Small Break LOCA Analysis of ACR-700 NPP

2006 ◽  
Author(s):  
Limin Zheng ◽  
Sen Shen ◽  
David Wright
Keyword(s):  
Nuclear Power ◽  
Sensitivity Study ◽  
Hydraulic Analysis ◽  
Reactor Outlet ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Subsequent Loss ◽  
Safety Systems ◽  
Limiting Case ◽  
Thermal Hydraulic Analysis

A small break loss of coolant accident (SB-LOCA) analysis to assess a preliminary conceptual design of the ACR-700 PHWR nuclear power plant (NPP) developed by AECL has been performed with CATHENA MOD 3.5d, a PHWR system thermal-hydraulic analysis code. The limiting break size has been found by performing a sensitivity study for three different break locations [i.e. reactor inlet header (RIH), HTS pump suction (PS) pipe and reactor outlet head (ROH)] under the limiting case (i.e. SB-LOCA with subsequent loss of class IV power with all safety systems available). The analysis results indicate that the SB-LOCA acceptance criteria are satisfied.

Steady-State Thermal-Hydraulic Analysis of a Particle-Bedded Boiling Water Reactor

2009 ◽  
Author(s):  
Xiaoyu Cai ◽  
Suizheng Qiu ◽  
Guanghui Su ◽  
Changyou Zhao
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Superheated Steam ◽  
Nuclear Reactor ◽  
Hydraulic Analysis ◽  
Light Water ◽  
Safety Systems ◽  
Fuel Elements ◽  
Thermal Hydraulic Analysis

The current Light Water Reactors both BWR and PWR have extensive nuclear reactor safety systems, which provide safe and economical operation of Nuclear Power Plants. During about forty years of operation history the safety systems of Nuclear Power Plants have been upgraded in an evolutionary manner. The cost of safety systems, including large containments, is really high due to a capital cost and a long construction period. These conditions together with a low efficiency of steam cycle for LWR create problems to build new power plants in the USA and in the Europe. An advanced Boiling Water Reactor concept with micro-fuel elements (MFE) and superheated steam promises a radical enhancement of safety and improvement of economy of Nuclear Power Plants. In this paper, a new type of nuclear reactor is presented that consists of a steel-walled tube filled with millions of TRISO-coated fuel particles (Micro-Fuel Elements, MFE) directly cooled by a light-water coolant-moderator. Water is used as coolant that flows from bottom to top through the tube, thereby fluidizing the particle bed, and the moderator water flows in the reverse direction out of the tube. The fuel consists of spheres of about 2.5 mm diameter of UO2 with several coatings of different carbonaceous materials. The external coating of steam cycle the particles is silicon carbide (SiC), manufactured with chemical vapor deposit (CVD) technology. Steady-State Thermal-Hydraulic Analysis aims at providing heat transport capability which can match with the heat generated by the core, so as to provide a set of thermal hydraulic parameters of the primary loop. So the temperature distribution and the pressure losses along the direction of flow are calculated for equilibrium core in this paper. The calculation not only includes the liquid region, but the two phase region and the superheated steam region. The temperature distribution includes both the temperature parameters of micro-fuel elements and the coolant. The results show that the maximum fuel temperature is much lower than the limitation and the flow distribution can meet the cooling requirement in the reactor core.

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.

scholarly journals SUBSTANTIATION OF MODERNIZED BLACKOUT & LOSS-OF-COOLANT ACCIDENT MANAGEMENT STRATEGY AT NUCLEAR POWER PLANTS WITH WWER

2020 ◽  
Vol 2 (61) ◽  
pp. 70-77
Author(s):  
V. Skalozubov ◽  
◽  
V. Spinov ◽  
D. Spinov ◽  
Т. Gablaya ◽  
...  
Keyword(s):  
Management Strategy ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Management Strategies ◽  
Computational Modelling ◽  
Loss Of Coolant Accident ◽  
Accident Management ◽  
Loss Of Coolant ◽  
Safety Systems

The analysis of the known results of RELAP5/V.3.2 simulation for loss of coolant & blackout accidents at WWER nuclear power plants showed that the design accident management strategies with design passive safety systems do not provide the necessary safety conditions for the maximum permissible temperature of fuel claddings, the minimum permissible level of coolant in the reactor and feed water in the steam generators. A conservative thermohydrodynamic model for a design and modernized blackout & loss-of-coolant accident management strategy at a nuclear power plant with WWER has been developed. Design passive safety systems carry out the design accident management strategy: pressurizer safety valves, secondary steam relief valves, and hydraulic reservoirs of the emergency core cooling system of the reactor. Promising afterheat removal passive systems and the reactor level and steam generator water level control systems carry out the modernized blackout & loss-of-coolant accident management strategy. The main conservative assumptions of the presented model of blackout & loss-of-coolant accidents: complete long-term failure of all electric pumps of active safety systems, the temperature of nuclear fuel in the central part of the fuel matrix is assumed as the maximum allowable one, effect of “run down” flow of a turbine feed pump and the coolant level in pressurizer on accident process is not considered. Computational modelling has found that violations of the safety conditions are over the entire range of leak sizes for the design blackout & loss-of-coolant accident management strategy. For the modernized blackout & loss-of-coolant accident management strategy, safety conditions are provided for 72 hours of the accident and more. The presented results of computational modelling of blackout accident management strategies for nuclear power plants can be used to modernize and improve symptom-informed emergency instructions and guidelines for the severe accident management at nuclear power plants with WWER. Application of the results of computational modelling of blackout accident management strategies is generally not substantiated for other types of reactor facilities. In this case, it is necessary to develop calculated models for blackout accident management taking into account the specifics of the structural and technical characteristics and operating conditions for safety related systems of nuclear power plants.

