Best Estimate Simulation and Uncertainty Analysis of LB LOCA for KudanKulam VVER-1000 NPP

2009 ◽  
Author(s):  
Luben Sabotinov ◽  
Abhishek Srivastava ◽  
Pierre Probst
Keyword(s):  
Uncertainty Analysis ◽  
Nuclear Power ◽  
Input Data ◽  
Power Plants ◽  
Computer Code ◽  
Uncertainty Evaluation ◽  
Statistical Tool ◽  
Best Estimate ◽  
Safety Margins ◽  
Input Uncertainties

In the accident analysis of the Nuclear Power Plants (NPP) nowadays the international licensing practice considers several acceptable options for demonstrating the safety i.e. use of conservative computer codes with conservative assumptions, best estimate codes combined with conservative assumptions and conservative input data and application of best estimate codes with assumptions and realistic input data but associated with uncertainty evaluation of the results. The last option is particularly attractive because it allows for more precise prediction of safety margins with respect to safety criteria and their future use for power up-rating. The best estimate simulation with uncertainty analysis constitutes the framework of the present study which is to apply the last version of the French best estimate computer code CATHARE 2 in order to predict the thermal-hydraulic phenomena in the Indian KudanKulam Nuclear Power Plant (KK NPP) with VVER-1000 reactors during LB LOCA and to evaluate uncertainty along with sensitivity studies using the IRSN methodology. The paper first describes the modeling aspects of LB LOCA with CATHARE and then it presents the basic results. It highlights the use of SUNSET statistical tool developed by IRSN for sampling, management of several runs using CATHARE and further post treatment for uncertainty and sensitivity evaluation. The paper also deals with the difficulties associated with the selection of input uncertainties, code applicability and discusses the challenges in uncertainty evaluation.

Uncertainty Analysis Becoming Common Practice in Safety Assessment

2010 ◽  
Author(s):  
Horst G. Glaeser
Keyword(s):  
Uncertainty Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Evaluation Model ◽  
Computer Code ◽  
Uncertainty Evaluation ◽  
Acceptance Criteria ◽  
Model Calculations ◽  
Uncertainty Range ◽  
Best Estimate

During the recent years an increasing interest in computational reactor safety analysis is to replace the conservative evaluation model calculations by best estimate calculations supplemented by uncertainty analysis of the code results. The evaluation of the margin to acceptance criteria, e.g. the maximum fuel rod clad temperature, should be based on the upper limit of the calculated uncertainty range. For example, due to power increase, licensing limits are approached. Therefore, regulators are looking closer on the way, how calculations are performed to meet these acceptance criteria. Methods have been developed and presented to quantify the uncertainty of computer code results. They are briefly presented in this paper. The present overview considers the international situation of development of uncertainty evaluation of computer code results and their application in licensing. Best estimate analysis plus uncertainty evaluation is used in licensing up to now in approximately seven countries. Demonstrations of applying uncertainty methods have been performed in nine additional countries at least. Most organizations use statistical methods. One statistical method is the GRS method proposing ordered statistics. Several demonstrations to apply the GRS method have been performed by GRS for courses of events in the nuclear steam supply system, calculating experiments as well as nuclear power plants. The method has also been applied for post test calculations of containment behavior, as well as severe accidents. One of the most important conclusions is that care must be exercised in determining ranges and probability distributions of the uncertain input parameters.

CANDU 6 Accident Analysis Using RELAP/SCDAPSIM With the Integrated Uncertainty Package

2018 ◽  
Author(s):  
Roxana-Mihaela Nistor-Vlad ◽  
Daniel Dupleac ◽  
Ilie Prisecaru ◽  
Chris Allison ◽  
M. Perez-Ferragut ◽  
...  
Keyword(s):  
Uncertainty Analysis ◽  
Nuclear Power ◽  
Natural Circulation ◽  
Probabilistic Approach ◽  
Uncertainty Evaluation ◽  
Accident Analysis ◽  
Complete Uncertainty ◽  
Best Estimate ◽  
Source Correlation ◽  
Main Steam

