Community Research and Development Information Service - CORDIS

Final Report Summary - SCWR-FQT (Supercritical Water Reactor - Fuel Qualification Test)

Executive Summary:
The major challenges for the Supercritical Water Reactor (SCWR) are to develop a viable core design, accurately estimate the heat transfer coefficient and develop materials for the fuel and core structures. Core, reactor and plant design concepts have been worked out in substantial detail, like the European High Performance Light Water Reactor (HPLWR) concept. As the next step, the system research plan prepared by the steering committee of the SCWR system in GIF proposes to test a small-scale fuel assembly under typical prototype conditions in a research reactor.
Design and licensing of the fuel qualification test facility, comprising of a four-rod fuel bundle including the required supercritical water loop and its safety and auxiliary systems, is subject of the Supercritical Water Reactor – Fuel Qualification Test (SCWR-FQT) project. The Research Centre Rez (Centrum Vyzkumu Rez) is offering the necessary infrastructure with their LVR-15 research reactor.
The objectives of the project are to make significant progress towards the design, analysis and licensing of the Fuel Qualification Test facility cooled with supercritical water in the research reactor LVR-15. Test of the fuel assembly is addressed with the following concept. A pressure tube is placed instead of a fuel assembly in the LVR-15 reactor. It contains 4 fuel rods with 8 mm diameter and 9.44 mm pitch, like the HPLWR assembly concept, inside a square assembly box. The heated length is limited to 600 mm to match with the core height of the reactor.
The SCWR-FQT project started on January 1, 2011 and finished on December 31, 2014. It has been structured into 3 interconnected blocks that ran in parallel; the first block containing all the design work and analyses of the FQT facility, the second one dealing with the design of a similar, electrically heated test section that shall serve for pre-qualification of the test section and that shall be designed and built in China and finally, the third one that deals with the choice of a suitable cladding material, including necessary corrosion and mechanical tests.
Final design and results of analyses of the test section including the supercritical water loop shall form the basis of the licensing documents for the Czech regulator. Data from the operation of the electrically heated test section shall serve as pre-qualification operation as well as for validation of the codes used for analyses. Corrosion and mechanical data shall be available for the selected materials and a choice of the cladding material shall be made during the project.

Project Context and Objectives:
The overall objective of the Supercritical Water Reactor – Fuel Qualification Test (SCWR-FQT) project is to design, analyse and license a small scale fuel assembly test facility with supercritical water in the research reactor LVR-15. The test section is a 4-rod fuel assembly located inside an active channel placed in the research reactor. The height of the test section matches the height of the reactor core, i.e. 600 mm. The fuel is uranium dioxide UO2 with 20% enrichment with total fissile power of the fuel assembly at 63.6 kW. A recuperator and cooler are placed inside the active channel above the test section. Top of the active channel is closed by a headpiece with inlet and outlet coolant lines connecting the active channel with the remaining loop. The loop shall be operated at evaporator conditions of the HPLWR, i.e. at temperature near or slightly above the pseudo-critical temperature 384°C, pressure 25 MPa and maximum linear heat rate reaching 39 kW/m.
Material tests of candidate cladding materials chosen among stainless steels (i.e. candidate materials immediately available commercially in nuclear quality) are performed in autoclaves and a choice is made early in the project in order to provide input material data into design and analyses. Additionally, autoclave tests of a fuel rod mock-up are performed once the final design of the fuel assembly is fixed.
Steady-state as well as transient tests are performed with an identical, but electrically heated test section, performed for the purpose of pre-qualification of the test section before actual testing with nuclear fuel. The existing supercritical water facility SWAMUP in China is offered to perform the tests within the parallel Chinese project SCRIPT. The experimental data are also intended to be used for validation of codes used for analyses in WP2 (steady-state analyses) and WP3 (safety analyses).

The project is divided into 4 technical work packages (out of the total of 6 work packages) as follows:
Work package 1 “Design” contains all the design work on the facility. Detailed design of the active channel with the tested fuel assembly including a 3D model, and a scheme of the remaining systems are provided. Design of the individual parts is refined according to analyses performed in the analytical work packages 2 and 3, as described further.
In work package 2 “Analyses of normal operation” flow and heat transfer analyses, structural analyses of the key components of the fuel assembly, active channel and primary system, and neutronic analyses of the reactor core containing the facility. A core configuration including the facility is proposed as realistically as possible (the facility will be built and operated beyond 2016). Transient analyses of start-up and shut-down of the facility are performed as well. Results of the analyses are used to refine the design in work package 1. CFD codes used for analyses are validated with experimental data available from the pre-qualification test in the SWAMUP facility in China.
Work package 3 “Safety analyses” is dedicated to safety analyses, the proposal of the safety systems, definition of the necessary safety instrumentation and the triggers and actions of the safety systems at off-normal conditions. Design basis accidents as well as an accident beyond the design basis and its consequences are studied. Finally, failure of the pressure tube and its consequences on the adjacent fuel assemblies in the reactor are studied. System codes used for analyses are validated with experimental data available from the pre-qualification test in the SWAMUP facility in China.
Work package 4 “pre-qualification” is divided into 2 parts; the first part deals with material tests of stainless steels identified as cladding materials suitable for the fuel qualification test at the given test conditions. Nuclear grade, commercially available materials are only considered. Tests for general corrosion and susceptibility to stress corrosion cracking are performed. A choice is made between the considered materials and the relevant material data are considered in the final design and analyses. Later on, when the final design of the fuel rods is fixed, fuel rod mock-ups are fabricated and tested in autoclave for mechanical stability. The second part of this work package is dedicated to the preparation and execution of the pre-qualification test with an electrically heated test section at conditions of the fuel qualification test during steady-state operation as well as during different transients, such as depressurization. Performance of the test section is evaluated and experimental data are collected and used for validation of codes used in WP2 and 3.

