RELAP5-3D Analysis of EBR‐II Shutdown Heat Removal Test SHRT‐17

 

D. De Luca, A. Petruzzi, M. Cherubini

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

 

NUTHOS-11: The 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety

Gyeongju, Korea, October 9-13, 2016

 

Abstract — Coordinated Research Project (CRP) on EBR-II Shutdown Heat Removal Tests (SHRT) was established by International Atomic Energy Agency (IAEA). The objective of the project is to support and to improve validation of simulation tools and projects for Sodium-cooled Fast Reactors (SFR). The Experimental Breeder Reactor II (EBR-II) plant was a uranium metal-alloy-fuelled liquid-metal-cooled fast reactor designed and operated by Argonne National Laboratory (ANL) for the U.S. Department of Energy at the Argonne-West site.

In the frame of this project, benchmark analysis of one of the EBR-II shutdown heat removal tests, protected loss-of-flow transient (SHRT-17), has been performed. The aim of this paper is to present modeling of EBR-II reactor design using RELAP-3D, to show the results of the transient analysis of SHRT-17, and to discuss the results of application of the Fast Fourier Transform Based Method (FFTBM) to perform a quantitative accuracy evaluation of the model developed.
Complete nodalization of the reactor was made from the beginning. Model is divided in primary side that contains core, pumps, reactor pool and, for this kind of reactor specific, Z pipe, and intermediate side that contains Intermediate Heat Exchanger (IHX).

After achievement of acceptable steady-state results, transient analysis was performed. Starting from full power and flow, both the primary loop and intermediate loop coolant pumps were simultaneously tripped and the reactor was scrammed to simulate a protected loss-of-flow accident. In addition, the primary system auxiliary coolant pump, that normally had an emergency battery power supply, was turned off. Despite early rise of the temperature in the reactor, the natural circulation characteristics managed to keep it at acceptable levels and cooled the reactor down safely at decay heat power levels.
Thermal-hydraulics characteristics and plant behavior was focused on prediction of natural convection cooling by evaluating the reactor core flow and temperatures and their comparison with experimental data that were provided by ANL.

Finally, the process of qualification of a system thermal-hydraulic code calculation was applied. It consists of three steps: 1) the geometrical fidelity of the nodalization, related with the evaluation and comparison of the geometrical data of the hardware respect to the estimated numerical values implemented in the nodalization; 2) the steady state level qualification, dealing with the capability of the nodalization to reproduce the steady state qualified conditions of the system; 3) the “on-transient” qualification, necessary to demonstrate the capability of the code and of the developed nodalization to reproduce the relevant thermal-hydraulic phenomena expected during the transient. The latter is a very complex step which foreseen different phases following our methodology of qualification (SCCRED, Standardized and Consolidated Calculated & Reference Experimental Database methodology). In the framework of the benchmark, the focus was only on the so called “Quantitative Accuracy Evaluation” that is performed by the FFTBM.

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Development and Qualification of RELAP5-3D Nodalization of the Core of OPAL RR

 

D. De Luca, A. Petruzzi, M. Eaton, V. Badalassi, J. Scott, V.Mottl

 

NUTHOS-11: The 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety

Gyeongju, Korea, October 9-13, 2016

 

Abstract — The Australian Radiation Protection and Nuclear Safety Agency (ARPANSA) is the country’s primary authority on radiation and nuclear safety. One of its licensed facilities is the Open Pool Australian Lightwater (OPAL) reactor, a 20 MW multipurpose Research Reactor (RR) for radioisotope production, irradiation services and neutron beam research. The OPAL RR uses Low Enriched Uranium (LEU) fuel in a compact core, cooled by light water and moderated by heavy water. It is Australia’s only operating research reactor. ARPANSA, the country’s independent nuclear regulator, issued a licence to operate OPAL RR in 2006.

The objective of present work is to evaluate the variations of coolant, cladding and fuel temperatures between two stationary conditions. The first corresponds to the nominal operating power of 20 MW, the level at which OPAL usually operates. The second condition is at a power of 28 MW. Although no such steady state power level increase is planned, the associated analysis provides insight into the margin of safety for increased reactor powers. The licence holder plans to request a minor change to reactor  power which effectively allows the reactor to achieve its design output of 20 MW (currently about 19.3 MW).

The SCCRED (Standardized Consolidated Calculated and Reference Experimental Database) methodology developed by NINE to qualify a thermal-hydraulic system code calculation is summarized and adopted to qualify the RELAP5-3D nodalization of OPAL RR and the two steady state calculations. The methodology consists of several steps including a) the creation of a Reference Data Set (RDS) to constitute a systematic reference for the development of the computer code model of the core of OPAL RR, b) the development of the nodalization sketches, c) the demonstration of the geometrical fidelity of the developed nodalization input and d) the subsequent qualifications at steady state and e) during transients.

Giving the project’s objective, a Best Estimate (BE) evaluation model of the core of OPAL RR was developed using the RELAP5-3D code. The relevant assumptions are connected with the boundary conditions of the developed model (limited to the core region) and, in particular, with the core inlet flow distribution and the outlet core pressure. The developed BE RELAP5-3D nodalization constitutes a very detailed model of the core of OPAL RR: a total of 875 hydraulic volumes, 184 heat structure components and 2035 axial meshes have been used to model the hydraulics and the heat structures parts of the core of OPAL RR. Giving the features of the developed model, it is currently used to predict stationary behavior inside the core and has yet to be qualified at transient level.

