Methodology of Internal Assessment of Uncertainty and Extension to Neutron Kinetics/Thermal-Hydraulics Coupled Codes

Petruzzi, A.; D'Auria, Francesco Saverio; Giannotti, W.; Ivanov, Kostadin

doi:10.13182/nse04-26

Cited by 16 publications

(12 citation statements)

References 9 publications

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“…The contemporaneous acceptability of the accuracy (step of the process connected with experiments in ITF) and of the similarity analysis (step of the process connected with NPP) constitutes the positive demonstration of the code capability and the end of the code assessment. The calculated accuracy is possibly included in the data base suitable for uncertainty evaluation (block 6 in Figure 2, [12,13]). If the accuracy is not in the range of acceptability or the code fails the similarity analysis, the code is considered not qualified and the code-development team will be informed in order to develop new code models or to improve the existing ones.…”

Section: Code Qualificationmentioning

confidence: 99%

“…(8) A special attention from the scientific community has always been given to the quantification of code uncertainty in predicting plant transients. Methodologies to evaluate the "uncertainty" have been proposed [12,13] and tested in several international activities, like UMS (uncertainty method study, [14]) and BEMUSE (bestestimate methods-uncertainty and sensitivity evaluation, [15,16]) that allowed the comparison of uncertainty results obtained from different methodologies.…”

Section: Introductionmentioning

confidence: 99%

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Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures

Petruzzi

D'Auria

2008

Science and Technology of Nuclear Installations

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In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Whereas the first system codes, developed at the beginning of the 1970s, utilized the homogenous equilibrium model with three balance equations to describe the two-phase flow, nowadays the more advanced system codes are based on the so-called “two-fluid model” with separation of the water and vapor phases, resulting in systems with at least six balance equations. The wide experimental campaign, constituted by the integral and separate effect tests, conducted under the umbrella of the OECD/CSNI was at the basis of the development and validation of the thermal-hydraulic system codes by which they have reached the present high degree of maturity. However, notwithstanding the huge amounts of financial and human resources invested, the results predicted by the code are still affected by errors whose origins can be attributed to several reasons as model deficiencies, approximations in the numerical solution, nodalization effects, and imperfect knowledge of boundary and initial conditions. In this context, the existence of qualified procedures for a consistent application of qualified thermal-hydraulic system code is necessary and implies the drawing up of specific criteria through which the code-user, the nodalization, and finally the transient results are qualified.

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Section: Code Qualificationmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures

Petruzzi

D'Auria

2008

Science and Technology of Nuclear Installations

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“…The second approach reviewed as the propagation of code calculation output errors, is representatively illustrated by the Uncertainty Method based upon Accuracy Extrapolation 22) (UMAE) 'embedded' into a Code with the capability of Internal Assessment of Uncertainty 23,24) (CIAU) in Fig. 4.…”

Section: Propagation Of Code Calculation Output Errorsmentioning

confidence: 99%

Role of Best Estimate Plus Uncertainty Methods in Major Nuclear Power Plant Modifications

Andrea¹,

Petruzzi²

2010

Journal of Nuclear Science and Technology

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Current industrial trends to increase power production challenge the initial safety design limits of plants, which were evaluated generally using conservative tools and hypotheses. Advanced numerical tools and methods allow demonstrating that safety margins are still respected. These tools are modern fully validated thermal-hydraulic codes, coupled thermal-hydraulic/neutron-kinetic codes, and methodologies that use realistic hypotheses (rather than conservative ones) and estimate the uncertainty. The paper illustrates modern computational tools and approaches for improving plant operation and control and for supporting design modifications on existing nuclear power plants. A survey performed by Joint Research Centre's Institute for Energy is also presented and summarizes the feedback received by several countries of the European Union on the impact of modern Thermal-Hydraulics and Neutron-Kinetics tools on Operational Limits and Conditions and Operating Instructions and Procedures.

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“…The second approach (Figure 8), reviewed as the propagation of code output errors, is representatively illustrated by the UMAE-CIAU (uncertainty method based upon accuracy extrapolation [15] "embedded" into the code with capability of internal assessment of uncertainty [16,17]). Note that this class of methods includes only a few applications from industry.…”

Section: Approaches For Computing Uncertaintymentioning

confidence: 99%

Approaches, Relevant Topics, and Internal Method for Uncertainty Evaluation in Predictions of Thermal-Hydraulic System Codes

Petruzzi¹,

D'Auria²

2008

Science and Technology of Nuclear Installations

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The evaluation of uncertainty constitutes the necessary supplement of best-estimate calculations performed to understand accident scenarios in water-cooled nuclear reactors. The needs come from the imperfection of computational tools, on the one side, and the interest in using such a tool to get more precise evaluation of safety margins. The paper reviews the salient features of three independent approaches for estimating uncertainties associated with predictions of complex system codes. Namely, the propagations of code input error and calculation output error constitute the keywords for identifying the methods of current interest for industrial applications, while the adjoint sensitivity-analysis procedure and the global adjoint sensitivity-analysis procedure, extended to performing uncertainty evaluation in conjunction with concepts from data adjustment and assimilation, constitute the innovative approach. Throughout the developed methods, uncertainty bands can be derived (both upper and lower) for any desired quantity of the transient of interest. For one case, the uncertainty method is coupled with the thermal-hydraulic code to get the code with capability of internal assessment of uncertainty, whose features are discussed in more detail.

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Methodology of Internal Assessment of Uncertainty and Extension to Neutron Kinetics/Thermal-Hydraulics Coupled Codes

Cited by 16 publications

References 9 publications

Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures

Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures

Role of Best Estimate Plus Uncertainty Methods in Major Nuclear Power Plant Modifications

Approaches, Relevant Topics, and Internal Method for Uncertainty Evaluation in Predictions of Thermal-Hydraulic System Codes

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