Thermo-Hydraulic Design of Westinghouse 4-Loop Reactor Core by COBRA-EN Code

2009 ◽  
Author(s):  
J. Jafari ◽  
B. Kalagar ◽  
E. Abdi Aghdam ◽  
F. D’Auria
Keyword(s):  
Heat Flux ◽  
Fuel Assembly ◽  
Nuclear Power ◽  
Reactor Core ◽  
Nucleate Boiling ◽  
Operating Conditions ◽  
Control Rod ◽  
Hydraulic Design ◽  
Fuel Assemblies

The Westinghouse 4-Loop PWR is a 3411MWth Nuclear Power Plant (NPP). The reactor core consists of 193 fuel assemblies within the core shroud. Each fuel assembly is arranged in 17×17 arrays and includes 264 fuel rods, 24 control rod guide tubes and one instrument tube. The objective of thermal and hydrodynamic design is to safely remove of the generated heat in the fuel without producing excessive fuel temperatures or steam void formations and without approaching the critical heat flux under steady-state operating conditions. This paper presents reactor core and fuel assembly modeling of the Westinghouse 4-Loop NPP using the thermo hydraulic subchannel analysis COBRA-EN code. The results of this modeling are compared with the VIPRE-01 thermal hydraulic code. The study involves the determination of the departure from nucleate boiling ratio (DNBR) in the hot channel of the reactor core, the temperature profiles, heat flux and pressure drop across the hottest channel of the hot assemblies. The obtained results shows that the good agreements are exist between the COBRA-EN and VIPRE-01 thermal hydraulic codes.

scholarly journals Method for Contactless Determination of the Height of A Fuel Assembly by Means of A 3-D Reconstruction of A Collinear Stereopair Images

2018 ◽  
Vol 7 (3.13) ◽  
pp. 51
Author(s):  
S Kravtsov ◽  
K Rumyantsev
Keyword(s):  
Fuel Assembly ◽  
Nuclear Power ◽  
Vision System ◽  
Reactor Core ◽  
System A ◽  
Fuel Assemblies ◽  
Nuclear Power Unit ◽  
Head Height ◽  
Set Of Points

A method for determining the head height of fuel assemblies in the reactor core of a nuclear power unit using a 3-D reconstruction of a stereopair of collinear images is considered. The method is based on the principle of statistical evaluation of the height of a set of points for a 3-D reconstruction of the contour of the head of the fuel assembly. To obtain a stereopair of images, it is suggested to use a collinear digital stereo-vision system. A model experiment was carried out. The results are compared with the known method for determining the height of the heads of fuel assemblies, based on an estimate of the height of the centers of gravity of the contours of fuel assembly heads. The proposed method shows a higher accuracy in solving the problem of determining the heights of fuel assembly heads in comparison with the known method.  

Questions Relative to CHF Testing for New Pressurized Water Reactor Advanced Fuel Assemblies

2010 ◽  
Author(s):  
Christophe Herer
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Heat Removal ◽  
Nucleate Boiling ◽  
Operating Conditions ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Fuel Assemblies ◽  
Advanced Fuel

One of the limiting conditions during operation of a Pressurized Water Reactor is cladding integrity in case of occurrence of any conditions I or II events. The decoupling criterion is the absence of Departure from Nucleate Boiling (DNB) during the full sequence of any of these transients. Heat transfer between the clad and the water is limited by the DNB phenomenon when local surface heat flux is greater than the so-called Critical Heat Flux (CHF). Heat production at the surface is higher than heat removal capacity by the coolant therefore a vapor blanket is formed around the clad; consequently the heat transfer will drastically drop resulting in a sudden significant increase of the local wall temperature and clad damage may appear if no corrective action is initiated. DNB can not be estimated with physical principles only. Experimental support is needed for evaluation. Occurrence of DNB is evaluated using the Departure from Nucleate Boiling Ratio (DNBR) which is a function of both core thermal hydraulic (T/H) parameters and design of the fuel assembly. Advanced fuel assemblies claim higher CHF values compared to previous designs. Along with increased DNB performances for advanced fuel assemblies, CHF correlation development and advanced methodologies enable to extend normal operating conditions of a nuclear plant. On the one hand, CHF performances really increased allow additional margin related to the loss of fuel cladding integrity whereas on the other hand optimized correlations and advanced methodologies reduce this margin. An accurate assessment of the CHF performance of the advanced fuel assemblies is therefore required. This paper will raise issues regarding the assessment of the CHF performance of new advanced fuel assemblies design. The issues will be focused on the reliability of the experimental assessment of the CHF values and the accuracy of the transposition of mock up geometries to plant core configuration (representativity of the experiments). The verification that the tests conditions (pressure, flowrate, quality, heat flux …) ensure a proper coverage of all core conditions encountered during any of the conditions I & II transients is closely linked to DNBR methods and will not be extensively covered in this paper. This paper suggests some thoughts about relevance of the demonstration carried out by vendors on these matters.

