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International fuel performance study of fresh fuel experiments for PCMI effects during RIA experiments Seokbin Seo a, Charles Folsom a, Colby Jensen a, David Kamerman a, Luana Giaccardi b, Marco Cherubini b, Pavel Suk b, Martin Sevecek c, Jerome Sercombe d, Isabelle Guenot-Delahaie d, Alessandro Scolaro e, Matthieu Reymond e, Katalin Kulacsy f, Luis Herranz g, Francisco Feria g, Pau Aragón g, Grigori Khvostov h, Imran Khan i, Anuj Kumar Deo j, Srinivasa Rao Ravva j, Rolando Calabrese k, Felix Boldt l, Jonathan Sappl l, Florian Falk l, Asko Arkoma m, Georgenthum Vincent n, Yudai Tasaki o, Kazuo Kakiuchi o, Yutaka Udagawa o, Gregory Delipei p, Charles Cheron p, James Corson q, Jinzhao Zhang r, Thomas Drieu r, Jan Klouzal s, Martin Dostal s, Vitezslav Matocha s, Tereza Kinkorová t, Carlo Fiorina u a Idaho National Laboratory (INL), USA |
Abstract - This paper presents the results of High-burnup Experiments for Reactivity-initiated Accident (HERA) Modeling & Simulation (M&S) exercise. The HERA project under the Nuclear Energy Agency (NEA) Second Framework for Irradiation Experiments (FIDES-II) program is focused on studying Light Water Reactor (LWR) fuel behavior during Reactivity-Initiated Accident (RIA) conditions. The Part I M&S cases are based on a series of tests in the Transient Reactor Test (TREAT) facility in the United States and the Nuclear Safety Research Reactor (NSRR) in Japan. The purpose of this work is to evaluate the test design to accomplish its goals in establishing clearer understanding of the effects of power pulse width during RIA conditions. The blind predictions using various computational tools have been performed and compared amongst to interpret the behaviors of high burnup fuels during RIA. While many international participants evaluate the thermal–mechanical behavior of fuel rod under different conditions, a considerable scatter of outputs comes out for the cases due to the disparity between codes in predicting mechanical behaviors. In general, however, the results of thermal–mechanical analysis elaborate that nominal design conditions the shorter pulse width tests in NSRR should cause cladding failures while the TREAT tests appear to have more split prediction of failure or not. Furthermore, the sensitivity analysis varying key testing parameters reveals the considerable effect of power pulse width and total energy deposition on prediction of fuel rod failure.
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Toward a Better Understanding of Reflood Thermal Hydraulics: A Summary of the OECD/NEA RBHT Project Stephen M. Bajorek a, Brian Lowery b, Fan-Bill Cheung c, Alessandro Del Ferraro d, Marco Cherubini d, Alessandro Petruzzi d, Jinzhao Zhang e, and Martina Adorni f a U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001 |
Abstract - Reflood thermal hydraulics remains a difficult and complex subject, and understanding thephysical phenomena that occur during a reflood transient is important to nuclear safety. The Organisation forEconomic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) Rod Bundle Heat Transfer(RBHT) project was designed to provide unique experimental data for code assessment and model development.Participants, which came from 21 international organizations, used analysis codes including APROS, ATHLET, CATHARE, CTF, MARS, RELAP5, TRACE, and SPACE to simulate the tests performed in the RBHT facility. The experimental campaign carried out within the OECD/NEA RBHT project produced data for a total of 16 reflood tests conducted in two test series. An “open” test series consisted of 11 experiments, and a “blind” test series consisted of 5 experiments. In the blind tests, only the initial and boundary conditions were providedto participants prior to simulation of those experiments. Reflood rates ranged from 0.5 to 15 cm/s, thusproducing data applicable to dispersed flow film boiling and inverted annular flow film boiling. Inlet subcoolingranged from 2.8 to 80 K. Tests with variable reflood rates and oscillatory reflood rates were included in the testmatrix. This paper describes the project and presents a summary of major experimental and analytical findings.
