Numerical simulation of natural circulation experiments conducted at the IIST facility

Natural circulation plays an important role in long-term cooling of pressurized water reactors (PWRs) under small break loss-of-coolant accidents. Recently, natural circulation experiments have been conducted at the Institute of Nuclear Energy Research integral system test (IIST) facility, which is used to simulate the Westinghouse three-loop Maanshan PWR. A numerical simulation is presented to investigate the natural circulation phenomena of the IIST facility with the RELAP5/MOD3 code. The calculated results are in good agreement with the experimental data of the single-phase natural circulation both quantitatively and qualitatively. The influences of power level and system pressure on natural circulation can also be predicted by the current model. Based on the two-phase natural circulation data, the calculated flow rate history is similar to that obtained from the experiment.     

Design,experiments and Relap5 code calculations for the perseo facility

Research on innovative safety systems for light water reactors addressed to heat removal by in-pool immersed heat exchangers, led to design, build-up and test the PERSEO facility at SIET laboratories.The research started with the CEA-ENEA proposal of improving the GE-SBWR isolation condenser system, by moving the triggering valve from the high pressure primary side of the reactor to the low pressure pool side. A new configuration of the system was defined with the heat exchanger contained in a small pool, connected at bottom and top to a large water reservoir pool, the triggering valve being located on the pool bottom connecting pipe.ENEA funded the whole activity that included the definition and build-up of a new heat exchanger pool, on the basis of the already existing PANTHERS IC-PCC facility, at SIET laboratories, and the new plant requirements. The heat exchanger connections to the pressure vessel were maintained.An experimental campaign was executed at full scale and full thermal-hydraulic conditions for investigating the behaviour and performance of the plant in steady and unsteady conditions. The Relap5 code was utilised during all phases of the research: for the heat exchanger pool dimension definition and from pre-test and post-test analyses. The Cathare code was applied too from pre-test and post-test analyses.This paper deals with the experimental and calculated results limited to the Relap5 code.     

Full-scale modelling of the MNSR reactor to simulate normal operation, transients and reactivity insertion accidents under natural circulation conditions using the thermal hydraulic code ATHLET

A full-scale ATHLET system model for the Syrian miniature neutron source reactor (MNSR) has been developed. The model represents all reactor components of primary and secondary loops with the corresponding neutronics and thermal hydraulic characteristics. Under the MNSR operation conditions of natural circulation, normal operation, step reactivity transients and reactivity insertion accidents have been simulated. The analyses indicate the capability of ATHLET to simulate MNSR dynamic and thermal hydraulic behaviour and particularly to calculate the core coolant velocity of prevailing natural circulation in presence of the strong negative reactivity feed back of coolant temperature. The predicted time distribution of reactor power, core inlet and outlet coolant temperature follow closely the measured data for the quasi steady and transient states. However, sensitivity analyses indicate the influence of pressure form loss coefficients at core inlet and outlet on the results. The analysis of reactivity accidents represented by the insertion of large reactivity, demonstrates the high inherent safety features of MNSR. Even in case of insertion of total available cold excess reactivity without scram, the high negative reactivity feedback of moderator temperature limits power excursion and avoids consequently the escalation of clad temperature to the level of onset of sub-cooled void formation. The calculated peak power in this case agrees well with the data reported in the safety analysis report. The ATHLET code had not previously been assessed under these conditions. The results of this comprehensive analysis ensure the ability of the code to test some conceptual design modifications of MNSR's cooling system aiming the improvement of core cooling conditions to increase the maximum continuous reactor operation time allowing more effective use of MNSR for irradiation purposes.     

Experiments on slug mixing under natural circulation conditions at the ROCOM test facility using high-resolution measurement techniques and numerical modeling

ROCOM is a four-loop test facility used for the investigation of coolant mixing in the primary circuit of pressurized water reactors. Recently, a new sensor was developed for an improved visualisation and quantification of the coolant mixing in the downcomer. This new sensor array spans a dense measuring grid and covers nearly the whole downcomer. In the presented work, special emphasis was given to the comparison of the data of this sensor with the results of calculations using the Computational Fluid Dynamics (CFD) code ANSYS CFX. A coolant mixing experiment during natural circulation conditions has been conducted. The underlying scenario of this experiment is based on a boron dilution scenario following a SBLOCA event. The corresponding CFD code solution has been obtained using the Best Practice Guidelines. All main effects observed in the measurement are described by the calculation. The detailed comparison reveals that the calculation underestimates the coolant mixing inside the reactor pressure vessel.The measurement data, boundary conditions of the experiment and facility geometry can be made available to other CFD code users for benchmarking.     

Capability of the RELAP5 code to simulate natural circulation behavior in test facilities

