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NT

KV-Scaling Analysis to support the Validation of Atucha-II Best Estimate Evaluation Model

 

 

 

 

A. Petruzzia, M. Cherubinia

a Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy

 

Nuclear Engineering and Design, Volume 110609, March 2020

 

Abstract — In the frame of the licensing process of the Atucha-II PHWR (Pressurized Heavy Water Reactor) the BEPU (Best Estimate Plus Uncertainty) approach has been selected for issuing the Chapter 15 of FSAR (Final Safety Analysis Report) dealing with accident analysis. Approaches based on Best Estimate Plus Uncertainties (BEPU) to perform accident analysis of a Nuclear Power Plant (NPP) for licensing purposes request the availability of validated tools and specific procedures suitable for the analysis of accident conditions envisaged in the concerned NPP. This implies the necessity to adopt and to prove an adequate quality of the so-called Evaluation Model (EM), i.e. the ensemble of data, assumptions, codes and nodalizations developed and validated to be used for carrying out the safety analysis. The purpose of the present paper is to outline key aspects and results of one of the steps of the BEPU process concerning the qualification of the     Atucha-II PHWR Evaluation Model for the licensing process. A special procedure, named “Kv-Scaled Analysis + Countepart Test Calculations” have been developed to take into account the specificities of Atucha-II NPP (CNA-2) and the fact that Atucha-II has not a prototype integral test facility. The proposed procedure uses a set of facilities which share some similar geometrical features and ranges of variations of relevant physical quantities for some subsystems or components of CNA-2, taking into account the different kind of accidents addressed in the Chapter 15 of the Final Safety Analysis Report (FSAR) of Atucha-II NPP. The final goal of the procedure is to qualify separately systems/sub-systems of the nodalization of the Atucha-II NPP while keeping the features of a full system analysis involving the whole NPP nodalization. A total of nine tests performed in six different scaled facilities have been considered and analyzed.

 


 

NT

Foreword: Selected papers from the 2018 Best Estimate Plus Uncertainty International Conference (BEPU 2018)

 

 

 

 

A. Petruzzia, K. Ivanovb, E. Ivanovc

Nuclear and INdustrial Engineering (NINE), Lucca, Italy

North Carolina State University (NCSU), Raleigh, North Carolina, USA

c   Institute for Radiological Protection and Nuclear Safety (IRSN), Paris, France


Volume 205, 2019 - Nuclear Technology


Abstract —
Approximately three hundred experts from more than 30 countries traveled to Lucca, Italy, to attend BEPU2018, which was sponsored by the American Nuclear Society, the Nuclear Energy Agency, and the International Atomic Energy Agency and was also cosponsored by a local organizing committee led by Nuclear and Industrial Engineering. Over 250 draft papers were reviewed, and finally, a grand total of over 170 full papers were accepted and presented in technical sessions. In addition, 21 invited keynote lectures, 13 plenary speeches, and 6 panel discussions addressed the state-of-the-art challenges in various areas of BEPU. The BEPU technical program committee and the special issue guest editors then coordinated efforts to select a limited number of papers and invited keynote and plenary lectures for consideration for archival publication in leading scientific journals. The authors were then invited to update their papers before submitting them for additional peer review for these journal special issues.
The papers in this special issue may be collected into four groups:

1. General considerations about the BEPU approach

2. Development of BEPU methods and techniques

3. Applications of BEPU methods

4. Development of multiphysics, multiscale approaches

We hope you enjoy this special issue of Nuclear Technology and look forward to seeing you at the next BEPU conference in Sicily, Italy, in May 2020.

https://www.tandfonline.com/eprint/EDCTUQBAVWVPKK4ANCM8/full?target=10.1080/00295450.2019.1676080 


NTThe CASUALIDAD Method for Uncertainty Evaluation of Best-Estimate System Thermal-Hydraulic Calculations


 

 

A. Petruzzi

Nuclear and INdustrial Engineering (NINE), Lucca, Italy

 

Volume 205, 2019 - Nuclear Technology

Abstract —
Predictive Modeling Methodology constitutes an innovative approach to perform uncertainty analysis, which reduces the subjectivity and the user-defined way to manage experimental data and derive uncertainty of input parameters that instead characterize the Propagation of Input Uncertainties and/or Propagation of Output Accuracies methods.