CFD-Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions: Influence of Temperature Safety Injection and Fluid-Structure Thermal Coupling

2005 ◽  
Author(s):  
A. Martin ◽  
D. Monfort ◽  
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 whose purpose is to understand the main phenomena which 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. On the whole, 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.

scholarly journals ANALYSIS OF AN EXTREME LOSS OF COOLANT IN THE IPR-R1 TRIGA REACTOR USING A RELAP5 MODEL

2013 ◽  
Vol 12 (2) ◽  
pp. 46
Author(s):  
P. A. L. Reis ◽  
A. L. Costa ◽  
C. Pereira ◽  
M. A. F. Veloso ◽  
H. V. Soares ◽  
...  
Keyword(s):  
Research Reactor ◽  
Technology Development ◽  
Nuclear Research ◽  
Hydraulic Analysis ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Transient Case ◽  
Fuel Elements ◽  
Thermal Hydraulic Analysis ◽  
Pool Type

The RELAP5/MOD3.3 code has been applied for thermal hydraulic analysis of power reactors as well as nuclear research reactors with good predictions. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA have been validated for steady state and transient situations. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. In this work, an extreme transient case of loss of coolant accident (LOCA) has been simulated. For this type of analysis, the automatic scram of the reactor was not considered because the main aim was to verify the evolution of the fuel elements heating in the absence of coolant. The temperature evolutions are presented as well as an analysis about the temperature safety limits.

Analysis on Impact of Fuel Thermal Conductivity Degradation (TCD) on Large Break Loss of Coolant Accidents of CAP1000

2017 ◽  
Author(s):  
Wang Weiwei ◽  
Lu Lu
Keyword(s):  
Thermal Conductivity ◽  
Nuclear Power ◽  
Power Plants ◽  
Sensitivity Study ◽  
Fuel Pellet ◽  
Steady State Condition ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Fuel Storage ◽  
High Burnup

Under high burnup conditions, thermal conductivity of fuel pellet degrades, which is referred to as thermal conductivity degradation (TCD). TCD phenomenon influences fuel average temperature and fuel storage energy under steady state condition before loss of coolant accident (LOCA) and further influences peak cladding temperature (PCT) during large break LOCA process. In this study, sensitivity study on double ended guillotine break of cold leg in CAP1000 at different burnup conditions was performed, using large break LOCA analysis code WCOBRA/TRAC and PCTs under different conditions were obtained. The modified NFI (Nuclear Fuels Institute) TCD model was adopted to model fuel conductivity after degradation in analysis and decrease of peaking factors including FQ and FΔh after 30GWD/MTU was also considered. Sensitivity analysis showed that: after considering the influence of TCD and peaking factor burndown, the PCT limiting case did not occur in low burnup range again, but occurred at burup of about 29GWD/MTU. Compared to other burnup points, the first and second peak values of PCT at that burnup point were all at the highest level. Performing of this study could prefer reference for analysis and estimation of large break LOCA of passive nuclear power plants under high burnup conditions.

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

2011 ◽  
Vol 133 (3) ◽  
Author(s):  
A. Martin ◽  
S. Benhamadouche ◽  
G. Bezdikian ◽  
F. Beaud ◽  
F. Lestang
Keyword(s):  
Heat Transfer ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Hydraulic Analysis ◽  
Loss Of Coolant Accident ◽  
Pressurized Thermal Shock ◽  
Loss Of Coolant ◽  
Thermal Hydraulic Analysis ◽  
Reactor Pressure

Integrity evaluation methods for nuclear reactor pressure vessels (RPVs) under pressurized 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 small break loss of coolant accident (SBLOCA) transients. This paper presents the research and development program started at EDF on the computational fluid dynamics (CFD) determination of the cooling phenomena of a PWR vessel during a pressurized thermal shock. The numerical results are obtained with the thermal-hydraulic tool Code̱Saturne, 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, this paper presents mainly a parametric study that 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 downcomer 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.

Fatigue Evaluation for Thermally Stratified DVI Piping

2012 ◽  
Author(s):  
Hwan Ho Lee ◽  
Joon Ho Lee ◽  
Dong Jae Lee ◽  
Seok Hwan Hur ◽  
Il Kwun Nam ◽  
...  
Keyword(s):  
Temperature Distribution ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Heat Removal ◽  
Local Stress ◽  
Piping System ◽  
Hydraulic Analysis ◽  
Fatigue Evaluation ◽  
Core Cooling ◽  
Thermal Hydraulic Analysis

A numerical analysis has been performed to estimate the effect of thermal stratification in the safety injection piping system. The Direct Vessel Injection (DVI) system is used to perform the functions of Emergency Core Cooling and Residual Heat Removal for an APR1400 nuclear power plant (Korea’s Advanced Power Reactor 1400 MW-Class). The thermal stratification is anticipated in the horizontally routed piping between the DVI nozzle of the reactor vessel and the first isolation valve. Non-axisymmetric temperature distribution across the pipe diameter induced by the thermal stratification leads to differential thermal growth of the piping causing the global bending stress and local stress. Thermal hydraulic analysis has been performed to determine the temperature distribution in the DVI piping due to the thermal stratification. Piping stress analysis has also been carried out to evaluate the integrity of the DVI piping using the thermal hydraulic analysis results. This paper provides a methodology for calculating the global bending stresses and local stresses induced by the thermal stratification in the DVI piping and for performing fatigue evaluation based on Subsection NB-3600 of ASME Section III.

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