Pre-test analysis of protected loss of primary pump transients in CIRCE-HERO facility

In the frame of LEADER project (Lead-cooled European Advanced Demonstration Reactor), a new configuration of the steam generator for ALFRED (Advanced Lead Fast Reactor European Demonstrator) was proposed. The new concept is a super-heated steam generator, double wall bayonet tube type with leakage monitoring [1]. In order to support the new steam generator concept, in the framework of Horizon 2020 SESAME project (thermal hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors), the ENEA CIRCE pool facility will be refurbished to host the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section to investigate a bundle of seven full scale bayonet tubes in ALFRED-like thermal hydraulics conditions. The aim of this work is to verify thermofluid dynamic performance of HERO during the transition from nominal to natural circulation condition. The simulations have been performed with RELAP5-3D© by using the validated geometrical model of the previous CIRCE-ICE test section [2], in which the preceding heat exchanger has been replaced by the new bayonet bundle model. Several calculations have been carried out to identify thermal hydraulics performance in different steady state conditions. The previous calculations represent the starting points of transient tests aimed at investigating the operation in natural circulation. The transient tests consist of the protected loss of primary pump, obtained by reducing feed-water mass flow to simulate the activation of DHR (Decay Heat Removal) system, and of the loss of DHR function in hot conditions, where feed-water mass flow rate is absent. According to simulations, in nominal conditions, HERO bayonet bundle offers excellent thermal hydraulic behavior and, moreover, it allows the operation in natural circulation

Post-test simulation of a PLOFA transient test in the CIRCE-HERO facility

CIRCE is a lead–bismuth eutectic alloy (LBE) pool facility aimed to simulate the primary system of a heavy liquid metal (HLM) cooled pool-type fast reactor. The experimental facility was implemented with a new test section, called HERO (Heavy liquid mEtal pRessurized water cOoled tubes), which consists of a steam generator composed of seven double-wall bayonet tubes (DWBT) with an active length of six meters. The experimental campaign aims to investigate HERO behavior, which is representative of the tubes that will compose ALFRED SG. In the framework of the Horizon 2020 SESAME project, a transient test was selected for the realization of a validation benchmark. The test consists of a protected loss of flow accident (PLOFA) simulating the shutdown of primary pumps, the reactor scram and the activation of the DHR system. A RELAP5-3D© nodalization scheme was developed in the pre-test phase at DIAEE of “Sapienza” University of Rome, providing useful information to the experimentalists. The model consisted to a mono-dimensional scheme of the primary flow path and the SG secondary side, and a multi-dimensional component simulating the large LBE pool. The analysis of experimental data, provided by ENEA, has suggested to improve the thermal–hydraulic model with a more detailed nodalization scheme of the secondary loop, looking to reproduce the asymmetries observed on the DWBTs operation. The paper summarizes the post-test activity performed in the frame of the H2020 SESAME project as a contribution of the benchmark activity, highlighting a global agreement between simulations and experiment for all the primary circuit physical quantities monitored. Then, the attention is focused on the secondary system operation, where uncertainties related to the boundary conditions affect the computational results

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

Transient analysis of a locked rotor/shaft seizure accident involving the EU-DEMO WCLL Breeding Blanket primary cooling circuits

The EU-DEMO Water-Cooled Lithium-Lead (WCLL) Breeding Blanket (BB) main subsystems to be cooled are the Breeder Zone (BZ) and the First Wall (FW). Each subsystem will be equipped with an independent Primary Heat Transfer System (PHTS). Within the framework of the EUROfusion Work Package Breeding Blanket research program, several accidents belonging to the category of “Decrease in Coolant System Flow Rate” were studied. The activity was aimed at evaluating the blanket and primary cooling systems thermal-hydraulic performances during such transient conditions. A complete model including the BB and related PHTS circuits has been developed at Sapienza University of Rome. A modified version of RELAP5/Mod3.3 system code has been used to perform the calculations. The simulation results showed that a locked rotor/shaft seizure of a BZ or a FW main coolant pump is the most challenging scenario. BZ and FW system behavior has been analyzed following this initiating event with the goal of the design improvement and to individuate the need for preventive measures. The influence of loss of off-site power on the accident evolution has also been investigated. Moreover, management strategies have been proposed for different reactor components. Calculations demonstrate that the current blanket and PHTS design is appropriate to cope with these kinds of accident scenarios