The CASUALIDAD (Code with the capability of Adjoint Sensitivity and Uncertainty AnaLysis by Internal Data ADjustment and assimilation) method can be developed as a fully deterministic method based on advanced mathematical tools for performing internally to the thermal-hydraulic system code the sensitivity and the uncertainty analysis.  The method is based upon powerful mathematical tools to perform sensitivity analysis and upon the Data Adjustment/Assimilation methodology by which experimental observations are combined with code predictions and their respective errors through the application of the Bayesian Theorem and of the Principle of the Maximum Likelihood, to provide an improved estimate of the system state and of the associated uncertainty considering all input parameters that affect any prediction.

The methodology has been structured in two main steps. The former has the aim to generate the database of improved estimations starting from the available set of experimental data and related qualified calculations; the latter is dealing with the use of the selected (from the obtained database) set of improved estimations for the uncertainty evaluation of the predicted Nuclear Power Plant (NPP) transient scenario.

The proposed methodology clearly interrelates in a consistent and robust framework, the code validation issue with the evaluation of the uncertainty of code responses passing through the quantification of input uncertainty parameters of code models and thus constituting a step forward respect to the subjectivity of the current methods based on Propagation of Input Uncertainties and/or Propagation of Output Accuracies.
https://www.tandfonline.com/eprint/89WX73M2IY35TTSPRHS2/full?target=10.1080%2F00295450.2019.1632092&


ned

Development of good practice guidance for quantification of thermal-hydraulic code model input uncertainty

 

 

 

A. Petruzzia et al.
a Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy

 

Nuclear Engineering and Design, Volume 354, December 2019


Abstract —  Taking into account uncertainties is a key issue in nuclear power plant safety analysis using best estimate plus uncertainty methodologies. It involves two main types of treatment depending on the variables of interest: input parameters or system response quantity. The OECD/NEA PREMIUM project devoted to the first type of variables has shown that inverse methods for input uncertainty quantification can exhibit strong user-effect. One of the main reasons was the lack of a clear guidance to perform a reliable analysis. This work is precisely devoted to the development of a first good practice guidance document for quantification of thermal-hydraulic code model input uncertainty. The developments have been done in the framework of the OECD/NEA SAPIUM project (January 2017–September 2019). This paper provides a summary of the main project outcome. Recommendations and open issues for future developments are also given.

https://www.sciencedirect.com/science/article/pii/S0029549319301839?dgcid=coauthor


ned Quantification of the uncertainty of the physical models in the system thermal-hydraulic codes – PREMIUM benchmark 

 

 

 

 

A. Petruzzia et al.
aNuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy

 

Nuclear Engineering and Design, Volume 354, December 2019

 

Abstract —  PREMIUM (Post BEMUSE Reflood Models Input Uncertainty Methods) was an activity launched with the aim of pushing forward the methods of quantification of physical model uncertainties in thermal-hydraulic codes. The benchmark PREMIUM was addressed to all who apply uncertainty evaluation methods based on input uncertainties quantification and propagation. The benchmark was based on a selected case of uncertainty analysis application to the simulation of quench front propagation in an experimental test facility. Applied to an experiment, enabled evaluation and confirmation of the quantified probability distribution functions on the basis of experimental data. The scope of the benchmark comprised a review of the existing methods, selection of potentially important uncertain input parameters, quantification of the ranges and distributions of the identified parameters using experimental results of tests performed on the FEBA test facility, verification of the performed quantification on the basis of tests performed at the FEBA test facility and validation on the basis of blind calculations of the Reflood 2-D PERICLES experiment. The benchmark has shown dependency of the results on the applied methodology and a strong user effect. The conclusion was that a systematic approach for the quantification of model uncertainties is necessary.

