Experimental and computational analyses in support to the design of a SG mock-up prototype for LFR technology applications

The configuration of the Lead cooled Fast Reactor (LFR) pointed out in the framework of the European Lead cooled SYstem (ELSY) and Lead-cooled European Advanced DEmonstration Reactor (LEADER) EC founded projects deals with the compact pool type reactor in which the steam generators (SG) are located inside the reactor tank. In this scenario, the steam generator design plays an important role since it represents the interface between the nuclear island and the secondary system and one of the most impacting incidents connected to this is the Steam Generator Tube Rupture (SGTR) that may propagate to the near-bough tubes (which are located in the same vessel of the reactor core). In the framework of the LEADER project, an innovative configuration of SG has been proposed: the super-heated steam double wall once through bayonet type with leakage monitoring. This conceptual design was studied since 60’ for Sodium Reactor application. An example of facility that operates with this concept is CIRCE (ENEA Brasimone), nevertheless the application is limited to the heat exchange function. The single tube vertical unit consists of three concentric tubes. Starting from the smallest one (feed-water tube), the water crosses it in down-flow. At the end of this tube, it enters the second concentric tube (annular riser) in up-flow where it starts to boil due to the heat exchange with the liquid lead that flows in counter-current at the tube outer surface. The tube design allows the achievement of super-heated steam. The liquid lead is not in direct contact with the second tube. A third concentric tube, that creates an annulus, separates it from the steam-water sides realizing the so called double wall. The study of this configuration is motivated by safety improvement. In fact, it allows the double physical separation between lead and water sides. Furthermore, by means of gap pressurization (with Helium), a leakage check system should be introduced between the double wall in order to prevent incident scenarios. On the other hands, the monitor-ability of the pressurized gap has to be demonstrated and the thermal efficiency of the unit has to be improved. In particular two main design issues are of great importance from the economic point of view. The first deals with monitor-ability of leakages: since a gas as helium is required and it has low conductivity, the gap between the tube should be minimized and a porous heat transfer enhancer should be introduced to reduce, as much as possible, the SG volume (tube length or tube number). The second deals with the thermal insulation of the feed-water descending tube. In fact, the water enters at its minimum temperature at the top of the SG, it is above the active length and it exchanges heat with the superheated stem that is leaving the SG unit. Therefore, in order to avoid steam condensation at the feed-water tube outer surface this tube should be designed with sandwich wall: steel - thermal insulator – steel. This doctorate is co-financed by ANSALDO and ENEA and is directly connected to LEADER and Accordo Di Programma (ADP) national project. The activity aims to support the design of the double wall bayonet tube bundle SG with leakage monitoring and to investigate its TH performance both analytically and experimentally (into a representative prototype scaled down in power). The activity started in 2011 and is still ongoing. At present time the following objectives are reached and are presented in this thesis: • Assessment of the TH performance of the bayonet tube by means of RELAP-5. • Design, construction, operation and disposal of the TxP (Tubes for Powders) facility devoted to powders conductivity measurement into an annular geometry. Its aim was to qualify the conductivity of powders for their application in the annular gap between the tubes that separate the fluids both for application as heat transfer enhancer (as for the ALFRED SG) and to both to accommodate a give temperature drop (as for the HX of facilities under operation at Brasimone). • Design and commissioning of the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section based on the results achieved during the experimental campaigns in TxP. This test section is actually located in CIRCE and aims to investigate the TH behavior of a bundle of seven tubes that represent, as much as possible, the ALFRED SG tubes (1:1 in length). Eight main sections and four attached appendixes constitute the structure of this work. The first three sections have to be intended as introduction to the activity. Justification of the activity, framework and objective are given in section one. GEN-IV systems and LEADER project are briefly described in section 2. The third section presents the ALFRED SG. It includes also the historical development of double wall SG in the nuclear technology and a brief description of the HXs that have been operated or are under construction at ENEA CR Brasimone that make use of double wall concept. The fourth and fifth sections are theoretical activities conducted in support to the R&D of the steam generator bayonet tube. In particular, section four is aimed to assess the TH performance of a single tube of the ALFRED SG by fixing, in a theoretical way (which is described in section five), some modeling issues still not defined in the ALFRED design (i.e. the powder material to be introduced between the double wall and its modeling). Section five aims to point out empirical models to treat the conductivity of powder media and to individuate candidate materials to be acquired and tested with the intent to design a double wall bayonet tube bundle test section. This last section includes an investigation on insulating materials for their application to the the feed-water tube. Section six constitutes the core of this activity and has to be intended as a first step on R&D in support to the design of the bayonet tube steam generator with particular reference to the heat transfer enhancer porous medium placed between the double wall. The Tubes for Powders (TxP) facility has been designed (by ENEA), constructed (by LIMAINOX), instrumented (by ENEA) and operated to test the conductivity of powders both under un-pressurized air environment and under pressurized helium atmosphere. This process takes more than two years and gives rise to two main experimental campaigns. The first set of tests have been conducted to qualify the HXs of the facilities under operation at the Brasimone Research center while the second campaign aimed to support the design of the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section. The design of the HERO test section is still on-going. At present time the SG bayonet tube unit has been designed, constructed (by CRIOTEC), instrumented and connected to CIRCE (section seven). The secondary loop is under design phase. Conclusions are finally given in section eight