The results of the stationary analysis were consistent with those provided by the licensee. In particular the maximum fuel (centerline and surface) and coolant temperatures showed only a modest increase from 20 MW to 28 MW. The results give confidence that the safety margin in the fuel temperature limit is not challenged by modest increases in steady state power.

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OECD RIA Benchmark Phase II – Towards a better understanding of the RIA fuel modelling

 

Olivier Marchand^a, Jinzhao Zhang^b, Marco Cherubini^c

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

^b Tractebel (Engie), avenue Ariane 7, 1200 Brussels, Belgium

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

 

TOP FUEL 2016
Boise, Idaho, USA, September 11-16, 2016

 

Abstract — Reactivity-initated accident (RIA) fuel rod codes have been developed for a significant period of time and validated against their own  available database. However, the high complexity of the scenarios dealt with has resulted in a number of different models and assumptions adopted by code developers; additionally, databases used to develop and validate codes have been different depending on the availability of the results of some experimental programmes. This diversity makes it difficult to find the source of estimate discrepancies, when they occur.
In the framework of OECD/NEA/WGFS activities, a RIA fuel-rod-code benchmark was organized in 2010-2013. It consisted of a consistent set of four experiments on very similar highly irradiated fuel rods tested under different experimental conditions (NSRR VA-1, VA-3, CABRI CIP0-1 and CIP3-1). This benchmark reveals the need to better understand the basic models in each fuel rod code for realistic simulation of the complicated integral RIA tests with high burnup fuel rods.
A second phase of the RIA fuel-rod-code benchmark was thus launched early in 2014. This RIA benchmark Phase-II has been organized as two complementary activities:
- The first activity is to compare the results of different simulations on simplified cases in order to provide additional bases for understanding the differences in modelling of the concerned phenomena.
- The second activity is focused on the assessment of the uncertainty of the results. In particular, the impact of the initial states and key models on the results of the transient are to be investigated.
The detailed comparison of the results from the first activity is presented in this paper.
Based on the Phase-I and Phase-II conclusions, some generic recommendations can be made:
- Fuel and clad thermomechanical models (with the associated material properties) should be further improved and validated more extensively against a sound RIA database.
- Build-up of a comprehensive and robust database consisting of both separate-effect tests and integral tests should be pursued in the short term. In this way, both individual model validation and model integration into codes would be feasible.
- An assessment of the uncertainty of fuel thermo-mechanics is of high interest (which is consistent with the second activity of this RIA benchmark Phase-II).
Finally, as RIA fuel codes are more and more likely to be used for reactor accident studies, particularly for those involving safety analyses, the fuel rod failure criteria (generally used in such analyses) will have to be carefully justified and validated. The current RIA fuel failure criteria are mainly based on the fuel thermal variables and the verification is based on “conservative” assumptions for the heat transfer conditions. As all codes give rather consistent evaluations of such variables, it appears possible, taking into account adequate provisions, to derive criteria based on thermal variables from experimental values or from an analytical approach. However, if in the future more mechanistic modelling is ever to be used to establish fuel-failure criteria based on mechanical variables, the codes will have to be further improved and validated for all the aspects identified above.

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Investigations on RELAP5-3D to RELAP5-3D Coupling Methodology by PVMEXEC

 

Valeria Parrinello^a , Marco Cherubini^a, Alessandro Petruzzi^a and Marco Lanfredini^b

^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

 

Embedded Topical Meeting on Advances in Thermal Hydraulics—2016 (ATH ’16)

June 12–16, 2016, New Orleans, LA, Hyatt Regency

 

Abstract — In the framework of a BEPU (Best Estimate Plus Uncertainty) approach within the licensing process of a nuclear power plant, the need to extend the resources of nuclear system thermal-hydraulics codes, such as RELAP5-3D, arises to allow more detailed simulations of the complex 3D reality of Nuclear Power Plants (NPPs), either under normal steady-state or during various accident scenarios.

Currently, it is not possible to achieve the same degree of detail for a whole nuclear system when it is simulated with RELAP5-3D and this is due to the inherent limitations in the number of components and volumes to be used for the analysis. For this reason, it is of extreme interest the use of tools for codes coupling that enable the use of different codes for the simulation of different portions of a system in a unified analysis.

In this paper the attention will be focused on the decomposition of the thermal-hydraulic domain of a system into subsystems to be simulated by different instances of the same code (e.g. RELAP5-3D) coupled together by means of PVMEXEC program and parallel virtual machine (PVM) technology. Explicit and semi-implicit solution algorithms were used for the analyses.

Among the analyzed cases, the following will be discussed in detail with the aim to provide additional guidelines for the use of the PVMEXEC tool: (i) the Edward’s pipe blowdown test, (ii) a simplified countercurrent heat exchanger, (iii) different hydraulics and heat structure coupling schemes for a shell-tube heat exchanger and (iv) a three-task coupled model of an adaptation of the Christensen subcooled boiling experiment.

