An Experimental Investigation on Thermal-Hydraulic Interaction Between Core Makeup Tank and Safety Injection Tank of an Integral Small Reactor During a Small Break Loss-of-Coolant Accident

2015 ◽  
Author(s):  
Hwang Bae ◽  
Sung Uk Ryu ◽  
Hyo Bong Ryu ◽  
Woo Shik Kim ◽  
Sung-Jae Yi ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Injection System ◽  
Test Facility ◽  
Passive Safety ◽  
Injection Test ◽  
Loss Of Coolant Accident ◽  
Test Loop ◽  
Loss Of Coolant ◽  
Reactor Pressure

A passive injection test was conducted using a core makeup tank (CMT), a safety injection tank (SIT) and an automatic depressurization system (ADS), which consists of a passive safety system (PSS) of the SMART reactor. This paper investigates the thermal-hydraulic interaction between CMT and SIT during sequential injections of coolant from these two tanks to a high-temperature and high-pressure reactor pressure vessel using an integral effect test facility of SMART-ITL (System-Integrated Modular Advanced ReacTor-Integral Test Loop). Both CMT and SIT were connected to the reactor pressure vessel by a pressure balance line (PBL) and injection line (IL). A steady-state condition was maintained for 1,000 seconds before the start of the injection. The major parameters agreed well with the target value. After one of safety injection system line was simulated to be broken, a transient injection test was conducted according to the small-break loss-of-coolant accident (SBLOCA) scenario. Coolant injections from a CMT and SIT were started sequentially by opening quick-opening valves installed on the IL and PBL piping, respectively. Several thermal-hydraulic phenomena such as direct contact condensation, thermal stratification, and coupling effects between the CMT and SIT were locally observed during the SBLOCA scenario. The results show that the adopted passive safety injection system functions well as an emergency core cooling system.

Safety Assessment of Reactor Pressure Vessel Integrity for Loss of Coolant Accident Conditions

2011 ◽  
Vol 134 (1) ◽  
Author(s):  
Dieter Beukelmann ◽  
Wenfeng Guo ◽  
Wieland Holzer ◽  
Robert Kauer ◽  
Wolfgang Münch ◽  
...  
Keyword(s):  
Finite Element ◽  
Pressure Vessel ◽  
Safety Margin ◽  
Reactor Pressure Vessel ◽  
Test Facility ◽  
Temperature Distributions ◽  
Loss Of Coolant Accident ◽  
System Pressure ◽  
Loss Of Coolant ◽  
Reactor Pressure

One of the critical issues for reactor pressure vessel (RPV) structural integrity is related to the pressurized thermal shock (PTS) event. Therefore, within the framework of safety assessments special emphasis is given to the effect of PTS-loadings caused by the nonuniform azimuthal temperature distribution due to cold water plumes or stripes during emergency coolant injection. This paper describes the method used to predict the thermal mechanic boundary conditions (system pressure and wall temperature). Using a system code the pressure and global temperature distributions were calculated, systematically varying the leak size and the location of the coolant water injection. Spatial and temporal temperature distributions in the main circulation pipes and at the RPV wall were predicted by mixing analyses with a computational fluid dynamics (CFD) code. The model used for these calculations was validated by post-test calculations of a UPTF (upper plenum test facility) experiment simulating cold leg injection during a small break loss of coolant accident (LOCA). Comparison with measured temperatures showed that the modeling used is suitable to obtain enveloping results. Fracture mechanics analyses were carried out for circumferential flaw sizes in the weld joint near the core region and between the RPV shell and the flange, as well as for axial flaws in the nozzle corner. Stress intensity factors KI were calculated numerically using the finite element program ansys and analytically on the basis of weight and polynomial influence functions using stresses obtained from elastic finite element analyses. Benchmark tests revealed good agreement between the results from numerical and analytical calculations. For all regions of the RPV investigated and the most severe transients it was demonstrated that a large safety margin against brittle crack initiation exists and brittle fracture of the RPV can be excluded.

