Standardized Consolidated Calculated and Reference Experimental Database (SCCRED): A Supporting Tool for V&amp;V and Uncertainty Evaluation of Best-Estimate System Codes for Licensing Applications

Petruzzi, A.; D'Auria, Francesco Saverio

doi:10.13182/nse15-79

Cited by 21 publications

(7 citation statements)

References 3 publications

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“…A consistent code assessment supported by a qualified experimental database is an important step for developing a solid ground for the uncertainty evaluation in the frame of Best Estimate and Uncertainty approach. Thus, the SCCRED (Standardized and Consolidated Calculated & Reference Experimental Database) methodology [9], embedded in NEMM [10], has been developed to generate a series of documents and tools to set up a qualified experimental and calculated database for Verification and Validation (V&V) purposes of Best Estimate and Uncertainty applications.…”

Section: Overview Of Sccred Methodologymentioning

confidence: 99%

International Benchmark Activity in the Field of Sodium Fast Reactors

Luca¹,

Pasquale²,

Cherubini³

et al. 2021

Recent Advances in Numerical Simulations

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Global interest in fast reactors has been growing since their inception in 1960 because they can provide efficient, safe, and sustainable energy. Their closed fuel cycle can support long-term nuclear power development as part of the world’s future energy mix and decrease the burden of nuclear waste. In addition to current fast reactors construction projects, several countries are engaged in intense R&D and innovation programs for the development of innovative, or Generation IV, fast reactor concepts. Within this framework, NINE is very actively participating in various Coordinated Research Projects (CRPs) organized by the IAEA, aimed at improving Member States’ fast reactor analytical simulation capabilities and international qualification through code-to-code comparison, as well as experimental validation on mock-up experiment results of codes currently employed in the field of fast reactors. The first CRP was focused on the benchmark analysis of Experimental Breeder Reactor II (EBR-II) Shutdown Heat Removal Test (SHRT-17), protected loss-of-flow transient, which ended in the 2017 with the publication of the IAEA-TECDOC-1819. In the framework of this project, the NINE Validation Process– developed in the framework of NEMM (NINE Evaluation Model Methodology) – has been proposed and adopted by most of the organizations to support the interpretation of the results calculated by the CRP participants and the understanding of the reasons for differences between the participants’ simulation results and the experimental data. A second project regards the CRP focused on benchmark analysis of one of the unprotected passive safety demonstration tests performed at the Fast Flux Test Facility (FFTF), the Loss of Flow Without Scram (LOFWOS) Test #13, started in 2018. A detailed nodalization has been developed by NINE following its nodalization techniques and the NINE validation procedure has been adopted to validate the Simulation Model (SM) against the experimental data of the selected test. The third activity deals with the neutronics benchmark of China Experimental Fast Reactor (CEFR) Start-Up Tests, a CRP proposed by the China Institute of Atomic Energy (CIAE) launched in 2018 the main objective of which is to improve the understanding of the start-up of a SFR and to validate the fast reactor analysis computer codes against CEFR experimental data. A series of start-up tests have been analyzed in this benchmark and NINE also proposed and organized a further work package focused on the sensitivity and uncertainty analysis of the first criticality test. The present chapter intends to summarize the results achieved using the codes currently employed in the field of fast reactor in the framework of international projects and benchmarks in which NINE was involved and emphasize how the application of developed procedures allows to validate the SM results and validate the computer codes against experimental data.

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Section: Overview Of Sccred Methodologymentioning

confidence: 99%

International Benchmark Activity in the Field of Sodium Fast Reactors

Luca¹,

Pasquale²,

Cherubini³

et al. 2021

Recent Advances in Numerical Simulations

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“…Table 1 provides an example of such a list. [17] developed the idea to consolidate qualified databases (both experimental and code calculation results) through standardization, aiming to support the V&V activities of system codes and uncertainty methodologies. The database, called SCCRED (Standardized Consolidated Calculated and Reference Experimental Database) includes documentation such as the reference data set of the facility (description of the facility, instrumentation, physical properties of the material, evaluation of the pressure losses...) and of the associated tests (main phenomena investigated during the test, configuration of the facility,..).…”