scholarly journals KYLIN-V2.0 CODE CALCULATION ABILITY VERIFICATION BASED ON VERA BENCHMAR

2021 ◽  
Vol 247 ◽  
pp. 10020
Author(s):  
Dongyong Wang ◽  
Yingrui Yu ◽  
Xingjie Peng ◽  
Chenlin Wang ◽  
Kun Liu ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Power Distribution ◽  
Nuclear Power ◽  
Neutron Transport ◽  
Monte Carlo Code ◽  
Neutron Transport Equation ◽  
Control Rod ◽  
Fuel Assemblies ◽  
Fuel Pin ◽  
Calculation Ability

Virtual Environmental for Reactor Analysis (VERA) benchmark was released by the Consortium for Advanced Simulation of Light water reactors (CASL) project in 2012. VERA benchmark includes more than ten problems at different levels, from 2D fuel pin case to 2D fuel assembly case to 3D core refuelling case, in addition, reference results and experimental measured data of some problems were provided by CASL. Fuel assemblies in VERA benchmark are various, including control rod assemblies, Pyrex assembly, IFBA assembly, WABA assembly and gadolinium poison assembly, and so on. In this paper, various fuel assembly models in the VERA benchmark have been built by using KYIIN-V2.0 code to verify its calculation ability from 2D fuel pin case to 2D fuel assembly case to 2D 3x3 fuel assembly case, and making a comparative analysis on the reference results in VERA benchmark, as well as the calculation results of the Monte Carlo code RMC. KYLIN-V2.0 is an advanced neutron transport lattice code developed by Nuclear Power Institute of China (NPIC). The subgroup resonance calculation method is used in KYIIN-V2.0 to obtain effective resonance selfshielding cross section, method of modular characteristics (MOC) is adopted to solve the neutron transport equation, and CRAM method and PPC method is adopted to solve the depletion equation. The numerical results show that KYLIN-V2.0 code has the reliable capability of direct heterogeneous calculation of 2D fuel assembly, and the effective multiplication factor, assembly power distribution, rod power distribution and control rod reactivity worths of various fuel assemblies that are calculated by KYLIN-V2.0 are in better agreement with the reference.

scholarly journals Reactor Core Conceptual Design for a Scalable Heating Experimental Reactor, LUTHER

2021 ◽  
Vol 2 (2) ◽  
pp. 207-214
Author(s):  
Thinh Truong ◽  
Heikki Suikkanen ◽  
Juhani Hyvärinen
Keyword(s):  
Low Temperature ◽  
Conceptual Design ◽  
Nuclear Power ◽  
Water Pressure ◽  
Reactor Core ◽  
Electricity Production ◽  
Design Parameters ◽  
Experimental Reactor ◽  
Fuel Assemblies ◽  
Enriched Uranium

In this paper, the conceptual design and a preliminary study of the LUT Heating Experimental Reactor (LUTHER) for 2 MWth power are presented. Additionally, commercially sized designs for 24 MWth and 120 MWth powers are briefly discussed. LUTHER is a scalable light-water pressure-channel reactor designed to operate at low temperature, low pressure, and low core power density. The LUTHER core utilizes low enriched uranium (LEU) to produce low-temperature output, targeting the district heating demand in Finland. Nuclear power needs to contribute to the decarbonizing of the heating and cooling sector, which is a much more significant greenhouse gas emitter than electricity production in the Nordic countries. The main principle in the development of LUTHER is to simplify the core design and safety systems, which, along with using commercially available reactor components, would lead to lower fabrication costs and enhanced safety. LUTHER also features a unique design with movable individual fuel assembly for reactivity control and burnup compensation. Two-dimensional (2D) and three-dimensional (3D) fuel assemblies and reactor cores are modeled with the Serpent Monte Carlo reactor physics code. Different reactor design parameters and safety configurations are explored and assessed. The preliminary results show an optimal basic core design, a good neutronic performance, and the feasibility of controlling reactivity by moving fuel assemblies.

