scholarly journals CADOR “Core with Adding DOppleR effect” concept application to sodium fast reactors

2019 ◽  
Vol 5 ◽  
pp. 1
Author(s):  
Alain Zaetta ◽  
Bruno Fontaine ◽  
Pierre Sciora ◽  
Romain Lavastre ◽  
Robert Jacqmin ◽  
...  
Keyword(s):  
Mechanical Energy ◽  
Core Size ◽  
Fuel Temperature ◽  
Fast Reactors ◽  
The Core ◽  
Loss Of Coolant ◽  
Neutron Moderator ◽  
Concept Application ◽  
Sodium Fast Reactors ◽  
Reactivity Insertion

Generation-IV sodium fast reactors (SFR) will only become acceptable and accepted if they can safely prevent or accommodate reactivity insertion accidents that could lead to the release of large quantities of mechanical energy, in excess of the reactor containment's capacity. The CADOR approach based on reinforced Doppler reactivity feedback is shown to be an attractive means of effectively preventing such reactivity insertion accidents. The accrued Doppler feedback is achieved by combining two effects: (i) introducing a neutron moderator material in the core so as to soften the neutron spectrum; and (ii) lowering the fuel temperature in nominal conditions so as to increase the margin to fuel melting. This study shows that, by applying this CADOR approach to a Generation-IV oxide-fuelled SFR, the resulting core can be made inherently resistant to reactivity insertion accidents, while also having increased resistance to loss-of-coolant accidents. These preliminary results have to be confirmed and completed to meet multiple safety objectives. In particular, some margin gains have to be found to guarantee against the risk of sodium boiling during unprotected loss of supply power accidents. The main drawback of the CADOR concept is a drastically reduced core power density compared to conventional designs. This has a large impact on core size and other parameters.

scholarly journals Analysis of thermal hydraulic behavior of KONVOI PWR during a design extension condition

2021 ◽  
pp. 4-4
Author(s):  
Salwa Helmy ◽  
Magy Kandil ◽  
Ahmed Refaey
Keyword(s):  
Power Plants ◽  
Pressurized Water Reactor ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Water Reactor ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Hydraulic Behaviour ◽  
Loss Of Coolant ◽  
Safety Systems

In Nuclear Power Plants the Design Extension Conditions are more complex and severe than those postulated as Design Basis Accidents, therefore, they must be taken into account in the safety analyses. In this study, many hypothetical investigated transients are applied on KONVOI pressurized water reactor during a 6-in. (182 cm2) cold leg Small Break Loss-of-Coolant-Accident to revise the effects of all safety systems ways through their availability/ nonavailability on the thermal hydraulic behaviour of the reactor. The investigated transients are represented through three cases of Small Break Loss-of-Coolant-Accident as, case-1, without scram and all of the safety systems are failure, case-2, the normal scram actuation with failure of all safety systems (nonavailability), and finally case 3, with normal actuation scram sequence and normal sequential actuation of all safety systems (availability). These three investigated transient cases are simulated by creation a model using Analysis of Thermal-Hydraulics of LEaks and Transient code. In all transient cases, all types of reactivity feedbacks, boron, moderator density, moderator temperature and fuel temperature are considered. The steady-state results are nearly in agreement with the plant parameters available in previous literatures. The results show the importance effects of the feedbacks reactivity at Loss-of-Coolant-Accident on the fallouts power, since they are considered the key parameters for controlling the clad and fuel temperatures to maintain them below their melting point. Moreover, the calculated results in all cases show that the thermal hydraulic parameters are in acceptable ranges and encounter the safety criterion during Loss-of-Coolant-Accident the Design Extension Conditions accidents processes. Furthermore, the results show that the core uncovers and fuel heat up do not occur in KONVOI pressurized water reactor in theses the Design Extension Conditions simulations since, all safety systems provide adequate core cooling by sufficient water inventory into the core to cover it.

