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.

Fluid-Structure Coupling Analysis for the Flow Distribution Device of Nuclear Reactor Internals

2013 ◽  
Author(s):  
Mingqian Zhang ◽  
Yuangang Duan ◽  
Xiaobing Ran ◽  
Yanwu Liu
Keyword(s):  
Fluid Flow ◽  
Flow Field ◽  
Pressure Vessel ◽  
Flow Distribution ◽  
Reactor Pressure Vessel ◽  
Turbulent Model ◽  
Pressurized Water ◽  
Fluid Structure ◽  
Ansys Cfx ◽  
Reactor Pressure

In order to better understand the stresses to which the flow distribution device (FDD) is subjected for the pressure fluctuation, we need to improve our knowledge of the fluid flow inside the reactor pressure vessel (RPV). The flow field of the reactor lower plenum which is associated with a typical pressurized water reactor (PWR) is simulated by using ANSYS CFX code. Calculations have been carried out from reactor pressure vessel inlet to the core outlet. Grid sizes of million nodes with the k-epsilon turbulent model have been used with a porous zone approach for the reactor core space. Predictions of the steady-state pressure and velocity field have been made. The results are compared with the scaled experiment data in order to verify the accurate description of the fluid flow. Based on this verified turbulent model, a sub-domain is extracted from the lower plenum for the transient two-way Fluid-Structure Interaction (FSI) simulation which is limited by the computer capability and computing time. This transient analysis of fluid-structure coupling system is conducted by using CFX and ANSYS in numeric calculation of flow field and structure, with an exchanging platform MFX-ANSYS/CFX which can transfer fluid pressure and structure displacement between computational fluid dynamics (CFD) and computational structure dynamics (CSD) grid systems. The loose coupling method is used to investigate the transient dynamic response of the flow distribution device which is immerged in the bottom plenum. Dynamic stress and strain of the flow distribution device are given and discussed. This analysis practice can be guidance for the optimization design of reactor and improve our understanding of reactor components flow induced vibration phenomena.

Analysis of a Concept for the Diagnosis of the Core State During Severe Accidents in Pressurized Water Reactors

2014 ◽  
Author(s):  
Sebastian Schmidt ◽  
Daniel Fiß ◽  
Alexander Kratzsch
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Core State ◽  
Process Technology ◽  
Pressurized Water ◽  
Non Invasive ◽  
The Core ◽  
Gamma Radiations ◽  
State Diagnosis ◽  
Reactor Pressure

In line with the cooperative project “Non-invasive Condition Monitoring of Nuclear Reactors for the Detection of Level Change and Deformation of the Core” between the Technical University Dresden and the Institute of Process Technology, Process Automation and Measuring Technology (IPM) of Zittau/Goerlitz University of Applied Sciences, a measuring system for the core state diagnosis during a core melt accident in the reactor pressure vessel of a pressurized water reactor is going to be developed. The operational principle of this system is based on the non-invasive measurement of continuously changing gamma radiations (caused by the shifting of melted materials and fission products) outside of the reactor pressure vessel by means of several gamma radiation sensors. The sensors are arranged over the height of the core and the lower head. By using computer based and real-time capable methods for evaluation of the measured gamma radiations conclusions about the core state can subsequently be drawn. This paper includes a description of a concept as well as several methods for the core state diagnosis during a core melt accident. The main part is the analysis of the methods for the core state diagnosis based on different evaluation criteria.