The main scientific and technical objectives of the project are summarized as follows:
• Design the experimental facility for the Fuel Qualification Test in supercritical water; the facility consists of an active channel containing the tested fuel assembly installed in the research reactor, and the safety and auxiliary systems placed outside the reactor. Auxiliary and safety systems shall reside in the experimental building directly adjacent to the reactor hall and the building shall be connected to the existing connecting systems (active ventilation, draining).
• Provide a core configuration and the active channel position inside the research reactor.
• Determine the total power and power profile of the fuel rods in the given position.
• Ensure the coolability of the in-pile test section under normal steady-state operation as well as during start-up and shut-down of the loop by performing necessary CFD analyses.
• Design shielding of the loop for radiation protection of staff during operation of the facility.
• Design the fuel handling system and specify the handling procedure to dismantle the test section for the case of failed experiment.
• Perform burn-up analyses of the reactor core with the active channel of the FQT facility in the defined position.
• Perform structural analyses to ensure integrity of the test section and key components of the active channel and the primary system.
• Perform analyses of heat transfer and flow distribution in the recuperator and cooler located inside the active channel.
• Perform blind predictions of the thermal-hydraulic tests planned for pre-qualification of the test section in the SWAMUP facility in China.
• Perform validation of CFD codes used for thermal-hydraulic analyses using experimental data from pre-qualification of the test section in the SWAMUP facility in China.
• Provide a scheme of the safety systems including safety instrumentation. Define warning levels and actions preventing accidents.
• Perform safety analyses to protect the first barrier during design basis accidents.
• Assess the risk of a pressure tube rupture by predicting the probability of a pressure tube failure and by studying the consequences of pressure tube rupture to the neighbouring structures (fuel assemblies) in the research reactor.
• Perform a probabilistic safety assessment of the facility.
• Perform analyses of pressure tube failure due to a seismic event.
• Perform analyses of a postulated accident beyond the design basis leading to failure of the first safety barrier (fuel cladding). Assess all potential source terms of the accident.
• Modify the APROS code to include the boiling crisis after depressurization.
• Perform validation of system codes used for safety analyses using experimental data from pre-qualification of the test section in the SWAMUP facility in China.
• Collect the available physical and mechanical properties of the selected cladding material candidates in a data sheet for use as input into design and analyses. The candidate materials are selected from commercially available nuclear-grade stainless steels: 316L, 347H and 08Cr18Ni10Ti (equivalent to AISI 321).
• Perform general corrosion tests of the candidate cladding materials in supercritical water at 25 MPa, at temperatures 500 and 550°C and two oxygen levels (low – 150 ppb and high – 2000 ppb).
• Perform mechanical tests for susceptibility to stress corrosion cracking of the candidate cladding materials at conditions specified above.
• Design, manufacture and test fuel rod mock-up in supercritical water autoclave.
• Provide design input and assist with the design and manufacturing of the pre-qualification test section with electrically heated fuel rods for testing in the SWAMUP supercritical water facility in China. Assist the Chinese project partners with the test program of the pre-qualification tests.

Project Results:
MAIN SCIENTIFIC AND TECHNICAL RESULTS
The overall project objectives have been achieved, with only minor deviations from the initial work plan, which have necessitated adjustments in the work distribution between partners or back-up solutions. The scientific and technical achievements and results of the project are summarized in the following sections, each dedicated to one work package.

RESULTS OF WORK PACKAGE 1 “DESIGN”
The major objectives of WP1 “Design” were to design all components of the test facility, comprising the active channel with the fuel assembly, the supercritical water loop (primary circuit), its safety, auxiliary and radiological protection systems. In addition, reactor building modifications and design of the new building housing all main components of the SCWR-FQT loop, except the active channel itself, was worked out. Additionally, the handling procedure of irradiated fuel inserted in the active channel of the SCWR-FQT experiment was developed.
The design of all main components was modeled with the software Autodesk Inventor. Created 3D models were then analyzed in other WPs. Moreover within WP1, the fuel, materials of all components, as well as limits and conditions of the SCWR-FQT loop for suppliers and for other WPs were specified.

Task 1.1 Main loop design
The designed experimental loop for fuel qualification in supercritical water consists of a closed pressurized water circuit. This pressurized primary circuit operating at pressure 25 MPa contains the main circulation pump which forces the circulation of the coolant (supercritical water) through the test section placed inside the active channel which is intended to be inserted into the existing research pool type reactor LVR-15.
The primary circuit ensures that the required physical conditions are achieved around the fuel rods located in the active channel. The design temperature around the fuel rods of the SCWR-FQT loop is 385°C – 450°C.

During start-up, and before the reactor LVR – 15 is started, the primary circuit is gradually heated up to ~300°C by the u-tube cooler, located in the upper part of the active channel, which transfers heat from the secondary circuit generated by electrical heater EO-1 to the primary circuit.
The required pressure in the primary circuit is controlled by pressurizers, to compensate thermal expansion of the coolant when heated, and by supply pumps. The pressure in the circuit is kept constant through control of volume changes induced by changes of temperature provided by pressurizers KO-1 and by supply pumps. Depressurization of the primary circuit is enabled either by on-line discharging through the sampling system or by fast opening of valve, which allows for the coolant to flow to the depressurization tank BN.
Over pressure protection is ensured by a relief valve with opening pressure 32 MPa. Discharged coolant flows to the depressurization tank BN where generated steam condensates. Non-condensable gasses are directed from the depressurization tank to the special venting system.
The mechanical parts of the SCWR-FQT loop are assembled in 2 blocks with dimensions 3x3.5x2.5 m. Inlet and outlet lines of the primary system are running inside a shielded duct through the reactor building to the primary block situated in the experimental building adjacent to the reactor hall.

The main component of the SCWR-FQT loop is the active channel containing the test section, and a recuperator and cooler, which is placed into the research reactor. The active channel is manufactured from stainless steel 08Ch18N10T, for which the temperature limit, if used for pressure boundary tubes, is around 450°C. The active channel contains four fuel rods with a total heating power of ~63.6 kW. The fuel rods will contain pellets of UO2 (enrichment of 19.7% U235) on a total height of 600 mm inserted in a stainless steel cladding tube with dimensions 8x0.5 mm. The cladding is filled with helium (required pressure approx. 7 MPa) and seal-welded on both ends. A spiral wire of 1.44 mm diameter is wrapped along the rod with 200 mm pitch and welded to both end caps. Above the fuel pellets a gas plenum with 40 mm height is situated to maintain space for fission gases generated inside the fuel rod during irradiation. In the gas plenum a spring is placed to compensate the temperature induced dilatation of the fuel which is expected to be greater than that of the cladding tube. Thus the plenum height will decrease as the spring will be gradually compressed by the expanding fuel.
Above the fuel assembly a recuperator is placed to achieve hot channel conditions as they are expected to occur in the evaporator of the HPLWR . The recuperator consists of 30 3x0.2 mm tubes closed by the tube sheets on each end and held by 18 spacers, which are distributed equally along the total length of 3.7 m of the recuperator. The segments are mutually turned by 120°. Thus, a spiral flow is created to enhance heat transfer and thermal efficiency.

