scholarly journals OECD/NRC PSBT Benchmark: Investigating the CATHARE2 Capability to Predict Void Fraction in PWR Fuel Bundle

2012 ◽  
Vol 2012 ◽  
pp. 1-10 ◽  
Author(s):  
A. Del Nevo ◽  
D. Rozzia ◽  
F. Moretti ◽  
F. D’Auria
Keyword(s):  
Void Fraction ◽  
Nuclear Power ◽  
Volume Fraction ◽  
Numerical Models ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Fuel Bundle ◽  
Hydraulic Conditions ◽  
Fraction Distribution ◽  
Engineering Corporation

Accurate prediction of steam volume fraction and of the boiling crisis (either DNB or dryout) occurrence is a key safety-relevant issue. Decades of experience have been built up both in experimental investigation and code development and qualification; however, there is still a large margin to improve and refine the modelling approaches. The qualification of the traditional methods (system codes) can be further enhanced by validation against high-quality experimental data (e.g., including measurement of local parameters). One of these databases, related to the void fraction measurements, is the pressurized water reactor subchannel and bundle tests (PSBT) conducted by the Nuclear Power Engineering Corporation (NUPEC) in Japan. Selected experiments belonging to this database are used for the OECD/NRC PSBT benchmark. The activity presented in the paper is connected with the improvement of current approaches by comparing system code predictions with measured data on void production in PWR-type fuel bundles. It is aimed at contributing to the validation of the numerical models of CATHARE 2 code, particularly for the prediction of void fraction distribution both at subchannel and bundle scale, for different test bundle configurations and thermal-hydraulic conditions, both in steady-state and transient conditions.

Influence of Sump on Containment Thermal Hydraulics: Synthesis of the TOSQAN Tests

2015 ◽  
Vol 1 (4) ◽  
Author(s):  
Emmanuel Porcheron ◽  
Pascal Lemaitre ◽  
Amandine Nuboer
Keyword(s):  
Nuclear Power ◽  
Severe Accident ◽  
Water Evaporation ◽  
Plant Water ◽  
Pressurized Water Reactor ◽  
Steam Condensation ◽  
Pressurized Water ◽  
Noncondensable Gases ◽  
Nuclear Core ◽  
Hydraulic Conditions

During the course of a severe accident in a nuclear power plant, water can be collected in the sump containment through steam condensation on walls, cooling circuit leak, and by spray systems activation. Therefore, the sump can become a place of heat and mass exchanges through water evaporation and steam condensation, which influences the distribution of hydrogen released in containment during nuclear core degradation. The objective of this paper is to present the analysis of semi-analytical experiments on sump interaction between containment atmosphere for typical accidental thermal hydraulic conditions in a pressurized water reactor (PWR). Tests are conducted in the TOSQAN facility developed by the Institut de Radioprotection et de Sûreté Nucléaire in Saclay. The TOSQAN facility is particularly well adapted to characterize the distribution of gases in a containment vessel. A tests’ grid was defined to investigate the coupled effect of the sump evaporation with wall condensation, for air steam conditions, with noncondensable gases (He, SF6), and for steady and transient states (two depressurization tests).

Results of Overpressurization Test of a 1:4-Scale Prestressed Concrete Containment Vessel Model

2002 ◽  
Author(s):  
Michael F. Hessheimer ◽  
Satoru Shibata ◽  
James F. Costello
Keyword(s):  
Prestressed Concrete ◽  
Nuclear Power ◽  
Limit State ◽  
Nuclear Regulatory Commission ◽  
Scale Model ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Containment Vessel ◽  
Engineering Corporation ◽  
Regulatory Commission

The Nuclear Power Engineering Corporation (NUPEC) of Japan and the U.S. Nuclear Regulatory Commission (NRC) have been co-sponsoring and jointly funding a Cooperative Containment Research Program at Sandia National Laboratories. The purpose of the program is to investigate the response of representative models of nuclear containment structures to pressure loading beyond the design basis accident and to compare analytical predictions with measured behavior. This is accomplished by conducting static, pneumatic overpressurization tests of scale models at ambient temperature. The first project in this program was a test of a mixed scale steel containment vessel (SCV). Next, a 1:4-scale model of a prestressed concrete containment vessel (PCCV), representative of a pressurized water reactor (PWR) plant in Japan, was constructed by NUPEC at Sandia National Laboratories from January 1997 through June, 2000. Concurrently, Sandia instrumented the model with over 1500 transducers to measure strain, displacement and forces in the model from prestressing through the pressure testing. The limit state test of the PCCV model was conducted in September, 2000 at Sandia National Laboratories. This paper describes the conduct and some of the results of this test.