Prediction of Void Fraction in PWR Subchannel by CATHARE2 Code

The current generation of thermal-hydraulic system codes benefits of about sixty years of experiments and forty years of development and are considered mature tools to provide best estimate description of phenomena and detailed reactor system representations. However, there are continuous needs for checking the code capabilities in representing nuclear system, in drawing attention to their weak points, in identifying models which need to be refined for best-estimate calculations. Prediction of void fraction and Departure from Nucleate Boiling (DNB) in system thermal-hydraulics is currently based on empirical approaches. The database carried out by Nuclear Power Engineering Corporation (NUPEC), Japan addresses these issues. It is suitable for supporting the development of new computational tools based on more mechanistic approaches (i.e. 3 field codes, 2 phase CFD, etc.) as well as for validating current generation of thermal-hydraulic system codes. Selected experiments belonging to this database are also used for the OECD/NRC PSBT benchmark. The paper presents the validation activity performed by CATHARE2 v2.5_1 (six equation, two field) code on the basis of the sub-channel experiments available in the database and performed in different test sections. Four sub-channel test sections are addressed in different thermal-hydraulic conditions (i.e. pressure, coolant temperature, mass flow and power). Sensitivity analyses are carried out addressing nodalization effect and the influence of the initial and boundary conditions of the tests

Capabilities of TRANSURANUS Code in Simulating BWR Super-Ramp Project

After one-two years of normal operation in a LWR, the fuel–cladding gap may close, as a result of as a result of several phenomena and processes, including the different thermal expansion and swelling of both the fuel and the cladding (Pellet Cladding Interaction). In this equilibrium state, a significant increase of local power (like a transient power ramp, i.e. power increase in the order of 100kW/m-h), induces circumferential stresses in the cladding. In presence of corrosive fission products (i.e. iodine) and beyond specific stress threshold, material dependent, cracks typical of stress corrosion may appear and grow-up: this phenomenon is called stress corrosion cracking (SCC). The cracks of the cladding may spread out from the internal surface, causing the fuel failure. The objective of the activity (performed in the framework of the IAEA CRP FUMEX III), is to validate the TRANSURANUS models relevant in predicting the fuel failures due to PCI/SCC during power ramps. Focus is given on the main phenomena, which are involved or may influence the cladding failure behavior. The database selected is the Studsvik BWR Super-Ramp Project, which belongs to the “public domain database on nuclear fuel performance experiments for the purpose of code development and validation – International Fuel Performance Experiments (IFPE) database” by OECD/NEA. It comprises the data of sixteen BWR fuel rods, that have been modeled and simulated with suitable input decks. The burn-up values range between 28 and 37 MWd/kgU. Eight rods, of KWU standard type, are subjected to fast ramps, the remaining rods experience slow ramps and are of standard GE type

OECD/NRC PSBT Benchmark: Investigating the CATHARE2 Capability to Predict Void Fraction in PWR Fuel Bundle