The internal flow is preventing creep conditions of the pressure tube, which would occur at around 450°C. The total mass flow rate at inlet of the active channel is guided down along the pressure tube and cools down the pressure tube, in which γ-heat is generated. After turning around at the bottom of the pressure tube, water flows upwards by passing the internal structures to the top of the active channel. At the top, the flow turns down and enters the shell side of the recuperator. In the core region, the coolant flows around the test section with the fuel rods located at the bottom of the active channel. Underneath the fuel rods, the flow is turned around at the bottom and enters the test section, cooling the fuel rods. Hot flow then enters the recuperator above which water is further cooled down in the U-tube cooler to outlet temperature which is equivalent to the inlet temperature.
The fuel rods are constrained axially by these spacers and radially by wire wraps. To allow for compensation of the different thermal expansion of wire and cladding, 50 disk springs placed on top of each other were designed.
The assembly is designed such that it presses the disk springs during operation when the cladding is hotter than the wire. During manufacturing of the fuel rods, particularly when adding the wire wrap, it is important to install the wire wrap tightly. Therefore, a pad enables to adjust the vertical position of the disk springs, ensuring that the wire will be tight before fixing (welding) it to the top endings. A tight wire minimizes the risk of possible fretting of the fuel cladding. For the first test series to be performed, the available stainless steel 316L, qualified for reactor applications, shall be used for fuel claddings, which implies that the peak cladding temperature must be kept below 550°C under normal operating conditions. The material of the disk springs, the end plugs and the wire is from 316L as well to mitigate possible negative interaction.

The active channel occupies one cell of the active zone of the reactor. This determines the actual space limitation for maximum dimensions of the pressure tube. The entire active channel is inserted in an aluminium displacer, which determines its position in the grid of the active zone of the reactor LVR-15. The aluminium displacer assembly consists of an aluminium tube with diameter 70x3 mm, to which the actual displacer with a shape identical to an ordinary fuel assembly of the reactor LVR-15 is welded. Grooves are machined at the bottom of the aluminium displacer, which enable the displacer to be placed in its position in the active zone of the reactor. The entire active channel with its wall temperature in the active zone of up to 400°C is isolated from the colder water of the reactor pool with 50°C by an air gap of 3.5 mm between pressure tube and aluminium displacer. The air gap prevents the reactor water from boiling on the outer side of the aluminium displacer.
The length of the active channel with the aluminium displacer is determined by the level of the reactor water, by the height of the reactor lid and by the sufficient space for mounting and dismounting operations. The maximum height from the active core to the bottom of the reactor lid is approximately 5500 mm. From the head part 1 between inlet and outlet of the primary circuit, tube 1 (the most external tube with dimensions 36x1.5 mm) is hanging down, to which all other internal structures including the fuel assembly are fixed. Tube 1 allows for thermal expansion by about 16.5 mm with respect to the pressure tube.
During emergency cases, water is lead from the pressurizers TZ through the emergency tube and cools down the fuel rods. The emergency tube is terminated right above the fuel assembly so that the fuel assembly can be flooded in case of emergency. Some of the tubes of the recuperator will be blinded to serve as guide tubes for thermocouples. Several thermocouples are planned to pass through the emergency tube as well.

Task 1.2 Operational conditions and limits
The intended operational conditions are defined. Maximum temperature limits are a consequence of the selected materials (maximum temperature of the 316L cladding material is set to 550°C), which must be qualified for such nuclear application and which must be available today. This constraint had sorted out some advanced material options for the first test runs.
The nominal power of the fuel rods (63.6 kW) and the γ-power (9.8 kW) are a consequence of the chosen enrichment of the test fuel rods and of the available neutron flux in the reactor core. The coolant mass flow of 900 kg/hour is simulating hot channel conditions of an evaporator fuel assembly of the HPLWR.

Task 1.3 Auxiliary and safety systems
Since the SCWR-FQT loop includes nuclear fuel, a safety system of the test loop is unconditionally needed. The aim is to reduce the risk that a single failure of a critical system could cause release of radioactive material. The structure of the fuel itself, the fuel claddings and the high pressure system form the first, second and the third barrier, respectively, against release of radioactive material. The reactor building of the LVR-15 test reactor acts as the fourth barrier. In an experimental nuclear facility, the reactor building serves as containment, although it is not constructed as leak tight. Instead, the shielded bridge and the hall with primary block are kept below atmospheric pressure by active ventilation (at -20 Pa). Analyses verify that the active ventilation system can keep the pressure inside the experimental hall at sub-atmospheric conditions even during a large break LOCA, so that any activation of the primary coolant prior to the accident will be kept inside the building. Therefore, leakage of activity potentially present in the SCWR-FQT loop to the outside environment is prevented.
The purification system maintains the desired chemical parameters of primary water. Maintaining these parameters at required level is secured by continuous filtration of water, which passes through filters connected by a bypass to HCC and uses a pressure gradient generated by the pump. Recovered water is gradually cooled in economizer RV-2 and cooler CH-4 from 300°C to 40°C; then it passes through mechanical filter MF-1 to 4 ionex filters IF-1 connected in parallel. In case of accidents, there are 2 additional ionex filters parallel to the 2 standard ionex filters used for normal operation. The volume of the filtering medium (monospheric nuclear grade mixbed) is designed as 1 dm3 (for 1 filter) with a height of 0,85 m. The filtration speed is expected to be 15 m/h for nominal flowrate 30 kg/h.
The system of on-line measurement provides continuous control of chemical and physical properties of primary water. Depending on the required type of measurement, the system must enable installation of appropriate measuring probes as well as to provide sufficient cooling and pressure reduction for their operation. The measuring probes are installed for measurement of conductivity and H2 and O2 concentration. Dissolved gasses, mostly H2 or O2, are released from water when the pressure of water declines (behind reduction valve in venting device). From there they are drawn off by an air vent machine to a recombining chamber, where they are diluted by air flow to non-explosive concentration.
The secondary circuit is used for heat removal from the primary circuit. Heat transfer to the secondary circuit is realized in the u-tube cooler situated in the top part of the active channel. Heat transfer from the secondary circuit to the tertiary circuit is realized via cooler CH-1. The system is pressurized to 25 MPa (pressurizer KO-2 similar to KO-1) since it remains on high temperatures (200-235°C) to mitigate boiling. The nominal flow rate is designed for 1800 kg/h and is maintained by a pump CS-1 (similar to primary pump). There is an electric heater EO-1 (maximum power 50kW) which heats up the secondary circuit, which then transfers heat to the primary circuit (during start-up operation). During normal operation the electric heater is switched off. During shut-down, the electric heater may be employed to prevent thermal shocks during fast shut-down.

Task 1.4 Radiological protection
Due to the passage of the primary (active) pipelines through the LVR-15 reactor hall, it is required to place a biological shielding around these pipelines in order to ensure radiation safety of the facility.
Additionally, most of the primary circuit components of the proposed SCWR-FQT facility will be located in a separate hall adjacent to the existing old reactor hall. As presence of operational staff is expected during operation, biological shielding design and construction is necessary around the primary circuit components as well, in order to lower the dose rates around the facility to reasonably low values. MCNP calculations have shown that a satisfactory solution for the primary circuit shielding assuring safe operation would be a lead shielding with a thickness of 10 cm around the entire primary circuit and certain measurement circuit components, an additional shielding thickness of 7 cm around the ion-exchange filters to reduce the radiation impact of the accumulated corrosion products, and an extra shielding of at least 2 cm around the main circulation pump (the dominant 16N activity component). The pipelines connecting the reactor with the primary block would require at least a 6 cm thick lead layer.
The radiation monitoring of the facility will be provided by current systems of the LVR-15 reactor and additionally by monitoring system in the new experimental hall 211/12.

