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.
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.
OECD RIA Benchmark Phase II – Towards a better understanding of the RIA fuel modelling
Olivier Marchanda, Jinzhao Zhangb, Marco Cherubinic
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.
Investigations on RELAP5-3D to RELAP5-3D Coupling Methodology by PVMEXEC
Valeria Parrinelloa , Marco Cherubinia, Alessandro Petruzzia and Marco Lanfredinib
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.
Coupling RELAP5-3D Models by PVMEXEC
Valeria Parrinelloa, Marco Cherubinia, Alessandro Petruzzia and Marco Lanfredinib
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.
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.