Severe Fuel Damage Experiments With Advanced Cladding Materials to be Performed in the QUENCH Facility (QUENCH-ACM)

2008 ◽  
Author(s):  
L. Sepold ◽  
M. Große ◽  
M. Steinbru¨ck ◽  
J. Stuckert
Keyword(s):  
Nuclear Power ◽  
Cooling Water ◽  
Test Series ◽  
Experimental Program ◽  
Limited Data ◽  
Standard Type ◽  
Operational Conditions ◽  
Loss Of Coolant Accident ◽  
Design Basis Accident ◽  
Accident Conditions

The QUENCH out-of-pile experiments are part of the Severe Fuel Damage (SFD) program at the Karlsruhe Research Center. They are to investigate the hydrogen source term that results from reflooding an uncovered core of a Light-Water Reactor (LWR) with emergency cooling water. In the QUENCH experimental program Zircaloy-4 was used as standard-type material for rod cladding and grid spacer. Up to the end of 2007, 12 QUENCH experiments have been performed with this type of cladding; two test bundles contained B4C and one AgInCd absorber. One experiment (QUENCH-12) was conducted with Zr1%Nb cladding (VVER-type). Due to the niobium-bearing cladding, the VVER-type test QUENCH-12 could be regarded as a precursor for the upcoming program “QUENCH-ACM” with advanced cladding materials, i.e. M5, Duplex, ZIRLO, to be tested under SFD or BDBA (beyond design basis accident) conditions. These materials were developed for longer operation times in nuclear power reactors and extended burnup. They are optimized regarding their corrosion behavior under operational conditions and were also tested for LOCA (loss of coolant accident) and RIA (reactivity-initiated accident) conditions by the manufacturers. However, there are only very limited data available on the behavior of the new alloys in the SFD/BDBA temperature range, i.e. above 1500 K. The QUENCH-ACM test series has been defined with three experiments, i.e. QUENCH-14 through QUENCH-16. As in the Zircaloy-4 experiments, fuel is represented by ZrO2 pellets. Also, the test section instrumentation will be as usual with thermocouples attached to the cladding, shroud, and cooling jacket at elevations between −50 mm and 1350 mm. The QUENCH-ACM test series is scheduled to be performed in the period of 2008–2010. Test matrix and test bundle arrangements are presented in this paper.

Radiation Dose Evaluation of Typical Design Basis Accident for Advanced PWR in China

2021 ◽  
Author(s):  
Haiying Chen ◽  
Shaowei Wang ◽  
Xinlu Tian ◽  
Fudong Liu
Keyword(s):  
Radiation Dose ◽  
Effective Dose ◽  
Half Life ◽  
Nuclear Power ◽  
Outer Boundary ◽  
Design Basis ◽  
Loss Of Coolant Accident ◽  
Design Basis Accident ◽  
Typical Design ◽  
Total Effective Dose

Abstract The loss of coolant accident (LOCA) is one of the typical design basis accidents for nuclear power plant. Radionuclides leak to the environment and cause harm to the public in LOCA. Accurate evaluation of radioactivity and radiation dose in accident is crucial. The radioactivity and radiation dose model in LOCA were established, and used to analyze the radiological consequence at exclusion area boundary (EAB) and the outer boundary of low population zone (LPZ) for Hualong 1. The results indicated that the long half-life nuclides, such as 131I, 133I, 135I, 85Kr, 131mXe, 133mXe and 133Xe, released to environment continuously, while the short half-life nuclides, such as 132I, 134I, 83mKr, 85mKr, 87Kr, 88Kr, 135mXe and 138Xe, no longer released to environment after a few hours in LOCA. 133Xe may release the largest radioactivity to environment, more than 1015Bq. Inhalation dose was the major contribution to the total effective dose. The total effective dose and thyroid dose of Hualong 1 at EAB and the outer boundary of LPZ fully met the requirements of Chinese GB6249.

