacity by the coolant. Several numerical codes such as COCOMO have been developed to simulate thermal hydraulics in multi-dimensional debris beds and predict the cooling limit, but they are computationally expensive and not suitable for probabilistic risk analysis. This paper aims to develop a surrogate model which can serve as a quick-estimate tool for the dryout power density of a heap-like debris bed in a saturated water pool. The dryout power density predicted from the COCOMO code is treated as the full model. A characteristic factor is introduced as the dryout power density ratio between the multi-dimensional debris bed (predicted by COCOMO code) and the corresponding one-dimensional debris bed (predicted by Lipinski 0-D model). The characteristic factor is correlated by the Kriging method with six parameters: bed porosity, particle diameter, debris mass, bed slope, cavity radius and containment pressure. After the surrogate model is trained and validated, it is employed to analyze the coolability of prototypical debris beds of a reference Nordic BWR, given the bed mass and containment pressure from MELCOR simulation. Coolability maps are produced as quick look-up diagrams for identification of coolable domain with the variation of porosity, particle diameter and slope angle. A preliminary uncertainty analysis is performed to demonstrate the effect of uncertain input parameters on non-coolable domain.
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| 4. |
- Chen, Yangli, et al.
(författare)
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Development of surrogate model for debris bed coolability analysis
- 2019
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Ingår i: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019. - : American Nuclear Society. ; , s. 6770-6779
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Konferensbidrag (refereegranskat)abstract
- The cornerstone of severe accident management (SAM) strategy of a Nordic boiling water reactor (BWR) is to flood the reactor cavity with water from the pressure suppression pool before failure of the reactor pressure vessel (RPV). The idea is to form a deep water pool which can accommodate the corium ejected from the RPV breach and cool the debris bed in the reactor cavity. Hence, assessment of debris bed coolant in the deep water pool is of paramount importance to the qualification of this SAM strategy. For the coolability analysis of a debris bed, one needs to estimate the dryout heat flux/power density of the particle bed, which is considered as the limit for heat removal capacity of coolant. For a multi-dimensional debris bed, the dryout power density can be assessed only by numerical simulation of two-phase flow and heat transfer in porous media. Since the numerical simulation is computationally expensive, it is neither suitable for massive calculations, nor feasible to be implemented into a system code (e.g. MELCOR). There is a clear need to develop a fast-running tool to estimate the dryout power density of a prototypical debris bed. The present study is concerned with development of a surrogate model which is sufficient for PSA study or capable of coupling with the MELCOR code without significant sacrifice of computational efficiency. The surrogate model is conceived from the coolability database predicted by COCOMO which is a mechanistic code for simulating thermal-hydraulic response of debris bed and has been extensively validated and applied in our previous studies [1][2]. The comparative results show that the surrogate model is not only able to predict the coolability limit of a debris bed, but also employed in the sensitivity study of bed’s characteristics (e.g., particle diameter, bed geometry and porosity) and the uncertainty and risk analysis.
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| 5. |
- Chen, Yangli
(författare)
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MELCOR Capability Development for Simulation of Debris Bed Coolability
- 2021
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Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- The severe accident management (SAM) strategy for a Nordic boiling water reactor (BWR) employs cavity flooding prior to vessel failure, so that the core melt (corium) discharged from the vessel could fragment and form a particulate debris bed. The key to the success of this SAM strategy is the coolability of ex-vessel debris beds.The safety analysis involves knowledge about the reactor response to severe accidents under this SAM strategy, which requires the integral simulation of a system code such as MELCOR. Since currently the MELCOR code lacks the modeling of ex-vessel particulate debris beds, the present study aims to develop the capability of MELCOR for the simulation of debris bed coolability through the coupling of MELCOR with other codes, which are dedicated to this phenomenon.The study is started from the qualification of a MELCOR model for severe accident analysis of a reference Nordic BWR, with the aim to help identify a proper core nodalization. For this purpose, three different core meshes (coarse, medium, and fine) are employed to obtain their impacts on corium release conditions. It is found the coarse mesh is sufficient in the present study, since it is not only computationally efficient, but also predicting earlier vessel failure and faster corium release, providing a more conservative condition for debris bed coolability analysis.Two couplings are then adopted: (i) coupling of MELCOR with the COCOMO code, which is a mechanistic code for simulation of thermal hydraulics in debris beds; and (ii) coupling of MELCOR with a surrogate model developed in the present study. The first method can simulate the transient behavior of a debris bed during quench process. The second method can efficiently predict the coolability limit (dryout power) required in safety analysis. The surrogate model is developed based on the COCOMO prediction of two-dimensional debris beds.The developed simulation tools, including the coupled codes and the surrogate model, are applied to the safety analysis of the reference Nordic BWR. The coupled MELCOR/COCOMO simulation is used to investigate the debris bed properties. The effective particle diameter is found as approximately 10% larger than the surface mean diameter of a debris bed with distributed sizes, quantified by the quench rate. For the effect of debris bed shape, it shows a faster quench process with a lower bed slope angle. The quench front propagation as well as the responses of local temperature and containment pressure are obtained.The coupled MELCOR/surrogate model simulation is performed to estimate the coolability of ex-vessel vessel debris beds. The results show that debris beds are coolable under prototypical conditions with probable bed properties. The surrogate model is used to generate coolability maps, which show the debris bed coolability with the variation of bed properties. The sensitivity analysis indicates that the porosity and the geometry are the most influential to coolability limit. An uncertainty analysis methodology is proposed to obtain the probability of non-coolable debris beds.
