The Application of Best Estimate and Uncertainty Analysis Methodology to Large LOCA Power Pulse in a CANDU 6 Reactor

2002 ◽  
Author(s):  
A. Abdul-Razzak ◽  
J. Zhang ◽  
H. E. Sills ◽  
L. Flatt ◽  
D. Jenkins ◽  
...  
Keyword(s):  
Statistical Analysis ◽  
Uncertainty Analysis ◽  
Large Loss ◽  
Power Pulse ◽  
Analysis Methodology ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Safety Margins ◽  
Loss Of Coolant ◽  
Two Phases

The paper describes briefly a best estimate plus uncertainty analysis (BE+UA) methodology and presents its prototyping application to the power pulse phase of a limiting large Loss-of-Coolant Accident (LOCA) for a CANDU 6 reactor fuelled with CANFLEX® fuel. The methodology is consistent with and builds on world practice [1], [2]. The analysis is divided into two phases to focus on the dominant parameters for each phase and to allow for the consideration of all identified highly ranked parameters in the statistical analysis and response surface fits for margin parameters. The objective of this analysis is to quantify improvements in predicted safety margins under best estimate conditions.

Identification and Ranking of Phenomena Leading to Peak Cladding Temperatures in Boiling Water Reactors During Large Break Loss of Coolant Accident Transients

2002 ◽  
Author(s):  
Ruwan K. Ratnayake ◽  
S. Ergun ◽  
L. E. Hochreiter ◽  
A. J. Baratta
Keyword(s):  
Uncertainty Analysis ◽  
Nuclear Reactors ◽  
Computer Code ◽  
Boiling Water Reactors ◽  
Validation Process ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Water Reactors ◽  
Accident Scenarios

In the licensing and validation process of best estimate codes for the analysis of nuclear reactors and postulated accident scenarios, the identification and quantification of the calculational uncertainty is required. One of the most important aspects in this process is the identification and recognition of the crucial contributing phenomena to the overall code uncertainty. The establishment of Phenomena Identification and Ranking Tables (PIRT) provides a vehicle to assist in assessing the capabilities of the computer code, and to guide the uncertainty analysis of the calculated results. The process used in this work to identify the phenomena was reviewing both licensing and best estimate calculations, as well as experiments, which had been performed for BWR LOCA analyses. The initial PIRT was developed by a group of analysts and was compared to existing BWR LOCA PIRTs as well as BWR LOCA analyses. The initial PIRT was then independently reviewed by a second panel of experts for the selected ranking of phenomena, identification of phenomena which were ignored, as well as the basis and rationale for the ranking of the phenomena. The differences between the two groups were then resolved. PIRTs have been developed for BWR types 4 and5/6 for the Large Break Loss of Coolant Accidents (LB-LOCA). The ranking and the corresponding rationale for each phenomenon is included in tables together with the assessed uncertainty of the code capability to predict the phenomena.

Response Surface Modeling of a Large Break Loss of Coolant Accident With Comparisons to Order Statistics Approaches

2010 ◽  
Author(s):  
Andrew B. French
Keyword(s):  
Order Statistics ◽  
Response Surface ◽  
Response Surface Model ◽  
Evaluation Methodology ◽  
Surface Model ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Safety Margins ◽  
Loss Of Coolant ◽  
Pipe Break

NUREG/CR-5249 “Quantifying Reactor Safety Margins Application of Code Scaling, Applicability, and Uncertainty Evaluation Methodology to a Large Break, Loss-of-Coolant Accident” provides the general methodologies to be used in the development of realistic loss of coolant safety analyses. The objective of this paper is to start with NUREG/CR-5249 and develop a modified methodology. The modified approach will include a response surface, model adequacy checks, and development of the 95/95% confidence peak clad temperature cumulative distributions function. The response surface model will then be used to develop simulated results and conclusions about the order statistics best estimate approach. All work is conducted using a verified safety analysis input deck and RELAP as the thermal hydraulic best estimate analysis code. The objective of the order statistics comparison is to investigate the number of cases in which the maximum PCT, in a simulated order statistics approach, falls below the 95th percentile value of the distribution and to assess the standard deviation in the maximum peak clad temperature of order statistics sets. Although order statistics may be a more economic approach to satisfying regulatory requirements, response surface models have several benefits that can complement the use of order statistics. The primary benefit is the insight gained into which parameters are most important in determining the peak clad temperature. This is of particular value to the licensee in convincing the regulator that its analysis is robust. The disadvantage is the number of runs required to develop the models. If we examine the main effects, the most significant input parameter is pipe break size. In support of a proposed modification to 10CFR50.46, the U.S. Nuclear Regulatory Commission undertook an expert elicitation to assess the change in frequency of pipe break accidents as a function of break size. The result of that elicitation was a probability density function that decreases approximately as (pipe diameter)−1.5 in the region of large pipe diameters. Because break diameter is shown to be such a large contributor to PCT by the response surface, it is evident that calculated PCT could be substantially reduced if credit were given for this form of the uncertainty distribution rather than for the flat distribution used in the analysis (and industry).