Determination of Thermal-Hydraulic Loads on Reactor Internals in a DBA-Situation

2006 ◽  
Author(s):  
Ville Lestinen ◽  
Timo Toppila ◽  
Antti Timperi ◽  
Timo Pa¨ttikangas ◽  
Markku Ha¨nninen
Keyword(s):  
Boundary Condition ◽  
Nuclear Power ◽  
Break Point ◽  
Mechanical Analysis ◽  
Element Analysis ◽  
Loss Of Coolant Accident ◽  
Design Basis Accident ◽  
Cfd Analyses ◽  
Reactor Internals

According to Finnish regulatory requirements, reactor internals have to stay intact in design basis accident (DBA) situations, so that control rods can always penetrate into the core. This is the basic motivation to study and develop more detailed methods for analyses of thermal-hydraulic loads on reactor internals during the DBA situation in the Loviisa Nuclear Power Plant (NPP) in Finland. In this work, the studied accident situation is Large Break Loss of Coolant Accident (LBLOCA). The objective of this work is to connect thermal-hydraulic and mechanical analysis methods with the goal to produce a reliable method for determination of thermal-hydraulic and mechanical loads on reactor internals in the accident situation. In the present model, the downcomer of a PWR is only included and the reactor internals will be added later. The tools studied are thermal-hydraulic system codes, computational fluid dynamics (CFD) codes and finite element analysis (FEA) codes. Both thermal-hydraulic and mechanical aspects are discussed in this paper. Firstly, the pressure boundary condition in the pipe break point was calculated with the system code. In the second step, CFD analyses were made. Finally, the full fluid-structure interaction coupling between the CFD and FEA codes was used. The codes used for development and simulations are APROS system code for boundary condition calculations, STAR-CD and FLUENT for CFD calculations and ABAQUS for FEA calculations.

The Sensitivity of the Addition of Low-Low Pressurizer Pressure in Coincidence With High Containment Pressure on M310 Type Nuclear Power Plant

2016 ◽  
Author(s):  
Qiu Yanfei
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Deterministic Analysis ◽  
Limit Value ◽  
Loss Of Coolant Accident ◽  
Stop Time ◽  
Design Basis Accident ◽  
Operator Action ◽  
Containment Pressure

Due to the new security system that the operator intervention is assumed to occur in 20 minutes is not acceptable, for the current M310 type nuclear power plant. The loss of coolant accident with Intermediate breaks in primary loop is the only one of design basis accident which need operator action in 20 minutes. For certain size break, the consequences are very sensitive to the pump stop time. According to deterministic analysis that for a certain size break, if stop the pump in 20 minutes after accident, the peak cladding temperature will exceed the limit value of 1204°C. Therefore, it is necessary to add low-low pressurizer pressure in coincidence with high containment pressure signal to stop pump automatically on M310 type nuclear power plant.

scholarly journals EXPERIMENTAL STUDY OF DAMAGED CR-COATED FUEL CLADDING IN POST-ACCIDENT CONDITIONS

2020 ◽  
Vol 28 ◽  
pp. 1-7
Author(s):  
Petr Červenka ◽  
Jakub Krejčí ◽  
Ladislav Cvrček ◽  
Vojtěch Rozkošný ◽  
František Manoch ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Experimental Methodology ◽  
Fuel Cladding ◽  
Loss Of Coolant Accident ◽  
Ring Compression ◽  
Loss Of Coolant ◽  
Accident Conditions ◽  
Cr Coating ◽  
The Impact ◽  
Positive Effect

To enhance the safety of nuclear power, the focus of researchers all around the world has recently mainly objected on the development of Accident Tolerant Fuels. Especially the Chromium coating of current Zirconium based cladding has been widely suggested and discussed for its immense positive effect on overall cladding properties. Nevertheless, it was observed that during the first stage of the Loss of Coolant Accident, cracks appear in the Cr coating due to its inability to tolerate higher plastic strain. Therefore, experimental methodology used in this article focuses on testing fuel cladding with damaged Cr coating after the high-temperature transient. The impact of cracks on degradation of cladding mechanical properties was observed using optical microscopy, ring compression test, microhardness, and evaluating hydrogen content and weight gain.