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| 6. |
- Chen, Yangli, et al.
(författare)
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Sensitivity and uncertainty analysis of trace simulation against FIx-II experiments
- 2016
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Ingår i: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017. - : Association for Computing Machinery (ACM).
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Konferensbidrag (refereegranskat)abstract
- In a previous study [1], the US NRC code TRACE was employed to simulate the FIX-II tests which were carried out to investigate the loss of coolant accident (LOCA) of a boiling water reactor (BWR). Results exhibited that the TRACE simulation was sensitive to modelling parameters. In order to further qualify the TRACE code for BWR safety analysis, and to increase our confidence in the simulation results, sensitivity and uncertainty analysis is performed in this paper for the possible uncertain parameters, so as to identify the most influential ones. 12 parameters related to the simulated physical phenomena are selected by resorting to phenomena identification and ranking tables (PIRTs) in relative references. The sensitivity analysis method chosen is based on Finite Mixture Models (FMM) together with Hellinger distance and Kullback-Leibler divergence. Kolmogorov-Smirnov test is first introduced to combine FMM, and it has better performance in screening. Sensitivity analysis results of FMM method show that decay power, choked flow multipliers and break area have the most important influence on calculating peak cladding temperature (PCT). Although previous study failed to predict PCT, uncertainty analysis provides a certain range that successfully covers experiment result.
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| 7. |
- Chen, Yangli, et al.
(författare)
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Uncertainty quantification for TRACE simulation of FIX-II No. 5052 test
- 2020
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Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 143
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Tidskriftsartikel (refereegranskat)abstract
- The Best Estimate Plus Uncertainty approach requires the knowledge of input uncertainties for the uncertainty propagation with best-estimate codes. Inaccurate judgement of some model parameter uncertainties related to the dominant physical phenomena may result in misestimation of the safety margin. This paper presents a framework of inverse uncertainty quantification (UQ) to quantify model parameter uncertainties in order to address this issue. It is applied to TRACE simulation of a large break loss of coolant accident conducted on the FIX-II facility, and peak cladding temperature (PCT) is the simulation output. Sensitivity analysis identifies the parameters of the critical flow model as the most influential to the PCT. The inverse UQ is performed based on Bayesian framework, which adopts Markov Chain Monte Carlo sampling and surrogate modelling algorithms. The quantified uncertainties of the model parameters are the desired results from the inverse UQ process, which are useful in BEPU studies.
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| 8. |
- Jin, Binghan, et al.