Safety Assessment of Reactor Pressure Vessel Integrity for Loss of Coolant Accident Conditions

2010 ◽  
Author(s):  
D. Beukelmann ◽  
W. Guo ◽  
W. Holzer ◽  
R. Kauer ◽  
W. Mu¨nch ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Finite Element ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Test Facility ◽  
Worst Case ◽  
Temperature Distributions ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Reactor Pressure

One of the critical issues for Reactor Pressure Vessel (RPV) structural integrity is related to the Pressurized Thermal Shock (PTS) event. Therefore, within the framework of safety assessments special emphasis is given to the effect of PTS-loadings caused by the non-uniform azimuthal temperature distribution due to cold water plumes or stripes during emergency coolant injection. The paper describes the method used to predict the thermal mechanic boundary conditions (system pressure, wall temperature). Using a system code the pressure and global temperature distributions were calculated, systematically varying the leak size and the location of the coolant water injection. Local and temporal temperature distributions in the main circulation pipes and at the RPV wall were predicted by mixing analyses with a Computational Fluid Dynamics (CFD) code. The model used for these calculations was validated by post-test calculations of a UPTF (Upper Plenum Test Facility) experiment simulating cold leg injection during a small break Loss of Coolant Accident (LOCA). Comparison with measured temperatures showed that the modelling used is suitable to obtain bounding results. Fracture mechanics analyses were carried out for circumferential flaw sizes in the weld joint near the core region and between the RPV shell and the flange, as well as for axial flaws in the nozzle corner. Stress intensity factors KI were calculated numerically using the finite element program ANSYS and analytically on the basis of weight and polynomial influence functions using stresses obtained from elastic finite element analyses. Benchmark tests revealed good agreement between the results from numerical and analytical calculations. In order to determine the worst case loading conditions a wide spectrum of thermal-hydraulic transients was considered. Since the resulting load paths decrease with lower temperatures after a maximum, the warm prestress (WPS) effect was employed. The fracture toughness curve determined by deeply notched specimens with high constraint is not representative of the nozzle corner due to the considerable loss of constraint at LOCA conditions. Hence the influence of constraint on fracture toughness was accounted applying the constraint modified master curve concept and the relationship between the T-stress and the reference temperature T0. According to ASME Code Cases N-629 and N-631 the reference temperatures RTNDT and RTT0 can be used alternatively for the adjustment of the KIC-curve. Therefore both the RTNDT- and the RTT0-concept were considered. For all regions of the RPV investigated and the most severe transients it was demonstrated that a large safety margin against crack initiation exists and brittle fracture of the RPV can be excluded.

Application of the BEREMIN Model for Evidencing the Warm Pre-Stress Effect in an Irradiated Reactor Pressure Vessel Containing a Semi-Elliptical Subclad Defect Under Small and Large LOCA Conditions

2008 ◽  
Author(s):  
Ph. Gilles ◽  
J.-P. Izard ◽  
J. Devaux
Keyword(s):  
Brittle Fracture ◽  
Pressure Vessel ◽  
Fracture Resistance ◽  
Thermal Fluctuations ◽  
Reactor Pressure Vessel ◽  
Stress Effect ◽  
Load History ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Reactor Pressure

The nuclear power plants lifetime is strongly dependent of the guarantee of the reactor pressure vessel (RPV) integrity. Therefore, the RPV integrity has to be demonstrated under the most severe configuration, namely the Pressurized Thermal Shock induced by the Loss of Coolant Accident induced by a large break in the primary loop. For such a transient, the apparent risk of failure is maximum when the load is decreasing; the fracture resistance decreasing rate being stronger. However, such type of loading generates an increase of the fracture resistance as shown by numerous studies (Chell, 1980 – BEREMIN, 1981 – Smith et al., 2004). This is known as the warm pre-stress (WPS) effect. This beneficial effect on the resistance to brittle fracture is not accounted for in the French RCCM and RSEM codes (RCCM, 2000 – RSEM, 2005). EDF has launched several R&D actions with CEA and AREVA as well as with European partners (SMILE, 2001) to validate and model the WPS effect under RPV representative conditions. Proving the existence of this beneficial load history effect (designated as Warm Pre Stress WPS), in the case of a defective RPV in emergency and faulted conditions is the aim of the present paper. The demonstration is conducted in the case of cleavage fracture using an improved version of the BEREMIN model. As opposite to the classical Fracture Mechanics methodology, this approach allows to account for load history effects on cleavage. The study analyzes the behavior of a semi-elliptical under clad crack in the EoL core shell of a 900 MWe RPV for two loading cases: the large break Loss Of Coolant Accident transient and a small break LOCA inducing thermal fluctuations on the vessel inner wall. The WPS effect is evidenced by comparing the plasticity corrected SIF levels of two loadings for the same value of failure probability: the considered WPS loading and a virtual monotonously increasing load applied at the temperature at which the brittle fracture risk is estimated.