https://www.sciencedirect.com/science/article/pii/S0029549319302080?dgcid=coauthor


Design-Basis Accident Analysis Methods for Light-Water Nuclear Power Plants
Modern Nuclear Energy Analysis Methods - Vol.3, 2019
World Scientific

book1    book2


 NTUncertainty calculation in small break LOCA in the emergency core cooling system connected to the hot leg of Angra 2 nuclear power plant

 

 

 

Eduardo Madeira Borgesa, Gaianê Sabundjiana, F. D'Auriab, A. Petruzzic


a
Instituto de Pesquisas Energéticas e Nucleares (IPEN-CNEN-SP), Av. Professor Lineu Prestes, 2242, São Paulo, Brazil

bUniversity of Pisa, Largo L. Lazzarino 2, Pisa, Italy
c
Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy

 

Nuclear Energy Science and Technology, Vol. 12, No. 2, 2018


Abstract — 
Owing to the occurrence of nuclear accidents, worldwide nuclear regulatory organisations included the analysis of accidents considered as design basis accidents – Loss of Coolant Accident (large and small-break, LBLOCA or SBLOCA) – in the safety analysis reports of nuclear facilities. In Brazil, the tool selected by the licensing authority, Comissão Nacional de Energia Nuclear (CNEN), is RELAP5 Code. The aim of this paper is the evaluation of the performance of the Emergency Core Cooling System (ECCS) of Angra 2 nuclear reactor during SBLOCA. In this study, the RELAP5 code and the Code Internal Assessment of Uncertainty (CIAU) were used to simulate and analyse the uncertainties of the results. The postulated accident is the SBLOCA in the hot leg connected to the ECCS described in the Final Safety Analysis Report of Angra 2 (FSAR/A2). The results from this study were satisfactory when compared with the FSAR/A2. 


 nuclear journal picUncertainty and sensitivity analysis in reactivity-initiated accident fuel modeling: synthesis of organisation for economic co-operation and development (OECD)/nuclear energy agency (NEA) benchmark on reactivity-initiated accident codes phase-II

 

 

Olivier Marchanda, Jinzhao Zhang b, Marco Cherubinic 


a
Institut de Radioprotection et de Sûrete Nucleaire (IRSN), PSN-RES, SEMIA, Cadarache, St Paul-Lez-Durance, 13115, France

bTractebel (ENGIE), Avenue Ariane 7, 1200 Brussels, Belgium
c
Nuclear and Industrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy

 

Nuclear Energy Science and Technology, 50 (2018) 280-291


Abstract — 
In the framework of OECD/NEA Working Group on Fuel Safety, a RIA fuel-rod-code Benchmark Phase I was organized in 2010e2013. It consisted of four experiments on highly irradiated fuel rodlets tested under different experimental conditions. This benchmark revealed the need to better understand the basic models incorporated in each code for realistic simulation of the complicated integral RIA tests with high burnup fuel rods. A second phase of the benchmark (Phase II) was thus launched early in 2014, which has been organized in two complementary activities: (1) comparison of the results of different simulations on simplified cases in order to provide additional bases for understanding the differences in modelling of the concerned phenomena; (2) assessment of the uncertainty of the results. The present paper provides a summary and conclusions of the second activity of the Benchmark Phase II, which is based on the input uncertainty propagation methodology. The main conclusion is that uncertainties cannot fully explain the difference between the code predictions. Finally, based on the RIA benchmark Phase-I and Phase-II conclusions, some recommendations are made. © 2018 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the 

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ned Atucha-1 NPP containment venting analysis following SBO and LBLOCA events by GOTHIC code

 

 

 

 

A. Popa,b, W. Giannottia, A. Petruzzia , R. Garberoc, O. Mazzantinic


a
Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy

bUniversity of Pisa, Largo L. Lazzarino 2, Pisa, Italy
c
NA-SA, Nucleoelectrica Argentina, Buonos Aires, Argentina

Nuclear Engineering and Design, October 2018


Abstract — 
Containment behaviour plays a key role in the safety framework of a Nuclear Power Plant (NPP). The GOTHIC thermal hydraulic code has been adopted to evaluate the Atucha-1 NPP containment responses during two postulated severe accident scenarios, Station Black Out and Large Break Loss of Coolant Accident without Safety Injection Pumps (SIPs), while assuming that the external cooling of the Reactor Pressure Vessel is carried out during the transients.