THE VALIDATION OF COMPREHENSIVE SOFTWARE FOR THE ASSESSMENT OF RADIOLOGICAL IMPACTS ON HUMANS THROUGH BENCHMARK ANALYSIS WITH RADIOLOGICAL CODES

2020 ◽  
Vol 190 (3) ◽  
pp. 250-268
Author(s):  
Ali Haghighi Shad ◽  
Mitra Athari Allaf ◽  
Darioush Masti ◽  
Kamran Sepanloo ◽  
Seyed Amir Hossein Feghhi
Keyword(s):  
Thyroid Gland ◽  
Effective Dose ◽  
Nuclear Power ◽  
Power Plants ◽  
Whole Body ◽  
The Public ◽  
Loss Of Coolant Accident ◽  
Initial Stage ◽  
Loss Of Coolant ◽  
Accident Conditions

Abstract In this paper, a novel domestic code called KIANA was developed for the assessment of radiological impacts on the population in normal and accident conditions including design basis accident (DBA) and beyond DBA (BDBA) for the nuclear power plants. The validation process of the KIANA code was performed using the results of the DOZA_M radiological code, whose results are presented in the Final Safety Analysis Report (FSAR) of the Bushehr Nuclear Power Plant Unit One (BNPP-1). The calculations of KIANA are performed based on the Gaussian diffusion model. The developed KIANA code has the potential of calculating the concentration and radionuclide doses due to the pathways such as airborne, foodstuff, marine (both one and two boxes models), soils, animals, vegetation (with and without tritium) and other pathways without any restriction. In the current research, the individual dose from a cloud to the member of the public in the region of BNPP-1 in normal condition was calculated. Moreover, the total effective dose to the member of the public from the primary to the secondary leakage inside steam generators, large break loss-of-coolant accident (LBLOCA) and small break loss-of-coolant accident was calculated. Thyroid gland equivalent dose for the infant (1–8 years) in the case of LBLOCA at the BNPP in DBA conditions was also evaluated. Finally, the prevented dose at the initial stage for the whole body of adults after BDBA, prevented dose at the initial stage for the thyroid gland of children after BDBA and the effective dose during the first year after the accident (external body irradiation from presence in the area) in the case of BDBA are assessed. The KIANA simulation results showed a good agreement with the FSAR data of BNPP.

VVER-440 Containment Thermal Hydraulic Analyses With MELCOR and CONTAIN Codes

2002 ◽  
Author(s):  
Gregory Gromov ◽  
Igor Lola ◽  
Stanislav Sholomitsky ◽  
Dmitry Gumenyuk ◽  
Valery Shikhabutinov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Design Basis ◽  
Loss Of Coolant Accident ◽  
Design Basis Accident ◽  
Sar Model ◽  
Release Data ◽  
Main Steam Line ◽  
Safety Features ◽  
Main Steam ◽  
Containment Pressure

In support of the analyses for the Rivne Nuclear Power Plant (RNPP) VVER-440/213 (Ukraine) Safety Analysis Report (SAR), detailed MELCOR and CONTAIN models of the containment were developed. The RNPP containment features a bubble condenser tower with air locks and active and passive spray systems. Code input models were developed to accurately represent the containment volumes, room interconnections, structural masses, and the engineering safety features. Although MELCOR 1.8.3 [1] was the primary tool for the SAR containment analysis, comparison calculations were performed using CONTAIN Version 1.12 [2]. Consequently, both the response of the VVER-440 containment to limiting design conditions as well as a comparison of the two codes is presented. In the context of SAR requirements, the present application was performed for design basis accidents with conservative assumptions to compare the containment temperature and pressure with design criteria. The peak containment pressure and temperature were evaluated using the most intensive release of the primary and secondary coolant into the hermetic compartments, in particular, for the large break loss of coolant accident and main steam line break. Conservative coolant release data were evaluated using the RELAP5/Mod3.2 SAR model. The selection of the accident scenario, initial and boundary conditions, and the major results are presented. The results of the analyses will be included in the design basis accident analysis chapter of the RNPP SAR.