Task 1.4 Radiological protection also included work on a system for dismounting the fuel, transportation from the active channel to the storage and final transportation to the post-irradiation examination has been finalized.
For the rather low probable scenario when the SCWR-FQT experiment ends unexpectedly due to failure of one or more fuel pins a Fuel Handling System (FHS) is designed. This is a gas-tight design (in case fission gasses are present in the experiment) while separating the internals from the pressure tube of the experiment. Out of the three options for the FHS proposed initially, a decision for the most suitable the design was made and this has been worked out into final design.

Task 1.5 Civil engineering
This task included design of the new experimental building that will house the FQT facility is a standalone building directly adjacent to the original LVR-15 reactor building. The experimental building is directly accessible from the reactor building. The controlled area in the reactor building will be extended into the experimental building. The dimensions of the building are 10.55*14.10 m with a height ~11.85-12.30 m (volume 1760 m3).The structure of the building is a steel skeleton with sandwich type (steel sheet, insulation, steel sheet) shell.
The building is kept below atmospheric pressure by active ventilation (at -20 Pa). It was verified that the active ventilation system can keep the pressure inside the experimental hall at sub-atmospheric conditions even during a large break LOCA, so that any activation of the primary coolant prior to the accident will be kept inside this containment.

RESULTS OF WORK PACKAGE 2 “ANALYSES OF NORMAL OPERATION”
The main objective of WP2 “Analyses of normal operation” was to ensure the integrity of the in-pile test section including the wire-wrapped fuel rods under normal steady-state operation, as well as during start-up and shut-down. This has been done by means of flow and heat transfer analyses, structural analyses and neutronics analyses. Continuous exchange of information with WP1 was necessary to integrate the results from WP2 into the design of the FQT loop and vice versa.
WP2 was divided in four interconnected tasks. In task 2.1 the (thermal) stresses and integrity of the fuel rods and in-pile components were analysed. The flow and heat transfer in the loop was analysed in task 2.2 to verify the coolability of the fuel rods and associated peak temperatures. Neutronics analyses were performed in task 2.3 to determine a proper core arrangement, the reactivity effects of the loop on the core, the power from the fuel rods and the burn-up of the fuel. Shielding calculations have also been performed in Task 2.3 to ensure safety of the personnel during operation of the loop. In Task 2.4 the CFD codes and models have been validated. The achievements and results the tasks within WP2 are summarized below.

Task 2.1 Structural assessment
Structural analyses of the pressure tube and bottom piece using OpenFoam indicated that the stresses resulting from inside overpressure and temperature gradients were within the allowable limits according Czech standards (NTD ASI).
Structural analyses of the pressure tube’s head piece (including sealing stones and screws) were performed with the SALOME MECA code. The head piece design was modified in order to satisfy the NTD ASI limits.
Structural analyses of the recuperator tube plates and the primary cooling pipes were performed with the SALOME MECA code. The predicted stresses stay well within the allowable limits.
Structural analyses of the wire wrapped fuel assembly, based on the temperature distribution calculated with CFD, have been carried out. The analyses show that the fuel rods will bend slightly towards the assembly box wall due to the non-uniform temperature distribution over the rods. As a result, the wire spacer wrapped around each rod is squeezed in-between the cladding and the assembly box at some points. At these locations, stresses higher than the yield strength of the construction material (SS 316L) have been predicted. Since these peak stresses will relax directly after local plastic deformation, they are not a source of concern for safety.

Task 2.2 Flow and Heat transfer analyses of in-pile section
CFD analyses based on the conceptual loop design have been performed for a total heating power of 53.2 kW. Local hot spots were observed on the cladding in the wake of the lower fuel rod support plate. These hot spots were avoided by changing the geometry of the lower fuel rod support plate.
CFD analyses of the entire 4-rod fuel assembly (including conjugate heat transfer through the cladding, the wire-wrap spacers, the fuel rod support plates and the assembly box) have been performed for the modified loop design for a total heating power of 63.6 kW. The predicted peak cladding temperature for the modified design stays well below the acceptable cladding temperature of 550°C, see Fig. 4.

Sensitivity analyses showed that the cladding temperatures of the fuel rods are not significantly affected by the rod-wire contact area or gaps between rod and wire. Conduction through the cladding and wire does play an important role in the heat transfer process and should be included in the CFD analyses.
CFD analyses of the cladding temperature for the bended rods were not successful because of the high resolution required to correctly reconstruct the deformed fuel assembly geometry (as obtained from the structural analyses) in the CFD code. However, the predicted deformation is so small that little or no impact on the flow and temperature distribution is expected. As shown by sensitivity analyses, the change in rod-wire contact has also little effect on the cladding temperature. Even in a worst-case scenario, the cladding temperatures will stay well below the acceptable peak cladding temperature of 550°C.
Comprehensive CFD analyses have been performed to inspect the inflow and outflow to and from the test section. Large vortices at the inlet were observed for the conceptual design, which were solved by changing the geometry of the bottom closure of the inner guide tube.
CFD analyses of the recuperator and cooler sections have been performed. The heat exchange capacity of these components is sufficient to heat-up or cool-down the supercritical water coolant to the desired operating temperature.
Transient analyses with the system code ATHLET 3.0A have shown that the recommended limit for temperature gradients in the loop (50 K/h) were exceeded for a period of about 30 minutes at start-up and at shut-down of the reactor. Additional transient thermal hydraulic calculations and stress analyses demonstrated, however, that the resulting thermal stresses during start-up and shut-down are well within the acceptable limits.

Task 2.3 Neutronic analyses
Neutronic calculations were performed for a defined reference core configuration. D3 was chosen as the most suitable position of the active FQT test-section.
The total fuel power and power profiles were calculated using GLOBUS and MCNP codes. The results of the codes are in good agreement and both predict a linear power close to the allowed limit of 39 kW/m set for the HPLWR.
Burn-up calculations have been performed for different core configurations, which showed that maximum values of linear power will be reached during irradiation. It is demonstrated that power reduction can be achieved by insertion of control rod 4, or else by lower fuel enrichment.
Shielding analyses of the activated components of the primary system located outside the reactor have been performed with the Monte Carlo method implemented in the MCNP code. A lead shielding of minimal 15 cm has to be installed around the primary system components.

