The CAREM project involves technological and engineering solutions, as well as several innovative design features that must be properly demonstrated during the design phase. Also specific codes used for modeling systems related with safety issues to obtain design parameters (primary cooling system, reactor core, fuel design, etc.) must be verified and validated against worldwide benchmarks and/or experimental data to build confidence in their results (Mazzi et al., 2012).

Within the CAREM project, the effort has been focused mainly on the nuclear island (inside containment and safety systems) where several innovative design solutions require developments of the first stage (to assure they comply with functional requirements). This comprises mainly: the reactor core cooling system (RCCS), the reactor core and fuel assembly, and the FSS. An extensive experimental plan has been prepared, including the design and construction of several experimental facilities to fulfill the project’s requirements.

A high-pressure natural circulation rig, CAPCN, was constructed to perform dynamic tests of RCCS. Its purpose was mainly to study the thermal-hydraulic dynamic response of the CAREM primary loop, including all the coupled phenomena that may be described by one-dimensional models. This includes the assessment of the calculation codes on models of the rig, and the extension of validated models to the analysis of the CAREM reactor.

The CAPCN rig resembles CAREM in the primary loop (self-pressurized natural circulation) and the SG (helical once-through), while the secondary loop is designed only to produce adequate boundary conditions. Operational parameters are reproduced for intensive quantities (pressure, temperature, void fraction, heat flux, etc.) and scaled for extensive quantities (flow, heating power, cross-sections, etc.). Height was kept approximately on a 1:1 scale. The heating power may be regulated up to 300 kW. The secondary loop pressure and cold leg temperatures are controlled through valves. The pump regulates the flow. The condenser is of air-cooled type with airflow control. The control of the actuators (heaters, valves, pumps, etc.), data acquisition and operating follow up were carried out from a control room, through a PC-based, multi-node software (flexible enough to define any feedback loop).

Many experiments were performed in order to investigate the thermal-hydraulic response of the system in conditions similar to CAREM operational states. The influence of different parameters like vapor dome volume, hydraulic resistance and dome nitrogen pressure was studied. Perturbations in the thermal power, heat removal and pressure relief were applied. The dynamic responses at low pressure and temperature, and with control feedback loops, were also studied. It was observed that around the operating point self-pressurized natural circulation was very stable, even with important deviations on the relevant parameters. A representative group of transients were selected, in order to check computer models.

The thermal-hydraulic design of CAREM reactor core was carried out using a 3-D two fluid model code. In order to take into account the strong coupling of the thermal-hydraulic and neutronic of the core, this code was linked with neutronic codes. This coupled model allows a 3-D evaluation of power and thermal-hydraulic parameters at any stage of the burnup cycle.

The mass flow rate in the core of the CAREM reactor is rather low compared with typical light-water reactors and therefore correlations or experimental data available should be assessed in the range of interest. In order to perform this assessment experiments were conducted at the thermal-hydraulic laboratories of the Institute of Physics and Power Engineering (IPPE, Obninsk, Russian Federation).

The main goal of the experimental program was to generate a substantial database to assess the prediction methodology for critical heat flux (CHF) applicable to the CAREM core, covering a wide range of thermal-hydraulic parameters around the point of normal operation. Most tests were performed using a low-pressure Freon rig, and results were extrapolated to water conditions through scaling models. Finally a reduced set of tests were performed in water at high pressure and temperature, to validate the method for scaling. Different test sections were assembled to simulate different regions in the fuel element as well as radial uniform and non-uniform power generations. A bundle with 35% of the full length was tested to obtain CHF data under average sub-cooled conditions. More than 250 experimental points under different conditions were obtained in the Freon loop and more than 25 point in the water loop.

The fuel assemblies and absorbing clusters were or will be subject to a series of tests, including standard mechanical evaluations, and hydraulic tests. The latter comprise:

• tests in a low-pressure rig evaluating pressure losses (performed), flow-induced vibrations

and general assembling behavior;

• endurance tests in a high-pressure loop points to wear-out and fretting issues.

A fuel rod irradiation test to be performed in the Halden boiling water reactor is under preparation. This test has the purpose of characterizing the most relevant performance aspects of the fuel, such as temperature behavior, dimensional stability and fission gas release.

Neutronic modeling validations were made against VVER reactor geometry using experimental data from a ZR-6 Research Reactor, Hungary, and a series of benchmark data for typical PWR reactors. The RA-8 critical facility has been designed and constructed as an experimental facility to measure neutronic parameters of CAREM. Further experimental data were obtained from RA-8.

One of the most innovative systems behind the CAREM concept is the hydraulic (in-vessel) control rod drive (HCRD) mechanism. The design embraces mechanical and thermal-hydraulic innovative solutions, so a complete experimental program is necessary to achieve the high reliability performance jointly with low maintenance. This development plan includes the construction of several experimental facilities.

Preliminary tests were first performed to prove the feasibility of the theoretical approach, to have an idea of some of the most sensitive controlling parameters and to determine spot points to be focused on during design. Tests were undertaken on a rough device with promising experimental results, and good agreement with first modeling data was obtained.

First prototype tests helped to determine preliminary operating parameters on a full-scale mechanism as a first approach towards detail engineering. These parameters include range of flow, ways to produce hydraulic pulses, etc. Manufacturing hints that simplified and reduce costs of the first design were also found. Tests were carried out in a specially built rig and as part of this experimental development it was decided to separate the regulating and fast-drop requirements in different devices.

Test on a low-pressure loop were carried out with the CRD at atmospheric pressure, and with feed-water temperature regulation up to low sub-cooling. The feed-water pipeline simulated alternative configurations of the piping layout with a second injection line (dummy) to test possible interference of pulses. The ad-hoc test loop (CEM) was designed to allow automatic control of flow, pressure and temperature, and its instrumentation produces information of operating parameters including pulse shape and timing. The tests included the characterization of the mechanism and the driving water circuit at different operating conditions, and the study of abnormal situations as increase in drag forces, pump failure, loss of control on water flow or temperature, saturated water injection, suspended particle influence, and pressure ‘noise’ in feeding line. The tests, carried out at turbulent regime, which are the closest conditions to operation obtained in this loop, showed good reliability and repetitiveness as well as sensitivity margins for the relevant variables within control capabilities of a standard system.

Finally, a high-pressure loop (CAPEM) was constructed in order to reach the actual operating conditions (P = 12.25 MPa, T ^ 326 °C). The main objectives are to verify the behavior of the mechanisms, to tune up the final controlling parameter values and to perform endurance tests. After this stage, the system under abnormal conditions, such as the behavior during RPV depressurization, simulated breakage of feeding pipes, etc., will be tested.

Since the HCRD design adopted has no movable parts outside the RPV, it was necessary to design a special probe to measure the rod position able to withstand primary environmental conditions. The proposed design consists in a coil wired around the HCRD cylinder with an external associated circuit that measure electric reluctance variations induced by the movement of the piston-shaft (made of magnetic steel) inside the cylinder. A cold test performed showed that the system is capable of sensing one-step movement of the regulating CRD, with an acceptable accuracy. In-furnace high-temperature tests were conducted to evaluate the behavior of the system against temperature changes similar to those occurring during operational transients.

To fulfill the Argentinian regulation, CAREM 25 has two independent shutdown systems. Two independent reactor protection systems exist to actuate these shutdown systems and the safety systems. The qualification of these reactor protection systems includes the development, construction and testing of prototypes.