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Coupling RELAP5-3D Models by PVMEXEC

 

Valeria Parrinello^a, Marco Cherubini^a, Alessandro Petruzzi^a and Marco Lanfredini^b

^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

 

24th International Conference on Nuclear Engineering (ICONE24)

Charlotte, North Carolina, USA, June 26-30, 2016

 

Abstract — In the framework of a Best Estimate Plus Uncertainty (BEPU) approach within the licensing process of a nuclear power plant (NPP), it is of great importance to achieve a realistic simulations of the complex 3D reality of NPPs, either under normal steady-state or during various accident scenarios.
Currently, it is not possible to achieve the same degree of detail for a whole nuclear system when it is simulated with RELAP5-3D and this is due to the inherent limitations of best estimate (BE) System Thermal-Hydraulic (SYS-TH) codes, i.e. approximations of equations and models implemented in the codes and maximum number of components to be used for the analysis. The need to extend the nuclear SYS TH codes’ resources, such as RELAP5-3D, arises to allow a more detailed representation of the system. For this reason, it is of extreme interest the use of tools for codes coupling that enable the use of different codes for the simulation of different portions of a system in a unified analysis.
In this paper the attention will be focused on the decomposition of the thermal-hydraulic domain of a system into subsystems to be simulated by different instances of RELAP5-3D coupled together by means of the PVMEXEC program and the parallel virtual machine (PVM) technology.
Different solution algorithms were used for carrying out the analyses of a large variety of sample models. Among the investigated cases, the following will be discussed in detail with the aim to provide additional guidelines for the use of the PVMEXEC tool: (i) the Edward’s pipe blowdown test, (ii) a simplified countercurrent heat exchanger, (iii) different hydraulics and heat structure coupling schemes for a shell-tube heat exchanger and (iv) a three-task coupled model of a simplified BWR model.

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ATLAS A5.1 Blind Test Calculation

 

V. Parrinello, M. Cherubini, A. Petruzzi

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

 

NUTHOS-11: The 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety

Gyeongju, Korea, October 9-13, 2016

 

Abstract — The main objective of the OECD/NEA ATLAS Project is to provide experimental data for resolving LWR thermal-hydraulics safety issues related to multiple high-risk failures by using the ATLAS facility (Advanced Thermal-Hydraulic Test Loop for Accident Simulation) at KAERI (Korean Atomic Energy Research Institution).
The experimental program comprises several tests in different research topics to be conducted at the ATLAS facility. The ATLAS A5.1 test pertains to a small break loss-of-coolant accident (SBLOCA). It is a counterpart test of LSTF 1% SBLOCA SB-CL-32, and its boundary conditions were derived from the aforementioned test, adopting proper scaling parameters.
The aims of ATLAS test are to demonstrate the progression of the cold leg line break accident and to contribute to the understanding of the behavior of nuclear reactor systems and to the assessment and improvement of the existing best estimate codes.
The present paper deals with the thermal-hydraulic analysis carried out for the ATLAS A5.1 blind test calculation in order to reproduce the phenomena occurring during the cold leg line break. The adopted system thermal-hydraulic code to perform the analysis is RELAP5Mod3.3.
A methodology was pursued for developing the nodalization and qualifying the predicted code calculations, i. e. the SCCRED methodology (Standardized Consolidated Calculated and Reference Experimental Database), a supporting tool for V&V and uncertainty evaluation of Best-Estimate system codes for licensing applications.

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THE CASUALIDAD METHOD FOR UNCERTAINTY EVALUATION OF BEST-ESTIMATE SYSTEM THERMAL-HYDRAULICS CALCULATIONS

 

A. Petruzzi

Nuclear and INdustrial Engineering (NINE), Borgo Giannotti, Lucca, Italy

 

17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17)
Xi'an, China, 03 - 08 September 2017

 

Abstract — The present paper deals with the description of the salient features of three independent approaches for estimating uncertainties associated with predictions of complex system codes. The 1st approach is the “standard” one and the most used at the industrial level: it is based upon the selection of input uncertain parameters, on assigning related ranges of variations and, possibly, PDF (Probability Density Functions) and on performing a suitable number of code runs to get the combined effect of variation on the results. In the 2nd approach the uncertainty derives from the comparison between relevant measured data and results of corresponding code calculations. The 3rd approach is based on the Bayesian inference technique and on the availability of experimental data by which computer model predictions can be improved and the ranges of variation of (in theory) „all‟ input parameters can be characterized. More details are provided in respect with the third approach that has been named CASUALIDAD (Code with the capability of Adjoint Sensitivity and Uncertainty AnaLysis by Internal Data ADjustment and assimilation).

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ATUCHA-1 NPP CONTAINMENT VENTING ANALYSIS FOLLOWING SBO AND LBLOCA EVENTS BY GOTHIC CODE

 

A. Pop^{a b}, W. Giannotti^a, F. Terzuoli^a, A. Petruzzi^a, N. Forgione^b, O. Mazzantini^c

^a Nuclear and INdustrial Engineering (NINE), Borgo Giannotti, Lucca, Italy

^b University of Pisa, Largo Lucio Lazzarino 2, 56126 Pisa, Italy

^c Nucleoeléctrica Argentina Arribeños 3619, C1429BKQ, Buenos Aires, Argentina

 

17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17)
Xi'an, China, 03 - 08 September 2017

 

Abstract — Containment behaviour plays a key role in the safety framework of a Nuclear Power Plant. The GOTHIC thermal hydraulic code has been adopted to evaluate the Atucha-1 NPP containment responses during two postulated accident scenarios, Station Black Out and Large Break Loss of Coolant Accident, while assuming 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 ex-vessel cooling, which could cause an excessive pressurization of the containment, 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 monitored 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 cells. While 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, e.g. all doors and blow off panels have been simulated to open with the designed differential pressure logic.
Both accident transients, for each type of nodalization, were simulated for 200,000 seconds. Results showed that for the Large Break Loss of Coolant Accident pressure is predicted to reach around 5.25 bar, while the Station Black Out Scenario reaches 4.4 bar.

