| 1. |
- Bandini, G., et al.
(författare)
-
Safety Analysis Results of Representative DEC Accidental Transients for the ALFRED Reactor
- 2013
-
Konferensbidrag (refereegranskat)abstract
- The conceptual design of the Advanced Lead Fast Reactor European Demonstrator (ALFRED) is under development within the LEADER project to meet the safety objectives of Gen IV nuclear energy systems. This paper presents the main results of the safety analysis for beyond design basis conditions, namely design extension conditions (DEC), which include the failure of prevention and mitigation systems, like the reactor scram in the so called unprotected transients. The main objective of this analysis is to evaluate the impact of the core and plant design features on the intrinsic safety behaviour of the ALFRED reactor. Several computer codes: SIM LFR, RELAP5, CATHARE, SPECTRA and TRACE are applied to evaluate the consequences of representative unprotected accident scenarios such as Loss of Flow, Loss of Heat Sink and Reactivity initiated accidents. Additionally, the consequences of steam generator tube rupture and partial sub assembly flow blockage events are assessed by means of appropriate fluid dynamic codes.
|
|
| 2. |
- Geffray, C., et al.
(författare)
-
Verification and validation and uncertainty quantification
- 2018
-
Ingår i: Thermal Hydraulics Aspects of Liquid Metal Cooled Nuclear Reactors. - : Elsevier. - 9780081019801 - 9780081019818 ; , s. 383-405
-
Bokkapitel (övrigt vetenskapligt/konstnärligt)abstract
- In this chapter, an overview of the verification, validation, and uncertainty quantification process is offered. First, the context of the dialog with the safety authorities is explained, and the need for a thorough code validation procedure able to meet the regulatory safety requirements is highlighted. Then, the concept of code verification is introduced, and the main steps are described. The validation process is depicted next. Emphasis is made upon the identification of the physical phenomena of interest and upon the choice of adequate computational tools to capture them. The targeted validity domain of these computational tools and its dependence on available and accurate experimental data are detailed with respect to the issue of scaling. Finally, an overview of selected techniques for uncertainty and sensitivity analysis is provided.
|
|
| 3. |
|
|
| 4. |
- Jeltsov, Marti, 1985-, et al.
(författare)
-
Parametric study of sloshing effects in the primary system of an isolated lead-cooled fast reactor
- 2015
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 190:1, s. 1-10
-
Tidskriftsartikel (refereegranskat)abstract
- Risks related to sloshing of liquid metal coolant due to seismic excitation need to be investigated. Sloshing effects on reactor performance include first, fluid-structure interaction and second, gas entrapment in the coolant with subsequent transport of void to the core region. While the first can hypothetically lead to structural damage or coolant spill, the second increases the risk of a reactivity insertion accident and/or local dryout of the fuel. A two-dimensional computational fluid dynamics study is carried out in order to obtain insights into the modes of sloshing depending on the parameters of seismic excitation. The applicability and performance of the numerical mesh and the Eulerian volume of fluid method used to track the free surface are evaluated by modeling a simple dam break experiment. Sloshing in the cold plenum free surface region of the European Lead-cooled SYstem (ELSY) conceptual pool-type lead-cooled fast reactor (LFR) is studied. Various sinusoidal excitations are used to imitate the seismic response at the reactor level. The goal is to identify the domain of frequencies and magnitudes of the seismic response that can lead to loads threatening the structural integrity and possible core voiding due to sloshing. A map of sloshing modes has been developed to characterize the sloshing response as a function of excitation parameters. Pressure forces on vertical walls and the lid have been calculated. Finally, insight into coolant voiding has been provided.
|
|
| 5. |
- Jeltsov, Marti, 1985-, et al.
(författare)
-
Pre-test analysis of an LBE solidification experiment in TALL-3D
-
Ingår i: Nuclear Engineering and Design. - 0029-5493 .- 1872-759X.