Analysis of Flow Blockage of a Single Fuel Assembly in the JRR-3 20MW Research Reactor

2018 ◽  
Author(s):  
Yuchuan Guo ◽  
Guanbo Wang ◽  
Dazhi Qian ◽  
Heng Yu ◽  
Bo Hu
Keyword(s):  
Fuel Assembly ◽  
System Analysis ◽  
Research Reactor ◽  
Natural Circulation ◽  
Reactor Core ◽  
Nucleate Boiling ◽  
Regulation System ◽  
Lower Power ◽  
Emergency Shutdown ◽  
Flow Blockage

The case of flow blockage of a single fuel assembly in the JRR-3 20MW open-pool-type research reactor is investigated without taking into account the effect of the power regulation system. The coolant system and multi-channel reactor core are modeled in detail using thermal hydraulic system analysis code RELAP5/MOD3.4. MDNBR (Minimum Departure From Nucleate Boiling Ratio) and the maximum fuel central temperature are investigated to assess the integrity of fuels. The fuel plates in blocked assembly are not damaged until the blockage ratio exceeds 70%. In addition, the mitigative effect of the assumed 18 MW lower power emergency shutdown operation on the accident is also discussed qualitatively. Results indicate that although the assumed lower power emergency shutdown operation cannot avoid the most severe operating condition, it can obviously mitigate the consequences of the accident. The reactor eventually remains in the long-term safe state when natural circulation is established.

scholarly journals Predicting Ex-core Detector Response in a PWR with Monte Carlo Neutron Transport Methods

2020 ◽  
Vol 225 ◽  
pp. 03007
Author(s):  
Tanja Goričanec ◽  
Domen Kotnik ◽  
Žiga Štancar ◽  
Luka Snoj ◽  
Marjan Kromar
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Reactor Core ◽  
Response Prediction ◽  
Detector Response ◽  
Design Calculation ◽  
Monte Carlo Code ◽  
Control Rod ◽  
The Core ◽  
Fuel Assemblies

An approach for calculating ex-core detector response using Monte Carlo code MCNP was developed. As a first step towards ex-core detector response prediction a detailed MCNP model of the reactor core was made. A script called McCord was developed as a link between deterministic program package CORD-2 and Monte Carlo code MCNP. It automatically generates an MCNP input from the CORD-2 data. A detailed MCNP core model was used to calculate 3D power distributions inside the core. Calculated power distributions were verified by comparison to the CORD-2 calculations, which is currently used for core design calculation verification of the Krško nuclea power plant. For the hot zero power configuration, the deviations are within 3 % for majority of fuel assemblies and slightly higher for fuel assemblies located at the core periphery. The computational model was further verified by comparing the calculated control rod worth to the CORD-2 results. The deviations were within 50 pcm and considered acceptable. The research will in future be supplemented with the in-core and ex-core detector signal calculations and neutron transport outside the reactor core.

AP1000 Plant CILC/CIPS Risk Assessment Using Advanced TH Methodology

2012 ◽  
Author(s):  
G. Wang ◽  
P. Sapienza ◽  
R. J. Fetterman ◽  
M. Y. Young ◽  
J. R. Secker ◽  
...  
Keyword(s):  
Risk Assessment ◽  
Nucleate Boiling ◽  
Operating Conditions ◽  
Localized Corrosion ◽  
Pressurized Water ◽  
Assembly Design ◽  
Power Shift ◽  
Fuel Assemblies ◽  
Water Reactors ◽  
The Relationship

Similar to many existing Pressurized Water Reactors (PWR), the AP1000® cores will undergo sub-cooled nucleate boiling in the upper grid spans of some fuel assemblies at normal operating conditions. Sub-cooled nucleate boiling may increase crud deposits on the fuel cladding surface which may increase the risk of Crud Induced Power Shift (CIPS) and/or Crud Induced Localized Corrosion (CILC). A CIPS/CILC risk assessment has been performed to support the AP1000 fuel assembly design finalization. In this paper, the advanced thermal-hydraulic (TH) methodology used in the AP1000 plant CIPS/CILC risk assessments are summarized and discussed, and the relationship between the CIPS/CILC mechanisms, fuel reliability, and plant operating conditions is also presented. Finally, acceptable AP1000 core CIPS/CILC risk assessment results are summarized and suggestions that specifically target reducing CIPS/CILC risks for AP1000 plants are described.