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

2020 ◽  
Vol 7 (3) ◽  
Author(s):  
Andrea Bachrata ◽  
Fréderic Bertrand ◽  
Nathalie Marie ◽  
Fréderic Serre
Keyword(s):  
Mitigation Strategies ◽  
Severe Accident ◽  
Pressurized Water Reactors ◽  
Physical Parameters ◽  
Fast Reactors ◽  
Pressurized Water ◽  
The Core ◽  
Sodium Fast Reactors ◽  
Water Reactors ◽  
Accident Scenarios

Abstract 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). First, 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. This 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.

A Study on Doppler Weighting Factor for Control Element Assembly Ejection Accident by Using Newly Developed Nuclear Design Code and Non-LOCA Methodology

2013 ◽  
Author(s):  
Kyungmin Yoon ◽  
Chansu Jang ◽  
Jooil Yoon
Keyword(s):  
Safety Analysis ◽  
Weighting Factor ◽  
Time Effect ◽  
Kinetics Model ◽  
Control Element ◽  
Fuel Temperature ◽  
Design Code ◽  
The Core ◽  
Point Kinetics ◽  
Reactivity Insertion

Among Reactivity Initiated Accidents (RIAs) for Pressurized Water Reactor (PWR), Control Element Assembly Ejection (CEAE) accident causes the rapid positive reactivity insertion to the core. It causes an asymmetric power distortion which results in the rising of local fuel temperature, fuel pellet thermal expansion and cladding ballooning or rupture. In the CEAE accident, Doppler feedback has a profound effect because the negative reactivity insertion due to the rise of fuel temperature reduces the core power after rapid power excursion. But the Doppler reactivity can’t be calculated properly in the safety analysis code, using point kinetics model, because the point kinetics model is not able to consider spatial-time effect of the sudden rise in local fuel temperature on Doppler feedback calculation during CEAE accident. And then the excessively high core power which results from the underestimated Doppler feedback would make more severe results such as PCMI fuel failure, fuel cladding rupture and serious DNB fuel failure. Therefore, Doppler Weighting Factor (DWF) is needed for the safety analysis of CEAE accident to compensate a missing spatial-time effect on Doppler feedback calculation. In this study, the adequacy of the application of DWF for APR1400 was evaluated by using nuclear design code called ASTRA (Advanced Static and Transient Reactor Analyzer)[1] and a methodology called ISAM (Integrated Safety Analysis Methodology)[2]. ASTRA is the 3D nuclear design code newly developed by KNF and has various functions such as the static core design, the transient core analysis and the operational support. ISAM is the methodology which is newly developed by KNF to perform the Non-LOCA safety analysis by using RETRAN[3] code which is widely used in the transient analysis and based on the point kinetics model.

scholarly journals Geometrical Scaling of Antiresonant Hollow-Core Fibers for Mid-Infrared Beam Delivery

Crystals ◽  
2021 ◽  
Vol 11 (4) ◽  
pp. 420
Author(s):  
Ang Deng ◽  
Wonkeun Chang
Keyword(s):  
Structural Parameters ◽  
Transmission Band ◽  
Confinement Loss ◽  
Hollow Core ◽  
Core Size ◽  
Core Diameter ◽  
Mid Infrared ◽  
The Core ◽  
Tubular Type ◽  
Beam Delivery

We numerically investigate the effect of scaling two key structural parameters in antiresonant hollow-core fibers—dielectric wall thickness of the cladding elements and core size—in view of low-loss mid-infrared beam delivery. We demonstrate that there exists an additional resonance-like loss peak in the long-wavelength limit of the first transmission band in antiresonant hollow-core fibers. We also find that the confinement loss in tubular-type hollow-core fibers depends strongly on the core size, where the degree of the dependence varies with the cladding tube size. The loss scales with the core diameter to the power of approximately −5.4 for commonly used tubular-type hollow-core fiber designs.