Task 2.4 CFD code validation and code to code comparison
A blind prediction exercise has been defined as back-up option for the CFD code validation against experimental data from the out-of-pile test performed in the SWAMUP facility in China. The blind prediction exercise, based on the SWAMUP geometry and nominal test conditions, has been performed and shows good comparison between the codes for predicting the mean cladding temperature. The agreement in predicted peak cladding temperature is less good and subject to further investigation by the involved partners outside the SCWR-FQT project.

RESULTS OF WORK PACKAGE 3 “SAFETY ANALYSES”
A safety system has been designed for the SCWR-FQT test facility to control and mitigate a number of design basis accidents. The primary system comprises the pressure tube, the main recirculation pump HCC and a check valve CV1; the coolant for this primary system is supplied from a tank DN and the pressure is built up with the boost pump VC. A bladder type compensator KO1 is minimizing pressure fluctuations due to density variations in the loop. The fissile and -power from the test section are removed by the secondary cooling system. Its recirculation pump CS is controlled by the outlet temperature of the primary system from the pressure tube, keeping it at 300°C. The secondary system is equipped with a cooler CH1 and a heater EO1 such that it can also heat up the primary system before the reactor is started.
In case of a loss of coolant from the primary system, the reactor is scrammed and two passive, bladder type accumulators TZ1 and TZ2 are adding their cold coolant inventory of 30 liters each to the feed line L1 of the primary loop and to the emergency cooling line L3, respectively. Their high coolant flow rates are quenching the test section during depressurization, avoiding a temporary boiling crisis or at least minimizing its consequences to acceptable peak cladding temperatures of less than 816°C. Afterwards, the residual heat is removed either by a feed line coolant injection system FLCI or by an emergency coolant injection system ELCI. Both systems are taking the required coolant from an emergency coolant reservoir HN1. As the exact position of the break can hardly be detected during such an accident, the decision criteria for activation of each system are based on pressures, temperatures and mass flow only. A system pressure of less than 22.5 MPa, measured in the feed line L1, or a coolant mass flow of less than 0.15 kg/s, measured with an orifice downstream of pump HCC, are automatically activating the FLCI system. A negative pressure difference between the feed line L1 and the emergency cooling line L3, or a coolant temperature of more than 500°C outlet of the test section are activating the ELCI system. In case of conflict, the ELCI system is the dominant system.
In case of a loss of flow, caused e.g. by a trip of pump HCC, by a flow blockage or by a shortcut of the coolant, thus bypassing the test section, the reactor is scrammed and the automatic depressurization valves ADS1 or ADS2 are opened with a delay of 1 sec. The choice between both valves is based on the decision criteria given above: valve ADS2 is opened together with activation of the FLCI system and valve ADS1 is opened together with the ELCI system, causing a reversed flow. Opening of any ADS valve is first causing a release of coolant from both accumulators TZ1 and TZ2. Once the pressure dropped to sub-critical conditions, pump HC1 of the FLCI system or pump HC2 of the ELCI system provide the required coolant flow rate to remove the residual heat.

Both ADS valves dump the coolant to the pressure suppression tank BN, which will increase its water level, while the one of the emergency coolant reservoir HN1 will decrease. A coolant pump HC3 between both tanks will refill coolant back into tank HN1 once the water level in tank BN has reached an upper limit.
In case of loss of the heat sink, e.g. in case of failure of the secondary cooling loop, the reactor is scrammed as well, the system is depressurized by valve ADS2 after a delay of 1 sec, and the FLCI system removes the residual heat to tank BN and from there to tank HN1 via pump HC3. As an independent heat sink, the residual heat can be removed to the reactor pool simply by flooding the insulation gap between the pressure tube and the aluminum displacer around it, using water from tank HV. Moreover, using this passive insulation gap flooding system IGFS, an active coolant flow would not be needed anymore after 12 days as heat can be removed just by natural convection then.
In case of loss of off-site power, these safety systems are powered by batteries and by a diesel generator.

Task 3.1 Design basis accidents
The performance of this safety system has been analyzed with the system code APROS. The reactor shut-down rods are released within 0.06 s after a scram signal has been given, and a driving time of 1 s is needed to fully insert them into the reactor core. The automatic depressurization valve, either ADS1 or ADS2, is assumed to open within 0.5 sec, but a delay of 1 sec between the scram signal and the signal for this valve to open is provided to reduce the fissile power while the pressure is still supercritical. All pumps have been modelled with a run-up or coast down time of 2 sec.
As an example for a loss-of-coolant-accident, let us consider a sudden, guillotine break of the emergency cooling line L3. The APROS result shows that this break will cause a rapid depressurization, immediately followed by scram of the reactor once the system pressure is less than 22.5 MPa. The accumulator TZ1 increases the initial, steady-state coolant mass flow of 0.25 kg/s to more than 1 kg/s, which reduces effectively the coolant temperature in the test section. The other accumulator TZ2 is just feeding into the break. The small contribution of the pressurizer KO1 is negligible. Within 30 sec after the break occurred, both accumulators will be empty. In the meantime, pump HC1 of the FLCI system is started and it continues to remove the residual heat with a small mass flow. Change over to the active system causes a short but small increase of the coolant temperature again, but the test section remains well cooled throughout the entire transient. Within 4.5 hours after the break happened, the coolant reservoirs BN and HN1 will be empty and missing water needs to be replaced then.
Results of these design basis accidents demonstrate how the safety system of the fuel qualification test loop is designed and how it is intended to work. The action taken after reactor scram is always depressurization of the primary system, either caused by a break which might happen at any position of the loop or, actively, by opening the automatic depressurization system. Immediate quenching of the hot test section is always provided by passive coolant injection from accumulators, of which two ones are needed to cover all kinds of accidents and break positions. Afterwards, two active pumps provide the required residual heat removal, which are designed for a small mass flow only to minimize the size of the required coolant reservoirs. These two pumps are not redundant but cover two different kinds of accidents. The large reactor pool can be used as an independent heat sink simply by flooding the insulation gap between the pressure tube and the pool.

