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.

Development and Application of Multi-Physics Safety Analysis Code for Advanced Liquid Metal Cooled Reactor

2018 ◽  
Author(s):  
Chi Wang ◽  
Xuebei Zhang ◽  
Jingchao Feng ◽  
Muhammad Shehzad Khan ◽  
Minyou Ye ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Safety Analysis ◽  
Coolant Temperature ◽  
Medium Size ◽  
High Tech ◽  
Fuel Temperature ◽  
Flow Phenomena ◽  
Nuclear Reactor Safety ◽  
Cfd Method ◽  
Point Kinetics

The simulation of 3D thermal-hydraulic problem for the pool type fast reactors, is one of the necessary and great importance. Most system codes can’t be used to simulate multi-dimensional thermal-hydraulics problems, whereas, the CFD method is suitable to deal with these type of simulation challenges. Based on the CFD method, a neutronics and thermohydraulic coupling code FLUENT/PK for nuclear reactor safety analysis by coupling the commercial CFD code FLUENT with the point kinetics model (PKM) and the pin thermal model (PTM) is developed by University of Science and Technology of China (USTC). The coupled code is verified by comparing with a series of benchmarks on beam interruptions in a lead-bismuth-cooled and MOX-fuelled accelerator-driven system. The variations of transient power, fuel temperature and outlet coolant temperature all agree well with the benchmark results. The validation results show that the code can be used to simulate the transient accidents of critical and sub-critical lead/lead-bismuth cooled reactors. Then this coupling code is used to evaluate the safety performance of MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) at unprotected beam over-power (UBOP) accident, and M2LFR-1000 (Medium-size Modular Lead-cooled Fast Reactor) at the unprotected transient over-power (UTOP) and unprotected loss of flow (ULOF) accident. The transient power, the temperature of coolant and fuel and multi-dimensional flow phenomena in upper plenum and lower plenum are presented and discussed in this paper.

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 THE LiCore POWER PLANT SIMULATOR OF THE MOLTEN SALT FAST REACTOR

2021 ◽  
Vol 247 ◽  
pp. 06030
Author(s):  
A. Laureau ◽  
E. Rosier ◽  
E. Merle ◽  
S. Beils ◽  
O. Bruneau ◽  
...  
Keyword(s):  
Power Plant ◽  
Molten Salt ◽  
Basic Principle ◽  
Kinetics Model ◽  
Simple Point ◽  
Delayed Neutron ◽  
Integrated Simulation ◽  
The Core ◽  
Definition Of ◽  
Point Kinetics

Molten salt reactors as liquid-fuelled reactors are flexible in terms of operation or design choices, but they are very different in terms of design, operation and safety approach compared to solid-fuelled reactors. Such reactors call for a new definition of their operating procedures and safety approach. Dedicated developments and studies have been performed in the frame of the European SAMOFAR project of Horizon2020 and in parallel in France involving CNRS, CORYS and Framatome to develop a system code called LiCore adapted to such reactors, corresponding to a basic-principle power plant simulator. The neutronic model LiCore, at the centre of the simulator, corresponds to an improved point-kinetics model to take into account the specificities of a MSR, notably the circulation of the delayed neutron precursors out of the core. Coupled to a simple piston model for the fuel motion in the core, this code can perform calculations faster than real time to simulate the behaviour of the fuel circuit. Transient calculations performed with LiCore are presented, together with comparisons first to a simple point-kinetics model and then to 3D calculations with the TFM-OpenFOAM coupled code. Finally, the LiCore code has recently been integrated in the ALICES platform, the integrated simulation toolset designed by CORYS for the development, maintenance and operation of major simulator such as power plant simulators.

Inaccurate Use of Boron Concentration for Reactivity Feedback

2002 ◽  
Author(s):  
Birol Aktas ◽  
Sule Ergun
Keyword(s):  
Boron Concentration ◽  
Solute Concentration ◽  
Kinetics Model ◽  
The Core ◽  
The Us ◽  
Point Kinetics

A deficiency has been identified with the use of boron concentration for formulating reactivity feedback while the power of a PWR in response to boron injection in the aftermath of a scram failure was simulated. The US NRC Consolidated Code (TRAC-M) was used to simulate this transient, in which the boron injection was expected to shutdown the reactor. The results of this study, which employed the point-kinetics model of TRAC-M, reveal that the use of core-average boron concentration for formulating the reactivity feedback is inadequate for transients when there is substantial coolant void inside the core. It is recommended that the macroscopic boron density be used for more accurate predictions of the boron feedback as the use of solute concentration is only adequate when no voiding occurs in the core.