Bounding Acoustic Load Development for Boiling Water Reactor Shroud and Shroud Supports

2020 ◽  
Author(s):  
Matthew Walter ◽  
Minghao Qin ◽  
Daniel Sommerville
Keyword(s):  
Pressure Vessel ◽  
Reactor Core ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Structural Evaluation ◽  
Loss Of Coolant Accident ◽  
Specific Analysis ◽  
Reactor Pressure

Abstract As part of the license basis of a nuclear boiling water reactor pressure vessel, a sudden loss of coolant accident (LOCA) event needs to be analyzed. One of the loads that results from this event is a sudden depressurization of the recirculation line. This leads to an acoustic wave that propagates through the reactor coolant and impacts several structures inside the reactor pressure vessel (RPV). The authors have previously published a PVP paper (PVP2015-45769) which provides a survey of LOCA acoustic loads on boiling water reactor core shrouds. Acoustic loads are required for structural evaluation of core shrouds; therefore, a defensible load is required. The previous research compiled plant-specific data that was available at the time. Since then, additional data has become available which will add to the robustness of the bounding load methodology that was developed. Investigations are also made regarding the shroud support to RPV weld, which was neglected from the previous study. This will allow a practitioner a convenient method to calculate bounding acoustic loads on all shroud and shroud support welds in the absence of a plant-specific analysis.

scholarly journals Integrity Evaluation of a Reactor Pressure Vessel Based on a Sequential Abaqus-FRANC3D Simulation Method

2021 ◽  
Vol 2021 ◽  
pp. 1-12
Author(s):  
M. Annor-Nyarko ◽  
Hong Xia
Keyword(s):  
Finite Element ◽  
Stress Intensity ◽  
Pressure Vessel ◽  
Simulation Method ◽  
Reactor Pressure Vessel ◽  
Injection System ◽  
Intensity Factors ◽  
3D Fem ◽  
Ageing Management ◽  
Reactor Pressure

The safety-risk pressurized thermal shock (PTS) have on a reactor pressure vessel (RPV) is one of the most important studies for the lifetime ageing management of a reactor. Several studies have investigated PTS induced by postulated accidents and other anticipated transients. However, there is no study that analyzes the effect of PTS induced by one of the most frequent anticipated operational occurrences—inadvertent operation of the safety injection system. In this paper, a sequential Abaqus-FRANC3D simulation method is proposed to study the integrity status of an ageing pressurized water reactor subjected to PTS induced by inadvertent actuation of the safety injection system. A sequential thermal-mechanical coupling analysis is first performed using a three-dimensional reactor pressure vessel finite element model (3D-FEM). A linear elastic fracture mechanics submodel with a postulated semielliptical surface crack is then created from the 3D-FEM. Subsequently, the submodel is used to evaluate the stress intensity factors based on the M-integral approach coupled within the proposed simulation method. Finally, the stress intensity factors (SIFs) obtained using the proposed method are compared with the conventional extended finite element method approach, and the result shows a good agreement. The maximal thermomechanical stress concentration was observed at the inlet nozzle-inner wall intersection. In addition, The ASME fracture toughness of the reactor vessel’s steel compared with SIFs show that the PTS event and crack configuration analysed may not pose a risk to the integrity of the RPV. This work serves as a critical reference for the ageing management and fatigue life prediction of reactor pressure vessels.

