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
b Amazônia Azul Tecnologias de Defesa S.A., Corifeu de Azevedo Marques, 1847 São Paulo, SP, Brazil
c Nuclear and Energy Research Institute, University of São Paulo, Professor Lineu Prestes, 2242 São Paulo, SP, Brazil
d Nuclear and Industrial Engineering, Via della Chiesa XXXII 759, 55100 Lucca, Italy
Nuclear Engineering and Design, Volume 411, September 2023, 112414
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.
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
b Nuclear and Industrial Engineering (NINE), Italy
c ALVEL, Czech Republic
d French Alternative Energies and Atomic Energy Commission(CEA), France
e EPFL, Switzerland
f HUN-REN Centre for Energy Research (HUN-REN EK-CER), Hungary
g Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas(CIEMAT), Spain
h Paul Scherrer Institute(PSI), Switzerland
i Bhabha Atomic Research Centre(BARC), India
j Atomic Energy Regulatory Board(AERB), India
k ENEA, Italy
l Gesellschaft für Anlagen und Reaktorsicherheit gGmbH(GRS), Germany
m Technical Research Centre of Finland(VTT), Finland
n Institut de Radioprotection et de Surete Nucleaire(IRSN), France
o Japan Atomic Energy Agency(JAEA), Japan
p North Carolina State University(NCSU), USA
q Nuclear Regulatory Commission(NRC), USA
r TRACTEBEL, Belgium
s UJV, Czech Republic
t Czech Technical University(CTU), Czech Republic
u Texas A&M University (TAMU), USA
Nuclear Engineering and Design, Volume 430, 15 December 2024, 113673
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.
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
b The Pennsylvania State University, Applied Research Laboratory, University Park, Pennsylvania 16802
c The Pennsylvania State University, Department of Mechanical Engineering and Nuclear Engineering, UniversityPark, Pennsylvania 16802
d Nuclear and Industrial Engineering (NINE), Via della Chiesa XXXII 759, Lucca, Italy
e Tractebel (ENGIE), Boulevard Simon Bolivar 34-36, 1000 Brussels, Belgium
f Organisation for Economic Co-operation and Development, Nuclear Energy Agency, Paris, France
Nuclear Technology, 18 October 2024
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.
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, M. Cherubini g, T. Nemec h, M. Sánchez i,
R. Mendizábal i, C. Queral j, L.E. Herranz k, M. Adorni l, M. Bales l
a Tractebel (ENGIE), Boulevard Simon Bolivar 36, 1000 Brussels, Belgium
b BEL V (subsidiary of FANC), 148 rue Walcourt, 1070 Brussels, Belgium
c Czech Technical University in Prague, Brehova 7, 11519 Praha 1, Czech Republic
d ÚJV Rez, a. s., Hlavní 130, 25068 Husinec – Rez, Czech Republic
e Centre for Energy Research (EK), Konkoly-Thege 25-33, Budapest 1121, Hungary
f ENEA, FSN-SICNUC-SIN, Via Martiri di Monte Sole, 4, 40129 Bologna, Italy
g N.IN.E.-Nuclear and INdustrial Engineering S.r.l., Via della Chiesa XXXII, 759, 55100 Lucca, Italy
h Slovenian Nuclear Safety Administration (SNSA), Litostrojska 54, 1000 Ljubljana, Slovenia
i Consejo de Seguridad Nuclear (CSN), Pedro Justo Dorado Dellmans 11, 28040 Madrid, Spain
j Universidad Politécnica de Madrid (UPM), Ramiro de Maeztu, 7, 28040 Madrid, Spain
k CIEMAT, Avda. Complutense, 40, 28040 Madrid, Spain
l OECD Nuclear Energy Agency (NEA), 46 quai Alphonse Le Gallo, 92100 Boulogne-Billancourt, France
Nuclear Engineering and Design, Volume 425, August 2024, 113320
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.