Supplemental Stress and Fracture Mechanics Analyses of Pressurized Water Reactor Pressure Vessel Nozzles

2012 ◽  
Author(s):  
Matthew Walter ◽  
Shengjun Yin ◽  
Gary L. Stevens ◽  
Daniel Sommerville ◽  
Nathan Palm ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Additional Stress ◽  
Corner Crack ◽  
Pressurized Water ◽  
The Core ◽  
Water Reactors ◽  
Reactor Pressure

In past years, the authors have undertaken various studies of nozzles in both boiling water reactors (BWRs) and pressurized water reactors (PWRs) located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Those studies described stress and fracture mechanics analyses performed to assess various RPV nozzle geometries, which were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-life (EOL) to require evaluation of embrittlement as part of the RPV analyses associated with pressure-temperature (P-T) limits. In this paper, additional stress and fracture analyses are summarized that were performed for additional PWR nozzles with the following objectives: • To expand the population of PWR nozzle configurations evaluated, which was limited in the previous work to just two nozzles (one inlet and one outlet nozzle). • To model and understand differences in stress results obtained for an internal pressure load case using a two-dimensional (2-D) axi-symmetric finite element model (FEM) vs. a three-dimensional (3-D) FEM for these PWR nozzles. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated. • To investigate the applicability of previously recommended linear elastic fracture mechanics (LEFM) hand solutions for calculating the Mode I stress intensity factor for a postulated nozzle corner crack for pressure loading for these PWR nozzles. These analyses were performed to further expand earlier work completed to support potential revision and refinement of Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G, “Fracture Toughness Requirements,” and are intended to supplement similar evaluation of nozzles presented at the 2008, 2009, and 2011 Pressure Vessels and Piping (PVP) Conferences. This work is also relevant to the ongoing efforts of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, Working Group on Operating Plant Criteria (WGOPC) efforts to incorporate nozzle fracture mechanics solutions into a revision to ASME B&PV Code, Section XI, Nonmandatory Appendix G.

Experimental Investigation on Boiling Flow Characteristics Under Passive IVR-ERVC Conditions

2018 ◽  
Author(s):  
Fan Wang ◽  
Bo Kuang ◽  
Pengfei Liu ◽  
Longkun He
Keyword(s):  
Pressure Vessel ◽  
Natural Circulation ◽  
Flow Characteristics ◽  
Reactor Vessel ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Test Facility ◽  
Circulation Flow ◽  
Boiling Flow ◽  
Reactor Pressure

In vessel retention (IVR) of molten core debris via water cooling at the external surface of the reactor vessel is an important severe accident management feature of advanced passive plants. During postulated severe accidents, the heat generated due to the molten debris relocation to the lower reactor pressure vessel head needs to be removed continuously to prevent vessel failure. Besides the local critical heat flux (CHF) of outer wall surface which is the first importance, a stable feature of natural circulation flow and an effective natural circulation capability within the external reactor vessel cooling (ERVC) channel tend to be rather crucial for the success of IVR. Under this circumstance, a full-height ERVC test infrastructure for large advanced pressurized water reactor (PWR) IVR strategy engineering validation, namely reactor pressure vessel external cooling II test facility (REPEC-II), has been designed and constructed in Shanghai Jiao Tong University (SJTU). And therefore, a brief introduction to the SJTU REPEC II facility as well as the experimental progress to date, is hereby given in the paper. During test campaign on the REPEC II facility, the one-dimensional natural circulation boiling flow characteristics during IVR-ERVC severe accident mitigation are investigated, with the experimental observation and measurement on natural circulation flow characteristics within the REPEC II test facility. Based on the abundant results acquired in the test campaign, it is attempted, in this paper, to summarize and evaluate the ERVC performances and trends under various practical engineered conditions. The main evaluation results includes: influence on ERVC flow characteristics of various non-uniform heat load distribution cooling limits, the observed sinusoidal oscillation is suggested to be flashing-induced density wave oscillations and the oscillation period correlated well with the passing time of single-phase liquid in the riser. It is expected that these conclusions may help designers to have a reliable estimate of the impact of some related engineered factors on real IVR-ERVC performance.

scholarly journals Analysis of the Technological Evolution of Materials Requirements Included in Reactor Pressure Vessel Manufacturing Codes

