Analysis of the Uncertainty of Calculations of Loss-of-Coolant Accidents for the No. 1 Unit of the Kursk Nuclear Power Plant

Afremov, D. A.; Zhuravleva, Yu. V.; Mironov, Yu. V.; Nazarov, V. S.; Radkevich, V. E.; Yashnikov, D. A.

doi:10.1007/s10512-005-0223-5

Cited by 4 publications

(3 citation statements)

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“…2 unit of the Smolensk nuclear power plant with an RBMK-1000 reactor [7] reaches ±8%, which corresponds to σ ≈ 0.05. To sum up, the lower estimate of the value of σ even neglecting other factors affecting the uncertainty of the temperature of the fuel element cladding is σ ≈ 0.058 or almost six times greater than obtained computationally in [18]. Since the physically predetermined width of the confidence interval is almost six times greater than the computed values, the computational experiment could not show what it was supposed to show -it could not give a prediction of the physically validated uncertainties of the maximum temperature of the fuel-element cladding.…”

Section: Prediction Of the Temperature Of Fuel-element Cladding With mentioning

confidence: 71%

“…In a subsequent work by essentially the same group of authors [18], the final result of the computational investigations of the maximum cladding temperature for the two accident processes studied became the estimates of the confidence intervals for the temperature whose mathematical expectation lies in the range 888-920 K. The relative width of the computed 90% confidence intervals (interval width divided by the mathematical expectation) does not exceed 0.033. Since the total width of the 90% confidence interval is 3.2σ with an error not exceeding 5%, where σ is the standard error [2,3], for a wide class of high-entropy probability density distributions, the estimated relative value of σ for all calculations performed does not exceed 0.01 or 1%.…”

Section: Prediction Of the Temperature Of Fuel-element Cladding With mentioning

confidence: 99%

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Safety prediction in nuclear power

Rumyantsev

2007

At Energy

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The practical operational safety of nuclear objects is of fundamental importance for assessing the future prospects under discussion and selecting a strategy for the development of nuclear power. It is shown that the methods currently being used for making safety predictions do not contain an analysis of the unavoidable errors and uncertainties of the models used or the initial and boundary conditions under which the physical processes that develop into serious accidents arise and develop. It is proposed that the method of quantile estimates of the uncertainties, which is free of the drawbacks of the Monte Carlo method and which increases the reliability of safety predictions in nuclear power, be used.Considering life expectancy and the threat of serious accidents, the safety of operating nuclear objects is of fundamental importance for making assessments of the prospects and selecting a strategy for the development of nuclear power [1]. The methods which are currently used to make safety predictions are based on stationary and dynamic deterministic models which generalize past experience and, as a rule, do not contain an analysis of the unavoidable errors and uncertainties in the models or the initial and boundary conditions under which the physical processes that develop into serious accidents arise and develop. In contrast to building airplanes or rockets, the development of full-scale prototypes to investigate the safety of nuclear systems experimentally and to eliminate the inaccuracies in the theory and computational models is still considered to be too expensive. For this reason, safety predictions in nuclear power must be based on a generalization of the past limited experience in building and operating its objects. To increase their reliability, safety predictions must include explicitly an analysis of the possible uncertainties of the models used and of the external events which influence the risk and are associated with this field of human activity. However, analysis of the uncertainties only relatively recently became a subject of systematic study in attempts to find validated methods for taking into account explicitly the incompleteness of our knowledge in predictive assessments of the safety of objects in nuclear power.The objective of the present article is to analyze some acceptable methods for predicting the safety of the objects in nuclear power and to demonstrate the need for improving the existing approaches to simulating the physical processes in a manner that explicitly takes account of the incompleteness of our knowledge of the initial and boundary conditions under which processes that turn into serious accidents can arise and develop. The analysis is based on results obtained using the method of quantile estimates of the uncertainties [2][3][4]. This method encompasses the range from almost exact to sparse knowledge with the possibility that the ratio of the width of the 90% confidence interval to the median of the distribution varies over several orders of magnitude. The method operates w...

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Section: Prediction Of the Temperature Of Fuel-element Cladding With mentioning

confidence: 71%

Section: Prediction Of the Temperature Of Fuel-element Cladding With mentioning

confidence: 99%

Safety prediction in nuclear power

Rumyantsev

2007

At Energy

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“…The confidence interval for the correction factor for the computed critical heat flux is 0.864-0.885 with reliability 0.95. This interval is used as the range for the correction to the computed critical heat flux in the statistical analysis of the uncertainty of the parameters of the thermohydraulic calculations [8,9].…”

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confidence: 99%

Verification of relap5/mod3.2 critical heat transfer computational method by experiments performed on the ks facility