CFD-Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions: Influence of Turbulence Model and Mesh Refinement on the Vessel Thermal Loading During PTS Transient

2006 ◽  
Author(s):  
A. Martin ◽  
S. Benhamadouche ◽  
G. Bezdikian ◽  
F. Beaud ◽  
F. Lestang
Keyword(s):  
Heat Transfer ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Fluid Temperature ◽  
Hydraulic Analysis ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Thermal Hydraulic Analysis ◽  
Reactor Pressure

Integrity evaluation methods for nuclear Reactor Pressure Vessels (RPVs) under Pressurised Thermal Shock (PTS) loading are applied by French Utility. They are based on the analysis of the behavior of relatively shallow cracks under loading PTS conditions due to the emergency cooling during SBLOCA (Small Break Loss of Coolant Accident) transients. This paper presents the Research and Development program started at E.D.F on the Computational Fluid Dynamic (CFD) determination of the cooling phenomena of a PWR vessel during a Pressurised Thermal Shock. The numerical results are obtained with the thermal-hydraulic tool Code_Saturne, in combination with the thermal-solid code SYRTHES to take into account the coupled effect of heat transfer between the fluid flow and the vessel. Based on the global and local Thermal-hydraulic analysis of a Small Break Loss of Coolant Accident transient, the paper presents mainly a parametric study which helps to understand the main phenomena that can lead to better estimating the margin factors. The geometry studied represents a third of a PWR pressure vessel and the configuration investigated is related to the injection of cold water in the vessel during a SBLOCA transient. Conservative initial and boundary conditions for the CFD calculation are derived from the global Thermal-hydraulic analysis. Both the fluid behavior and its impact on the solid part formed by cladding and base metal are considered. The main purpose of the numerical thermal-hydraulic studies is to accurately estimate the distribution of fluid temperature in the down comer and the heat transfer coefficients on the inner RPV surface for a fracture mechanics computation which will subsequently assess the associated RPV safety margin factors.

Buoyancy Driven Coolant Mixing Studies of Natural Circulation Flows at the ROCOM Test Facility Using ANSYS CFX

2006 ◽  
Author(s):  
Thomas Ho¨hne ◽  
So¨ren Kliem ◽  
Ulrich Rohde ◽  
Frank-Peter Weiß
Keyword(s):  
Pressure Vessel ◽  
Natural Circulation ◽  
Reactor Pressure Vessel ◽  
Density Difference ◽  
Test Facility ◽  
Mean Flow ◽  
Pressurized Water ◽  
The Core ◽  
Ansys Cfx ◽  
Reactor Pressure

Coolant mixing in the cold leg, downcomer and the lower plenum of pressurized water reactors is an important phenomenon mitigating the reactivity insertion into the core. Therefore, mixing of the de-borated slugs with the ambient coolant in the reactor pressure vessel was investigated at the four loop 1:5 scaled ROCOM mixing test facility. Thermal hydraulics analyses showed, that weakly borated condensate can accumulate in particular in the pump loop seal of those loops, which do not receive safety injection. After refilling of the primary circuit, natural circulation in the stagnant loops can re-establish simultaneously and the de-borated slugs are shifted towards the reactor pressure vessel (RPV). In the ROCOM experiments, the length of the flow ramp and the initial density difference between the slugs and the ambient coolant was varied. From the test matrix experiments with 0 resp. 2% density difference between the de-borated slugs and the ambient coolant were used to validate the CFD software ANSYS CFX. To model the effects of turbulence on the mean flow a higher order Reynolds stress turbulence model was employed and a mesh consisting of 6.4 million hybrid elements was utilized. Only the experiments and CFD calculations with modeled density differences show a stratification in the downcomer. Depending on the degree of density differences the less dense slugs flow around the core barrel at the top of the downcomer. At the opposite side the lower borated coolant is entrained by the colder safety injection water and transported to the core. The validation proves that ANSYS CFX is able to simulate appropriately the flow field and mixing effects of coolant with different densities.

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