(författare)
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Abdominal Adiposity and Total Body Fat as Predictors of Cardiometabolic Health in Children and Adolescents With Obesity
- 2020
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Ingår i: Frontiers in Endocrinology. - : FRONTIERS MEDIA SA. - 1664-2392. ; 11
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Tidskriftsartikel (refereegranskat)abstract
- Objective:We aimed to assess the role of adipose tissue distribution in cardiometabolic risk (in particular insulin sensitivity) in a population of children and adolescents with obesity. Methods:In this cross-sectional study, participants were 479 children and adolescents with obesity (322 boys and 157 girls) aged 3 to 18 years attending the Children's Hospital at Zhejiang University School of Medicine (Hangzhou, China). Clinical assessments included anthropometry, body composition (DXA scans), carotid artery ultrasounds, and OGTT. Insulin sensitivity was assessed using the Matsuda index. Participants were stratified into groups by sex and pubertal stage. Key predictors were DXA-derived android-to-gynoid-fat ratio (A/G) and total body fat percentage (TBF%). Results:Irrespective of sex and pubertal stage, there was a strong association between increasing A/G (i.e., greater abdominal adiposity) and lower insulin sensitivity. In multivariable models, every 0.1 increase in A/G was associated with a reduction in insulin sensitivity in prepubertal boys [-29% (95% CI -36%, -20%);p< 0.0001], pubertal boys [-13% (95% CI -21%, -6%);p= 0.001], and pubertal girls [-16% (95% CI -24%, -6%);p= 0.002]. In contrast, TBF% was not associated with insulin sensitivity when A/G was adjusted for, irrespective of pubertal stage or sex. In addition, every 0.1 increase in A/G was associated with increased likelihood of dyslipidemia in prepubertal boys [adjusted odds ratio (aOR) 1.62 (95% CI 1.05, 2.49)], impaired glucose tolerance in pubertal boys [aOR 1.64 (95% CI 1.07, 2.51)] and pubertal girls [aOR 1.81 (95% CI 1.10, 2.98)], and odds of NAFLD in both prepubertal [aOR 2.57 (95% CI 1.56, 4.21)] and pubertal [aOR 1.69 (95% CI 1.18, 2.40)] boys. In contrast, higher TBF% was only associated with higher fasting insulin and ALT in pubertal boys, being also predictive of NAFLD in this group [aOR 1.15 per percentage point (95% CI 1.06, 1.26)], but was not associated with the likelihood of other cardiometabolic outcomes assessed in any group. Conclusions:A/G is a much stronger independent predictor of cardiometabolic risk factors in children and adolescents with obesity in China, particularly glucose metabolism.
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| 9. |
- Wang, Hongdi, et al.
(författare)
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Thermo-mechanical behavior of an ablated reactor pressure vessel wall in a Nordic BWR under in-vessel core melt retention
- 2021
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Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 379, s. 111196-
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Tidskriftsartikel (refereegranskat)abstract
- The reactor pressure vessel (RPV) of a nuclear reactor is one of the key safety barriers preventing radioactive environmental releases during a severe accident. One of the promising strategies of severe accident management (SAM) is to retain the molten core having continuous decay heat inside the RPV by natural water cooling of the external vessel surface. The feasibility of such a strategy relies on complex safety analyses including accurateprediction of vessel thermo-mechanical behavior which can be assessed by mechanical stresses and strains. In this paper, we present the stress–strain response of an ablated RPV of a Nordic boiling water reactor (BWR) to dynamic thermomechanical loads set by expanding volumetrically heated molten pool inside the RPV cooled by water at the external surface. MELCOR 2.2.9541 severe accident code is used to simulate the in-vessel behavior and provides the input conditions for dedicated structural analysis of the RPV using ANSYS® Mechanical APDL 19.2. A creep model of the SA533B1 vessel steel is validated against uniaxial creep tests carried out by INEL (Idaho National Engineering Laboratory) and creep tests performed at CEA (French AlternativeEnergies and Atomic Energy Commission) as part of the OLHF (OECD Lower Head Failure) Project. Two generic severe accident scenarios are considered: (i) Station Blackout (SBO) and (ii) Station Black-out and Loss-of-coolant Accident (SBO + LOCA). In both scenarios, we found that the RPV has maintained structural integrity considering two failure criteria: stress-based and strain-based.
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| 10. |
- Wang, Hongdi, et al.
(författare)
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Vessel Failure Analysis of a Boiling Water Reactor During a Severe Accident
- 2022
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Ingår i: Frontiers in Energy Research. - : Frontiers Media SA. - 2296-598X. ; 10
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Tidskriftsartikel (refereegranskat)abstract
- In a postulated severe accident, the thermo-mechanical loads from the corium debris that has relocated to the lower head of the reactor pressure vessel (RPV) can pose a credible threat to the RPV's structural integrity. In case of a vessel breach, it is vital to predict the mode and timing of the vessel failure. This affects the ex-vessel accident progression and plays a critical role in the development of mitigation strategies. We propose a methodology to assess RPV failure based on MELCOR and ANSYS Mechanical APDL simulations. A Nordic-type boiling water reactor (BWR) is considered with two severe accident scenarios: i) SBO (Station Blackout) and ii) SBO + LOCA (Loss of Coolant Accident). In addition, the approach considers the dynamic ablation of the vessel wall due to a high-temperature debris bed with the use of the element kill function in ANSYS. The results indicate that the stress failure mechanism is the major cause of the RPV failure, compared to the strain failure mechanism. Moreover, the axial normal stress and circumferential normal stress make the dominant contributions to the equivalent stress sigma at the lower head of RPVs. As expected, the region with high ablation is most likely the failure location in both SBO and SBO + LOCA. In addition, comparisons of the failure mode and timing between SBO and SBO + LOCA are described in detail. A short discussion on RPV failure between ANSYS and MELCOR is also presented.
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