The Atucha-1 NPP has a containment designed to work at full pressure, constituted by a steel sphere enveloped by a concrete shell, and having an annular gap of air in between.

The target of the analysis is the evaluation of the effects caused by the additional production of steam in the reactor cavity as a consequence of the external vessel cooling, which could cause an increase in containment pressure, and lead to pressure values above the safety limit. The containment pressure and temperature, the distribution of hydrogen in the containment atmosphere and the water hold-up in the most relevant rooms have been analysed as target variables. Each accident scenario was simulated using two different nodalizations, characterized by a different level of refinement. The “detailed” nodalization is meant to be the most refined nodalization according to the available computational resources; having high fidelity three dimensional details, with a high number of cells. Taking into consideration that several sensitivities were performed, the “coarse” nodalization was developed in order to lower the demand for computational resources without significantly compromising the global scenario response. Both nodalizations are characterized by high complexity in the representation of rooms and their connections.

Both accident transients, for each type of nodalization, were simulated for 200,000 s. At the end of the simulated transient, results showed that for the Large Break Loss of Coolant Accident pressure is predicted to surpass 5 bar, while the Station Black Out scenario is calculated to reach 4.4 bar. The performed sensitivities were simulated for 100,000 s and were meant to understand and characterize the impact of the different nodalization parameters (geometrical aspects, material properties, BCs). In addition, due to several code anomalies identified, several other sensitivity calculations were performed in order to find a way to analyse and mitigate the issues.


ned Instrumenting full-scale Boron Injection Test Facility to support Atucha-2 NPP licensing

 

 

 

 

F. Morettia, F. Terzuolia, F. D’Auria,b O. Mazzantinic


a
Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy

bGRNSPG—University of Pisa, Via Livornese 1219, San Piero a Grado (PI), Pisa, Italy
c
NA-SA, Nucleoelectrica Argentina S.A., UG-CNAII, 2806 Lima, Argentina

 

Nuclear Engineering and Design, February 2018

 

Abstract —  The Atucha-2 Pressurized Heavy Water Reactor is equipped with a back-up shutdown system based on the fast injection of boron into the moderator tank. Such system had initially been designed to cope with a 10%-area (0.1A) break Loss Of Coolant Accident (LOCA) scenario, but based on upgraded licensing requirements the design had to be revised and possibly improved against a double-ended guillotine (2A) break LOCA. In particular, the boron injection had to be proven fast enough to allow a timely shutdown of the reactor, even in the case of a failure of the primary shutdown system (control rods).

A full-scale test facility was built for such “design validation” purpose, in the framework of a cooperation program between the University of Pisa – San Piero a Grado Nuclear Research Group (GRNSPG) and the utility Nucleoeléctrica Argentina S.A. (NA-SA). A special instrumentation system, based on conductivity probes designed on purpose by the Helmholtz Zentrum Dresden-Rossendorf (HZDR), was adopted for the measurement of the injection delay, as well as for the monitoring of pressure at several key locations. Care was taken to reproduce the relevant NPP conditions as closely as possible to those expected on the basis of extensive safety analyses performed adopting a Best Estimate Plus Uncertainty (BEPU) approach. In this respect, not only the test facility is full-scale, but also the key components (such as the fast opening air valves, the boric acid tanks, the rupture device, the injection lance) were directly borrowed from the Atucha-2 NPP.