Insulation Material Deposition and Distribution in a PWR Fuel Assembly Cluster

2012 ◽  
Author(s):  
Stefan Renger ◽  
Sören Alt ◽  
Wolfgang Kästner ◽  
André Seeliger ◽  
Frank Zacharias
Keyword(s):  
Fuel Assembly ◽  
Reactor Core ◽  
Visual Observation ◽  
Test Facility ◽  
Insulation Material ◽  
Control Rod ◽  
Free Jet ◽  
Laser Sensors ◽  
Material Deposition ◽  
Loss Of Coolant Accident

Background of experimental and methodical work is the loss of coolant accident (LOCA) with release of fibrous pipe insulation material. Latest investigations were focused on material deposition and distribution (cross mixing) in the reactor core. Therefore, a 2×2 PWR fuel assembly (FA) cluster was constructed. Four shortened PWR-FA-dummies are provided with separated in- and outlets. Every 16×16 fuel rod dummy consists of 20 control rod simulators, two spacers, FA-head and FA-bottom with a 3.5×3.5 mm integrated debris-screen filter (IDF). The cluster is encased in an acrylic housing for visual observation. It is connected with the test facility “Zittau Flow Tray” (ZFT), a simplified sump model, which allows inclusion and investigation of complex phenomena like material sedimentation in the sump and strainer blockages. A well mixing of air in the fluid was also considered by free jet expansions and flows through full cone-nozzles as well as marginal air entrainments. This Paper includes descriptions of applied measuring techniques (digital image processing, thrubeam laser sensors etc.) and an overview of all considered boundary conditions. Experimental results, aiming at the development, implementation and verification of multiphase flow and strainer models, are presented.

Sensitivity of CTF Solution to Subchannel Window Size

2018 ◽  
Author(s):  
Roman Mukin ◽  
Marcus Seidl ◽  
Ivor Clifford ◽  
Hakim Ferroukhi
Keyword(s):  
Heat Flux ◽  
Window Size ◽  
Local Heat ◽  
Nucleate Boiling ◽  
Pressurized Water Reactor ◽  
Reference Case ◽  
Pressurized Water ◽  
Local Flow ◽  
Power Maps ◽  
Fuel Assemblies

In this work, a so-called mini-core consisting of a 3 × 3 array of 17 × 17 pressurized water reactor (PWR) fuel assemblies (FA) is considered with the aim of identifying the most conservative window size for hot channel analysis of bowed fuel assemblies. Overall, five different mini-core configurations are analyzed: one is the reference case, i.e. without FA displacement and four different cases with diagonal and parallel FA displacements. Rod power maps for these mini-cores were exported from neutronic calculations with CASMO-SIMULATE codes. Subchannel modelling with COBRA-TF code of all five mini-cores allows one to identify the rod position with a minimum departure from the nucleate boiling ratio (DNBR) and to construct input decks with different rod window sizes around the previously identified rod position. Overall, eight different window sizes are considered: 3 × 3, 5 × 5, 7 × 7, 9 × 9, 11 × 11, 13 × 13, 15 × 15 and 17 × 17. Results of subchannel analysis for a mini-core and different subchannel window configurations are compared with the help of DNBR parameter, which is the ratio between the critical heat flux (CHF) and the actual local heat flux on a rod. An assessment of three different CHF models is applied in this work: Groeneveld CHF look-up table (LUT), W3 CHF correlation, and Doroschuk CHF LUT. The general conclusion of this work is that for deformed core configurations, an appropriate rod window size needs to be determined to adequately capture the local flow redistribution. For large displacements (the largest displacement considered in this work is 10 mm), the DNBR ratio can drop to one. DNBRs obtained with the W3 CHF correlation give the most conservative results.

Heat Transfer Characteristics of a Radial Jet Reattachment Flame

1997 ◽  
Vol 119 (2) ◽  
pp. 258-264 ◽  
Author(s):  
J. W. Mohr ◽  
J. Seyed-Yagoobi ◽  
R. H. Page
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Combustion Process ◽  
Local Heat ◽  
Operating Conditions ◽  
Recirculation Region ◽  
Heat Transfer Characteristics ◽  
Transfer Characteristics ◽  
Impingement Surface

A Radial Jet Reattachment Combustion (RJRC) nozzle forces primary combustion air to exit radially from the combustion nozzle and to mix with gaseous fuel in a highly turbulent recirculation region generated between the combustion nozzle and impingement surface. High convective heat transfer properties and improved fuel/ air mixing characterize this external mixing combustor for use in impingement flame heating processes. To understand the heat transfer characteristics of this new innovative practical RJRC nozzle, statistical design and analysis of experiments was utilized. A regression model was developed which allowed for determination of the total heat transfer to the impingement surface as well as the NOx emission index over a wide variety of operating conditions. In addition, spatially resolved flame temperatures and impingement surface temperature and heat flux profiles enabled determination of the extent of the combustion process with regards to the impingement surface. Specifically, the relative sizes of the reaction envelope, high temperature reaction zone, and low temperature recirculation zone were all determined. At the impingement surface in the reattachment zone very high local heat flux values were measured. This study provides the first detailed local heat transfer characteristics for the RJRC nozzle.

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