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Analyses of Design Extension Conditions Without Significant Fuel Degradation for Operating Nuclear Power Plants: an OECD/NEA Review J. Zhang a, M. Havet a, J. Zheng a, A. Bousbia Salah b, M. Ševeček c, P. Kral d, J. Krhounkova d, A. Guba e, Z. Hozer e, A. Bersano f, F. Mascari f, a Tractebel (ENGIE), Boulevard Simon Bolivar 36, 1000 Brussels, Belgium |
Abstract - Since 2012, many NEA member countries have implemented deterministic safety analyses for operating nuclear power plants under design extension conditions without significant fuel degradation or core melt (DEC-A). However, variations persist among these countries in defining DEC-A scenarios and acceptance criteria, validation and application of computer codes, development and application of deterministic safety analysis methods. Furthermore, there is a dearth of shared international experience and methodologies among various stakeholders, including regulatory authorities, technical safety or support organizations, utilities, engineering and consulting companies. To address these gaps, the OECD/NEA initiated a project in 2021, titled “Good Practices for Analyses of Design Extension Conditions without Significant Fuel Degradation for Operating Nuclear Power Plants” (or “DEC-A”), under the auspices of the Working Group on Accident Management and Analysis (WGAMA) and the Working Group on Fuel Safety (WGFS). The DEC-A project aims to review and summarize the current requirements, knowledge status, and best practices in NEA member countries. This paper outlines the objectives and scope of the OECD/NEA DEC-A project, and presents the findings from the review and discussions for each task.
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Optimisation of Accident Management Measures to Reduce Iodine Releases During SGTR Bernd Hrdy a, Raphael Zimmerl b, Marco Cherubini c, Nikolaus Müllner a a University of Natural Resources and Life Sciences, Vienna, Department of Water, Atmosphere, and Environment, Institute of Safety and Risk Sciences, Peter-Jordan-Straße 76/1, 1190 Vienna, Austria |
Abstract - Steam generator tube rupture (SGTR) accidents create a bypass of the containment of a pressurised water reactor (PWR) and can therefore result in the release of primary system coolant to the atmosphere via the steam relief or safety valves. In general, primary system coolant will transport radionuclides such as iodine-131. Accident management strategies for SGTR accidents therefore aim to reduce releases to the environment while ensuring core cooling.
The Downhill Simplex algorithm is used in this paper to optimise the timing of accident management measures during a SGTR accident. The secondary system steam relief and safety valves (SRV) are assumed to fail in the stuck open position at the first opening. Depressurisation of the primary system by opening the pressuriser pilot operated relief valve (PORV) and keeping the primary pressure low by shutting down two of the three trains of the high pressure injection system (HPIS) is assumed as the accident management procedure. The success of the measures is evaluated by a Relap5-3D simulation, which calculates the thermal hydraulic behaviour of the system. One of the key parameters used to assess success is the amount of iodine-131 released into the environment. The algorithm varies the timing of a set of three operator actions — opening the PORV and shutdown of HPIS trains one and two. In addition to iodine release, two other parameters are evaluated — reactor core coolant level and primary system pressure. Three normalisation functions are used to convert these parameters into a single target value, which is low when the core is covered and both primary system pressure and iodine release are low. The simplex algorithm then modifies the timing of operator actions to achieve a local minimum of the target value.
The results show that the Downhill Simplex algorithm can be used to optimise the timing of operator actions. Although timing cannot be directly implemented in Accident Management Procedures (AMPs), it is important to be aware of time sensitivity when designing AMPs. In addition, the algorithm can be adapted to optimise design parameters such as valve sizes, hydro accumulator nominal pressure levels.
The work was performed within the EURATOM R2CA project.