Validation of the ATHROC Code With NUPEC Tests

2016 ◽  
Author(s):  
Yongzheng Chen ◽  
Yingwei Wu ◽  
Wenxi Tian ◽  
Suizheng Qiu ◽  
Guanghui Su
Keyword(s):  
Nuclear Power ◽  
Spray Cooling ◽  
Pressurized Water Reactor ◽  
Experiment Data ◽  
Flow Paths ◽  
Pressurized Water ◽  
Hot Gas ◽  
Engineering Corporation ◽  
Relative Errors ◽  
Peak Pressures

Xi’an Jiaotong University (XJTU) has developed a containment code called ATHROC (Analysis of Thermal Hydraulic Response of Containment) for the design of a new advanced PWR (Pressurized Water Reactor) in China. In the present work, the well-known NUPEC (NUclear Power Engineering Corporation) tests are simulated to validate the code. NUPEC test M-7-1 and M-4-3 were separately simulated to investigate the mixing effect of hot gas and steam being injected at a low elevation with or without spray cooling in the dome. An analytical model consisting of 26 cells and 67 flow paths was established. In addition, 62 distributed heat structures and 1 lumped heat structure were also modeled in the simulation. The predictions of both simulated tests agree relatively well with the experiment data. The maximum relative errors of predicted temperature for M-7-1 test and M-4-3 test were about 10% and 3%, respectively. The peak pressures of the two tests were about 3.8% and 6.7% overestimated, respectively, and the peak helium concentrations were about 25% and 20% overestimated, respectively. On the other hand, thermal and helium stratification within the containment were well predicted.

The Failure Analysis and Processing of Digital Reactor Protection System

2013 ◽  
Author(s):  
Jun Zhao ◽  
Xing Zhou ◽  
Jin Hu ◽  
Yanling Yu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Phase 1 ◽  
Protection System ◽  
Pressurized Water Reactor ◽  
Communication Failure ◽  
Pressurized Water ◽  
Power Failure ◽  
Self Test ◽  
Reactor Protection System

The Qinshan Nuclear Power Plant phase 1 unit (QNPP-1) has a power rating of 320 MWe generated by a pressurized water reactor that was designed and constructed by China National Nuclear Corporation (CNNC). The TELEPERM XS I&C system (TXS) is to be implemented to transform analog reactor protection system (RPS) in QNPP-1. The paper mainly describes the function, structure and characteristic of RPS in QNPP-1. It focuses on the outstanding features of digital I&C, such as strong online self-test capability, the degradation of the voting logic processing, interface improvements and CPU security. There are some typical failures during the operation of reactor protection system in QNPP-1. The way to analyze and process the failures is different from analog I&C. The paper summarizes typical failures of the digital RPS in the following types: CPU failure, communication failure, power failure, Input and output (IO) failure. It discusses the cause, risk and mainly processing points of typical failure, especially CPU and communication failures of the digital RPS. It is helpful for the maintenance of the system. The paper covers measures to improve the reliability of related components which has been put forward effective in Digital reactor protection system in QNPP-1. It will be valuable in nuclear community to improve the reliability of important components of nuclear power plants.