2021 ◽  
Vol 13 (10) ◽  
pp. 5498
Author(s):  
Alvaro Rodríguez-Prieto ◽  
Mariaenrica Frigione ◽  
John Kickhofel ◽  
Ana M. Camacho
Keyword(s):  
Sustainable Development ◽  
Pressure Vessel ◽  
Nuclear Regulatory Commission ◽  
Green Energy ◽  
Reactor Pressure Vessel ◽  
Technical Literature ◽  
Pressurized Water ◽  
Water Reactors ◽  
Regulatory Commission ◽  
Reactor Pressure

The growth of green energy technologies within the frame of the 7th Sustainable Development Goal (SDG) along with the concern about climatic changes make nuclear energy an attractive choice for many countries to ensure energy security and sustainable development as well as to actively address environmental issues. Unlike nuclear equipment (immovable goods), which are often well-catalogued and analyzed, the design and manufacturing codes and their standardized materials specifications can be considered movable and intangible goods that have not been thoroughly studied based on a detailed evaluation of the scientific and technical literature on the reactor pressure vessel (RPV) materials behavior. The aim of this work is the analysis of historical advances in materials properties research and associated standardized design codes requirements. The analysis, based on the consolidated U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.99 Rev.2 model, enables determination of the best materials options, corresponding to some of the most widely used material specifications such as WWER 15Kh2MFAA (used from the 1970s and 1980s; already in operation), ASME SA-533 Grade B Cl.1 (used in pressurized water reactor-PWR 2nd–4th; already in operation), DIN 20MnMoNi55 and DIN 22NiMoCr37 (used in PWR 2nd–4th) as well as ASTM A-336 Grade F22V (current designs). Consequently, in view of the results obtained, it can be concluded that the best options correspond to recently developed or well-established specifications used in the design of pressurized water reactors. These assessments endorse the fact that nuclear technology is continually improving, with safety being its fundamental pillar. In the future, further research related to the technical heritage from the evolution of materials requirements for other clean and sustainable power generation technologies will be performed.

Reactor Pressure Vessel Integrity Assessment: French Simplified Analysis Method Applied to Underclad Postulated Defect in Nozzle

2016 ◽  
Author(s):  
Adolfo Arrieta-Ruiz ◽  
Eric Meister ◽  
Stéphane Vidard
Keyword(s):  
Fracture Mechanics ◽  
Fracture Risk ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Core Area ◽  
Integrity Assessment ◽  
Elastic Plastic ◽  
The Core ◽  
Simplified Method ◽  
Reactor Pressure

Structural integrity of the Reactor Pressure Vessel (RPV) is one of the main concerns regarding safety and lifetime of Nuclear Power Plants (NPP) since this component is considered as not reasonably replaceable. Fast fracture risk is the main potential damage considered in the integrity assessment of RPV. In France, deterministic integrity assessment for RPV vis-à-vis the brittle fracture risk is based on the crack initiation stage. As regards the core area in particular, the stability of an under-clad postulated flaw is currently evaluated under a Pressurized Thermal Shock (PTS) through a dedicated fracture mechanics simplified method called “beta method”. However, flaw stability analyses are also carried-out in several other areas of the RPV. Thence-forward performing uniform simplified inservice analyses of flaw stability is a major concern for EDF. In this context, 3D finite element elastic-plastic calculations with flaw modelling in the nozzle have been carried out recently and the corresponding results have been compared to those provided by the beta method, codified in the French RSE-M code for under-clad defects in the core area, in the most severe events. The purpose of this work is to validate the employment of the core area fracture mechanics simplified method as a conservative approach for the under-clad postulated flaw stability assessment in the complex geometry of the nozzle. This paper presents both simplified and 3D modelling flaw stability evaluation methods and the corresponding results obtained by running a PTS event. It shows that the employment of the “beta method” provides conservative results in comparison to those produced by elastic-plastic calculations for the cases here studied.