Task 3.2 Safety assessment of the pressure tube
The mechanical consequences of a rupture of the primary system have been studied only for the pressure tube inside the reactor. Having a high pressure component inside a nuclear reactor core could become a severe problem if parts of this pressure tube or the pressure tube as a whole are driven around in the reactor pool in case of an accident, hindering either the control rods from shutting down the reactor or the coolant from cooling the other, ordinary fuel elements. If such cases occur, the rupture of the pressure tube would cause a severe secondary damage.
A ductile failure of the pressure tube at design pressure could happen in case of low cycle fatigue, thermal fatigue or creep of the pressure tube, including its welds, causing a steam jet which drives the pressure tube by its thrust. The aluminum displacer around the pressure tube will most likely survive this accident without secondary failure, but the pressure tube would be pushed upwards then. Similarly, a ductile rupture near the head piece could drive the pressure tube radially, with the risk of hindering or even blocking the control rods of the reactor. To avoid such accident, the top attachment of the pressure tube has been checked such that the pressure tube is kept in place under such accident scenarios.
A brittle failure of the pressure tube at design pressure is unlikely to occur, as the tube material is expected to remain sufficiently ductile during the test runs. Embrittlement of the pressure tube is to be expected only with high dose rates after long term exposure, which can easily be excluded, or in case of hydrogen embrittlement. Even in these unlikely cases, however, the aluminum displacer can be designed as an effective fragment protection.
Fragments of different size and shape have been assumed to be broken out of the pressure tube, accelerated by the coolant jet, which has been modelled with APROS again. For simulation of fragment impact, a section of pressure tube, displacer and outside water environment has been modelled with ANSYS.
As a conclusion, the surrounding aluminum displacer can indeed stop fragments resulting from a brittle failure of the pressure tube. In general, the surrounding water contributes significantly to the resistance of the aluminum displacer: A higher safety margin for the retention of brittle fragments could be obtained using a displacer material with a higher ductility, even if it has a reduced strength then.

Task 3.3 Potential consequences of a pressure tube rupture
The investigated accidental scenario beyond the design basis, which may possess a very low probability of occurrence, is a double-ended guillotine rupture of the feed line L1. All coolant lines are confined inside a shielded duct between the reactor and the separate experimental hall, so that the reactor hall would not be contaminated, but leaking fluids might enter the experimental hall, which is operated at 20 Pa sub-atmospheric pressure. Therefore, the hall is equipped with a filtering system, which keeps the pressure sub-atmospheric even in case of loss-of-coolant accidents. Thus, the reactor, the shielded duct and the experimental hall form a containment.
The initiating event is assumed to be further aggravated by a station black-out in combination with failure of all onsite emergency power supply systems. This would prevent the start-up of the emergency pumps and would stop the flow of the secondary circuit. However, for the first 25 seconds after the accident initiation, cooling of the rods is ensured by the passive pressure accumulator TZ1. After this period, heat is removed solely by conduction over the flooded insulation gap.
The accident scenario was modelled with APROS. For the first 25 seconds of the accident scenario, a similar coolant temperature progression is predicted as for a design basis accident. As long as accumulator TZ1 injects coolant into the system, the temperatures in the test section decrease rapidly. Different from these simulations, however, no subsequent active cooling was assumed to be available. Therefore, the stagnant water inside the assembly box would be heated up by the residual power of the fuel until it reaches saturation temperature. Since the coolant inventory of the test section is small, the complete coolant inventory of the assembly box is evaporated already after approximately 400 seconds, causing a rapid temperature increase. Short-term failure of the stainless steel claddings must be expected due to ballooning or rupture at a temperature of more than 816 °C. The predicted peak cladding temperature is almost reaching the melting point of the claddings. From that moment on, the shielding and retention capacity of the cooling system would be lost and airborne radioactive material would be free to enter the pressure tube. As the pressure tube would mostly be voided then, the radioactive material would enter the shielded duct without filtering. After reaching a maximum cladding temperature of 1400 °C, the temperatures decreases slowly due to decreasing decay power. According to the progression of cladding and fuel centerline temperatures, damage of the core and failure of the claddings must be expected. However, melting of the UO2 pellets (Tmelt = 2865 °C) and failure of the pressure tube are unlikely.
These and other, similar analyses of unlikely accident scenarios demonstrate that a secondary failure of the research reactor can be excluded. They indicate, however, that the use of active fuel inside a high pressure test loop requires a closed containment around the loop and its supply systems to avoid radioactive release to the environment under worst case assumptions. Moreover, diverse and redundant safety injection systems are advisable to minimize the risk of failure of all active safety systems, as postulated here. The SCWR fuel qualification test has been designed accordingly with a closed confinement around the entire test facility, which is strong enough to withstand even a postulated explosion of hydrogen, which could be produced during such postulated severe accidents.

Task 3.4 System code validation
The system codes APROS and ATHLET have both been used for one-dimensional, time dependent predictions of thermal-hydraulic phenomena to be expected in the FQT loop and its safety systems. Numerical results of both codes have been compared with each other to exclude input errors and improper use of these codes as well as errors of the physical models. The input of both codes was programmed independently by different users, using the same design data of the FQT facility. In general, a good agreement was found. Differences were rather due to uncertainties in the assumptions and in the selection of model parameters than in the code itself.
A significant difference was found, however, in predicting the two-phase flow heat transfer during during depressurization from supercritical to sub-critical pressure. A physical model has been derived to explain the phenomena, which have to be expected, based on observations in boiler tubes of supercritical fossil fired power plants: If the heated surface is hotter than the Leidenfrost temperature under supercritical pressure, we expect a boiling crisis during depressurization with significant temperature peaks. Up to now, however, these model equations have not been included yet in the APROS code. The ATHLET code includes the heat-up of dry zones when passing the critical pressure, but equations describing the subsequent progression of the quench front are still not yet included.
Transient accident analyses with APROS are showing short and intense peaks of the mass flow if a sudden break of the primary system is assumed. Such effects are known as water hammer phenomena. They can eventually result in surprisingly high pressure peak. Therefore, the APROS code has been validated to confirm that these effects are properly predicted. Similarly, predictions of oscillations of the mass flow have been validated, which are expected during some transients, indicating that the density wave oscillations might have happened.
Besides system codes, the sub-channel code MATRA, the safety anlysis code SIMMER, and the finite element code ANSYS, which had been used as well for prediction of certain phenomena during accidents, have been validated against experimental data.

RESULTS OF WORK PACKAGE 4 “PRE-QUALIFICATION”
The objectives of WP4 are to test and pre-qualify selected candidate materials suitable for fuel cladding, as well as the pre-qualification of the test section and critical components in a similar electrically heated facility.

Task 4.1: Material characterization
At the beginning of the project, selection of potential cladding materials for general corrosion and SCC tests in autoclaves has been done; four austenitic stainless steels: 316L, 316Ti, Nb stabilized 347 H and Ti stabilized 08Cr18Ni10Ti (321 SS equivalent) were chosen and purchased. Mechanical and physical properties of these materials, needed for design and analyses, were collected from literature and available material data sheets. The cutting schemes were prepared, specimens manufactured in JRC-IET (coupons) and in CVR (SSRT specimens) and then distributed to VTT and JRC-IET for testing. Coupons were manufactured according to ASTM G1-03 with milled surface finish, while SSRT specimens manufactured according to ASTM G129.
No additional heat treatment has been performed. The basic microstructural analyses by SEM and EDX showed typical microstructure corresponding to this type of stainless steel described in detail in materials data sheets.