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RELAP5-3D ANALYSIS AND QUALIFICATION OF EBR-II SHRT-17

 

D. De Luca, A. Petruzzi, M. Cherubini

Nuclear and INdustrial Engineering (NINE), Borgo Giannotti, Lucca, Italy

17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17)
Xi'an, China, 03 - 08 September 2017

 

Abstract — Coordinated Research Project (CRP) on EBR-II Shutdown Heat Removal Tests (SHRT) was established by International Atomic Energy Agency (IAEA) in June 2012. The objective of the project is to support and to improve validation of simulation tools and projects for Sodium-cooled Fast Reactors (SFR).  In the frame of this project, benchmark analysis of one of the EBR-II shutdown heat removal tests, protected loss-of-flow transient (SHRT-17), has been performed.
The aim of this paper is to present modeling of EBR-II reactor design using RELAP-3D, to show the results of the transient analysis of SHRT-17, and to discuss the results of application of the qualification process, in particular of the Fast Fourier Transform Based Method (FFTBM) to perform a quantitative accuracy evaluation of the model developed.
Complete nodalization of the reactor was made from the beginning. After achievement of acceptable steady-state results, transient analysis of the protected (with scram) loss-of-flow test was performed. Thermal-hydraulics characteristics and plant behavior was focused on prediction of natural convection cooling by evaluating the reactor core flow and temperatures and their comparison with experimental data that were provided by ANL.
Finally, the process of qualification of a system thermal-hydraulic code calculation was applied. It consists of three steps: 1) the geometrical fidelity of the nodalization, related with the evaluation and comparison of the geometrical data of the hardware respect to the estimated numerical values implemented in the nodalization; 2) the steady state level qualification, dealing with the capability of the nodalization to reproduce the steady state qualified conditions of the system; 3) the “on-transient” qualification, necessary to demonstrate the capability of the code and of the developed nodalization to reproduce the relevant thermal-hydraulic phenomena expected during the transient. The latter is a very complex step which foreseen different phases following our methodology of qualification (SCCRED, Standardized and Consolidated Calculated & Reference Experimental Database methodology). In the framework of the benchmark, the focus was only on the so called “Quantitative Accuracy Evaluation” that is performed by the FFTBM.

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IAEA's Coordinated Research Project on EBR-II Shutdown Heat Removal Tests

 

 

Chirayu Batra (IAEA), Vladimir Kriventsev (IAEA), Laural L. Briggs (ANL), Stefano Monti (IAEA), Wenjun HU (CIAE), Dan-ting Sui (North ChinaElectric Power Uniersity), Guanghui Su (Xi'an Jiaotong University), Ludovic Maas (IRSN (French Institute for Nuclear Safety)), Barbara Vezzoni (KIT), Partha Sarathy UPPALA (IGCAR), Alessandro Del Nevo (ENEA), Alessandro Petruzzi (Nuclear and INdustrial Engineering (NINE)), Roberto Zanino (Politecnico di Torino-Italy), Hiroaki Ohira (JAEA), Willem F. van Rooijen (University of Fukui), K. Morita (Kyushu Univ), Chiwoong CHOI (KAERI), Andong Shin (KINS), M. Stempniewicz (Nuclear Research & Consultancy Group), Nikita Rtischev (IBRAE), Konstantin Mikityuk (Scherrer Inst), Ethan A. Bates (TerraPower)

 

17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17)
Xi'an, China, 03 - 08 September 2017

 

Abstract — A Coordinated Research Project (CRP) on “Benchmark Analysis of EBR-II Shutdown Heat Removal Tests (SHRT)” was launched by the International Atomic Energy Agency (IAEA) in 2012. A series of transient tests were conducted on the EBR-II reactor at Argonne National Laboratory (ANL) to improve the understanding of thermal hydraulics and neutronics of fast reactors. Shutdown heat removal tests conducted in 1984 and 1986 demonstrated mechanisms by which fast reactors can survive severe accident initiators with no core damage. Two SHRT tests, SHRT-17 representing Protected Loss of Flow (PLOF) transients and SHRT-45R representing Unprotected Loss of Flow (ULOF) transients, were studied in the IAEA CRP.  
The objectives of the CRP were to improve design and simulation capabilities in fast reactor thermal hydraulics, neutronics and safety analyses through benchmark analysis of these two important tests. At the first stage of the benchmark, ANL provided the input data on EBR-II geometry, as well as initial and boundary conditions for the SHRT-17 and SHRT-45R tests to perform “blind” calculations. At the second stage, ANL released the experimental observations and participants had the chance to analyze the difference and refine the models. At the third stage, a methodology to systematically analyze and compare the models and the results of each participant was applied. Nineteen organizations from eleven countries participated in the CRP, making it one of the largest CRP coordinated by the IAEA fast reactor team.  
The paper provides a general CRP overview, gives the basics of the EBR-II reactor design, describes the shutdown heat removal tests, the benchmark setup, and results of numerical simulations, followed by a detailed discussion on the EBR-II CRP.