-
Tidskriftsartikel (refereegranskat)abstract
- This paper presents the process of development of a solidification test section design for TALL-3D experimental facility (lead-bismuth eutectic (LBE) thermal-hydraulic loop of prototypic height). Coolant solidification is a phenomenon of potential safety importance for liquid metal cooled fast reactors (LMFRs). Solidification, e.g. in case of excessive performance of the passive decay heat removal systems, can affect local heat transfer and even lead to partial or complete blockage of the coolant flow paths. This might lead to failure of decay heat removal function. In case of reduced flow circulation, the temperature of the coolant will increase, which might prevent complete blockage of the flow. Prediction of possible outcomes of such scenarios with complex interactions between local physical phenomena of solidification and system scale natural circulation behavior is subject to modelling (epistemic) uncertainty. Development and validation of adequate models requires validation grade experimental data. In this work we discuss results of analysis carried out in support experiment development. The aim of the experimental design is to satisfy requirements stemming from the process of qualification of the model that can be used for addressing the safety-related concerns. In this work we focus on two aspects: (i) design of solidification test section (STS) for investigation of local solidification phenomena of lead-bismuth eutectic (LBE), and (ii) effect of the STS on the system scale behavior of the experimental facility. We discuss selection of the STS characteristics and experimental test matrix using computational fluid dynamic (CFD) and system thermal-hydraulic (STH) codes.
|
|
| 6. |
- Jeltsov, Marti, 1985-, et al.
(författare)
-
Pre-test analysis of an LBE solidification experiment in TALL-3D
- 2018
-
Ingår i: Nuclear Engineering and Design. - : Elsevier Ltd. - 0029-5493 .- 1872-759X. ; 339, s. 21-38
-
Tidskriftsartikel (refereegranskat)abstract
- Coolant solidification is a phenomenon of potential safety importance for Liquid Metal Cooled Fast Reactors (LMFRs). Coolant solidification can affect local flow, heat transfer and lead to partial or complete blockage of the coolant flow paths jeopardizing decay heat removal function. It is also possible that reduced flow circulation may increase coolant temperature, counteract solidification and prevent complete blockage of the flow. Complex interactions between local physical phenomena of solidification and system scale natural circulation make modelling of solidification uncertain. Development and validation of adequate models requires validation grade experimental data. In this work we discuss results of analysis carried out in support of experiment development, specifically, design of a solidification test section and a test matrix for TALL-3D experimental facility (lead-bismuth eutectic (LBE) thermal-hydraulic loop). The aim of the analysis is experimental design that satisfies requirements stemming from the process of model qualification. We focus on two aspects: (i) design of solidification test section (STS) for investigation of solidification phenomena in lead-bismuth eutectic (LBE), and (ii) effect of the STS pool on the system scale behavior of the TALL-3D facility. Selection of the STS characteristics and experimental test matrix is supported using computational fluid dynamic (CFD) and system thermal-hydraulic (STH) codes.
|
|
| 7. |
- Jeltsov, Marti, 1985-, et al.
(författare)
-
Risk of sloshing in the primary system of a lead cooled fast reactor
- 2014
-
Ingår i: PSAM 2014 - Probabilistic Safety Assessment and Management.
-
Konferensbidrag (refereegranskat)abstract
- Pool-type designs of Lead-cooled Fast Reactor (LFR) aim for commercial viability by simplified engineering solutions and passive safety systems. However, such designs carry the risks related to heavy coolant sloshing in case of seismic event. Sloshing can cause (i) structural damage due to fluid-structure interaction (FSI) and (ii) core damage due to void induced reactivity insertion or due to local heat transfer deterioration. The main goal of this study is to identify the domain of seismic excitation characteristics at the reactor vessel level that can lead to exceedance of the safety limits for structural integrity and core damage. Reference pool-type LFR design used in this study is the European Lead-cooled SYstem (ELSY). Liquid lead sloshing is analyzed with Computational Fluid Dynamics (CFD) method. Outcome of the analysis is divided in two parts. First, different modes of sloshing depending on seismic excitation are identified. These modes are characterized by wave shapes, loads on structures and entrapped void. In the second part we capitalize on the framework of Integrated Deterministic-Probabilistic Safety Assessment (IDPSA) to quantify the risk. Specifically, statistical parameters pertaining to mechanical loads and void transport are quantified and combined with the deterministically obtained data about consequences.
|
|
| 8. |
- Jeltsov, Marti, 1985-, et al.