scholarly journals GRS Method for Uncertainty and Sensitivity Evaluation of Code Results and Applications

2008 ◽  
Vol 2008 ◽  
pp. 1-7 ◽  
Author(s):  
Horst Glaeser
Keyword(s):  
Uncertainty Analysis ◽  
Evaluation Model ◽  
Computer Code ◽  
Model Calculations ◽  
Uncertainty Range ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Sensitivity Evaluation ◽  
Loss Of Coolant ◽  
Acceptance Limits

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, for example, the maximum fuel rod clad temperature, should be based on the upper limit of the calculated uncertainty range. Uncertainty analysis is needed if useful conclusions are to be obtained from “best estimate” thermal-hydraulic code calculations, otherwise single values of unknown accuracy would be presented for comparison with regulatory acceptance limits. Methods have been developed and presented to quantify the uncertainty of computer code results. The basic techniques proposed by GRS are presented together with applications to a large break loss of coolant accident on a reference reactor as well as on an experiment simulating containment behaviour.

scholarly journals PARAMETRIC STUDY OF LOCA IN TRIGA-2000 USING RELAP5/SCDAP CODE

2017 ◽  
Vol 19 (2) ◽  
pp. 59 ◽  
Author(s):  
Anhar Riza Antariksawan ◽  
Surip Widodo ◽  
Hendro Tjahjono
Keyword(s):  
Cooling System ◽  
Fuel Temperature ◽  
Blow Down ◽  
Beam Tube ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling ◽  
Dry Out

A postulated loss of coolant accident (LOCA) shall be analyzed to assure the safety of a research reactor. The analysis of such accident could be performed using best estimate thermal-hydraulic codes, such as RELAP5. This study focuses on analysis of LOCA in TRIGA-2000 due to pipe and beam tube break. The objective is to understand the effect of break size and the actuating time of the emergency core cooling system (ECCS) on the accident consequences and to assess the safety of the reactor. The analysis is performed using RELAP/SCDAPSIM codes. Three different break size and actuating time were studied. The results confirmed that the larger break size, the faster coolant blow down. But, the siphon break holes could prevent the core from risk of dry out due to siphoning effect in case of pipe break. In case of beam tube rupture, the ECCS is able to delay the fuel temperature increased where the late actuation of the ECCS could delay longer. It could be concluded that the safety of the reactor is kept during LOCA throughout the duration time studied. However, to assure the integrity of the fuel for the long term, the cooling system after ECCS last should be considered.  Keywords: safety analysis, LOCA, TRIGA, RELAP5 STUDI PARAMETRIK LOCA DI TRIGA-2000 MENGGUNAKAN RELAP5/SCDAP. Kecelakaan kehilangan air pendingin (LOCA) harus dianalisis untuk menjamin keselamatan suatu reaktor riset. Analisis LOCA dapat dilakukan menggunakan perhitungan best-estimate seperti RELAP5. Penelitian ini menekankan pada analisis LOCA di TRIGA-2000 akibat pecahnya pipa dan tabung berkas. Tujuan penelitian adalah memahami efek ukuran kebocoran dan waktu aktuasi sistem pendingin teras darurat (ECCS) pada sekuensi kejadian dan mengkaji keselamatan reaktor. Analisis dilakukan menggunakan program perhitungan RELAP/SCDAPSIM. Tiga ukuran kebocoran dan waktu aktuasi ECCS berbeda dipilih sebagai parameter dalam studi ini.  Hasil perhitungan mengonfirmasi bahwa semakin besar ukuran kebocoran, semakin cepat pengosongan tangki reaktor. Lubang siphon breaker dapat mencegah air terkuras dalam hal kebocoran pada pipa. Sedang dalam hal kebocoran pada beam tube, ECCS mampu memperlambat kenaikan temperatur bahan bakar. Dari studi ini dapat disimpulkan bahwa keselamatan reaktor dapat terjaga pada kejadian LOCA, namun pendinginan jangka panjang perlu dipertimbangkan untuk menjaga integritas bahan bakar.Kata kunci: analisis keselamatan, LOCA, TRIGA, RELAP5

scholarly journals Fuel R&D Needs and Strategy towards a Revision of Acceptance Criteria