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SUMMARY OF SWINTH-2016: A SPECIALISTS WORKSHOP ON ADVANCED INSTRUMENTATION AND MEASUREMENT TECHNIQUES FOR NUCLEAR REACTOR THERMAL-HYDRAULICS EXPERIMENTATION

 

F. Moretti^a, F. D'Auria^b, N. Aksan^b, S. Lutsanych^b, K. Umminger^c, S. Gupta^d, D. Bestion^e, U. Hampel^f, K.Y. Choi^g, H. Purhonen^h, A. Gubaⁱ, D. Paladinoi ^j

^aNuclear and INdustrial Engineering (NINE), Borgo Giannotti, Lucca, Italy

^bSan Piero a Grado Nuclear Research Group (GRNSPG) – University of Pisa, Italy

^cAREVA GmbH, Germany

^dBecker Technologies GmbH, Germany

^eCommissariat à l’Energie Atomique et aux Energies Alternatives (CEA) – Grenoble, France

^fHelmholtz Zentrum Dresden Rossendorf (HZDR), Germany

^gKorean Atomic Energy Research Institute (KAERI), Republic of Korea

^hLappeenranta University of Technology (LUT), Finland

ⁱHungarian Academy of Science Centre for Energy Research (MTA EK), Hungary

^jPaul Scherrer Institut (PSI), Switzerland

 

17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17)
Xi'an, China, 03 - 08 September 2017

 

Abstract — A “Specialists Workshop on Advanced Instrumentation and Measurement Techniques for Nuclear Reactor Thermal-Hydraulics (SWINTH)” was held in Italy on June 2016, promoted and organized by SILENCE, the “Significant Light and Heavy Water Reactor Thermal Hydraulic Experiments Network for the Consistent Exploitation of the Data”. The workshop covered the technology of experimentation in nuclear thermal-hydraulics, including both separate effect and integral test facilities, with coarse as well as CFD-grade measurements (although most contributions were actually devoted to the latter type), and not limited to light water-cooled reactor technology. The last international workshop with the same scope was an OECD/NEA/CSNI meeting held in Santa Barbara, CA, in 1997.
The main objectives of the workshop were to gain an up-to-date picture of the current trends in experimental programs and in technological advancement as to reactor thermal-hydraulics, to promote the technical-scientific exchange between experimentalists and scientists operating on the code development and validation side, and to help identify the existing gaps between the code validation needs and the available technology. The response of the international community to this initiative was very positive and brought valuable contribution to meeting the above-mentioned objectives through generally high-quality papers and presentations, many of which focusing on innovative solutions and state-of-the-art applications. The present paper provides an overview of the contents of the workshop papers and tries to identify lessons learned and recommendations.
In addition, questionnaires “to identify measurement needs and main gaps for further system code and CFD code development and validation” had been distributed prior to the workshop among a number of people active in this area, and the preliminary outcomes of the survey were presented at the workshop. This paper also gives an account of such outcomes. Any feedback to this paper will be taken into account for the organization of the future SWINTH workshops being planned, possibly within an OECD/NEA framework and with a scope extended to in-plant instrumentation (e.g. for Severe Accident Management).

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Application of Integrated RELAP5-Transuranus Approach to a RIA Benchmark Exercise Including Uncertainty and Sensitivity

 

F. Moretti, M. Cherubini, A. Pop

Nuclear and INdustrial Engineering (NINE), Borgo Giannotti, Lucca, Italy

 

17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17)
Xi'an, China, 03 - 08 September 2017

 

Abstract — A Reactivity Initiated Accident (RIA) Fuel Codes Benchmark was organized by the OECD/NEA Working Group on Fuel Safety. While the phase-I of the exercise focused on the analysis of power pulse tests performed on research reactors, the phase-II dealt with the benchmarking and comparison of participants’ modelling approaches: (task 1) to the analysis of several idealized RIA cases, and (task 2) with sensitivity analysis and uncertainty quantification (through the analysis of one particular RIA case). This paper describes the work performed by NINE as one of the participants to phase-II, and particularly on task 2. Such work involved the integrated use of two different codes, namely RELAP5 and TRANSURANUS for the thermal hydraulic and the fuel performance parts of the analysis respectively, the integration consisting in an off-line one-way sequential coupling. According to the benchmark specifications, the uncertainty analysis was performed by adopting a Monte Carlo statistical approach, requiring 200 runs with a number of uncertain input parameters randomly sampled from a specified Gaussian distribution, and by applying a 95/95 statistics to the selected output variables. Moreover, the sensitivity analysis was performed by calculating the Spearman’s rank correlation coefficients of specified output variables versus specified input parameters. The whole analysis process was handled through a Python-based “Statistical RELAP5-TRANSURANUS” (StaRT) coupled analysis programme developed on purpose. StaRT takes care of input parameter sampling, update of RELAP5 and TRANSURANUS input decks, post-processing of both codes’ results, uncertainty and sensitivity analyses. All the results produced at key steps of the process are stored in a single HDF5 file. StaRT turned out to be an effective tool for the purposes of the benchmark as well as for future applications to similar studies.