Film Boiling Behavior of a Pressurized Water Reactor Type Fuel Bundle

1981 ◽  
Vol 52 (1) ◽  
pp. 86-99 ◽  
Author(s):  
F. S. Gunnerson ◽  
D. T. Sparks ◽  
D. K. Kerwin
Keyword(s):  
Film Boiling ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Fuel Bundle ◽  
Reactor Type ◽  
Type Fuel

Thermal-Hydraulic Studies of the Steam Separation in Horizontal Steam Generator at PGV Test Facility

2009 ◽  
Author(s):  
Yuriy V. Parfenov ◽  
Oleg I. Melikhov ◽  
Vladimir I. Melikhov ◽  
Ilya V. Elkin
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Test Facility ◽  
Evolutionary Development ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Tube Bundles ◽  
Horizontal Steam Generator ◽  
Calculation Results ◽  
Steam Separation

A new design of nuclear power plant (NPP) with pressurized water reactor “NPP-2006” was developed in Russia. It represents the evolutionary development of the designs of NPPs with VVER-1000 reactors. Horizontal steam generator PGV-1000 MKP with in-line arrangement of the tube bundles will be used in “NPP-2006”. PGV test facility was constructed at the Electrogorsk Research and Engineering Center on NPP Safety (EREC) to investigate the process of the steam separation in steam generator. The description of the PGV test facility and tests, which will be carried out at the facility in 2009, are presented in this paper. The experimental results will be used for verification of the 3D thermal-hydraulic code STEG, which is developed in EREC. STEG pretest calculation results are presented in the paper.

Study of the Ageing Effects on the Lower Head Failure in a PWR Reactor

2020 ◽  
Author(s):  
R. Lo Frano ◽  
S. Paci ◽  
P. Darnowski ◽  
P. Mazgaj
Keyword(s):  
Nuclear Power ◽  
Quantitative Estimation ◽  
Computer Code ◽  
Severe Accident ◽  
Symmetric Model ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Term Operation ◽  
Ageing Effects ◽  
Lower Head

Abstract The paper studies influence the ageing effects on the failure of a Reactor Pressure Vessel (RPV) during a severe accident with a core meltdown in a Nuclear Power Plant (NPP). The studied plant is a generic high-power Generation III Pressurized Water Reactor (PWR) developed in the frame of the EU NARSIS project. A Total Station Blackout (SBO) accident was simulated with MELCOR 2.2 severe accident integral computer code. Results of the analysis, temperatures in the lower head and pressures in the lower plenum were used as initial and boundary conditions for the Finite Element Method (FEM) simulations. Two FEM models were developed, a simple two-dimensional axis-symmetric model of the lower head to study fundamental phenomena and complex 3D model to include interactions with the RPV and reactor internals. Ageing effects of a lower head were incorporated into the FEM models to investigate its influence onto lower head response. The ageing phenomena are modelled in terms of degraded mechanical material properties as σ(T), E(T). The primary outcome of the study is the quantitative estimation of the influence of ageing process onto the timing of reactor vessel failure. Presented novel methodology and results can have an impact on future consideration about Long-Term Operation (LTO) of NPPs.

Environmental Fatigue Test of CF8M With a Small Autoclave Simulating PWR Conditions

2008 ◽  
Author(s):  
Il-Seok Jeong ◽  
Gag-Hyeon Ha ◽  
Tae-Ryoung Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Low Cycle Fatigue ◽  
Test Procedure ◽  
Cyclic Strain ◽  
Fatigue Tests ◽  
Operating Conditions ◽  
Measuring Instruments ◽  
Pressurized Water Reactor ◽  
Pressurized Water

To develop a fatigue design curve of cast stainless steel CF8M used in primary piping material of nuclear power plants, low-cycle fatigue tests have been conducted by Korea Electric Power Research Institute (KEPRI). A small autoclave simulated the environment of a pressurized water reactor (PWR), 15 MPa and 315 °C. Fatigue life was measured in terms of the number of cycles with the variation of strain amplitudes at 0.04%/s strain rate. A small autoclave of 1 liter and cylindrical solid fatigue specimens were used for the strain-controlled low cycle environmental fatigue tests to make the experiments convenient. However, it was difficult to install displacement measuring instruments at the target length of the specimens inside the autoclave. To mitigate the difficulty displacement data measured at the shoulders of the specimen were calibrated based on the data relation of the target and shoulder length of the specimen during hot air test conditions. KEPRI developed a test procedure to perform low cycle environmental fatigue tests in the small autoclave. The procedure corrects the cyclic strain hardening effect by performing additional tests in high temperature air condition. KEPRI verified that the corrected test result agreed well with that of finite element method analysis. The process of correcting environmental fatigue data would be useful for producing reliable fatigue curves using a small autoclave simulating the operating conditions of a PWR.

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