Section: Step 4 -Establishment Of a List Of Available Experiments Andmentioning

confidence: 99%

SAPIUM: A Generic Framework for a Practical and Transparent Quantification of Thermal-Hydraulic Code Model Input Uncertainty

Baccou

Zhang

Fillion

et al. 2020

Nuclear Science and Engineering

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SAPIUM: a generic framework for a practical and transparent quantification of thermal hydraulic code model input uncertaintyUncertainty analysis (UA) is a key element in nuclear power plant (NPP) deterministic safety analysis using best-estimate thermal hydraulic codes and best estimate plus uncertainty (BEPU) methodologies. If forward uncertainty propagation methods have now become mature for industrial applications, the input uncertainties quantification (IUQ) on the physical models still requires further investigations. The OECD/NEA PREMIUM project attempted to benchmark the available IUQ methods, but observed a strong user-effect due to lack of best practices guidance.The SAPIUM project has been proposed towards the construction of a clear and shared systematic approach for input uncertainty quantification. The main outcome of the project is a first "good practices" document that can be exploited for safety study in order to reach consensus among experts on recommended practices as well as to identify remaining open issues for further developments. This paper describes the systematic approach that consists in five elements in a step by step approach to perform a meaningful model input uncertainty quantification and validation as well as some "good practice guidelines" recommendations for each step.

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“…A typ i cally huge da ta base [12] and the knowledge man age ment con sti tute key term and ac tiv ity, needed for a BEPU ap pli ca tion, re spec tively. The house of knowl edge orig i nally pro duced when is su ing the S-SOAR, (scal ing state of art re port), [8], can easily be ex tended to the BEPU.…”

Section: The Knowl Edge and Knowl Edge Man Age Mentmentioning

confidence: 99%

The Best Estimate Plus Uncertainty approach in nuclear reactor safety and licensing: Brief history and the elements after licensing

Francesco

Giorgio

2019

Nucl Technol Radiat Prot

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The best es ti mate plus un cer tainty is, at the same time, an ap proach, a pro ce dure and a framework in nu clear ther mal-hy drau lics and nu clear re ac tor safety and li cens ing. The mo ti va tion at the ba sis of the best es ti mate plus un cer tainty is the lack of knowl edge in the ar eas of sin gle and, mainly, two-phase tran sient ther mal-hy drau lics. In other terms and in tro duc ing some sim pli fi ca tions, the in suf fi cient knowl edge of tur bu lence im poses the de sign of roadmaps for the ap pli ca tion of im per fect (ther mal-hy drau lic) mod els to the eval u a tion of de sign fea tures and of safety for com plex tech no log i cal in stal la tions or sys tems like the nu clear power plants and, more spe cif i cally, the wa ter cooled nu clear re ac tors. Fur ther more, the le gal coun ter part of nu clear re ac tor safety, or the li cens ing, is con cerned: there fore the best es ti mate plus un certainty must ac count for rules and reg u la tions de rived from the fun da men tal radioprotection prin ci ple which im poses the minimization of the im pact of ra di a tions upon hu mans and the en vi ron ment un der any cir cum stance. In the pres ent pa per, the key el e ments of the ap proach are iden ti fied and char ac ter ized. These shall be seen as the sup port for a con sis tent ap pli ca tion of ther mal-hy drau lics to the de sign and safety of wa ter-cooled nu clear re ac tors. Key words: best es ti mate, un cer tainty eval u a tion, best es ti mate plus un cer tainty, li cens ing pro cess, val i da tion, scal ing F. S. D'Auria, et al.: Best Es ti mate Plus Un cer tainty Ap proach in Nu clear Re ac tor .

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Standardized Consolidated Calculated and Reference Experimental Database (SCCRED): A Supporting Tool for V&V and Uncertainty Evaluation of Best-Estimate System Codes for Licensing Applications

Cited by 21 publications

References 3 publications

International Benchmark Activity in the Field of Sodium Fast Reactors

International Benchmark Activity in the Field of Sodium Fast Reactors

SAPIUM: A Generic Framework for a Practical and Transparent Quantification of Thermal-Hydraulic Code Model Input Uncertainty

The Best Estimate Plus Uncertainty approach in nuclear reactor safety and licensing: Brief history and the elements after licensing

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