RELAP/SCDAPSIM is a best-estimate nuclear tool designed to analyze the behaviour of reactor systems during normal and accident conditions. Three main versions of RELAP/SCDAPSIM are currently used by program members and licensed users to support a variety of activities. RELAP/SCDAPSIM/MOD3.4 is the version of the code used by licensed users and program members for critical applications such as research reactors and nuclear power plant applications. Even though the code was initially designed for LWRs (Light Water Reactors), Politehnica University of Bucharest demonstrated the applicability of the RELAP/SCDAPSIM code for CANDU (CANada Deuterium Uranium) reactors analyses, by simulating some of the most important postulated accident transients (i.e. large break LOCA, main steam line break, the natural circulation in the heat transport system). Current trends refer to the BEPU (Best Estimate Plus Uncertainty) approach in the safety analysis of nuclear reactors. BEPU is a modern and technically consistent approach has been built upon best estimate methods including an evaluation of the uncertainty in the calculated results. ISS and UPC started the development of an uncertainty evaluation package to RELAP/SCDAPSIM/MOD4.0 code version which is currently implemented in MOD3.4 version of the code also. The uncertainty evaluation capability is implemented as an alternative run mode, the “uncertainty” mode, which allows the automatic execution of an uncertainty analysis based on the probabilistic approach. A complete uncertainty analysis using RELAP/SCDAPSIM/MOD3.4 code requires the execution of three related phases, namely the “setup” phase, the “simulation” phase consisting of several executions, and the “postprocessing” phase. The uncertainty data has to be supplied for the two types of parameters, the “input treatable” and the “source correlation” quantities. The required information is the probability distribution function and its characteristic parameters. This paper is mainly focused on the application of the uncertainty package in CANDU reactors accident analysis and it describes the steps to perform the uncertainty analysis, the uncertainty selected parameters (input treatable and source correlation parameters), the calculation results with RELAP/SCDAPSIM and some conclusions.

Comparison of Thermal Hydraulic Simulations of Beyond Design Basis Events Using the MAAP4 and CENTS Computer Codes

2009 ◽  
Author(s):  
N. Reed LaBarge ◽  
Barbara R. Baron ◽  
Raymond E. Schneider ◽  
Mathew C. Jacob
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Computer Code ◽  
Hydraulic Modeling ◽  
Severe Accident ◽  
Pressurized Water ◽  
Design Basis ◽  
Best Estimate ◽  
Water Reactors ◽  
Thermal Hydraulic Modeling

The MAAP4 computer code (Reference 1) is often used to perform thermal hydraulic simulations of severe accident sequences for nuclear power plant Probabilistic Risk Assessments (PRAs). MAAP4 can be used to simulate accidents for both Boiling Water Reactors (BWRs) as well as Pressurized Water Reactors (PWRs). This assessment employs MAAP 4.0.6a for PWRs (References 1 and 5), which incorporates explicit thermal hydraulic modeling of the Reactor Coolant System (RCS) and Steam Generators (SGs), along with a nodalized integrated containment model. In the PRA environment, MAAP4 has been used for applications such as the development of PRA Level 1 and Level 2 success criteria and human action timings. The CENTS computer code (Reference 2) is a simulation tool that is typically used to analyze non-Loss of Coolant Accident (non-LOCA) events postulated to occur in nuclear power plants incorporating Combustion Engineering (CE) and Westinghouse Nuclear Steam Supply System (NSSS) designs. It is licensed by the NRC perform design basis non-LOCA safety analyses. It is a best estimate code which uses detailed thermal hydraulic modeling of the RCS and SGs; however, it does not model the containment performance. It is used to perform a wide spectrum of licensing and best estimate non-LOCA event analysis and has the capability to simulate operator actions. The CENTS models are the basis for several full scope simulators in the industry. The purpose of the analyses described in this paper is to compare MAAP4 and CENTS predictions for the Station Blackout (SBO) and Total Loss of Feedwater (TLOFW) scenarios for a representative PWR in the Westinghouse fleet that employs a CE NSSS design. The results of this comparison are used to highlight postulated MAAP4 user challenges and assist in developing guidance on selecting MAAP4 parameters for use in these scenarios. The results of the analyses presented in this paper indicate several useful insights. Overall, this paper shows that when care is taken to normalize the MAAP4 and CENTS primary side natural circulation flowrate and SG modeling, the trends of the MAAP4 and CENTS predictions of core uncovery agree reasonably well.

