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

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

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

Post-BEMUSE Reflood Model input uncertainty methods (PREMIUM) Benchmark Phase II: identification of influential Parameters

The objective of the Post-BEMUSE Reflood Model Input Uncertainty Methods (PREMIUM) benchmark is to progress on the issue of the quantification of the uncertainty of the physical models in system thermal-hydraulic codes by considering a concrete case: the physical models involved in the prediction of core reflooding. The PREMIUM benchmark consists of five phases: - Phase I: mainly, definition of the different uncertainty methods; - Phase II: determination of the physical models influential in reflooding using the test 216 of the FEBA programme; - Phase III: quantification of the uncertainties of the parameters associated with the physical models identified as influential within Phase II, using FEBA/SEFLEX experimental results; - Phase IV: validation of the found uncertainties within Phase III by propagating them in the 2-D PERICLES reflood experiment; this phase will be performed blindly except for the coordinators; - Phase V: synthesis report. This report presents the results of Phase II. Phase II is dedicated to the identification of the uncertain code parameters associated with physical models used in the simulation of reflooding conditions. This identification is made on the basis of the Test 216 of the FEBA/SEFLEX programme according to the following steps: - identification of influential phenomena; - identification of the associated physical models and parameters, depending on the used code; - quantification of the variation range of identified input parameters through a series of sensitivity calculations. A procedure for the identification of potentially influential code input parameters has been set up in the Specifications of Phase II of PREMIUM benchmark. A set of quantitative criteria has been as well proposed for the identification of influential IP and their respective variation range. Thirteen participating organisations, using 8 different codes (7 system thermal-hydraulic codes and 1 sub-channel module of a system thermal-hydraulic code) submitted Phase II results. The base case calculations show spread in predicted cladding temperatures and quench front propagation that has been characterized. All the participants, except one, predict a too fast quench front progression. Besides, the cladding temperature time trends obtained by almost all the participants show oscillatory behaviour which may have numeric origins. Adopted criteria for identification of influential input parameters differ between the participants: some organisations used the set of criteria proposed in Specifications “as is”, some modified the quantitative thresholds proposed in Specifications, and others used their own methodologies. This fact was a partial reason for the different ranges of input parameter variation identified by participants, in addition to differences of the physical models adopted by the different codes. Therefore, such different variation ranges of IP and, correspondingly, such different variation ranges of cladding temperature and time of rewet, make rather difficult the task of meaningful and easy-comprehendible comparison of Phase II results. Out of a total of 72 input parameters, initially considered by all participants, only 6 were identified as influential by more than 4 participants that are: - bundle power; - wall heat transfer coefficient; - interphase friction coefficient; - interphase heat transfer coefficient; - heat transfer (enhancement) at the quench front; - droplet diameter. It should be noted that actual parameters considered in parameter “Heat transfer (enhancement) at the quench front” are code-specific and may have different influence on calculation results. Several participants discarded some identified influential parameters (e.g., droplet diameter) due to existing relation between this kind of parameters so-called “Input Coefficient Parameters” and more global parameters (e.g. interfacial friction coefficient and interphase heat transfer coefficient which use the droplet diameter) so-called “Input Global Parameters”. Some participants also discarded identified influential so-called “Input Basic Coefficients” (e.g. bundle power) since their uncertainty has not to be determined in the Phase III but will be provided by the coordinator from experimental data. The behaviour of the variation of the responses at the extremes of IP range of variation greatly depends on the type of input parameter and on the code used. Mainly, the following two different behaviours can be characterized: - For some parameters, like power, wall heat transfer and interphase heat transfer coefficients, a qualitative (but not quantitative) agreement between different codes is observed. - For other parameters, like interphase friction coefficient and droplet diameter, a contrary behaviour (i.e. in correspondence of one of the extreme of the IP range, the direction of change of the responses is different) between different codes and even between different selected models within the same code can be observed. This suggests that the effect of such parameters on the cladding temperatures is quite complex, probably because it involves a lot of physical models (e.g., via interphase friction and interphase heat transfer coefficients for the droplet diameter). It shall be noted that the analysis of differences between the reflood models of different codes is out of scope of the PREMIUM benchmark. Nevertheless, it is recommended to take into account the physical models/ input parameters found as influential by the other participants in order to select the influential input parameters for which uncertainties are to be quantified within the Phase III of PREMIUM. In particular, input parameters identified as influential by other participants using the same code should be considered

Instrumenting Full scale Boron Injection Test Facility to support Atucha-2 NPP licensing