Validation of Trio_U: Numerical Simulations of a ROCOM Buoyancy Driven Test Case

2004 ◽  
Author(s):  
T. Ho¨hne ◽  
U. Bieder ◽  
S. Kliem ◽  
H.-M. Prasser
Keyword(s):  
Turbulent Flows ◽  
Natural Circulation ◽  
Density Difference ◽  
Test Facility ◽  
Test Case ◽  
Pressurized Water ◽  
Constant Flow Rate ◽  
Sensor Position ◽  
Nuclear Systems ◽  
Water Reactors

A generic investigation of the influence of density differences between the primary loop inventory and the ECC water on the mixing in the downcomer was made at the ROCOM Mixing Test Facility at Forschungszentrum Rossendorf (FZR)/Germany. ROCOM is designed for experimental coolant mixing studies over a wide variety of possible scenarios. It is equipped with advanced instrumentation, which delivers high-resolution information characterizing either temperature or boron concentration fields in the investigated pressurized water reactor. For the validation of the Trio_U code an experiment with 5% constant flow rate in one loop (magnitude of natural circulation) and 10% density difference between ECC and loop water was taken. Trio_U is a CFD code developed by the CEA France, aimed to supply an efficient computational tool to simulate transient thermal-hydraulic single-phase turbulent flows encountered in the nuclear systems as well as in the industrial processes. For this study a LES approach was used for mesh sizes according to between 300000–2 million control volumes. The results of the experiment as well as of the numerical calculations show, that a streak formation of the water with higher density is observed. At the upper sensor, the ECC water covers a small azimuthal sector. The density difference partly suppresses the propagation of the ECC water in circumferential direction. The ECC water falls down in an almost straight streamline and reaches the lower downcomer sensor position directly below the affected inlet nozzle. Only later, coolant containing ECC water appears at the opposite side of the downcomer. The study showed, that density effects play an important role during natural convection with ECC injection in pressurized water reactors. Furthermore it was important to point out, that Trio_U is able to cope the main flow and mixing phenomena.

Break Size Effects on CET Response in an Upper Head SBLOCA Transient

2012 ◽  
Author(s):  
S. Gallardo ◽  
A. Querol ◽  
G. Verdú
Keyword(s):  
Pressure Vessel ◽  
Large Scale ◽  
Injection System ◽  
Test Facility ◽  
Energy Agency ◽  
Pressurized Water ◽  
The Core ◽  
Upward Direction ◽  
Water Reactors ◽  
Accident Conditions

In the transients produced during Small Break Loss-Of-Coolant Accidents (SBLOCA), the maximum Peak Cladding Temperature (PCT) in the core could suffer rapid excursions which might strongly affect the core integrity. Most Pressurized Water Reactors (PWR) have Core Exit Thermocouples (CETs) to detect core overheating by considering that superheated steam flows in the upward direction when core uncovery occurs during SBLOCAs. Operators may start Accident Management (AM) actions to mitigate such accident conditions when the CET temperature exceeds a certain value. However, in a Vessel Upper Head SBLOCA, a significant delay in time and temperature rise of CETs from core heat-up can be produced. This work is developed in the frame of OECD/NEA ROSA Project Test 6-1 (SB-PV-9 in JAEA) handled in the Large Scale Test Facility (LSTF) of the Japan Atomic Energy Agency (JAEA). Test 6-1 simulated a PWR pressure vessel Upper-Head SBLOCA with a break size equivalent to 1.9% of the cold leg break under the assumption of total failure of High Pressure Injection System (HPIS). The paper shows several analyses about the geometry variables (size, location, flow paths and Upper Head nodalization) which can influence on the pressure vessel Upper Head SBLOCA model performed using the thermal-hydraulic code TRACE5.

Dosimetry Assessments for the Reactor Pressure Vessel and Core Barrel in UK Pressurized Water Reactor (PWR) Plant

2012 ◽  
Vol 9 (4) ◽  
pp. 104016 ◽  
Author(s):  
D. A. Thornton ◽  
D. A. Allen ◽  
A. P. Huggon ◽  
D. J. Picton ◽  
A. T. Robinson ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Reactor Pressure
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