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ATLAS A5.1 POST-TEST CALCULATION

 

Valeria Parrinello, Marco Cherubini

Nuclear and INdustrial Engineering (NINE), Borgo Giannotti, Lucca, Italy

 

17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17)
Xi'an, China, 03 - 08 September 2017

 

Abstract — The ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation) is a full pressure, full temperature integral test facility operated by KAERI (Korean Atomic Energy Research Institution). The experimental program to be conducted at the ATLAS facility in the framework of the OECD-ATLAS Project comprises about ten tests pertaining to different research topics.
Among these tests, the A5.1 experiment was performed targeting a 1% horizontal small break loss of coolant accident (SBLOCA) at the cold leg with cold leg injection of the emergency core cooling system (ECCS) and accident management (AM) action by means of secondary side depressurization. This test is the counterpart test of the SB-CL-32 test conducted using the Large Scale Test Facility (LSTF) of the Rig-of-Safety Assessment-V (ROSA-V) Program.
The two rigs are designed in accordance to different scaling approaches and the counterpart test design implied a preliminary scaling analysis for deriving the boundary conditions of the ATLAS A5.1 test by means of proper scaling parameters. The aims of ATLAS experiment are to demonstrate the progression of the cold leg line break accident and contribute to the understanding of the behavior of nuclear reactor systems and to the assessment and improvement of the existing best estimate codes.
The present paper deals with the thermal-hydraulic analysis carried out for the ATLAS A5.1 post-test calculation. The adopted system thermal-hydraulic code to perform the analysis is RELAP5Mod3.3.
A qualification procedure was pursued for developing the nodalization to be used for the analysis and demonstrating that the code results constitute a realistic approximation of the A5.1 test behavior. The bases of the procedure are embedded into the SCCRED (Standardized and Consolidated Calculated & Reference Experimental Database) methodology, a supporting tool for V&V and uncertainty evaluation of Best-Estimate system codes for licensing applications.

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ATLAS A5.1 TEST BENCHMARK ACTIVITY

 

V. Parrinello^a, M. Cherubini^a, K.Y. Choi^b

^a Nuclear and INdustrial Engineering (NINE), Borgo Giannotti, Lucca, Italy
^b Korean Atomic Energy Research Institute (KAERI), Republic of Korea

 

17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17)
Xi'an, China, 03 - 08 September 2017

 

Abstract — The Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS) was operated by Korea Atomic Energy Research Institute (KAERI) for accident simulations of advanced Pressurized Water Reactors (PWRs). In the framework of the OECD-ATLAS Project, the ATLAS A5.1 test was performed, targeting a 1% horizontal small break loss of coolant accident (SBLOCA) at the cold leg with cold leg injection of the emergency core cooling system (ECCS) and accident management (AM) action by means of secondary side depressurization. This test is the counterpart test of the SB-CL-32 test conducted using the Large Scale Test Facility (LSTF) of the Rig-of-Safety Assessment-V (ROSA-V) Program.
Fourteen institutions from ten different countries took part in the analytical benchmark on the ATLAS A5.1 test. The participants were asked to provide a total of 128 parameters and 54 time trends. This large amount of data was used to characterize the code models developed by the benchmark participants and to determine how accurately these computational models are able to represent the ATLAS behavior.
Qualitative and quantitative assessments are the two pillars of the qualification process adopted within the benchmark activity, whose bases are embedded into the SCCRED (Standardized and Consolidated Calculated & Reference Experimental Database) methodology typically adopted by NINE.
The comparison among participants’ data included a comparison of the developed nodalization features, the nodalization geometrical fidelity, and the simulation of the steady state and transient behavior of the facility. A quantitative evaluation of the results was performed too by the adoption of the FFTBM (Fast Fourier Transform Based Method).
The complex work performed by NINE for the analytical benchmark activity and the adopted qualification process will be described and the outcomes of the comparison between the experimental data and the calculation results submitted by the participants will be presented.

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Defence in Depth and practical elimination of early and large releases – concepts and practice.

 

S. Michael Modro^a , Tomislav Bajs^b, Artur Lyubarskiy^c

^a Nuclear and INdustrial Engineering (NINE), Borgo Giannotti, Lucca, Italy
^b ENCONET d.o.o., Miramarska 20, 10000 Zagreb, Croatia
^c Atomenergoprojekt, Moscow, Podolskich Kursantov 1 st , Russian Federation

 

TopSafe 2017
Vienna, Austria, 12 - 16 February 2017

 

Abstract — One of the most important requirements of the International Atomic Energy Agency Safety Standard on design (SSR-2/1) is the requirement for practical elimination of accident sequences that would lead to large or early radioactive releases. This requirement is accompanied by requirements for extension of the plant design basis to include conditions considered earlier as “beyond design basis” which may include plant conditions with multiple failures of systems, including safety systems, without and with severe core damage. In this paper we are discussing the concept of practical elimination of large or early radioactive releases in the context of defence in depth and we are addressing some challenges for demonstration of this concept. The demonstration of the “practical elimination” is a complex process for which no specific guidance and criteria are available and not sufficient experience is so far generated. The “practical elimination” shall be considered as integral part of the defence in depth.