The Atucha-2 Pressurized Heavy Water Reactor is equipped with a back-up shutdown system based on the fast injection of boron into the moderator tank. Such system had initially been designed to cope with a 10%-area (0.1A) break Loss Of Coolant Accident (LOCA) scenario, but based on upgraded licensing requirements the design had to be revised and possibly improved against a double-ended guillotine (2A) break LOCA. In particular, the boron injection had to be proven fast enough to allow a timely shutdown of the reactor, even in the case of a failure of the primary shutdown system (control rods). A full-scale test facility was built for such “design validation” purpose, in the framework of a cooperation program between the University of Pisa – San Piero a Grado Nuclear Research Group (GRNSPG) and the utility Nucleoeléctrica Argentina S.A. (NA-SA). A special instrumentation system, based on conductivity probes designed on purpose by the Helmholtz Zentrum Dresden-Rossendorf (HZDR), was adopted for the measurement of the injection delay, as well as for the monitoring of pressure at several key locations. Care was taken to reproduce the relevant NPP conditions as closely as possible to those expected on the basis of extensive safety analyses performed adopting a Best Estimate Plus Uncertainty (BEPU) approach. In this respect, not only the test facility is full-scale, but also the key components (such as the fast opening air valves, the boric acid tanks, the rupture device, the injection lance) were directly borrowed from the Atucha-2 NPP. The experimental campaign carried out by NA-SA on such test facility allowed to improve the design of the boron injection system (especially as to some fluid-structure interaction issues) and finally to achieve the main goal, i.e. the demonstration that the system’s performance is fast enough to assure a timely and safe shutdown of the reactor. This was a key contribution to the successful completion of the NPP licensing process

Integrated Nuclear Knowledge Management System – NUTEMA

Knowledge Management has become one of the most important issues for the nuclear industry. On the one hand, the amount of codifiable knowledge in the nuclear technology area saw a steep increase in the last years; on the other hand, due to the “generation gap” in the nuclear industry it became very challenging to organize the seamless transfer of the noncodifiable knowledge from one generation of engineers to the other. Computer aided systems so far where aiming at to preserve the codifiable knowledge. The present paper introduces a system that should support knowledge management not only dealing with the codifiable part of it, but also address broader aspects (which includes the management of non-codifiable knowledge).- NUTEMA. The integrated nuclear knowledge management system NUTEMA should provide interactive combination of information and methods, but also identifying competences which more adequately fit to a given task, keeping track of keeping skills of the experts within an organization. Application is foreseen in nuclear engineering fields as system design, operation and maintenance plant and process modifications, standardization, certification and even for licensing-related tasks. The system combines an extensively diverse and modular database with computer based simulations including a scientific software platform. NUTEMA is conceived to operate in different modes, for example collecting and retrieving database knowledge, training applications, NPP operations support, computer code applications, and as plant analyzer. This paper will present two examples; one acting as at a supporting tool for typical NPP plant modification: In a second case, application on review and optimization of operational process is described. Despite the provided examples deal with different objectives and methods associated with different stages of an NPP lifetime, (design and operation) both are supported by the integrated nuclear knowledge management system

Proposal of a BEPU-FSAR

The accident analysis performance consists of a fundamental part of the licensing of the Nuclear Power Plants (NPP). There are conservative and best estimated methods to perform this analysis. Although Best Estimated Plus Uncertainty (BEPU) is used for qualified computational tools and methods of the accident analysis, it can be used in other parts of the Final Safety Analysis Report (FSAR), which require Analytical Techniques (AT). The need for uncertainty quantification and harmonization of the approaches to use the computer codes is an important issue constituting the background to perform a BEPU-FSAR. The objective of this paper is to present the BEPU-FSAR concept and discuss how-to and why-to perform it

Coupling of Thermal-Hydraulics and I&C for Licensing Analyses

The BEPU (Best Estimate Plus Uncertainty) approach constitutes a valuable and, under some circumstances, an unavoidable tool to demonstrate the safety of NPP (Nuclear Power Plants). Within the licensing process of the Atucha II PHWR (Pressurized Heavy Water Reactor) the BEPU approach has been followed for issuing the Chapter 15 of the FSAR (Final Safety Analysis Report). Namely, the BEPU approach replaced the classical conservative approach. The selection of PIE (Postulated Initiating Events) and, the analysis of each PIE by best estimate models supported by uncertainty evaluation constitute key elements for BEPU. An outline of the BEPU approach is included in the paper, which, otherwise focuses on the simulation needs for Instrumentation and Control (I & C). Sample results from the analysis of PIE are included in the paper. It is demonstrated that the simulation of I&C is necessary to evaluate the safety of the concerned NPP; furthermore, the simulation shall be part of the accident analysis in Chapter 15 of FSA