System thermal-hydraulic modelling of the phénix dissymmetric test benchmark

Phénix is a French pool-type sodium-cooled prototype reactor; before the definitive shutdown, occurred in 2009, a final set of experimental tests are carried out in order to increase the knowledge on the operation and the safety aspect of the pool-type liquid metal-cooled reactors. One of the experiments was the Dissymmetric End-of-Life Test which was selected for the validation benchmark activity in the frame of SESAME project. The computer code validation plays a key role in the safety assessment of the innovative nuclear reactors and the Phénix dissymmetric test provides useful experimental data to verify the computer codes capability in the asymmetric thermal-hydraulic behaviour into a pool-type liquid metal-cooled reactor. This paper shows the comparison of the outcomes obtained with six different System Thermal-Hydraulic (STH) codes: RELAP5-3D©, SPECTRA, ATHLET, SAS4A/SASSYS-1, ASTEC-Na and CATHARE. The nodalization scheme of the reactor was individually achieved by the participants; during the development of the thermal-hydraulic model, the pool nodalization methodology had a special attention in order to investigate the capability of the STH codes to reproduce the dissymmetric effects which occur in each loop and into pools, caused by the azimuthal asymmetry of the boundary conditions. The modelling methodology of the participants is discussed and the main results are compared in this paper to obtain useful guide lines for the future modelling of innovative liquid metal pool-type reactors

Consistent Posttest Calculations for LOCA Scenarios in LOBI Integral Facility

Integral test facilities (ITFs) are one of the main tools for the validation of best estimate thermalhydraulic system codes. The experimental data are also of great value when compared to the experiment-scaled conditions in a full NPP. The LOBI was a single plus a triple-loop (simulated by one loop) test facility electrically heated to simulate a 1300 MWe PWR. The scaling factor was 712 for the core power, volume, and mass flow. Primary and secondary sides contained all main active elements. Tests were performed for the characterization of phenomenologies relevant to large and small break LOCAs and special transients in PWRs. The paper presents the results of three posttest calculations of LOBI experiments. The selected experiments are BL-30, BL-44, and A1-84. They are LOCA scenarios of different break sizes and with different availability of safety injection components. The goal of the analysis is to improve the knowledge of the phenomena occurred in the facility in order to use it in further studies related to qualifying nodalizations of actual plants or to establish accuracy data bases for uncertainty methodologies. An example of procedure of implementing changes in a common nodalization valid for simulating tests occurred in a specific ITF is presented along with its confirmation based on posttests results

Experimental characterization of leak detection systems in HLM pool using LIFUS5/Mod3 facility

In the framework of the European Union MAXSIMA project, the safety of the steam generator (SG) adopted in the primary loop of the Heavy Liquid Metal Fast Reactor has been studied investigating the consequences and damage propagation of a SG tube rupture event and characterizing leak rates from typical cracks. Instrumentation able to promptly detect the presence of a crack in the SG tubes may be used to prevent its further propagation, which would lead to a full rupture of the tube. Application of the leak-before-break concept is relevant for improving the safety of a reactor system and decreasing the probability of a pipe break event. In this framework, a new experimental campaign (Test Series C) has been carried out in the LIFUS5/Mod3 facility, installed at ENEA Centro Ricerche Brasimone, in order to characterize and to correlate the leak rate through typical cracks occurring in the pressurized tubes with signals detected by proper transducers. Test C1.3_60 was executed injecting water at about 20 bars and 200°C into lead-bismuth eutectic alloy. The injection was performed through a laser microholed plate 60 μm in diameter. Analysis of the thermohydraulic data permitted characterization of the leakage through typical cracks that can occur in the pressurized tubes of the SG. Analysis of the data acquired by microphones and accelerometers highlighted that it is possible to correlate the signals to the leakage and the rate of release

RELAP5/SIMMER-III code coupling development for PbLi-water interaction

A major safety issue in the Water-Cooled Lead-Lithium Breeding Blanket (WCLL-BB) system foreseen for fusion reactor is the interaction concerning the primary coolant (water) and the neutron multiplier (PbLi), due to a hypothetical tube rupture in the coolant circuit. This scenario involves an exothermic chemical reaction between PbLi and water with the production of hydrogen, in addition to critical interactions in a complex multiphase system in non-thermal equilibrium. In recent years the PbLi/water reaction was successfully implemented in the SIMMER-III code and validated against data from the LIFUS5/Mod3 experimental campaign. However, due to limitations of SIMMER-III, this work was restricted to the prediction of the phenomena inside the vessel, neglecting the simulation of the injection line. Nevertheless, since the injection line may actually have an important effect on the development of the transient, the simulation of the whole facility would be highly desirable. Indeed, the University of Pisa recently developed a coupling methodology between the SIMMER-III and RELAP5/Mod3.3 codes and applied it to simple single-phase cases. In this paper the complete simulation of the LIFUS5/Mod3 facility is presented, with the injection line modelled through RELAP5. Furthermore, all the complex aspects of the phenomena inside the reaction tank were included: the multiphase system and the interaction between water and PbLi with the chemical reaction and the production of hydrogen were modelled by SIMMER. Preliminary results are presented, showing that the coupling methodology can be effectively employed for the prediction of the chemical and thermal-hydraulic behaviour of complex loop experimental facilities