scholarly journals Framework and Strategies for the Introduction of Best Estimate Models Into the Licensing Process

2009 ◽  
Author(s):  
Calogero Sollima ◽  
Gianni Petrangeli ◽  
Francesco D’Auria ◽  
Jozef Misˇa´k
Keyword(s):  
Nuclear Power ◽  
Input Data ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Best Estimate ◽  
Computer Codes ◽  
The University ◽  
Phd Thesis ◽  
Uncertainty Of Results ◽  
Made In

The current efforts to assure stable, safe and competitive operation of nuclear power plants go together with advances made in accident analysis domain where the deterministic safety analysis is an important instrument for confirming the adequacy and efficiency of provisions for the safety of nuclear power plants. Recently made advances offer two acceptable options for demonstrating that the safety is ensured with sufficient margin: use of best estimate (BE) computer codes either combined with conservative input data or with realistic input data but associated with evaluation of uncertainty of results. The objective, proposed for the present work, is the implementation of the best estimate plus uncertainty (BEPU) method into the licensing process, as developed by the University of Pisa in a PhD thesis. In the thesis activity, more emphasis is given to the study of “input and method” of BE calculations than on “uncertainty evaluation”. In the present paper, a summary of the results achieved is reported.

scholarly journals Mathematical Formulation and Analytic Solutions for Uncertainty Analysis in Probabilistic Safety Assessment of Nuclear Power Plants

Energies ◽  
2021 ◽  
Vol 14 (4) ◽  
pp. 929
Author(s):  
Gyun Seob Song ◽  
Man Cheol Kim
Keyword(s):  
Monte Carlo ◽  
Uncertainty Analysis ◽  
Probability Density ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Mathematical Formulation ◽  
Analytic Solutions ◽  
Probabilistic Safety Assessment ◽  
Probabilistic Safety

Monte Carlo simulations are widely used for uncertainty analysis in the probabilistic safety assessment of nuclear power plants. Despite many advantages, such as its general applicability, a Monte Carlo simulation has inherent limitations as a simulation-based approach. This study provides a mathematical formulation and analytic solutions for the uncertainty analysis in a probabilistic safety assessment (PSA). Starting from the definitions of variables, mathematical equations are derived for synthesizing probability density functions for logical AND, logical OR, and logical OR with rare event approximation of two independent events. The equations can be applied consecutively when there exist more than two events. For fail-to-run failures, the probability density function for the unavailability has the same probability distribution as the probability density function (PDF) for the failure rate under specified conditions. The effectiveness of the analytic solutions is demonstrated by applying them to an example system. The resultant probability density functions are in good agreement with the Monte Carlo simulation results, which are in fact approximations for those from the analytic solutions, with errors less than 12.6%. Important theoretical aspects are examined with the analytic solutions such as the validity of the use of a right-unbounded distribution to describe the uncertainty in the unavailability/probability. The analytic solutions for uncertainty analysis can serve as a basis for all other methods, providing deeper insights into uncertainty analyses in probabilistic safety assessment.

Enhanced Verification for RELAP5-3D Parameter and Sensitivity Studies

2016 ◽  
Author(s):  
George L. Mesina ◽  
Nolan Anderson
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
User Interaction ◽  
Computer Code ◽  
Unintended Consequences ◽  
Physical Models ◽  
Case Processing ◽  
Time Step ◽  
Sensitivity Studies ◽  
Code Changes

The RELAP5-3D1 program solves a complex system of governing, closure and special process equations to model the underlying physics of nuclear power plants. For SQA (software quality assurance), the code must be verified and validated (V&V) to ensure proper performance before release to users. The physical models are validated against data from experiments and plants and verified against specifications for the computer code. In addition to physics, programs such as RELAP5-3D perform numerous other functions and processes that should also be checked to guarantee correct results. Functions include input, output, data management, and user interaction, while processes include restart, time-step backup, code coupling, and multi-case processing. Previous articles have covered the verification of the physical models, restart, and backup through extremely accurate and automated sequential verification applied on a comprehensive suite of test cases to ensure that code changes produced no unintended consequences. New developments have enabled the verification of multi-case and multi-deck processing. These features are frequently used in parameter and code sensitivity studies and therefore must be verified as working correctly. Both theory and application are presented.