(författare)
-
Steam generator leakage in lead cooled fast reactors : Modeling of void transport to the core
- 2018
-
Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 328, s. 255-265
-
Tidskriftsartikel (refereegranskat)abstract
- Steam generator tube leakage and/or rupture (SGTL/R) is one of the safety issues for pool type liquid metal cooled fast reactors. During SGTL/R, water is injected from high-pressure secondary side to low-pressure primary side. The possible consequences of such an event include void transport to the core that has adverse effects on the reactor performance including heat transfer deterioration and reactivity insertion. This paper addresses the potential transport of steam bubbles to the core and subsequent void accumulation in the primary system in ELSY conceptual reactor. A CFD model of the primary coolant system for nominal operation is developed and verified. Bubble motion is simulated using Lagrangian tracking of steam bubbles in Eulerian flow field. The effects of uncertainties in the bubble size distribution and bubble drag are addressed. A probabilistic methodology to estimate the core and primary system voiding rates is proposed and demonstrated. A family of drag correlations by Tomiyama et al. (1998) provide the best agreement with the available experimental data. Primary system and core voiding analysis demonstrate that the smallest (sub-millimeter) bubbles have the highest probability to be entrained and remain in the coolant flow. It is found that leaks at the bottom region of the SG result in larger rates of void accumulation.
|
|
| 9. |
- Jeltsov, Marti, 1985-
(författare)
-
Validation and Application of CFD to Safety-Related Phenomena in Lead-Cooled Fast Reactors
- 2018
-
Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- Carbon-free nuclear power production can help to relieve the increasing energy demand in the world. New generation reactors with improved safety, sustainability, reliability, economy and security are being developed. Lead-cooled fast reactor (LFR) is one of them.Lead-cooled reactors have many advantages related to the use of liquid metal coolants and pool-type primary system designs. Still, several technological challenges are to be overcome before their successful commercial deployment. The decisions on new designs and licensing of LFRs require use of code simulations, due to lack of operational experience and the fact that full scale experimental investigations are impractical. Code development and validation is necessary in order to reduce uncertainty in predictions for risk assessment and decision support. In order to make the decisions robust, i.e. not sensitive to remaining uncertainty, the process of collecting necessary evidences should iteratively converge. Extensive sensitivity and uncertainty analysis is necessary in order to make the process user-independent.The pool-type design introduces 3D phenomena that require application of advanced modeling tools such as Computational Fluid Dynamics (CFD). The focus of this thesis is on the development and application of CFD modeling and general code validation methodology to the following issues of safety importance for LFR: (i) pool thermal stratification and mixing, (ii) steam generator tube leakage, (iii) seismic sloshing, and (iv) solidification.Thermal mixing and stratification in a pool-type reactor can affect natural circulation stability and thus decay heat removal capability and pose thermal stresses on the reactor structures. Standalone System Thermal-hydraulics (STH) and CFD codes are either inadequate or too computationally expensive for analysis of such phenomena in prototypic conditions. In this work an approach to coupling of STH and CFD is proposed and implemented. For validation of standalone and coupled codes TALL-3D facility (lead-bismuth eutectic (LBE) loop) and test matrix was designed featuring significant two-way thermal-hydraulic feedbacks between the pool-type test section and the rest of the loop.The possibility of core voiding due to bubble transport in case of steam generator tube leakage/rupture (SGTL/R) is a serious safety concern that can be a showstopper for LFR licensing. Voiding of the core increases risks for reactivity insertion accident (RIA) and/or local fuel damage due to deteriorated heat transfer. Bubble transport analysis for the ELSY (European Lead-cooled SYstem) reactor design demonstrated the effect of uncertainty in drag correlation, bubble size distributions and leak location on the core and primary system voiding. Possible design solutions for prevention of core voiding in case of SGTL/R are discussed.The effect of seismic isolation (SI) system on the sloshing of the heavy metal coolant and respective mechanical loads on the vessel structures is studied for design basis and beyond design basis earthquakes. It was found that SI shifts the frequencies closer to resonance conditions for sloshing, thus increasing the loads due to slamming of coolant waves onto the vessel internal structures including reactor lid. Possible mitigation measures, such as partitioning baffles have been proposed and their effectiveness was analyzed. Coolant solidification is another serious safety issue in liquid metal cooled reactors that can cause partial or even full blockage of the coolant flow path and thermo-mechanical loads on the primary system structures. Validation of melting/solidification models implemented in CFD is a necessary pre-requisite for its application to risk analysis. In this work, the parametric analysis in support of the design of a solidification test section (STS) in TALL3D facility is carried out. The goals and requirements for the validation experiment and how they are satisfied in the design are discussed in the thesis.
|
|
| 10. |
|
|