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
b Vienna Ombuds Office for Environmental Protection, Muthgasse 62, Vienna, 1190, Austria
c Nuclear and Industrial Engineering NINE Srl, Via della Chiesa XXXII, 759, Lucca, 55100, Italy
Annals of Nuclear Energy, Volume 203, August 2024, 110507
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.
Studsvik R2 Materials Test Reactor Ad Hoc Depletion Strategy for the Derivation of the Fuel Isotopic Composition of the MPCMIV Benchmark
L. Giaccardi a, S. Di Pasquale a, S. Dulla b, M. Cherubini a and A. Petruzzi a
aNuclear and Industrial Engineering (NINE), Via della Chiesa XXXII 759, 55100 Lucca, Italy
bPolitecnico di Torino, Dipartimento Energia, NEMO group, Corso Duca degli Abruzzi 24, 10129 Torino, Italy
International Conference on Physics of Reactors 2022 (PHYSOR 2022)
Pittsburgh (PA), USA, May 15–20, 2022
Abstract — The Ad Hoc Depletion Strategy elaborated by the NINE company, developed in support of the organization of the MPCMIV (Multi-physics Pellet Cladding Mechanical Interaction Validation) benchmark input and output specifications, is presented. This work aims at illustrating the strategy itself and then showing the results obtained with its application over the Studsvik R2 Testing Reactor, which is analyzed in the benchmark. The objective of the application of the strategy is to compute the fuel elements isotopic compositions at the beginning of some core loadings of interest for the benchmark. To this objective, it is necessary to implement first the simulation model of the three single assembly types and perform the infinite lattice depletions, then, to build the full core model and to perform the simulation of the core cycle. All the models and simulations were carried out with the use of the Monte Carlo particle transport code Serpent 2. Finally, the simulations results are assessed against Studsvik isotopic compositions of the fuel elements discharged from the R2 Reactor at the end of the core loading. Several assumptions were necessary during all the steps of the strategy, to overcome the lack of information regarding the core management. For this reason, the solution found at the end of the current analysis may not be completely optimized and further improvements regarding the model assumptions will be tested in a future work.
KEYWORDS: MPCMIV, Serpent 2, infinite lattice depletion, Core cycle, R2 Testing Reactor
Analysis of the Reactivity Effects Exercises of the Neutronics Benchmark of the CEFR Start-Up Tests
S. Di Pasquale a, M. Cherubini a, A. Petruzzi b and V. Giusti c
aNuclear and Industrial Engineering (NINE), Via della Chiesa XXXII 759, 55100 Lucca, Italy
bDipartimento di Ingegneria Civile ed Industriale (DICI), Università di Pisa, Italy, Largo Lucio Lazzarino (accanto all’edificio C), 56122 Pisa, Italy
International Conference on Physics of Reactors 2022 (PHYSOR 2022)
Pittsburgh (PA), USA, May 15–20, 2022
Abstract — The “Neutronics Benchmark of the CEFR Start-Up Tests” is an IAEA coordinated research project based on the simulation of the CEFR start-up tests. The main goals of the project are to improve the participant capabilities in SFR analysis and to perform an international validation of codes for Sodium Fast Reactor simulation. NINE-UNIPI work together on the creation of the Serpent 2 model and on the simulation of all the start-up tests proposed in the benchmark. In this work the three experiments related to the reactivity measurements are discussed. The geometry model is briefly described and the simulation set-up is presented. In particular, the geometry has been modeled considering the thermal expansion at the experimental temperatures. The nuclear data libraries used are the ENDF/B-VIII.0, pre-processed at the experimental temperatures and provided to the benchmark participants from SCK-CEN. The obtained results show a good agreement with the experimental data, except for the assembly-swap reactivity effect, which shows a small shift for all the considered cases. The results presented in this work could contribute to the validation of Serpent 2 for SFR criticality calculations.