This paper provides an overview of the test facility, with particular emphasis on the Authors’ contributions to its design, implementation and operation. Then, it highlights the final outcomes of the experimental campaign carried out by NA-SA, namely: allowing to improve the design of the boron injection system (especially as to some fluid–structure interaction issues) and – what was the main goal – demonstrating that the system’s performance is fast enough to assure a timely and safe shutdown of the reactor, thus contributing to the successful completion of the NPP licensing process.


NTUncertainty and sensitivity analysis in reactivity-initiated accident fuel modeling: synthesis of organisation for economic co-operation and development (OECD)/nuclear energy agency (NEA) benchmark on reactivity-initiated accident codes phase-II

 

 

O. Marchanda, J. Zhangb, M. Cherubinic

 

aInstitut de Radioprotection et de Sûrete Nucleaire (IRSN), PSN-RES, SEMIA, Cadarache, St Paul-Lez-Durance, 13115, France
bTractebel (ENGIE), Avenue Ariane 7, 1200 Brussels, Belgium
c
Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy

 

Nuclear Engineering and Technology, December 2017

 

Abstract —  In the framework of OECD/NEA Working Group on Fuel Safety, a RIA fuel-rod-code Benchmark Phase I was organized in 2010e2013. It consisted of four experiments on highly irradiated fuel rodlets tested under different experimental conditions. This benchmark revealed the need to better understand the basic models incorporated in each code for realistic simulation of the complicated integral RIA tests with high burnup fuel rods. A second phase of the benchmark (Phase II) was thus launched early in 2014, which has been organized in two complementary activities: (1) comparison of the results of different simulations on simplified cases in order to provide additional bases for understanding the differences in modelling of the concerned phenomena; (2) assessment of the uncertainty of the results. The present paper provides a summary and conclusions of the second activity of the Benchmark Phase II, which is based on the input uncertainty propagation methodology. The main conclusion is that uncertainties cannot fully explain the difference between the code predictions. Finally, based on the RIA benchmark Phase-I and Phase-II conclusions, some recommendations are made.
© 2018 Korean Nuclear Society, Published by Elsevier Korea LLC.
This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/).


IAEA CRP Project on Benchmark Analysis of EBR-II Shutdown Heat Removal Tests

This publication presents the results and main achievements of an IAEA coordinated research project to verify and validate system and safety codes used in the analyses of liquid metal thermal hydraulics and neutronics phenomena in sodium cooled fast reactors. The publication will be of use to the researchers and professionals currently working on relevant fast reactors programmes. In addition, it is intended to support the training of the next generation of analysts and designers through international benchmark exercises.


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annalsAnalysis of a small PWR core with the PARCS/Helios and PARCS/Serpent code systems

 

 

 

 

G. Baiocco,a A. Petruzzi,a S. Bznunib, T. Kozlowskic

 

aNuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy

bNuclear and Radiation Safety Center (NRSC), Yerevan, Armenia

cDepartment of Nuclear, Plasma and Radiological Engineering, University of Illinois at Urbana-Champaign, IL, USA


Annals of Nuclear Energy Volume 107, September 2017, Pages 42–48


Abstract — 
Lattice physics codes are primarily used to generate cross-section data for nodal codes. In this work the methodology of homogenized constant generation was applied to a small Pressurized Water Reactor (PWR) core, using the deterministic code Helios and the Monte Carlo code Serpent. Subsequently, a 3D analysis of the PWR core was performed with the nodal diffusion code PARCS using the two-group cross section data sets generated by Helios and Serpent. Moreover, a full 3D model of the PWR core was developed using Serpent in order to obtain a reference solution. Several parameters, such as keff, axial and radial power, fission and capture rates were compared and found to be in good agreement.