Accurate prediction of steam volume fraction and of the boiling crisis (either DNB or dryout) occurrence is a key safety-relevant issue. Decades of experience have been built up both in experimental investigation and code development and qualification; however, there is still a large margin to improve and refine the modelling approaches. The qualification of the traditional methods (system codes) can be further enhanced by validation against high-quality experimental data (e.g., including measurement of local parameters). One of these databases, related to the void fraction measurements, is the pressurized water reactor subchannel and bundle tests (PSBT) conducted by the Nuclear Power Engineering Corporation (NUPEC) in Japan. Selected experiments belonging to this database are used for the OECD/NRC PSBT benchmark. The activity presented in the paper is connected with the improvement of current approaches by comparing system code predictions with measured data on void production in PWR-type fuel bundles. It is aimed at contributing to the validation of the numerical models of CATHARE 2 code, particularly for the prediction of void fraction distribution both at subchannel and bundle scale, for different test bundle configurations and thermal-hydraulic conditions, both in steady-state and transient conditions

OECD/NRC PSBT Benchmark: Investigating the CATHARE2 Capability to Predict Void Fraction in PWR Fuel Bundle

Accurate prediction of steam volume fraction and of the boiling crisis (either DNB or dryout) occurrence is a key safety-relevant issue. Decades of experience have been built up both in experimental investigation and code development and qualification; however, there is still a large margin to improve and refine the modelling approaches. The qualification of the traditional methods (system codes) can be further enhanced by validation against high-quality experimental data (e.g., including measurement of local parameters). One of these databases, related to the void fraction measurements, is the pressurized water reactor subchannel and bundle tests (PSBT) conducted by the Nuclear Power Engineering Corporation (NUPEC) in Japan. Selected experiments belonging to this database are used for the OECD/NRC PSBT benchmark. The activity presented in the paper is connected with the improvement of current approaches by comparing system code predictions with measured data on void production in PWRtype fuel bundles. It is aimed at contributing to the validation of the numerical models of CATHARE 2 code, particularly for the prediction of void fraction distribution both at subchannel and bundle scale, for different test bundle configurations and thermalhydraulic conditions, both in steady-state and transient conditions

Modelling of pellet-clad interaction during power ramp in water reactor fuel

The comprehensive understanding of fuel rod behavior and accurate prediction of the lifetime in normal operation and in accident conditions are part of the defense in depth concept. The research in the field involves the availability of experimental data and the development of advanced computational tools suitable for a reliable, best estimate simulation of the fuel behavior. In this connection, OECD NEA sets up the “public domain database on nuclear fuel performance experiments for the purpose of code development and validation – International Fuel Performance Experiments (IFPE) database”, with the aim of providing a comprehensive and well-qualified database on UO2 fuel with Zr cladding for models development and codes validation. This database includes the data sets of carried out at Studsvik in the framework of the Inter-Ramp BWR and Super-Ramp PWR Projects. The phenomenon addressed in this activity is the pellet cladding interaction (PCI). In particular, special attention is given to the cladding failure during power ramps (stress corrosion cracking), caused by the combined effects of the clad stress occurring in the region of the pellets ends and the presence of aggressive fission products (e.g. iodine). The first part of this work is focused in the descriptions of the phenomena occurring in fuel and cladding during normal irradiation and during power ramps in light water reactors. In the second part the code TRANSURANUS version “v1m1j08” is assessed against the databases Inter-Ramp and Super-Ramp in order to verify the capability of the code in predicting the failures due to stress corrosion cracking and the associated phenomena prior and after ramps. The results presented include the complete set of simulations of all rods irradiated in the Studvisk R2 reactor and the corresponding comparisons with the experimental data. Sensitivity calculations are also performed to address the influence of the boundary conditions implemented and the choice of the different code options on results. Finally conclusions are presented with the aim to compare the results obtained from the two simulations

Modeling the Heat Transfer of Helical Coil Tubes Steam Generator in SMR by RELAP5 Code and Validation