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Best Estimate Plus Uncertainty Analysis of Metal-Water Reaction Transient Experiment Alan Matias Avelar a, Camila Diniz a, Fábio de Camargo b, Claudia Giovedi c, Alfredo Abe c, Marco Cherubini d, Alessandro Petruzzi d, Marcelo Breda Mourão a a Department of Metallurgical and Materials Engineering, University of São Paulo, Professor Mello Moraes, 2463 São Paulo, SP, Brazil |
Abstract - Uncertainty analysis is applied in the licensing process for nuclear installations to complement best estimate analysis and to verify that the upper bound value is less than the threshold corresponding to the safety parameter of interest. Metal-water reaction is a critical safety phenomenon of water-cooled nuclear reactors at accident conditions, e.g. Loss-Of-Coolant Accidents (LOCA). AISI 348 cladding is able to increase the accident tolerance comparing to Zr-based alloys and differently from other accident tolerant fuel cladding options, there is operational experience of nuclear power plants with stainless steel. In this study, a transient oxidation experiment of AISI 348 by steam was conducted and the major sources of uncertainty were addressed. An evaluation model was developed to calculate the evolution of mass gain during the experiment. Meanwhile, uncertainty propagation of experimental data was performed. The results show that the mass gain predicted by the transient metal-water reaction model lays within the experimental data uncertainty band. Furthermore, the selection of the oxidation kinetics model seems to be important whether the analysis wills to provide conservative results.
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Towards modeling Defective Fuel Rods in TRANSURANUS: Benchmark and Assessment of Gaseous and Volatile Radioactive Fission Product Release L. Giaccardi a, M. Cherubini a, G. Zullo b, D. Pizzocri b, A. Magni b, L. Luzzi b a Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII 759, 55100, Lucca, Italy |
Abstract - This work presents the results of a collaborative benchmark activity between different organizations towards the use of the TRANSURANUS code to estimate the release of gaseous and volatile radioactive fission products from defective fuel rods, into the primary coolant of pressurized water reactors. First, the radioactive release from the fuel to the gap is evaluated according to three approaches: the coupling between TRANSURANUS and SCIANTIX, the development of TRANSURANUS devoted subroutines, and the use of the ANS 5.4-2010 methodology. Fuel-to-gap release calculations are benchmarked and assessed against measured data from the CONTACT1 irradiation experiment. Then, TRANSURANUS has been used to estimate the radioactive release into the primary coolant by applying a first-order phenomenological rate theory and tested against measured data of fission product coolant concentrations from irradiation experiments of the CRUSIFON program.
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Preliminary Safety Analysis at the Decommissioning of the WWR-M Research Reactor Yu. M. Lobach1, S. Yu. Lobach2, V. M. Shevel1 1 Institute for Nuclear Research, National Academy of Sciences of Ukraine, Kyiv, Ukraine |
Abstract - Following the demands established by the current Ukrainian legislation, the Decommissioning Concept for the WWR-M research reactor was recently approved. The Concept envisages a strategy of immediate dismantling; it identifies and justifies the main technical and organizational measures for the preparation and implementation of decommissioning, the sequence of planned works and activities, as well as the necessary conditions and infrastructure. Decommissioning requires proper planning and demonstration that all planned dismantling works will be carried out safely. Presented safety assessment is a mandatory component of the Concept and the most important element of the overarching technological scheme. The purpose of the safety analysis is to provide input for detailed planning on how to ensure safety during decommissioning. Based on the results of the safety analysis, the measures to ensure radiation protection are defined while justifying their necessity and sufficiency.