scholarly journals The hourglass effect in hierarchical dependency networks

2017 ◽  
Vol 5 (4) ◽  
pp. 490-528 ◽  
Author(s):  
KAESER M. SABRIN ◽  
CONSTANTINE DOVROLIS
Keyword(s):  
Low Complexity ◽  
Core Size ◽  
Original Network ◽  
Hierarchical Network ◽  
Modular Systems ◽  
The Core ◽  
Dependency Networks ◽  
Dependency Network ◽  
Hierarchical Dependency ◽  
Almost All

AbstractMany hierarchically modular systems are structured in a way that resembles an hourglass. This “hourglass effect” means that the system generates many outputs from many inputs through a relatively small number of intermediate modules that are critical for the operation of the entire system, referred to as the waist of the hourglass. We investigate the hourglass effect in general, not necessarily layered, hierarchical dependency networks. Our analysis focuses on the number of source-to-target dependency paths that traverse each vertex, and it identifies the core of a dependency network as the smallest set of vertices that collectively cover almost all dependency paths. We then examine if a given network exhibits the hourglass property or not, comparing its core size with a “flat” (i.e., non-hierarchical) network that preserves the source dependencies of each target in the original network. As a possible explanation for the hourglass effect, we propose the Reuse Preference model that captures the bias of new modules to reuse intermediate modules of similar complexity instead of connecting directly to sources or low complexity modules. We have applied the proposed framework in a diverse set of dependency networks from technological, natural, and information systems, showing that all these networks exhibit the general hourglass property but to a varying degree and with different waist characteristics.

scholarly journals Analysis of mass transfer processes in a reactor during a loss-of-coolant accident

2018 ◽  
Vol 4 (2) ◽  
pp. 149-154
Author(s):  
Aleksey Kulikov ◽  
Andrey Lepyokhin ◽  
Vitaly Polunichev
Keyword(s):  
Time Interval ◽  
Initial Water ◽  
Pressurized Water ◽  
The Core ◽  
Maximum Time ◽  
Loss Of Coolant Accident ◽  
Safe State ◽  
Loss Of Coolant ◽  
Water Gas ◽  
Gas Ratio

The purpose of the work was to optimize the parameters of the spillage system equipped with a gas pressure hydroaccumulator for a ship pressurized water reactor in a loss-of-coolant accident. The water-gas ratio in the hydroaccumulator and the hydraulic resistance of the path between the hydroaccumulator and the reactor were optimized at the designed hydroaccumulator geometric volume. The main dynamic processes were described using a mathematical model and a computational analysis. A series of numerical calculations were realized to simulate the behavior dynamics of the coolant level in the reactor during the accident – by varying the optimized parameters. Estimates of the minimum and maximum values of the coolant level were obtained: depending on the initial water-gas ratio in the hydroaccumulator at different diameters of the flow restrictor on the path between the hydroaccumulator and the reactor. These results were obtained subject to the restrictive conditions that, during spillage, the coolant level should remain above the core and below the blowdown nozzle. The first condition implies that the core is in safe state, the second excludes the coolant water blowdown. The optimization goal was to achieve the maximum time interval in which these conditions would be satisfied simultaneously. The authors propose methods for selecting the optimal spillage system parameters; these methods provide the maximum time for the core to be in a safe state during a loss-of-coolant accident at the designed hydroaccumulator volume. Using these methods, it is also possible to make assessments from the early stages of designing reactor plants.

scholarly journals PARAMETRIC STUDY OF LOCA IN TRIGA-2000 USING RELAP5/SCDAP CODE

2017 ◽  
Vol 19 (2) ◽  
pp. 59 ◽  
Author(s):  
Anhar Riza Antariksawan ◽  
Surip Widodo ◽  
Hendro Tjahjono
Keyword(s):  
Cooling System ◽  
Fuel Temperature ◽  
Blow Down ◽  
Beam Tube ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling ◽  
Dry Out

