Thoughts in nuclear thermal hydraulics

Open thoughts for progressin in nuclear thermal hydraulics - brainstorming meeting (selected participants

Perspectives in System Thermal-Hydraulics

The paper deals with three main topics: a) the definition of System Thermal-Hydraulics (SYS TH), b) a historical outline for SYS TH and, c) the description of elements for reflection when planning research projects or improvement activities, this last topic being the main reason for the paper. Distinctions between basic thermal-hydraulics and computational Fluid-Dynamics (CFD) on the one side and SYS TH on the other side are considered under the first topic; stakeholders in the technology are identified. The proposal of Interim Acceptance Criteria for Emergency Core Cooling Systems in 1971 by US NRC (AEC at the time) is recognized as the starting date or the triggering event for SYS TH (second topic). The complex codes and the main experimental programs (list provided in the paper) constitute the pillars for SYS TH. Caution or warning statements are introduced in advance when discussing the third topic: a single person (or a researcher) has little to no possibility, or capability, of streamlining the forthcoming investments or to propose a roadmap for future activities. Nevertheless, the ambitious attempt to foresee developments in this area has been pursued without constraints connected with the availability of funds and with industrial benefits or interests. Demonstrating the acceptability of current SYS TH limitations and training in the application of those codes are mentioned as the main challenges for forthcoming research activities

Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures

In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Whereas the first system codes, developed at the beginning of the 1970s, utilized the homogenous equilibrium model with three balance equations to describe the two-phase flow, nowadays the more advanced system codes are based on the so-called "two-fluid model" with separation of the water and vapor phases, resulting in systems with at least six balance equations. The wide experimental campaign, constituted by the integral and separate effect tests, conducted under the umbrella of the OECD/CSNI was at the basis of the development and validation of the thermal-hydraulic system codes by which they have reached the present high degree of maturity. However, notwithstanding the huge amounts of financial and human resources invested, the results predicted by the code are still affected by errors whose origins can be attributed to several reasons as model deficiencies, approximations in the numerical solution, nodalization effects, and imperfect knowledge of boundary and initial conditions. In this context, the existence of qualified procedures for a consistent application of qualified thermal-hydraulic system code is necessary and implies the drawing up of specific criteria through which the code-user, the nodalization, and finally the transient results are qualified

the best estimate plus uncertainty challenge in the current licensing process of present reactors

Within the licensing process of the KWU Atucha II PHWR (Pressurized Heavy Water Reactor), the BEPU (Best Estimate Plus Uncertainty) approach has been selected for issuing of the Chapter 15 on FSAR (Final Safety Analysis Report). The key steps of the entire process are basically two: (a) the selection of PIE (Postulated Initiating Events) and (b) the analysis by best estimate models supported by uncertainty evaluation. Otherwise, key elements of the approach are (1) availability of qualified computational tools including suitable uncertainty method, (2) demonstration of quality, and (3) acceptability and endorsement by the licensing authority. The effort of issuing Chapter 15 is terminated at the time of issuing of the present paper, and the safety margins available for the operation of the concerned NPP (Nuclear Power Plant) have been quantified

Design of a Schlieren system for low enthalpy hypersonic flow visualization in GHIBLI facility and development of image processing and quantitative analysis codes with preliminary application to sonic free jet

GHIBLI is a 2 MW arc-jet hypersonic facility located at CIRA premises in Capua (Italy), designed for testing candidate TPS materials for re-entry vehicles. Measured data during cold flow tests, i.e. with arc-heater off, showed the achievement of hypersonic conditions at the nozzle exit. Hence, a Schlieren system has been designed to investigate qualitatively and quantitatively such a low enthalpy flow-jet. The apparatus is a classical Toepler’s double lens one. CFD analyses of the free jet were performed to determine the density gradients. On the basis of these results, the limit of sensitivity of the system was determined and the components of the apparatus were dimensioned. A COBLED extended white light source, along with slits made of high reflecting material were experimented. Schlieren images, projected on opaque screen are acquired by a CMOS monochromatic sensor. An image processing code was developed in MATLAB to obtain contrast and clearness enhancement. Quantitative analysis was approached by developing a density-contrast relation, based on schlieren phase-shift effects, modeled under the wave theory of light, and the CMOS tension-charge characteristics. For this purpose, a code named Density from Contrast was developed in MATLAB to measure the luminous intensity of each pixel of captured images and thus compute the density field

Nodalization effects on RELAP5 results related to MTR research reactor transient scenarios

The present work deals with the anal y sis of RELAP5 results obtained from the evaluation study of the total loss of flow transient with the deficiency of the heat removal system in a research reactor using two different nodalizations. It focuses on the effect of nodalization on the thermal-hydraulic evaluation of the re search reactor. The analysis of RELAP5 results has shown that nodalization has a big effect on the predicted scenario of the postulated transient. There fore, great care should be taken during the nodalization of the reactor, especially when the avail able experimental or measured data are insufficient for making a complete qualification of the nodalization. Our analysis also shows that the research reactor pool simulation has a great effect on the evaluation of natural circulation flow and on other thermal-hydraulic parameters during the loss of flow transient. For example, the on set time of core boiling changes from less than 2000 s to 15000 s, starting from the beginning of the transient. This occurs if the pool is simulated by two vertical volumes in stead of one vertical volume