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Assessment of the Dose loadduring the Dismantling of the WWR-M Reactor Yu. M. Lobach1, S. Yu. Lobach2, E. D. Luferenko1, V. M. Shevel1 1 Institute for Nuclear Research, National Academy of Sciences of Ukraine, Kyiv, Ukraine |
Abstract - The WWR-M is a light-water-cooled and moderated heterogeneous research reactor with a thermal output of 10 MW.The final decommissioning planning is in progress now. The general decommissioning strategy consists of the dismantling and separate removal of the bulky elements as a whole (in one piece) without preliminary segmentation. The dismantling of the primary and secondary cooling loops is considered as one of the key tasks; a separate dismantling design has been developed. The baseline principles for the technical solution and safety are presented in the given paper. Results of the dose assessment showed that the work can be performed at a collective dose of less than 20 man-mSv.Keywords: WWR type research reactor, decommissioning, cooling loops, dismantling, exposure dose.
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International Benchmark Activity in the Field of Sodium Fast Reactors Domenico De Luca, Simone Di Pasquale, Marco Cherubini, Alessandro Petruzzi and Gianni Bruna Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy |
Abstract - 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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External function for GOTHIC code to estimate critical heat flux conditions for in-vessel retention assessment A. Pop a b , A. Petruzzia,W. Giannottia a Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy b Università di Pisa, Largo Lucio Lazzarino 2, Pisa, PI, Italy Nuclear Engineering and Design, Volume 380, 15 August 2021, 111301 |
Abstract - GOTHIC is an integrated, general purpose thermal hydraulic software package for design, licensing, safety and operating analysis of Nuclear Power Plant containments, confinement buildings and system components. It bridges the gap between the lumped parameter codes frequently used for containment analysis (such as MELCOR, MAAP, COCOSYS, ASTEC codes) and Computational Fluid Dynamics codes. Within a single model, GOTHIC can include regions treated in conventional lumped parameter mode and regions with three-dimensional flows in complex geometries. The heat transfer correlations built into GOTHIC cover the portion of the boiling curve which spans single phase heat transfer up to pre-Critical Heat Flux (CHF) heat transfer. The implemented boiling curve is truncated to exclude post-CHF heat transfer as it has not been adequately verified and was considered by the developers to have little application in general containment analysis. As such, one area that the code is not currently qualified for is post-CHF heat transfer, which could occur for example in the case of In-Vessel Corium Retention, where cooling water enters in contact with the high temperature of the Reactor Pressure Vessel wall. The presented research focuses on creating an external subroutine that solves this limitation, enabling the GOTHIC code to account for CHF phenomena. The modeling of CHF would be very useful in order to enable the code to simulate the external Reactor Pressure Vessel (RPV) Cooling , as well as other types of severe accidents or analyses where post-CHF simulation is required. Several subroutine function switches were implemented in order to facilitate its usage for different types of heat structures and correlations. The subroutine determines the CHF values based on either: the 2006 Groeneveld Look-up Tables, Lookup Tables for Large Diameter Vertical Tubes, Look-up Tables for Large Diameter Horizontal Tubes, or the correlation used by the MELCOR code for critical heat flux situations. It shall be noted that the developed subroutine and its implementation were performed without the need to have access to the GOTHIC source code. In a previous paper, the GOTHIC code was used to perform a containment safety analysis for the Atucha-I NPP (CNA-I) for an in-vessel retention type of analysis. Highly conservative vapour generating boundary conditions were used in order to simulate the boiling between the cavity water and the RPV surface, and to bypass the GOTHIC limitation. The newly developed subroutine was used for the analysis of two postulated Atucha-I in-vessel retention scenarios, a Large Break Loss Of Coolant Accident (LBLOCA) and a Station Black-Out (SBO), with the simulation of heat transfer between RPV and cavity water. Specifically for in-vessel retention situations, a separate user selectable option for the subroutine was developed, in which the Critical Heat Flux is determined based on experiments performed at the ULPU facility from the University of California, Santa Barbara, USA, which were combined with the 2006 Groeneveld Look-up Tables primarily in order to have a pressure dependence. This was performed because for the Atucha-I analysis, the pressure was higher than in the ULPU facility.
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KV-Scaling Analysis to support the Validation of Atucha-II Best Estimate Evaluation Model A. Petruzzi, M. Cherubini Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy Nuclear Engineering and Design, Volume 363, July 2020, Article number 110609 |
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.