VHTR Core Preliminary Analysis Using NEPHTIS3 / CAST3M Coupled Modelling

2008 ◽  
Author(s):  
Fre´de´ric Damian
Keyword(s):  
Fuel Element ◽  
System Analysis ◽  
Hot Spot ◽  
Operating Conditions ◽  
Maximum Temperature ◽  
Design Parameters ◽  
Block Type ◽  
Fuel Temperature ◽  
Coupled Modelling ◽  
The Core

Along with the GFR another gas-cooled reactor identified in the Gen IV technology roadmap, the VHTR is studied in France. Some models have been developed at CEA relying on existing computational tools essentially dedicated to the prismatic block type reactor. These models simulate normal operating conditions and accidental reactor transients by using neutronic [1], thermal-hydraulic, system analysis codes [2], and their coupling [3, 4]. In the framework of the European RAPHAEL project, this paper presents the results of the preliminary investigations carried out on the VHTR design. These studies aimed at understanding the physical aspects of the annular core and to identify the limits of a standard block type VHTR with regard to a degradation of its passive safety features. Analysis was performed considering various geometrical scales: fuel cell and fuel column located at the core hot spot, 2D and 3D core configurations including the coupling between neutronic and thermal-hydraulic. From the thermal analysis performed at the core hot spot, the capability to reduce the maximum fuel temperature by modifying the design parameters such as the fuel compact and the fuel block geometry was assessed. The best performances are obtained for an annular fuel compact geometry with coolant flowing inside and outside the fuel compact (ΔT > 50°C). The reliability of such design option should however be addressed with respect to its performance during the LOFC transient (the residual decay heat will be evacuated by radiation during the transient instead of conduction through graphite). As far as the fuel element geometry is concerned, a gain of approximately 50°C can be achieved by making limited changes on the fuel compact distribution in the prismatic block: reduction of the number of fuel compact in the outer ring of the fuel element where the average ratio between coolant channels and fuel compact is smaller. On the other hand, the adopted modifications should also be evaluated with respect to the maximum temperature gradient achieved in the fuel (amoeba effect). In the end, calculations performed on the full core configuration taking into account the thermal feedback showed that the radial positioning of the fuel elements allows to reduce significantly the power peaking factor and the maximum fuel temperature. The gain on the fuel temperature, which varies during the core irradiation, is in the range 100 – 150°C. Several modifications such as the increase of the bypass fraction and the replacement of a part of the graphite reflector by material with better thermal properties were also addressed in this paper.

Parallel Computation of the Point Neutron Kinetic Equations Using Parallel Revisionist Integral Deferred Correction

2018 ◽  
Author(s):  
Yun Cai ◽  
Xingjie Peng ◽  
Qing Li ◽  
Zhizhu Zhang ◽  
Zhumin Jiang ◽  
...  
Keyword(s):  
Time Integration ◽  
Kinetic Equations ◽  
Accurate Method ◽  
Small Time ◽  
Time Step ◽  
Deferred Correction ◽  
Time Integration Method ◽  
Speed Up ◽  
Point Kinetics ◽  
Reactivity Insertion

The point kinetics is very important to the safety of the reactor operation. However, these equations are stiff and usually solved with very small time step. These equations are solved by Revisionist integral deferred correction (RIDC), which is a parallel time integration method. RIDC is a highly accurate method, and it reduces the error by iteration. Based on C++ and MPI, a four-core fourth-order RIDC is implemented and tested by several cases, such as step, ramp, and sinusoidal reactivity insertion. Compared with other methods, the time step of RIDC in the step reactivity insertion case is smaller, but it’s larger in the case of the sinusoidal reactivity insertion. RIDC can keep high accuracy while the time step is appropriately large. The numerical results also show that the speed-up ratio can achieve 2 when 4 processors are used.

scholarly journals THERMAL-HYDRAULIC ANALYSIS OF IRT-4M IN REACTIVITY INSERTION ACCIDENT AT VR-1 REACTOR

2016 ◽  
Vol 4 ◽  
pp. 22
Author(s):  
Filip Fejt
Keyword(s):  
Atmospheric Pressure ◽  
Room Temperature ◽  
Nuclear System ◽  
Hydraulic Analysis ◽  
Research Reactors ◽  
The Core ◽  
Fuel Assemblies ◽  
Coolant Circulation ◽  
Thermal Hydraulic Analysis ◽  
Reactivity Insertion

The paper deals with thermal-hydraulic analysis during reactivity insertion accident, i.e. a step increase of nuclear system reactivity by 0.7 eff, at VR-1 Reactor. The reactor utilizes IRT-4M type of fuel assemblies, and even though these fuel assemblies are designed for an operation at the high-power research reactors, they might be also used for zero-power reactors. The thermal-hydraulic analyses must take into account several specific assumptions that are derived from VR-1 reactor specifications. The reactor does not require a forced water flow for a fuel cooling, the core is placed in an open vessel with atmospheric pressure, and amount of coolant water in the vessel is sufficient for providing the inlet water at room temperature for the whole event. Coolant circulation is expected to be formed only by natural convection.

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