NURETH-15, Pisa (Italy), May 12-17, 2013 – Summary

Summary information about NURETH-15 Conference is provided in the present document. This intends to cover the conduct of the Conference and to provide highlights about the planning. Insights from the technical-scientific content of the papers and financial report shall be the subject of forthcoming documents. More details can be found on the website www.NURETH15.org. 1. Conference Conduct The NURETH-15 Conference has been held in Pisa May 12-17, 2013. F. D’Auria (General Chair) opened the Conference. The Welcome Address Session was chaired by M. Cumo (Honorary Chair together with N. Todreas). Key design features of the Conference are listed below: The Conference venue included three cities in Tuscany, Florence (‘Corsini Palace’), Lucca (‘Regio Collegio’) and Pisa with four locations (Verdi Theater, Congress Palace, Former Railway Station ‘Leopolda’ and Church ‘Del Carmine’). This needed special logistic arrangements including close attention at each cross-point. Specific Conference plans included: - Student Sessions. - Selection of NURETH Fellows (among around 6000 authors of all NURETH Conferences, since NURETH-1 in 1980, see below). - Poster Session. - Invited Speakers as distinguished Lecturers and Chairmen introducing the key Plenary Sessions. - Pre-Conference and Post-Conference Workshops. Committees were formed previous to Conference dates and worked during the Conference: - To select the Best Poster papers, Chair Y. Hassan (three awards given, see below). - To select the Best Oral papers, Chair N. Aksan (three awards given, see below). - To select the Best Student papers, Chair N. Cavlina (three awards given, see below). - To plan Technical Journal publication of selected papers, Chair N. Aksan (also expected to work after the Conference). A NURETH poster was created including the names of all NURETH authors (from NURETH-1 to NURETH-15, around 6000 scientists as already mentioned) and pictures from each NURETH Conference. About five hundred scientists, 432 registered participants with payment, NURETH Fellows (the list of those attending is provided below), selected Invited Speakers and undergraduate students from University of Pisa (who also served as stewards in each of the Conference sessions and places), attended the NURETH-15 Conference

Nuclear Energy for Sustainable Economic Development

The discovery and the application of nuclear energy constitute the most important technological achievement of the past century. However, the development and the exploitation of this technology have been remarkably smaller than foreseeable. An overview of the significant features of the nuclear technology including the comparison with competitive energy sources is made. The “embedded” safety engineering and the pollution are discussed and the main features are mentioned. Indeed, nuclear technology can be applied for the sustainable society development by producing substantial amount of clean water from the ocean. The idea is to build up nuclear power plant sites that produce desalinated water and pump it several tens of kilometers away to form a lake into a desert region. This could help to establish the conditions for an agriculture-based civilization

Validation of CATHARE TH-SYS Code Against Experimental Reflood Tests

This paper presents results of a code validation activity that has been carried out at the University of Pisa within the EC-funded NURESAFE project, aimed to assess CATHARE2 v2.5_3 Mod3.1 code capabilities to simulate scenarios featuring reflood conditions. For such purpose, experimental data available from FEBA and ACHILLES separate-effect test facilities was used. In order to set-up a reference calculation model, rigorous sensitivity studies have been performed for every of the selected experimental test facilities. Quantitative analysis of the results has been carried out for all of the considered tests, using the Fast Fourier Transform Based Method (FFTBM) for accuracy quantification of code predictions. The calculations of experimental tests of ACHILLES facility have been performed with CATHARE2 v2.5_3 mod 3.1 using both 1-D and 3-D models. The no-regression of the results predicted by such code was successfully checked through qualitative and quantitative comparison with results obtained by the one of previous code versions: CATHARE2 v2.5_2 mod 7.1. An assessment of the capabilities of the new CATHARE3 v1.3.13 code to simulate reflood phenomena using both two- and three-field 1-D models has then been carried out, based on the same ACHILLES tests. Simulations by CATHARE3 (three-field) exhibit faster quenching than CATHARE2, mainly due to the presence of the droplet field enhancing the heat exchange from the fuel rod simulators. The performed qualitative analysis has shown the ability of CATHARE2 code to capture the main features of the reflood phenomena using appropriate modeling. Nonetheless, the quantitative analysis shows a systematic underprediction of the PCT and faster quenching in the majority of tests

Assessment of NEPTUNE_CFD Code Capabilities to Simulate Two-Phase Flow in the OECD/NRC PSBT Subchannel Experiments

This paper deals with the validation of the multifield computational fluid dynamics code NEPTUNE_CFD v2.0.1 against experimental data available from the OECD/NRC NUPEC PWR subchannel and bundle tests (PSBT) international benchmark. The present work is performed in the framework of the NURESAFE European collaborative project and focuses on the steady-state single subchannel void fraction tests. From overall 126 PSBT experiments covering wide range of test conditions and 4 different geometrical configurations of PWR subchannel, 42 tests have been selected and simulated using NEPTUNE_CFD. Following the NEA/CSNI (Nuclear Energy Agency / Committee on the Safety of Nuclear Installations) best practice guidelines about computational grid design and grid quality, mesh sensitivity analysis has been performed using axial and radial grid refinement. Both axial and radial mesh sensitivity studies do not exhibit any significant change in the predicted results, which thus result to be grid-converged. Besides, a series of sensitivity calculations have been performed in order to investigate the influence of uncertainties of the experimental boundary conditions on the code predictions. The influence of code physical and closure models on the void fraction prediction has been studied and discussed in detail. Generally, the calculated cross-sectional averaged void fraction at the measurement plane differs from the measured one by maximum of +/- 8%. This discrepancy is comparable to the 2σ experimental uncertainty range on void fraction measurement. The performed investigations have shown the ability of NEPTUNE_CFD to predict reasonably the void fraction in PSBT subchannel using appropriate modelling