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Foreword: Selected papers from the 2018 Best Estimate Plus Uncertainty International Conference (BEPU 2018) A. Petruzzia, K. Ivanovb, E. Ivanovc a Nuclear and INdustrial Engineering (NINE), Lucca, Italy b North Carolina State University (NCSU), Raleigh, North Carolina, USA c Institute for Radiological Protection and Nuclear Safety (IRSN), Paris, France Nuclear Technology, Volume 205, 08 November 2019 |
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
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The CASUALIDAD Method for Uncertainty Evaluation of Best-Estimate System Thermal-Hydraulic Calculations A. Petruzzi Nuclear and INdustrial Engineering (NINE), Lucca, Italy Nuclear Technology , Volume 205, 06 Aug 2019 |
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&
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Development of good practice guidance for quantification of thermal-hydraulic code model input uncertainity Jean Baccoua, Jinzhao Zhangb, Philippe Fillionc , Guillaume Damblinc , Alessandro Petruzzid , Rafael Mendizábale , Francesc Reventósf, Tomasz Skorekg , Mathieu Coupleth , Bertrand Ioossh , Deog-Yeon Ohi , Takeshi Takedaj
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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
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Quantification of the uncertainty of the physical models in the system thermal-hydraulic codes – PREMIUM benchmark Tomasz Skoreka , Agnèsde Crécyb , Andriy Kovtonyukcm, Alessandro Petruzzic q, Rafael Mendizábald, Elsade Alfonsoe, Francesc Reventóse, Jordi Freixae, Christine Sarrettef, Milos Kynclg, Rostislav Pernicag, Jean Baccouh, Fabrice Foueth, Pierre Probsth, Bub-Dong Chungi, Tran TranhTrami, Deog-Yeon Ohj, Alexey Gusevk, Yuri Shvestovk, Dong Lil, Xiaojing Liul, Jinzhao Zhangm, Torsti Alkun, Joona Kurkin, Wadim Jägero, Victor Sánchezo, Damar Wicaksonop, Omar Zerkakp , Andreas Pautzp
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
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Uncertainty 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 Nuclear Energy Science and Technology, Vol. 12, No. 2, January 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.
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Uncertainty 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 Zhangb, Marco Cherubinic |
Abstract — In the framework of OECD/NEA Working Group on Fuel Safety, a RIA fuel-rod-code Benchmark Phase I was organized in 2010-2013. 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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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 Nuclear Engineering and Design, Volume 337, October 2018
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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.
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Instrumenting full-scale Boron Injection Test Facility to support Atucha-2 NPP licensing F. Morettia, F. Terzuolia, F. D’Auriab, O. Mazzantinic Nuclear Engineering and Design, Volume 336, 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.
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Uncertainty 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
Nuclear Engineering and Technology, Volume 50, March 2018 |
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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Analysis of a small PWR core with the PARCS/Helios and PARCS/Serpent code systems G. Baioccoa, A. Petruzzia, 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
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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.
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The BEPU Evaluation Model with RELAP5-3D for the Licensing of the Atucha-II NPP A. Petruzzia , M. Cherubinia , M. Lanfrediniab , F. D’Auriab, and O. Mazzantinic aNuclear 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
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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.
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RELAP5 Applications at GRNSPG-NINE: 30 Years of Activities A. Petruzzia, M. Cherubinia, F. D’Auriab aNuclear 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
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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.
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Assessment of stainless steel 348 fuel rod performance against literature available data using TRANSURANUS code C. Giovedia, M. Cherubinib, A. Abec, F. D’Auriad aLabRisco, 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. , Volume 2, May 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.
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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 A. Petruzzia, F. D’Auriab
bGRNSPG—University of Pisa, Via Livornese 1219, San Piero a Grado (PI), Pisa, Italy
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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.