NTThe BEPU Evaluation Model with RELAP5-3D for the Licensing of the Atucha-II NPP

 

 

 

 

A. Petruzzi,a M. Cherubini,a M. Lanfredini,a,b F. D’Auria,b and O. Mazzantinic

a
Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy

bUniversity of Pisa, GRNSPG, Via Livornese 1219, San Piero a Grado, Pisa, Italy

cNA-SA, Nucleoelectrica Argentina S.A., UG-CNAII, 2806 Lima, Argentina


Nuclear Technology, 193, January 2016

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Abstract —
 Within the licensing process of the Atucha-II pressurized heavy water reactor, the best-estimate plus uncertainty (BEPU) approach has been selected for issuing Chapter 15 of the Final Safety Analysis Report. The RELAP5-3D code developed by Idaho National Laboratory has been adopted as the best estimate system thermal-hydraulic code to perform the accident analyses. The complexity of a nuclear power plant (NPP) and of the accident scenarios may be a challenge for a conservative analysis and may justify the choice of a BEPU approach in the licensing process. This implies two main needs: (1) the need to adopt and to prove (to the regulatory authority) an adequate quality for the computational tools and (2) the need to account for the uncertainty. The purpose of the present paper is to outline key aspects of the BEPU process aimed at the licensing of the Atucha-II (CNA-II) NPP in Argentina operated by Nucleoeléctrica Argentina (NA-SA). Among the general attributes of a methodology to perform accident analysis of a NPP for licensing purposes, the very first one should be compliance with the established regulatory requirements. A second attribute deals with the adequacy and the completeness of the selected spectrum of events that should consider the combined contributions of deterministic and probabilistic methods. The third attribute is connected to the availability of qualified tools and analytical procedures suitable for the analysis of accident conditions envisaged for the NPP of concern. The execution of the overall analysis and the evaluation of results in relation to slightly fewer than 100 scenarios revealed the wide safety margins available for the NPP of concern, which was licensed on May 29, 2014.


NTRELAP5 Applications at GRNSPG-NINE: 30 Years of Activities

 

 

 

 

A. Petruzzi,a M. Cherubini,aand F. D’Auriab

a
Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy

bGRNSPG—University of Pisa, Via Livornese 1219, San Piero a Grado (PI), Pisa, Italy


Nuclear Technology, 193, January 2016

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Abstract — The application of RELAP5 at GRNSPG-NINE (Nuclear Research Group of San Piero a Grado–Nuclear and Industrial Engineering) started more than 30 years ago during which several versions of the code have been applied for the analysis of a very large variety of scenarios occurring in facilities and nuclear installations.
The present paper has two goals: (1) to summarize the results and main outcomes achieved through the application of RELAP5 to international projects and benchmarks in which GRNSPG-NINE was involved and (2) to qualify the system’s thermal-hydraulic code calculations through the systematic application of a set of developed procedures.
Among the analyses performed, this paper will provide insights into the code results and, whenever possible, into the comparison with the reference/experimental data of scenarios measured in (1) integral test facilities: PSB-VVER, ATLAS, PKL, LOBI, LOFT, SPES, PACTEL; (2) separate effect test facilities: BFBT, Neptun, PANDA; (3) research reactors: Experimental Breeder Reactor (EBR); and (4) nuclear power plants: Atucha-II [pressurized heavy water reactor (PHWR)-Konvoi], VVER-1000, Darlington (CANDU).
In relation to the methodology developed for qualifying a system thermal-hydraulic code calculation, this paper provides a short description and spot results of the systematic application to the cases mentioned above in respect to some of the following steps: (1) demonstration of the geometrical fidelity; (2) demonstration of the steady-state achievement; (3) qualification at the on-transient level, which implies the characterization of (a) phenomenological windows and (b) relevant thermal-hydraulic aspects; and (4) quantitative analysis to evaluate the accuracy of the code calculation using the fast Fourier transform based method.


epjn coverAssessment of stainless steel 348 fuel rod performance against literature available data using TRANSURANUS code

 

 

 

 

C. Giovedi,a M. Cherubini,b A. Abe c, F. D’Auria d

a
LabRisco, University of São Paulo, Av. Prof. Mello Moraes 2231, São Paulo, SP, Brazil

bNuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy

cNuclear and Energy Research Institute - IPEN/CNEN, Nuclear Engineering Center – CEN, Av. Prof. Lineu Prestes 2242, São Paulo, SP, Brazil

dGRNSPG—University of Pisa, Via Livornese 1219, San Piero a Grado (PI), Pisa, Italy