KEYWORDS: CEFR. Serpent 2, SFR, Start-Up Tests, Validation
Thermal-Hydraulics Analysis of the IAEA CRP FFTF LOFWOS Test #13
Domenico De Luca, Kaiyue Zeng, Marco Cherubini, Alessandro Petruzzi
Nuclear and Industrial Engineering (NINE), Via della Chiesa XXXII 759, 55100 Lucca, Italy
HND2022 - 13th International Conference of the Croatian Nuclear Society
Zadar, Croatia, June 5 – 8, 2022
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. Within this framework, the IAEA organized a Coordinated Research Projects (CRP) on FFTF Loss of Flow Without Scram (LOFWOS) Test #13, aimed at improving Member States’ fast reactor analytical simulation capabilities, international validation, and qualification of codes currently employed in the field of fast reactor. The Fast Flux Test Facility (FFTF) was a 400 MW thermal powered, oxide-fueled, liquid sodium cooled test reactor built to assist development and testing of advanced fuels and materials for fast breeder reactors. The present paper shows the work performed by NINE for the CRP focused on benchmark analysis of one of the unprotected passive safety demonstration tests performed at the FFTF. In particular, a detailed nodalization was developed following the NEMM (NINE Evaluation Model Methodology) already applied for LWR safety analysis. After achievement of acceptable steady-state results, transient analysis was performed. In addition, the NINE validation procedure was adopted in order to validate the Simulation Model (SM) against the experimental data. Two system thermal-hydraulic codes, namely RELAP5 and TRACE, were used to analyse the selected test and the comparison between the two SM results is also presented in this paper. The final goal of the activity is to present the main outcomes achieved through the use of codes currently employed in the field of fast reactor, and how the application of the NEMM procedures allows to develop and qualify the SM results and validate the computer codes against experimental data.
MELCOR-To-MELCOR Coupling Method in Severe Accident Analysis Involving Core and Pent Fuel Pool
Hector Lopez, Alessandro Petruzzi, Walter Giannotti, Domenico De Luca
Nuclear and Industrial Engineering (NINE), Via della Chiesa XXXII 759, 55100 Lucca, Italy
HND2022 - 13th International Conference of the Croatian Nuclear Society
Zadar, Croatia, June 5 – 8, 2022
Abstract — A lot of effort has been spent to prevent the occurrence of SA in nuclear plant and to develop Severe Accidents (SA) Management to mitigate the consequences of a SA. Those consequences are mainly related to limit the release of fission product to the environment. The core in the vessel is not the only source of fission products as the Spent Fuel Pool (SFP) hosting the fuel removed by the core is, in some NPP, inside the containment and SA conditions can also occur. This is especially important in reactors having proximity between the RPV and SFP such as the VVER-1200. This close proximity implies that any SA occurring in the SFP potentially affects the RPV and vice-versa. This potential combination might cause unexpected evolution in the SA progression to whom the safety systems are not able to contain. MELCOR code is a widely used, flexible powerful SA code but it is incapable (due to the uniqueness of the COR package use inside the same input) to reproduce a situation in which both the fuel in vessel core and the fuel in the SFP, inside the same containment, are going to experience a severe accident scenario. The current study presents a MELCOR-to-MELCOR coupling method to simulate simultaneously scenarios with both, core and SFP, as sources capable of H2 generation, fuel damage and FP release in a VVER-1200 NPP. The coupling is performed by running two simulations in parallel and with the data exchange supervised and managed by a dedicated Python coupling supervising script developed at NINE.