The Best-Estimate Plus Uncertainty (BEPU) Challenge in the Licensing of Current Generation of Reactors

2011 ◽  
Author(s):  
A. Petruzzi ◽  
N. Muellner ◽  
F. D’Auria ◽  
O. Mazzantini
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Current Generation ◽  
Computational Tools ◽  
Best Estimate ◽  
Safety Margins ◽  
Heavy Water Reactor ◽  
Key Steps ◽  
Uncertainty Method ◽  
Selection Of

Within the licensing process of the Atucha II PHWR (Pressurized Heavy Water Reactor) the BEPU (Best Estimate Plus Uncertainty) approach has been selected for issuing of the Chapter 15 on FSAR (Final Safety Analysis Report). The key steps of the entire process are basically two: a) the selection of PIE (Postulated Initiating Events) and, b) the analysis by best estimate models supported by uncertainty evaluation. Otherwise, key elements of the approach are: 1) availability of qualified computational tools including suitable uncertainty method; 2) demonstration of quality; 3) acceptability and endorsement by the licensing authority. The effort of issuing Chapter 15 is terminated at the time of issuing of the present paper and the safety margins available for the operation of the concerned NPP (Nuclear Power Plant) have been quantified.

Best Estimate Plus Uncertainty Analysis of LBLOCA for a Pickering B CANDU Reactor

2010 ◽  
Author(s):  
Larry Blake ◽  
George Gavrus ◽  
Jack Vecchiarelli ◽  
J. Stoklosa
Keyword(s):  
Power Plants ◽  
Safety Margin ◽  
Reactor Power ◽  
Natural Uranium ◽  
Nuclear Industry ◽  
Initial Increase ◽  
Power Pulse ◽  
Pulse Phase ◽  
Best Estimate ◽  
Safety Margins

The Pickering B Nuclear Generating Station consists of four CANDU reactors. These reactors are horizontal pressure tube, heavy water cooled and moderated reactors fuelled with natural uranium. Under a postulated large break loss of coolant accident (LOCA), positive reactivity results from coolant void formation. The transient is terminated by the operation of the safety systems within approximately 2 seconds of the start of the transient. The initial increase in reactor power, terminated by the action of the safety system, is termed the power pulse phase of the accident. In many instances the severity of an LBLOCA can be characterized by the adiabatic energy deposited to the fuel during this phase of the accident. Historically, Limit of Operating Envelope (LOE) calculations have been used to characterize the severity of the accident. LOE analyses are conservative analyses in which the key operational and safety related parameters are set to conservative or limiting values. Limit based analyses of this type result in calculated transient responses that will differ significantly from the actual expected response of the station. As well, while the results of limit calculations are conservative, safety margins and the degree of conservatism is generally not known. As a result of these factors, the use of Best Estimate Plus Uncertainty (BEPU) analyses in safety analyses for nuclear power plants has been increasing. In Canada, the nuclear industry has been pursuing best estimate analysis through the BEAU (Best Estimate Analysis and Uncertainty) methodology in order to obtain better characterization of the safety margins. This approach is generally consistent with those used internationally. Recently, a BEAU analysis of the Pickering B NGS was completed for the power pulse phase of a postulated Large Break LOCA. The analysis comprised identification of relevant phenomena through a Phenomena Identification and Ranking (PIRT) process, assessment of the code input uncertainties, sensitivity studies to quantify the significance of the input parameters, generation of a functional response surface and its validation, and determination of the safety margin. The results of the analysis clearly demonstrate that the Limit of Operating Envelope (LOE) results are significantly conservative relative to realistic analysis even when uncertainties are considered. In addition, the extensive sensitivity analysis performed to supplement the primary result provides insight into the primary contributors to the results.

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