 

EPJ Nuclear Sci. Technol. 2, 27 (2016)

 

Abstract — This Early pressurized water reactors were originally designed to operate using stainless steel as cladding material, but during their lifetime this material was replaced by zirconium-based alloys. However, after the Fukushima Daiichi accident, the problems related to the zirconium-based alloys due to the hydrogen production and explosion under severe accident brought the importance to assess different materials. In this sense, initiatives as ATF (Accident Tolerant Fuel) program are considering different material as fuel cladding and, one candidate is iron-based alloy. In order to assess the fuel performance of fuel rods manufactured using iron-based alloy as cladding material, it was necessary to select a specific stainless steel (type 348) and modify properly conventional fuel performance codes developed in the last decades. Then, 348 stainless steel mechanical and physics properties were introduced in the TRANSURANUS code. The aim of this paper is to present the obtained results concerning the verification of the modified TRANSURANUS code version against data collected from the open literature, related to reactors which operated using stainless steel as cladding.
Considering that some data were not available, some assumptions had to be made. Important differences related to the conventional fuel rods were taken into account. Obtained results regarding the cladding behavior are in agreement with available information. This constitutes an evidence of the modified TRANSURANUS code capabilities to perform fuel rod investigation of fuel rods manufactured using 348 stainless steel as cladding.

 


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

 

 

 

A. Petruzzi,a F. D’Auriab


a
Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy

bGRNSPG—University of Pisa, Via Livornese 1219, San Piero a Grado (PI), Pisa, Italy


Nuclear Science and Engineering, 182, January 2016

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Abstract — This paper discusses the role and the depth of the analysis required for merging, on one hand, suitable experimental data and, on the other hand, qualified code calculation results. The availability of an experimental and calculated qualified database is of outmost importance for the validation and qualification of codes. Such a database can be used not only for demonstrating that the code results are reliable and for performing an independent code assessment but also as a basis for developing and validating an uncertainty methodology. As discussed in several other papers and guidelines, an uncertainty methodology must rely on the availability of a qualified code and qualified procedures. The development of a Standardized Consolidated Calculated and Reference Experimental Database (SCCRED) that includes documentation such as the reference data set of the facility and of the tests, the qualification report of the code calculations, and the engineering handbook constitutes an approach envisaged also by the International Atomic Energy Agency to set up a qualified experimental and calculated database for verification and validation (V&V) purposes of computational tools and uncertainty methodologies.

In order to frame and to outline the role of a qualified database for performing a best-estimate and uncertainty analysis (UA), a summary of the approaches for performing the uncertainty evaluation is provided distinguishing among the methods based on propagation of input uncertainties, the methods based on code output accuracy propagation, and the predictive modeling methodology. The main issues from the review of the uncertainty methods are the following: (a), the identification of the uncertainty-method-user effect to be considered in addition to the more well-known code-user effect when the BEPU (Best Estimate Plus Uncertainty) approach is selected to perform accident analysis and (b) (partially connected also with the first issue) the need for validation of the uncertainty methods (in the same way as validation of a computer code is a fundamental prerequisite for application of the code).

To address the two issues above, the Code with the capability of Internal Assessment of Uncertainty (CIAU) is discussed in detail. In particular, it shall be noted that the CIAU method (which belongs to the methods based on code output accuracy propagation) needs a qualified set of experimental and code calculation results as input for performing a qualified intrinsically validated UA, which has a more limited uncertainty-method-user effect in comparison to the methods based on propagation of input uncertainties.

Following the above consideration, the creation of SCCRED as a standardized consolidated reference experimental and calculated database constitutes a prerequisite for the development and application of the CIAU method, but at the same time such a qualified database can also be used for the V&V process of methods based on propagation of input uncertainties contributing to limit the uncertainty-method-user effect.