Reactor Physics and Thermal Hydraulics Analyses for the OECD/NEA MPCMIV Benchmark
Luana Giaccardi, Domenico De Luca, Simone Di Pasquale, Marco Cherubini,Alessandro Petruzzi
Nuclear and Industrial Engineering (NINE), Via della Chiesa XXXII 759, 55100 Lucca, Italy
HND2022 - 13th International Conference of the Croatian Nuclear Society
Zadar, Croatia, June 5 – 8, 2022
Abstract — In order to complete the Multi-physics Pellet Cladding Mechanical Interaction Validation (MPCMIV) benchmark technical specifications, reactor physic and thermal hydraulic analyses have been performed. The work presented in this paper aims in particular to evaluate some of the missing Boundary and Initial Conditions necessary to complete the technical specifications, and also to perform some of the benchmark exercises connected with thermal hydraulic simulations. A far as the thermal hydraulic area is concerned, the analysis is carried out with the RELAP5 code. It is focused on the modelling of the in pile loop 1 located inside the R2 reactor core, in which a test fuel rodlet is inserted to perform some power ramp tests. The activity consists in the development of the simulation model of the in pile tube, the demonstration of the steady state achievement and the transient analysis of the first selected test, validating the simulation results against the benchmark experimental data. Considering the reactor physic area, the Monte Carlo code Serpent 2 is used to perform some single assemblies burn up calculations. The aim is to evaluate the initial composition of the fuel assemblies loaded in the core loadings of interest of the benchmark. Moreover, the temperature values to be used in the Serpent simulations are derived with thermal hydraulic simulations of the single assemblies. Further developments of the work will include the full core cycle analysis to validate the isotopic compositions and the complete model of the main circuit, using the gamma heating from the reactor physics calculations. Finally the TRANSURANUS fuel performance code will be adopted to compare the results against the available experimental data. A multi-physics effort is required to carry out the MPCMIV benchmark and appropriate coupling approach will be investigated and tested against the benchmark experimental results.
Simulation of the OECD/NEA Rod Bundle Heat Transfer (RBHT) Benchmark with RELAP5
Alessandro Del Ferraro, Domenico De Luca, Marco Cherubini, Alessandro Petruzzi
Nuclear and Industrial Engineering (NINE), Via della Chiesa XXXII 759, 55100 Lucca, Italy
HND2022 - 13th International Conference of the Croatian Nuclear Society
Zadar, Croatia, June 5 – 8, 2022
Abstract — The OECD/NEA RBHT (Rod Bundle Heat Transfer) Project is an International three-year NEA Joint Project whose objective is to conduct new experiments and evaluate system hydraulics and sub-channel codes in the simulation of reflood tests. Such tests are performed in a full height rod bundle facility equipped with advanced instrumentations capable to measure the real-time droplet field, cladding and steam/fluid temperatures, water carryover fraction and pressure drops. The test matrix encompasses both steady and oscillatory reflood inlet flow conditions. Within the RBHT project, a challenging benchmark exercise is conducted, including an open and a blind test phase providing a unique opportunity to project’s participants to validate codes and nodalization techniques. This paper presents a validation study of the RELAP5 code on the RBHT open test series. The simulations’ results generally well agree with the measured data, according to the accuracy metrics proposed by the benchmark team. A larger discrepancy is detected for experimental tests characterized by higher flooding rates with low subcooling degree. Several model’s parameters have been investigated including also different nodalization schemes to characterize the impact on the predicted results during the sensitivity analysis.
TOWARDS MODELLING DEFECTIVE FUEL RODS IN TRANSURANUS: BENCHMARK AND ASSESSMENT OF GASEOUS AND VOLATILE RADIOACTIVE FISSION PRODUCT RELEASE
L. Giaccardia, M. Cherubinia, G. Zullob, D. Pizzocrib, A. Magnib, L. Luzzib
a Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII 759, 55100, Lucca, Italy
b Politecnico di Milano, Department of Energy, Nuclear Engineering Division, Via La Masa 34, 20156, Milan, Italy
ANNALS OF NUCLEAR ENERGY, volume 197, March 2024, Article number 110249
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.
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
2 Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy
АТОМНА ЕНЕРГЕТИКА ATOMIC ENERGY, Volume 110609, March 2020
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.
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
2 Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy
ЯДЕРНА ФІЗИКА ТА ЕНЕРГЕТИКА / NUCL. PHYS. AT. ENERGY 23 (2022) 234-244
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.
NINE ha implementato una serie di interventi di adeguamento della propria infrastruttura informatica per una più efficiente ed efficace gestione e protezione dei dati, sia con investimenti in hardware che in software. Allo stesso tempo sono stati sostenuti costi per far fronte alle nuove esigenze lavorative causate dall’emergenza epidemiologica COVID-19, inclusi arredi da ufficio.
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