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DEVELOPMENT OF A BEST-ESTIMATE THERMAL HYDRAULICS MODEL OF THE HPR-1000 NPP FOR DEVELOPING/VERIFYING EOP

 

D. De Luca¹, V. Parrinello¹, S. Huang², M. Cherubini¹, A. Petruzzi¹ and C. Yang²

¹Nuclear and INdustrial Engineering (NINE), Borgo Giannotti, Lucca, Italy

²China Nuclear Power Engineering Company, Beijing 100840, China

 

BEPU 2018 - ANS Best Estimate Plus Uncertainty International Conference
Lucca, Italy, May 13-18, 2018

 

Abstract — Emergency Operating Procedures (EOPs) are essential for maintaining fundamental safety functions and preventing core damage during design basis accidents and beyond design basis accidents in a nuclear power plant. For the development and maintenance of EOPs and accident management procedures, best estimate codes together with realistic assumptions should be used. In this activity, a best estimate model of the HPR-1000 NPP has been developed and the preliminary simulations of four selected accident scenario (i.e. LOFW, MBLOCA, LBLOCA and SGTR) have been performed in order to investigate the system behaviour. Among these, the LOFW accident scenario with the feed and bleed procedures to remove the core decay heat is discussed in this paper. From the analysis of the results it can be seen that the effectiveness of the procedures mostly depends on the starting time of the operator actions and on the number of the PRZ safety valves that are request to operate. However, the present work has to be considered as a part of a more general framework of activities whose final goal is the development of a best estimate evaluation model of the selected NPP to be used for performing the accident safety analysis (including independent safety analysis activities to support the licensing) and the development of Emergency Operational Procedure (EOP). In order to reach this final goal, several additional activities must still be done, such as the validation of the evaluation model and the quantification of the uncertainties.

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VALIDATION OF MULTI-PHYSICS SIMULATION TOOLS USING FUEL RAMP TEST IN R2 REACTOR: THE MPCMIV BENCHMARK

 

A. Petruzzi¹, D. De Luca¹, J. Karlsson², T. Valentine³

¹Nuclear and INdustrial Engineering - NINE, Via della Chiesa XXXII 759, Lucca, Italy
²Studsvik Nuclear AB, SE-611 82 Nyköping, Sweden
³Oak Ridge National Laboratory, Oak Ridge, TN, USA

 

BEPU 2018 - ANS Best Estimate Plus Uncertainty International Conference
Lucca, Italy, May 13-18, 2018

 

Abstract — High-fidelity, multi-physics Modeling and Simulation (M&S) tools are being developed and utilized for a variety of applications in nuclear science and technology and show great promise in their abilities to reproduce observed phenomena for many applications. Even with the increasing fidelity and sophistication of coupled multi-physics M&S tools, the underpinning models and data still need to be validated against experiments that may require a more complex array of validation data because of the great breadth of the time, energy and spatial domains of the physical phenomena that are being simulated. The Expert Group on Multi-Physics Experimental Data, Benchmarks and Validation (MPEBV) of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) was formed to address the challenges with the validation of such tools. In this framework, NINE and STUDSVIK promoted a benchmark titled Multi-physics Pellet Cladding Mechanical Interaction Validation (MPCMIV). The MPCMIV exercise is based upon cold ramp tests conducted at the Studsvik R2 reactor test loop that requires coupling of reactor physics, thermal hydraulics, and fuel performance phenomena. The aim of the experiment was to investigate the fuel response at cold criticality conditions to examine whether or not the potential fuel failure mechanisms might differ at temperatures below 100 °C than at normal operating conditions. The validation exercise is structured into four tiers in order to maximize participation by various groups. The first tier is targeted for novel M&S tools that have the capability to model the 3D heterogeneous model for both the reactor core domain and the fuel rod domain; the second tier involves the use of a simplified model for novel M&S tools that utilizes boundary conditions for the reactor physics models of the R2 reactor core; the third tier involves the same simplified model of tier 2 but allows for the use of traditional M&S tools; and the four tier is target for the application of only fuel performance tools. For each tier, the MPCMIV exercise is structured into four phases: the model development phase, the pre-qualification phase, the blind simulation phase, and the post-test phase. For each of these phases the participants will establish their validation requirements and assumptions that are made whether it be for steady state models or transient models.

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MCNP6 UNCERTAINTY QUANTIFICATION APPLIED TO UAM-EXSERCISE I-1

 

L. Lampunio¹, V. Giusti², V. Parrinello¹ and A. Petruzzi¹

¹Nuclear and INdustrial Engineering - NINE, Via della Chiesa XXXII 759, Lucca, Italy
²University of Pisa, Largo Lucio Lazzarino 2, Pisa (PI), Italy

 

BEPU 2018 - ANS Best Estimate Plus Uncertainty International Conference
Lucca, Italy, May 13-18, 2018

 

