The pbenomenological analyses and thermal hydraulic characterization of a nuclear reactor arc the basis for its design and safety evaluation. In light of the impossibility and huge cost of performing meaningful experiments at full scale, sealed down expenmental tests - Integral Effect Test (IET) and Separate Effect Test (SET) • arc more feasible in developing "assessment database". The data arc useful in characterizing the prototype design and in the validation of computational tools for safety analysis. The analyses of system behaviors including component interactions in the Reactor Coolant System (RCS). the Containment System (PCV) and the RCS.'PCV coupled system have been extensively investigated using lETs in the past decades. Though several scaling methods, e.g. Linear, Power/Volume, Three level scaling, H2TS..., have been developed and applied in the IF.T and SET design, a direct extrapolation of the data to the prototype, i.e. the scalability, is in general not possible due to unavoidable scaling distortions. The scaling distortions arc related to many factors, mainly the complex geometry, multiple component interactions and two phase thermal hydraulic phenomena in steady state and transient condition of a nuclear reactor. The complex nature of scaling a nuclear reactor requires a large number of scaling parameters to be simultaneously fulfilled. In addition, physical construction and funding constraints demand that a scaling compromise is inevitable. Therefore a scaling approach, e.g. time preserved'not preserved, full height/reduced height, full pressure/reduced pressure, full power/reduced power..., has to be adopted in accordance with the objective of the IET or SET. Together with the scaling analysis. Best Estimate (BE) thermal hydraulic system code has been used for supporting experiment activity (design facilities, interpretation of results, etc) and for extrapolating results to full scale prototype conditions. Since the closure laws in the system code arc mainly based on scaled test data, the extrapolation of code results remains a challenging and open issue. Starting from a brief analysts of the main characteristics of IF.Ts and SETFs. the main objective of this paper is to analyze some IET scaling approaches used to the simulation of RCS responses which characterize the main scaling limits. The scaling approaches and their constraints in ROSA-III. FIST and PIPER-ONE facility will be used to analyze their impact to the experimental prediction in Small Break LOC'A counterpart tests. The liquid level behavior in the core and the core cladding temperature analysis are discussed used as judging criteria for the facilities sealing-up limits.

Scaling issues for the experimental characterization of reactor coolant system in integral test facilities and role of system code as extrapolation tool

De Rosa, F.;Mascari, F.
2015

Abstract

The pbenomenological analyses and thermal hydraulic characterization of a nuclear reactor arc the basis for its design and safety evaluation. In light of the impossibility and huge cost of performing meaningful experiments at full scale, sealed down expenmental tests - Integral Effect Test (IET) and Separate Effect Test (SET) • arc more feasible in developing "assessment database". The data arc useful in characterizing the prototype design and in the validation of computational tools for safety analysis. The analyses of system behaviors including component interactions in the Reactor Coolant System (RCS). the Containment System (PCV) and the RCS.'PCV coupled system have been extensively investigated using lETs in the past decades. Though several scaling methods, e.g. Linear, Power/Volume, Three level scaling, H2TS..., have been developed and applied in the IF.T and SET design, a direct extrapolation of the data to the prototype, i.e. the scalability, is in general not possible due to unavoidable scaling distortions. The scaling distortions arc related to many factors, mainly the complex geometry, multiple component interactions and two phase thermal hydraulic phenomena in steady state and transient condition of a nuclear reactor. The complex nature of scaling a nuclear reactor requires a large number of scaling parameters to be simultaneously fulfilled. In addition, physical construction and funding constraints demand that a scaling compromise is inevitable. Therefore a scaling approach, e.g. time preserved'not preserved, full height/reduced height, full pressure/reduced pressure, full power/reduced power..., has to be adopted in accordance with the objective of the IET or SET. Together with the scaling analysis. Best Estimate (BE) thermal hydraulic system code has been used for supporting experiment activity (design facilities, interpretation of results, etc) and for extrapolating results to full scale prototype conditions. Since the closure laws in the system code arc mainly based on scaled test data, the extrapolation of code results remains a challenging and open issue. Starting from a brief analysts of the main characteristics of IF.Ts and SETFs. the main objective of this paper is to analyze some IET scaling approaches used to the simulation of RCS responses which characterize the main scaling limits. The scaling approaches and their constraints in ROSA-III. FIST and PIPER-ONE facility will be used to analyze their impact to the experimental prediction in Small Break LOC'A counterpart tests. The liquid level behavior in the core and the core cladding temperature analysis are discussed used as judging criteria for the facilities sealing-up limits.
9781510811843
Facility scaling-up limits;Integral test facility;Reactor coolant system phenomena;Containment phenomena;Nuclear safety analysis;Uncertainty;Scaling
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Utilizza questo identificativo per citare o creare un link a questo documento: https://hdl.handle.net/20.500.12079/4233
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