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.

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.

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.

The BEMUSE Programme: Results of the First Part Concerning the LOFT L2-5 Test

2006 ◽  
Author(s):  
Agne`s de Cre`cy ◽  
Pascal Bazin ◽  
Francesco D’Auria ◽  
Alessandro Petruzzi ◽  
Yong-Ho Ryu
Keyword(s):  
Probabilistic Approach ◽  
Sensitivity Analyses ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Sensitivity Evaluation ◽  
Loss Of Coolant ◽  
Regression Techniques ◽  
Uncertain Input ◽  
The University ◽  
Range Of Variation

This paper is aimed at describing results of the first part of the BEMUSE (Best Estimate Methods – Uncertainty and Sensitivity Evaluation) programme. The purpose of BEMUSE is the comparison of best-estimate calculations, followed by the comparison of uncertainty and sensitivity analyses for a Large Break Loss of Coolant Accident (LB-LOCA). The first part of the programme is devoted to the study of the LOFT L2-5 experiment. After a general presentation of the programme, which implies more than ten participants, this paper describes the qualification process and the results of the best-estimate calculations. The results are significantly less dispersed than those of the ISP-13, concerning already LOFT L2-5 more than 20 years ago. Then, it presents extensively the methods and the results of uncertainty and sensitivity analyses. All the participants, apart from the University of Pisa with the CIAU method, use a fully probabilistic approach, based on Wilks’ formula. However, differences appear for the choice of the uncertain input parameters to be considered and for their associated range of variation. Sensitivity analysis is performed with regression techniques, and the results are also compared. As a conclusion, main lessons learnt from BEMUSE and recommendations are presented.

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.

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

Effect of Nitrogen Release From Accumulators on PWR LOCA Analysis

2002 ◽  
Author(s):  
Woon-Shing Yeung ◽  
Ramu K. Sundaram
Keyword(s):  
Coming Out ◽  
Computer Code ◽  
Nitrogen Release ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Concentration Difference ◽  
Dissolved Nitrogen ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Release Model

The accumulator in a Pressurized Water Reactor (PWR) is generally pressurized with inert nitrogen cover gas, and the accumulator water will be saturated with nitrogen. Nitrogen released due to system depressurization during a Loss-of-Coolant Accident (LOCA) transient, consists of the nitrogen that is in the gas phase as well as nitrogen coming out of the liquid from a dissolved state. The effect of nitrogen release from the accumulator on the accident sequence is generally not explicitly addressed in typical LOCA analyses. This paper presents an analytical nitrogen release model and its incorporation into the RELAP5/MOD3 computer code. The model predicts the amount of nitrogen release as a function of concentration difference between the actual and equilibrium conditions, and can track its subsequent transport through the downstream reactor coolant system in a LOCA transient. The model is compared to data from discharge tests with a refrigerant type fluid, pressurized with nitrogen. The results demonstrate that the model is able to calculate the release of the dissolved nitrogen as designed. The modified computer code has been applied to analyze the discharge from a typical PWR accumulator. The results are compared to those obtained without the nitrogen release model. The effect of nitrogen release on major system parameters, including accumulator level, accumulator flow rate, and accumulator pressure, is discussed.

Large-Break Loss-of-Coolant Accident Phenomena Identification and Ranking Table (PIRT) for the Advanced CANDU Reactor

2004 ◽  
Author(s):  
N. Popov ◽  
H. E. Sills ◽  
V. G. Snell ◽  
B. Boyack ◽  
V. J. Langman
Keyword(s):  
Heavy Water ◽  
Computer Code ◽  
Natural Uranium ◽  
Summary Table ◽  
Regulatory Review ◽  
Loss Of Coolant Accident ◽  
Unit Energy ◽  
Wide Range ◽  
Loss Of Coolant ◽  
Enriched Uranium

The Advanced CANDU Reactor (ACR™)* is an evolutionary advancement of the current CANDU 6® reactor, aimed at producing electrical power for a capital cost and unit-energy cost significantly less than that of current reactor designs. The ACR retains the modular concept of horizontal fuel channels surrounded by heavy water moderator, as with all CANDU reactors. However, ACR uses slightly enriched uranium (SEU) fuel, compared to the natural uranium used in CANDU 6. This achieves the twin goals of improved economics (e.g., via reductions in the heavy water requirements and the use of a light water coolant), as well as improved safety. This paper is focused on the double-ended guillotine critical inlet header break (CRIHB) loss-of-coolant accident (LOCA) in an ACR reactor, which is considered as a large break LOCA. Large Break LOCA in water-cooled reactors has been used historically as a design basis event by regulators, and it has attracted a very large share of safety analysis and regulatory review. The LBLOCA event covers a wide range of system behaviours and fundamental phenomena. The Phenomena Identification and Ranking Table (PIRT) for LBLOCA therefore provides a good understanding of many of the safety characteristics of the ACR design. The paper outlines the design characteristics of the ACR reactor that impact the PIRT process and computer code applicability. It also describes the LOCA phenomena, lists all components and systems that have an important role during the event, discusses the PIRT process and results, and presents the final PIRT summary table.

Performance of a Spray Pond for Nuclear Power Plant Ultimate Heat Sink

1977 ◽  
Vol 99 (4) ◽  
pp. 650-656
Author(s):  
V. E. Schrock ◽  
G. J. Trezek ◽  
L. R. Keilman
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Thermal Performance ◽  
Heat Sink ◽  
Nuclear Power ◽  
Field Tests ◽  
Computer Code ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Transient Thermal

Spray ponds have become an attractive method of providing the “ultimate heat sink”, i.e., the assured means of dissipating heat from a nuclear power plant. Two redundant spray ponds were the choice for this service in the Rancho Seco Nuclear Generating Station owned by Sacramento Municipal Utility District. This paper describes the results of full scale field tests of the Rancho Seco ponds which were conducted to verify the thermal performance, drift loss characteristics, and the capability to sustain the cooling requirements for a period of 30 days following a loss-of-coolant accident (LOCA). Correlations of local and average nozzle efficiency and of the drift loss are presented. A computer code was developed for the transient thermal performance of the pond. After verification the code was used to predict performance following LOCA under adverse meteorological conditions based on weather records.

Sign in / Sign up

Export Citation Format

Share Document