Task 4.2: Corrosion experiments in low and high oxygen SCW environment
Long term (2000 h) general corrosion tests at different SCW temperatures 500 and 550oC (p = 25 MPa) at two different concentrations of dissolved oxygen (2000 ppb in JRC-IET and 150 ppb in VTT) were performed.
JRC-IET performed tests in their autoclave connected to a recirculation loop at 500 and 550oC in SCW with 2000 ppb O2 using all four materials included in the portfolio. The tests were interrupted in regular time periods (after 600, 1400 and 2000 h), therefore, both the information on kinetics of corrosion process was obtained and the effect of cooling-down/heating-up on integrity of protective oxide layers was evaluated. Post-test examination of the exposed specimens consisted of weight change calculations and detailed macro- and microscopic investigation of oxide layers using optical microscopy, SEM and EDX. Finally, linear extrapolation to one year of operation in SCW was used as the most conservative approach. Experiments performed in VTT focused on long corrosion exposures in supercritical water at 550°C and concentration of 150 ppb dissolved oxygen. General corrosion tests were performed with weight gain coupon specimens and the coupons were exposed up to 3610 hours. After exposure, oxide layers have been examined using SEM/EDS. In addition, 3D-profilometer analyses were performed in order to obtain more information on sample surface morphology.
Additional corrosion tests were performed in both JRC-IET, VTT and SJTU (within the Chinese project SCRIPT) with the main target to investigate how to further increase corrosion resistance of selected materials by optimization of surface finish procedure such as shot-peening and/or sand blasting. Finally, as part of the parallel Chinese project SCRIPT, the effect of hydrogen water chemistry (HWC) on corrosion resistance of selected materials was evaluated at SJTU.

Task 4.3: SCC tests using cylindrical or plate type tensile specimens (i.e. Slow Strain Rate Tests – SSRTs)
This task contributes to the understanding of SCC susceptibility of the selected materials in SCW. The SSRT tests have been performed at two different temperatures 500 and 550oC (p = 25 MPa) in SCW with either 2000 ppb (JRC-IET) or 150 ppb O2 (VTT).
In JRC-IET, slow strain rate tensile tests (SSRT) have been performed at 550oC using a step-motor controlled loading device in an autoclave. Susceptibility to SCC was examined by SSRT tests with constant elongation rate of 5.2x10-7 s-1 in an autoclave connected to a recirculation loop allowing continual water chemistry control during the test. Moreover, reference tests under the same temperature and loading conditions have been conducted in an inert nitrogen atmosphere, in order to distinguish the effect of environment on mechanism of failure. During all tests, the loaded tensile specimens have been continually monitored by acoustic emission (AE) measurement with 1-D linear location mode in order to determine the time to SCC crack initiation. The obtained stress-strain curves did not indicate any significant change of mechanical properties of the examined materials in 550°C SCW compared to those measured in inert atmosphere. Nevertheless, post-test SEM fractographical analysis showed evidences of SCC initiation areas on main fracture surface and elevated number of secondary cracks along gage section of the tensile specimen. Sudden release of significant number of high energy AE signals, observed during the SSRT tests in SCW, was probably caused by initiation of SCC cracks because such AE signals were not detected within reference tests.
SCC susceptibility SSRT tests were performed by VTT simultaneously with general corrosion tests on three candidate materials. The plate type tensile specimens were exposed to the same environment as the coupons. The tests were performed using a step motor controlled loading device. SSRT tests were carried out according to ASTM G129 but with lower strain rate (1.0x10-7s-1) to make the effect of SCW and thus corrosion process more pronounced. Fractured specimens were studied with SEM to determine the mechanism of fracture. Of the tested steels, 316L had the lowest susceptibility to SCC. No clear indication of SCC was observed with SEM in 316L whereas transgranular SCC and possible intergranular SCC were observed in 316Ti and 347H. Due to the time consuming very slow SSRT tests, only 550°C tests have been conducted by VTT.

Task 4.4: Transfer of material tests results
On the basis of the general corrosion and SCC SSRT tests the selection of the cladding material was arranged in the following order, ranging from the most to the least suitable for the given test conditions: 316L - 347H, 08Cr18Ni10Ti and 316Ti.
Based on assessment of experimental results and calculations, fuel rod mock-ups were designed and manufactured in CVR. Two different accident scenarios of fuel cladding failure, namely a LOCA (potentially leading to ballooning of the cladding) and loss of internal pressure (leading to buckling) were evaluated both theoretically and experimentally. It must be stressed that these experiments were conducted at laboratory scale using a facility primarily designed for material testing; therefore, certain limitations had to be taken into account. These included significantly lower flow rates and subsequently lower heat flux given by limitations of the high pressure pump, pre-heater and autoclave heater.

Task 4.5: Thermal-hydraulic pre-qualification of the test section using SWAMUP facility
The objective of this task is to validate operational and safety analyses of the designed facility. For this purpose, a test section, as designed in WP1 for the fuel qualification test, was designed with electrically heated rods such that the same flow conditions around the fuel rods and in surrounding flow channels can be expected. Using this design, full scale out-of-pile thermal-hydraulic test of the fuel test bundle shall be carried out at the SWAMUP test facility of SJTU/CGNPC, to verify the fuel bundle design. However, due to several delays accumulated during the course of the project (delayed start of the counterpart SCRIPT project in China, long delivery times for the heating rods, manufacturing possibilities in China) experimental data have not been provided before the end of the project. The tests are expected to be completed in early 2015 and yield the expected results and data. The project partners intend to perform the validation of codes when the data are available.

Potential Impact:
POTENTIAL IMPACT OF THE PROJECT
Current uncertainties in core design for supercritical water cooled reactors have been expected to be reduced by the pre-qualification tests performed in China in work package 4. By the end of the project, the tests were in progress but have not been finalized and the data were therefore unavailable. Hence, limited conclusions may be drawn as to the impact on possible improvement of codes used for numerical predictions. However, as the experimental plan for these tests and the design of the electrically heated test section including the measurement points have been well prepared during the project, potential impact of the awaited results may be expected as follows:
In addressing uncertainties of coolant heat transfer in the vicinity of the pseudo-critical point, the pre-qualification test facility SWAMUP measures cladding surface temperatures of the electrically heated fuel assembly under realistic pressures, temperatures, heat and mass flux such that numerical predictions of CFD or sub-channel codes can be compared directly with the experimental results. 8 thermocouples are placed along the height of each rod (i.e. total of 32 thermocouples for the entire test section) at different axial positions to measure the surface temperatures. 48 steady-state and 34 transient tests with varying thermal-hydraulic parameters are performed. Test conditions with outlet temperature close to the pseudo-critical point, i.e. in the most sensitive range, are included.
Starting with supercritical conditions, the SWAMUP facility can be depressurized easily to subcritical conditions, making it possible to measure fuel rod temperatures throughout the boiling crisis during slow depressurization transients. Currently, neither the APROS code nor the ATHLET code can predict transient heat transfer phenomena during depressurization; and given that the APROS code needs deeper modifications in order to predict these phenomena correctly, as shown in work package 3, the tests are expected to yield a large amount of data, otherwise unavailable in literature, that could be beneficial in this respect.
Uncertainties of the corrosion resistance of stainless steels used as fuel cladding material is reduced by the autoclave tests performed in work package 4 in both Europe as well as China. Data are also available from tests with samples treated by surface cold work, which has been shown to have positive effect on corrosion resistance, although more tests and standardization of the used surface treatment methods will be necessary.
The European – Chinese collaboration is therefore most helpful in the areas mentioned above as expensive tests are shared and human resources are duplicated.