2010 ◽  
Vol 2010 ◽  
pp. 1-7 ◽  
Author(s):  
François Barré ◽  
Claude Grandjean ◽  
Marc Petit ◽  
Jean-Claude Micaelli
Keyword(s):  
Loss Of Coolant Accident ◽  
Safety Margins ◽  
Power Load ◽  
Higher Power ◽  
Loss Of Coolant ◽  
Advanced Fuel ◽  
Water Reactors ◽  
Improved Performance ◽  
Irradiated Fuel ◽  
Cladding Material

The study of fuel behaviour under accidental conditions is a major concern in the safety analysis of the Pressurised Water Reactors. The consequences of Design Basis Accidents, such as Loss of Coolant Accident and Reactivity Initiated Accident, have to be quantified in comparison to the safety criteria. Those criteria have been established in the 1970s on the basis of experiments performed with fresh or low irradiated fuel. Starting in the 1990s, the increased industrial competition and constraints led utilities to use fuel in more and more aggressive conditions (higher discharge burnup, higher power, load follow, etc.) and create incentive conditions for the development of advanced fuel designs with improved performance (new fuel types with additives, cladding material with better resistance to corrosion, etc.). These long anticipated developments involved the need for new investigations of irradiated fuel behaviour in order to check the adequacy of the current criteria, evaluate the safety margins, provide new technical bases for modelling and allow an evolution of these criteria. Such an evolution is presently under discussion in France and several other countries, in view of a revision in the next coming years. For this purpose, a R&D strategy has been defined at IRSN.

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.

scholarly journals Safety Evaluation of Silicon Carbide and Zircaloy-4 Cladding during a Large-Break Loss-of-Coolant Accident

Energies ◽  
2018 ◽  
Vol 11 (12) ◽  
pp. 3324 ◽  
Author(s):  
Kwangwon Ahn ◽  
Kyohun Joo ◽  
Sung-Pil Park
Keyword(s):  
Silicon Carbide ◽  
Ultimate Strength ◽  
Safety Margin ◽  
Safety Evaluation ◽  
Loss Of Coolant Accident ◽  
Safety Margins ◽  
Loss Of Coolant

In this study, we aim to conduct structural analyses of cladding materials, such as silicon carbide and zircaloy-4, during a Large-Break Loss-of-Coolant Accident. The safety margin is the key consideration regarding the performance of the cladding materials. Our study shows that, in terms of primary stresses, SiC has a greater safety margin than zircaloy-4 due to SiC having a higher yield and ultimate strength; the cladding outer pressure is not affected by the cladding materials and, thus, the primary stresses of all cladding materials are the same. However, for secondary stresses, zircaloy-4 has the smallest fluctuation and irradiated SiC recorded the largest; secondary stresses and temperature histories are material-dependent. Ultimately, both cladding materials were found to have sufficient safety margins with respect to primary and secondary stresses.

Investigation of Warm Prestress for the Case of Small ΔT During a Reactor Loss-of-Coolant Accident

1979 ◽  
Vol 101 (4) ◽  
pp. 298-304 ◽  
Author(s):  
F. J. Loss ◽  
R. A. Gray ◽  
J. R. Hawthorne
Keyword(s):  
Experimental Investigation ◽  
Statistical Analysis ◽  
Vessel Wall ◽  
Crack Extension ◽  
Crack Tip ◽  
Material Behavior ◽  
Virgin Material ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Failure Temperature

An experimental investigation was conducted to characterize the benefits of warm prestress (WPS) in limiting crack extension in the wall of a nuclear vessel during a LOCA-ECCS. The present research emphasized material behavior under conditions of a small ΔT between the temperature of WPS and the failure temperature as might occur during a LOCA. The results have demonstrated that fracture will not occur during a simultaneous unloading and cooling of the crack-tip region following WPS even though the critical KIc of the virgin material is achieved. Based on a statistical analysis, it is concluded that WPS produces an “effective” elevation in KIc; furthermore, it is suggested that this elevation will limit crack extension in the vessel wall so as to retain the coolant.

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