Single Passive Condenser Failure During a Large-Break LOCA in the SBWR-1200 Reactor

2003 ◽  
Author(s):  
S. T. Revankar ◽  
M. Ishii ◽  
Y. Xu ◽  
H. J. Yoon ◽  
L. Cheng
Keyword(s):  
Safety System ◽  
Integral Test ◽  
Cooling Systems ◽  
Loss Of Coolant Accident ◽  
Design Basis Accident ◽  
Accident Conditions ◽  
Main Steam Line ◽  
Main Steam ◽  
Core Cooling ◽  
Safety Systems

The performance of the safety systems of a new design of the 1200-MWe simplified boiling water reactor during a large-break, loss-of-coolant accident transient was investigated when one of three passive condenser system failed. The accident considered was a break in the main steam line which is the major design basis accident. The integral test was performed to assess the safety systems and the response of the emergency core cooling systems to accident conditions in a scaled facility called PUMA. The loss of single condenser system did not affect the LOCA transient indicating that the existing safety system will adequately handle the large break LOCA with single condenser failure. The details of the safety system behavior are presented.

scholarly journals Risk due to creep of prestressed concrete at moderate temperature

2019 ◽  
Vol 281 ◽  
pp. 01007
Author(s):  
Thierry Vidal ◽  
Hugo Cagnon ◽  
Nam Nguyen ◽  
Jean-Michel Torrenti ◽  
Alain Sellier
Keyword(s):  
Internal Pressure ◽  
Prestressed Concrete ◽  
Nuclear Power ◽  
Power Plants ◽  
Severe Accident ◽  
Experimental Program ◽  
Strong Increase ◽  
Creep Tests ◽  
Accident Conditions ◽  
Creep Deformations

This study is a part of a French national project dealing with the mechanical behaviour of the containment vessel of French Nuclear Power Plants in case of a severe accident. The accident conditions are characterized by the increases of internal pressure, +0.5 MPa, and of temperature, up to 180°C, during two weeks. Heating can induce a strong increase of creep deformations kinetics leading to prestressing losses of concrete. Associated to internal pressure, tensile stress could occur in some areas of the structure and the potential cracking could affect the containment capacity of the vessel. One of the objectives of the project was thus to provide original creep data to develop accurate models, taking into account the coupled effects of temperature, desiccation and damage, and able to predict the behaviour of prestressed concrete structures in such insitu conditions. A wide experimental program consisted of numerous creep tests under various thermo-hydro-mechanical conditions in the values range of the accident. The presented results concern uniaxial compressive and flexural creep tests respectively performed on concrete cylinders and prestressed concrete beams, at 20°C and 40°C without desiccation.

Thermal Hydraulics Analysis on a Simplified Model of Spent Fuel Pool

2013 ◽  
Author(s):  
Min Liang ◽  
Daogang Lu ◽  
Huining Xia ◽  
Yuhao Zhang
Keyword(s):  
Nuclear Power ◽  
Reactor Core ◽  
Cooling Water ◽  
Measuring Instruments ◽  
Spent Fuel ◽  
Circulating Water ◽  
Experimental Process ◽  
Spent Fuel Pool ◽  
Accident Conditions ◽  
Hydraulic Process

The spent fuel still has massive decay heat and is temporarily stored in the spent fuel pool, after unloaded from the reactor core; it is cooled by the circulating water in the spent fuel pool. The present study generally agreed that the spent fuel pool is relatively safe when the cooling water works, even if under accident conditions. It is generally considered that the thermal-hydraulic process in the spent fuel pool is very slow, and does not endanger the safety of nuclear power plant. But the boiling in the pool has been given more and more attention after the Fukushima nuclear accident; especially after the failure of core safety injection and the cooling water in the spent fuel pool is lost. School of Nuclear Science and Engineering, North China Electric Power University intends to perform overall test bench to simulate the AP1000 spent fuel pool, the experimented data can be used to validate the results of the COSINE software calculations. In this paper, the numerical simulation is performed with CFD software to calculate the temperature, velocity, pressure field of the scaled spent fuel pool experimental bench; it can provide the guideline for the selection of the test section measuring instruments and the arrangement of the measuring points. Also, it’s essential to verify the COSINE software and guide the experimental process.

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