Moreover, the final design and results of analyses of the fuel qualification test facility form the basis of the licensing documentation, through which completeness of safety studies is ensured for an exemplary nuclear facility with supercritical water, although with limited scope and limited number of components. This procedure is a generic test case to identify which problems are to be expected with licensing of a prototype SCWR in the next decade. The fuel qualification test features already the basic principles of a small test reactor, such as:
- Fuel at realistic temperatures and realistic linear heat rate;
- Coolant at realistic pressure, heat and mass flux, and in the most problematic temperature range;
- Safety systems with depressurization systems and low pressure coolant injection pumps like in SCWR;
- Thermal fatigue of components exposed to large temperature differences;
- Design details of realistic fuel rods for the SCWR, like end caps, spacers and assembly boxes.
As the facility will be built from an ongoing project in the Czech Republic, financed from structural funds, and licensed by the Czech nuclear authority, the design available from work package 1 and analyses from work packages 2 and 3 undergo a very critical review, which is an excellent exercise for later prototype design. Through discussion with members of the Technical Advisory Board from both Europe and China, high quality and thoroughness of predictions is achieved. Members of the Technical Advisory Board were from Europe – from the Halden Reactor in Norway with experience in testing of fuel, and from the Czech regulator authority SUJB, and from China – from the State Nuclear Power Technology Corporation and China Power Investment Corporation-Power Engineering Company.
Safety analyses and licensing of the facility, as mentioned above, are unavoidable steps before the fuel qualification test facility can be built and operated. Building of the facility follows this project right after its end and is financed by the Czech Sustainable Energy (SUSEN) project using structural funds. Further continuation of the project in the sense of the SCWR system research plan is thus already ongoing.
The scope of the fuel qualification test, as carried out in this project, is enlarged by a contribution from Canada where the test is intended to be carried out at higher temperature, using the designed facility, but with a modified active channel. Negotiations have been ongoing throughout the project and will continue beyond it to sign a project arrangement in the Generation IV International Forum with such an objective. In 2014, China has also signed the SCWR System Arrangement in GIF and, together with Russia, have joined the first GIF PPMB meeting of the project.

MAIN DISSEMINATION ACTIVITIES AND EXPLOITATION OF RESULTS
Given the nature of the project the most important means of communication and dissemination of the generated knowledge has been through the GIF community. The complexity of the fuel qualification test calls for an experimental facility flexible enough to carry out similar tests for the different concepts of the SCWR in GIF, whether it be the European HPLWR, the Canadian CANDU SCWR or the Japanese JSCWR. A project within GIF, dedicated to the fuel qualification test in supercritical water, has been negotiated; signature of the project arrangement is foreseen to take place after the end of this project. During the course of the project, China has also signed the SCWR System Arrangement and a representative has been present at the last PPMB meeting which was held in October 2014. The same meeting has also been newly attended by a representative from Russia.
A broader communication route is established through informing the wider scientific community and through involving students of Doctorate programs in the research work. Numerous papers have been presented at conferences, topical meetings and workshops, such as:
• International Symposium on Supercritical Water-Cooled Reactors;
• International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety;
• International Topical Meeting on Nuclear Reactor Thermal Hydraulics;
• Nordic Nuclear Materials Forum for Generation IV Reactors;
• 10th SCWR Information Exchange Meeting;
• International Conference on Nuclear Engineering;
• Joint HZDR & ANSYS Conference;
• The European Nuclear Conference;
• Siempelkamp Workshop “Kompetenzerhaltung in der Kerntechnik” (Maintaining Competence in the Nuclear Technology);
• European conference on Euratom research and training in reactor systems;
• European Research Reactor Conference;
• STAR Global Conference;
• Pacific Basin Nuclear Conference;
• Annual Meeting on Nuclear Technology.
Results from the project were also published through articles in the following journals:
• Progress in Nuclear Energy;
• International Journal of Heat and Mass Transfer;
• Nuclear Engineering and Design;
• Safety of Nuclear Energy (journal published by the Czech regulator – in Czech).

In the Karlsruhe Institute of Technology (KIT), two doctorate students have completed their theses within this project, namely:
• Manuel Raqué: Safety Analysis for a Fuel Qualification Test with Supercritical Water;
• Tobias Zeiger: Auswirkungen eines Druckrohrversagens während des SCWR Fuel Qualification Tests.
List of Websites:
The project consortium consists of the following partners:

1. Centrum vyzkumu Rez (CVR)
contact persons:
Mariana Ruzickova, Mariana.Ruzickova@cvrez.cz;
Ales Vojacek, Ales.Vojacek@ujv.cz.

2. Karlsruhe Institute of Technology (KIT)
contact persons:
Thomas Schulenberg, schulenberg@kit.de.

3. Nuclear Research and Consultancy Group (NRG)
contact persons:
Dirk Visser, visser@nrg.eu.

4. Magyar Tudomanyos Akademia KFKI Atomenergia Kutatointezet (MTA-EK, former KFKI)
contact persons:
Csaba Maraczy, maraczy.csaba@energia.mta.hu.

5. Valtion Teknillinen Tutkimuskeskus (VTT)
contact persons:
Aki Toivonen, aki.toivonen@vtt.fi;
Joona Kurki, joona.kurki@vtt.fi.

6.Budapesti Muszaki es Gazdasagtudomanyi Egyetem (BME)
contact persons:
Attila Kiss, kissa@reak.bme.hu.

7. Joint Research Centre – European Commission, Institute for Energy and Transport (JRC-IET)
contact persons:
Radek Novotny, Radek.NOVOTNY@ec.europa.eu.

Related information

Reported by

CENTRUM VYZKUMU REZ S.R.O.
Czech Republic

Subjects

Nuclear Fission
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