Afremov

Yashnikov

2011

At Energy

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The computational results obtained with the RELAP5/mod3.2 code for critical heat transfer are compared with experimental data obtained on the KS facility for RBMK fuel assembly models. The confidence interval is constructed for the correction factor to the computed critical heat flux.The method used in the RELAP5/mod3.2 code to calculate critical heat transfer for the RBMK-1000 reactor is verified. The models established on the KS facility (Kurchatov Institute) and used in 1969-1974 to study a crisis under RBMK conditions are closest to natural conditions [1]. For verification, 290 experimental points obtained on five sections were used.The method of [2] is used in RELAP5/mod3.2 to calculate the critical heat flux. This method consists of interpolating the critical heat flux in a three-dimensional space: pressure, mass flow, and relative entropy of the flow. The interpolation is performed according to a table with experimental data for tubes in terms of the diameter 8 mm. The table was constructed using 15000 experimental points and consists of a three-dimensional array of 4410 points, 15 values of the pressure in the range 0.1-20 MPa, 14 values of the mass flow in the range 0-7500 kg/(m 2 ·sec), and 21 values of the relative enthalpy of the flow --0.5-1. After interpolation the critical heat flux is multiplied by correction factors which take account of the following factors: diameter of the tube, difference of the critical heat flux for bundles of rods from the critical heat flux for tubes, the axial profile of the power release, and the change in the boundary layer at the entrance into a bundle and behind the spacing lattices.There are older versions of a simplified table for the critical heat flux that also take account of the difference in the critical heat fluxes of the first and second kinds [3][4][5]. The modified tables should be included in the computational codes. Unfortunately, a simplified 1986 table is still used even in the latest modifications of the RELAP5 code. However, it should be noted that the correction factor taking account of the difference of the critical heat flux for rod bundles from the critical heat flux for tubes has a larger effect on the discrepancy between the computed and experimental values of the critical heat flux than the simplified table mentioned above.The model of a full-scale fuel assembly was used in the experiments studying critical heat emission. The pressure vessel holding the electrically directly heated model comprised a Kh18H10T steel tube with outer diameter 121 mm and wall thickness 6 mm. A collection of electrically insulating bushings consisting of the natural mineral talcochlorite was placed inside the tube. The bushings were 120-mm high and had an inner diameter of 80 mm.The model used in the experiments consisted of 19 electrically heated Kh18H10T steel fuel-element simulators with outer diameter 13.5 mm, wall thickness 3.5 mm, and length (zone of heat release) 7 m. The model was heated by passing a constant electric current from thyristor units directly a...

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Library of thermophysical electronic databases

et al. 2007

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A library of electronic databases for thermophysical data is developed using Microsoft Access. This library includes 13 databases containing experimental data. Existing experimental data and their reduction to a unified electronic form for maintaining a large volume of data in their real state were collected and systematized to create the library. The proposed organization of the information greatly facilitates statistical analysis of the experimental data and the computational uncertainty for assessing the safety of operating nuclear power plants and verifying computer codes. Data on the thermohydraulics of RBMK fuel assemblies, critical efflux, and transient regimes in operating power-generating units and data for three-dimensional thermohydraulic calculations of rod assemblies have been placed into the library.A library of electronic databases for thermophysical data has been under construction for the last several years at the Research and Design Institute of the Electric Technology (NIKIÉT) as part of an agreement with Rosénergoatom. The team doing this work has set as its goal collecting and systematizing experimental data and reducing them to a unified electronic form, since many scientific archives exist only on paper (in the form of reports which are difficult to access).A common deficiency of the experimental data which are ordinarily available to anyone doing computations is the lack of exhaustive information on the experimental setup, the measurement apparatus (measurement errors), and the initial and boundary conditions for performing an experiment. To eliminate these deficiencies, we have drawn upon the most diverse sources (documents and reports) including direct contact with the experimenters (oral information).Another goal of this work was unification of the experimental data and maintaining a large volume of data in their real form. The fact that different sets of experimental data are used in computational modeling makes it more difficult to compare the computational results with one another, since the investigators performing an analysis of experiments exclude data which show the largest discrepancies between calculations and experiments or because of disagreement with a proposed expression. The goal of the library is to preserve the primary experimental information, which will not only make it possible to unify the computational modeling of each separate experiment but also facilitate statistical analysis of the data and analysis of the computational uncertainties [1, 2] when making a safety assessment of operating nuclear power plants and verifying computational codes, which answers the requirements of departmental documents [3].First of all, the data used for additional verification of domestic and foreign computational codes in application to RBMK reactors are placed in the library. The computational modeling in this case is concentrated on transient processes. The following phenomena are characteristic for such regimes:• outflow from a rupture in the circulation loop, including critical o...

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Analysis of the Uncertainty of Calculations of Loss-of-Coolant Accidents for the No. 1 Unit of the Kursk Nuclear Power Plant

Cited by 4 publications

References 3 publications

Safety prediction in nuclear power

Safety prediction in nuclear power

Verification of relap5/mod3.2 critical heat transfer computational method by experiments performed on the ks facility

Library of thermophysical electronic databases

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