A postulated loss of coolant accident (LOCA) shall be analyzed to assure the safety of a research reactor. The analysis of such accident could be performed using best estimate thermal-hydraulic codes, such as RELAP5. This study focuses on analysis of LOCA in TRIGA-2000 due to pipe and beam tube break. The objective is to understand the effect of break size and the actuating time of the emergency core cooling system (ECCS) on the accident consequences and to assess the safety of the reactor. The analysis is performed using RELAP/SCDAPSIM codes. Three different break size and actuating time were studied. The results confirmed that the larger break size, the faster coolant blow down. But, the siphon break holes could prevent the core from risk of dry out due to siphoning effect in case of pipe break. In case of beam tube rupture, the ECCS is able to delay the fuel temperature increased where the late actuation of the ECCS could delay longer. It could be concluded that the safety of the reactor is kept during LOCA throughout the duration time studied. However, to assure the integrity of the fuel for the long term, the cooling system after ECCS last should be considered.  Keywords: safety analysis, LOCA, TRIGA, RELAP5 STUDI PARAMETRIK LOCA DI TRIGA-2000 MENGGUNAKAN RELAP5/SCDAP. Kecelakaan kehilangan air pendingin (LOCA) harus dianalisis untuk menjamin keselamatan suatu reaktor riset. Analisis LOCA dapat dilakukan menggunakan perhitungan best-estimate seperti RELAP5. Penelitian ini menekankan pada analisis LOCA di TRIGA-2000 akibat pecahnya pipa dan tabung berkas. Tujuan penelitian adalah memahami efek ukuran kebocoran dan waktu aktuasi sistem pendingin teras darurat (ECCS) pada sekuensi kejadian dan mengkaji keselamatan reaktor. Analisis dilakukan menggunakan program perhitungan RELAP/SCDAPSIM. Tiga ukuran kebocoran dan waktu aktuasi ECCS berbeda dipilih sebagai parameter dalam studi ini.  Hasil perhitungan mengonfirmasi bahwa semakin besar ukuran kebocoran, semakin cepat pengosongan tangki reaktor. Lubang siphon breaker dapat mencegah air terkuras dalam hal kebocoran pada pipa. Sedang dalam hal kebocoran pada beam tube, ECCS mampu memperlambat kenaikan temperatur bahan bakar. Dari studi ini dapat disimpulkan bahwa keselamatan reaktor dapat terjaga pada kejadian LOCA, namun pendinginan jangka panjang perlu dipertimbangkan untuk menjaga integritas bahan bakar.Kata kunci: analisis keselamatan, LOCA, TRIGA, RELAP5

scholarly journals All-fiber passively Q-switched ytterbium doped double-clad fiber lasers: experiment And modeling.

2021 ◽  
Author(s):  
Yi Lu
Keyword(s):  
Pulse Energy ◽  
Traveling Wave ◽  
Pulse Width ◽  
Fiber Lasers ◽  
Peak Power ◽  
Wave Model ◽  
Core Size ◽  
The Core ◽  
Q Switching ◽  
Output Characteristics

All-fiber passively Q-switched lasers were demonstrated using ytterbium-doped double cladding fiber (YDF) as an active medium. The laser was pumped by three 25W, 975nm fiber coupled diodes and Q-switching was initiated when the amplified spontaneous emission generated in the core of the gain fiber bleached the saturable absorber (SA). A piece of samarium-doped fiber was used as SA in first configuration and pulses with 68μJ pulse energy and 210ns pulse width were obtained. In second configuration, a piece of ytterbium-doped fiber with much smaller core size was used as SA to produce pulse energy of 86μJ. The last configuration incorporated a 9m-long YDF as gain fiber. The far end from pump was acting as SA in this case and pulses with 82μJ pulse energy and 148ns pulse width were observed. The peak power was estimated at 554W. Traveling wave model was implemented to numerically simulate the output characteristics versus pump power.

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