Abstract — Uncertainty quantification has proved itself to be a more and more important aspect of the nuclear society for best estimate predictions to be provided with their confidence bounds. Besides, the ability to compute the multiplication factor sensitivity coefficients for uncertainty estimation is also useful for code validation, the development of benchmark suites applicable to specific sets of applications and the design of critical (integral) experiments. In the context of the OECD/NEA project on Benchmark for Uncertainty Analysis in Modelling (UAM), the MCNP 6.1 code has been used for sensitivity and uncertainty analysis in criticality calculations for the Exercise 1 “Cell physics”, Phase I. MCNP 6.1 is the first version of MCNP with the possibility to calculate the sensitivity coefficients of the multiplication factor k for nuclear data. The methods employed are based upon linear-perturbation theory using adjoint weighting, performed in a single forward calculation by means of the Iterated Fission Probability method. The test case under consideration is the Three Mile Island Pressurized Water Reactor fuel pin in Hot Zero Power and Hot Full Power conditions. The results of the MCNP 6.1 code are presented and compared with the ones already available obtained using Serpent and SCALE 6.0 codes. In particular, the results used for the present comparison had been produced by the TSUNAMI-1D module of SCALE and Serpent version 2.1.22. It was found a good agreement between the aforementioned codes in k uncertainty, though there are differences when the sensitivity profiles are analysed. The discrepancy obtained with the three codes in the k values are also presented.

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APPLICATION OF THE TRANSURANUS CODE TO HIGH BURN-UP LOCA TEST IN VIEW OF 10 CFR 50.46c

 

M. Cherubini, L. Lampunio

Nuclear and INdustrial Engineering (NINE), Borgo Giannotti, Lucca, Italy

TopFuel 2018
Prague, Czech Republic, 30 September - 04 October 2018

 

Abstract — Nowadays, the implementation of new cladding materials and higher rod burn-ups have led to the necessity of re-examining the LOCA safety criteria and verifying their validity regarding the importance of hydrogen content on cladding embrittlement. U.S – NRC 10 CFR 50.46, “Acceptance criteria for emergency core cooling systems for light water nuclear power reactors,” is currently under revision to account for the influence of hydrogen on cladding embrittlement under LOCA conditions. In this work, the tests 191 and 192 conducted at the Hot Cell Laboratory of Studsvik Nuclear AB were simulated with TRANSURANUS code. Two base irradiation simulations were performed for both the father rod and the fuel rodlet to achieve consistent results. Consequently, to deal with the LOCA test, the restart option of TRANSURANUS was used to set the appropriate boundary conditions. The simulated data proved to be in agreement the experimental values for both phases of the exercise.

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3-D SYS-TH : an OECD/NEA activity on multi-dimensional capabilities of thermalhydraulic system

 

C. Herer^a, D. Bestion^b, P. Fillion^b, R. Prea^b, V.Parrinello^c, A. Bousbia Salah^d, K. Kim^e, J. J. Jeong^f

^aIRSN Institut de Radioprotection et de Sureté Nucléaire BP 17 92262 Fontenay-aux-Roses Cedex France

^bCEA, STMF, Université Paris-Saclay, 91191 Gif sur Yvette Cedex, France

^cNINE N uclear and INdustrial Engineering, Borgo Giannotti, Lucca, Italy

^dBelV, Belgium

^eKAERI, South Korea

^fPusan National University, South Korea

 

ICAPP 2019 - International Congress on Advances in Nuclear Power Plants
Juan-les-pins, France, 12 - 15 May 2019

 

Abstract — The evaluation model and computational capabilities required for engineering design and safety analyses of nuclear installations have shown significant progress compared to the first tools established in the 60s. Regarding thermalhydraulics, first generation system codes were based on simple one-dimensional models associated with conservatisms intended to cover lack of knowledge, simplifications and limited computational capabilities and limited experimental support available at that time.
The second generation, mainly one-dimensional with some limited multi-dimensional capabilities, implemented the more advanced two-fluid sixequation model, and adopted the best–estimate approach, benefited from an extensive experimental program, and showed large improvement compared to the first generation tools.
However, the current tools still have limitations that both industries and regulatory bodies would like to address. Next generation codes are being developed to achieve an improved thermal-hydraulic analysis capacity. Within the Working Group on Analysis and Management of Accidents (WGAMA) of the OECD/NEA, an activity has been initiated in 2016 aiming at establishing a state of the art of current 3D capabilities in thermal hydraulic system codes which covers all aspects and limitations, from the equations and simplifications considered, time and space averaging hypotheses with unavoidable use of relatively coarse meshes, closure models up to available or needed experimental support. The main findings of this activity are presented in this paper.

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Studsvik R2 Materials Test Reactor Ad Hoc Depletion Strategy for the Derivation of the Fuel Isotopic Composition of the MPCMIV Benchmark

 

Giaccardi L.^a, Di Pasquale S.^a, Dulla S.^b, Cherubini M.^a and Petruzzi A.^a

^aNuclear and Industrial Engineering (NINE), Via della Chiesa XXXII 759, 55100 Lucca, Italy

^aPolitecnico 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

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Analysis of the Reactivity Effects Exercises of the Neutronics Benchmark of the CEFR Start-Up Tests

 

Di Pasquale S.^a, Cherubini M.^a, Petruzzi A.^b and Giusti V.^c
a Nuclear and Industrial Engineering (NINE), Via della Chiesa XXXII 759, 55100 Lucca, Italy
^b Dipartimento 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

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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.

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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.

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