OSU MASLWR integral test facility, owned and operated at Oregon State University, represents a SMR cooled and moderated by light water. The main features of the experimental facility are based on the preliminary design of the Multi-Application Small Light Water Reactor carried out at Idaho National Labs. This reactor relies on natural circulation and on passive and inherently safe engineering safety features. It is featured by a helical-coil steam generator. The primary system and the containment are inherently coupled, in some anticipated operation occurrences and postulated accident conditions. Reliable simulations of such system requires the demonstration that a thermal-hydraulic system code, such as RELAP5 in this paper, can predict or, at least, bound some phenomena, which are outside its standard area of application. Challenging phenomena for the code simulation are: heat exchange for helical-coil SG and two phase instabilities in parallel tubes (SG secondary side); mixing and thermal stratification in a pool system; the condensation on the free surface in presence of noncodensable gas; chocked flow, the coupling primary system and containment. This paper presents also the performance of RELAP5 code in simulating the heat transfer in tube bundle with crossflow and the validation performed using two experiments performed in OSU MASLWR integral test facility. They are: 1) a natural circulation experiment aimed at characterizing the system performances at different power levels, and 2) a total loss of feedwater flow postulated accident scenario. The activity has been performed in the framework of an International Collaborative Standard Problem under the auspices of the IAEA

Predictability of CNEA PHWR MOX Experiments by mean of TRANSURANUS Code, from the IFPE Database

Investigations of fuel behavior are carried out in close connection with experimental research, operation feedback and computational analyses. The comprehensive understanding of fuel rod behavior and accurate prediction of the lifetime in normal operation and in accident conditions are part of the defense in depth concept. In this connection, OECD NEA sets up the public domain database on nuclear fuel performance experiments – International Fuel Performance Experiments (IFPE) database, with the aim of providing a comprehensive and well-qualified database on UO2 fuel with Zr cladding for models development and codes validation. The CNEA’s PHWR MOX Experiment belong to this database. This experiment was carried out in the High Flux Reactor (HFR) of Petten, Holland. It involves six MOX rods prepared and controlled in the CNEA’s “Alpha” Facility (Argentina). The objective of the experiment was to verify the fabrication processes and the study of the fuel behavior with respect to cladding failure due to Stress Corrosion Cracking (SCC) under Pellet Cladding Interaction (PCI) conditions. These rods were irradiated from 0 until 15 MWd/kgU. The code TRANSURANUS version “v1m1j09” is assessed against the database CNEA PHWR MOX in order to verify the capability of the code in predicting the cladding failures due to SCC. Comparisons with the experimental data and the results obtained with BACO code developed by “CNEA” group of Bariloche are presented. Sensitivity calculations are also performed for supporting the analyses of the results, improving the level of understanding of the code capabilities. The main conclusion is that the clad failure propensity of the rods belonging to the PHWR CNEA MOX experiment is conservatively assessed

Investigation of PCI phenomenon in PWR fuel with Transuranus code

The fuel matrix and the cladding constitute the first barrier against radioactive fission product release. Therefore a defense in depth concept requires also the comprehensive understanding of fuel rod behavior and accurate prediction of the lifetime in normal operation and in accident condition as well. Investigations of fuel behavior are carried out in close connection with experimental research operation feedback and computational analyses. In this connection, OECD NEA sets up the “public domain database on nuclear fuel performance experiments for the purpose of code development and validation – International Fuel Performance Experiments (IFPE) database”, with the aim of providing a comprehensive and well-qualified database on UO2 fuel with Zircaloy cladding for model development and code validation. This paper describes the application of the TRANSURANUS code against power ramp tests executed in the research reactor R2 at Studsvik, Sweden. The datasets are part of the International Fuel Performance Experiments (IFPE) database. The experiments address the behavior of PWR fuel rods, including preceding base irradiation, during the over-power ramping. The burnup values range between 28 and 45 MWd/kgU. Pre-, during-, and post- irradiation, non destructive and destructive examinations were executed, in order to determine and understand the behavior of the fuel rods, but also to provide suitable data, useful for code validation. The experimental data were used for assessing the TRANSURANUS capabilities in predicting the phenomenon of the pellet clad interaction (PCI). The objective of the activity has been fulfilled developing input decks suitable for the assessment of TRANSURANUS code version “v1m1j08”. The current paper reports the main outcome of the assessment of the calculations