Many advanced reactor designs incorporate passive systems mainly to enhance the operational safety and possible elimination of severe accident condition. Some reactors are even designed to remove the nominal fission heat passively by natural circulation without using mechanical pumps e.g. ESBWR, AHWR, CHTR, CAREM, etc. while in most other new reactor concepts, the decay heat is removed passively by natural circulation following the pump trip conditions. The design and safety analysis of these reactors are carried out using the best estimate codes such as RELAP5, TRAC and CATHARE, etc. These best estimate codes have been developed for pumped circulation systems and it is not proven about their adequacy or applicability for natural circulation systems wherein the driving mechanism is completely different. Some of the key phenomena which are difficult to model but are significantly important to assess the natural circulation system performances are – low flow natural circulation mainly because the flow is not fully developed and can be multi-dimensional in nature; flow instabilities; critical heat flux under oscillatory condition; flow stratification particularly in large diameter vessel; thermal stratification in large pools; effect of non-condensable gases on condensation, etc. Though, these best estimate codes use a six equation two-fluid model formulation for the thermal-hydraulic calculation which is considered to be the best representative of two-phase flows, but their accuracies depend on the accuracies of the models for interfacial relationships for mass, energy and momentum transfer which are semi-empirical in nature. The other problem with two-fluid models is the effect of ill-posedness which may cause numerical instability. Besides, the numerical diffusion associated due to truncation of higher order terms can affect the prediction of flow instabilities. All these effects may lead to inability to capture the important physical instability in natural circulation systems and instability characteristics i.e. amplitude and frequency of flow oscillation. In view of this, it is essential to test the capability of these codes to simulate natural circulation behavior under single and two-phase flow conditions before applying them to the future reactor concepts.In the present study, one of the extensively used best estimate code RELAP5 has been used for simulation of steady state, transient and stability behavior of natural circulation based experimental facilities, such as the High-Pressure Natural Circulation Loop (HPNCL) and the Parallel Channel Loop (PCL) installed and operating at BARC. The test data have been generated for a range of pressure, power and subcooling conditions. The computer code RELAP5/MOD3.2 was applied to predict the transient natural circulation characteristics under single-phase and two-phase conditions, thresholds of flow instability, amplitude and frequency of flow oscillations for different operating conditions of the loops. This paper presents the effect of nodalisation in prediction of natural circulation behavior in test facilities and a comparison of experimental data in with that of code predictions. The errors associated with the predictions are also characterized.     

Experimental and post-test calculation results of the integral reflood test QUENCH-12 with a VVER-type bundle

The QUENCH-12 experiment was carried out to investigate the effects of VVER materials (niobium-bearing alloys) and bundle geometry on core reflood, in comparison with test QUENCH-06 using western PWR materials (Zircaloy-4) and bundle geometry. The test protocol involved pre-oxidation to a maximum of about 150 μm oxide thickness at a temperature of about 1450 K, followed by a power ramp until a temperature of 2050 K was reached, then reflood with water at room temperature was initiated. The total hydrogen production was 58 g (QUENCH-06: 36 g), 24 g of which were released during reflood (QUENCH-06: 4 g). Reasons for the increased hydrogen production may be extensive damaging of the cladding surfaces due to the breakaway oxidation and local melt formation with subsequent melt oxidation. Post-test videoscope observations and metallographic investigations showed an influence of the breakaway oxidation with extensive spalling of oxide scales of rod claddings, shroud and auxiliary corner rods. The hydrogen content in the corner rods, withdrawn from the bundle during the test, reached more than 30 at% at the bundle elevations of 850 and 1100 mm. Post-test calculations were performed with local versions of SCDAP/RELAP5 following on from pre-test analyses with SCDAP/RELAP5 and SCDAPSIM.     

An integral effect test of a complete loss of reactor coolant system flow rate for the SMART design using the VISTA-ITL facility and its simulation with the MARS-KS code

An integral effect test was successfully performed to provide data to assess the capability of the system analysis code to simulate a complete loss of reactor coolant system (RCS) flow rate (CLOF) scenario for the SMART (System-integrated Modular Advanced ReacTor) design. The steady-state conditions were achieved to satisfy initial test conditions presented in the test requirement, its boundary conditions were accurately simulated, and the CLOF scenario in the SMART design was reproduced properly using the VISTA-ITL facility. The natural circulation flow rate in the RCS was about 12.0% of the rated RCS flow rate and the flow rate in the passive residual heat removal system (PRHRS) loop was about 10.6% of its rated value in the early stage of the PRHRS operation. In this paper, the major experimental results of the CLOF test are discussed. The test results were analyzed using the best-estimate system analysis code, MARS-KS, to assess its capability to simulate a CLOF scenario for the SMART design.     

Development of the FHR advanced natural circulation analysis code and application to FHR safety analysis

The University of California, Berkeley (UCB) is performing thermal hydraulics safety analysis to develop the technical basis for design and licensing of fluoride-salt-cooled, high-temperature reactors (FHRs). FHR designs investigated by UCB use natural circulation for emergency, passive decay heat removal when normal decay heat removal systems fail. The FHR advanced natural circulation analysis (FANCY) code has been developed for assessment of passive decay heat removal capability and safety analysis of these innovative system designs. The FANCY code uses a one-dimensional, semi-implicit scheme to solve for pressure-linked mass, momentum and energy conservation equations. Graph theory is used to automatically generate a staggered mesh for complicated pipe network systems. Heat structure models have been implemented for three types of boundary conditions (Dirichlet, Neumann and Robin boundary conditions). Heat structures can be composed of several layers of different materials, and are used for simulation of heat structure temperature distribution and heat transfer rate. Control models are used to simulate sequences of events or trips of safety systems. A proportional-integral controller is also used to automatically make thermal hydraulic systems reach desired steady state conditions. A point kinetics model is used to model reactor kinetics behavior with temperature reactivity feedback. The underlying large sparse linear systems in these models are efficiently solved by using direct and iterative solvers provided by the SuperLU code on high performance machines. Input interfaces are designed to increase the flexibility of simulation for complicated thermal hydraulic systems. This paper mainly focuses on the methodology used to develop the FANCY code, and safety analysis of the Mark 1 pebble-bed FHR under development at UCB is performed.     

The light water integral reactor with natural circulation of the coolant at supercritical pressure B-500 SKDI

Pressure increase in the primary circuit over the critical value gives a possibility to construct the B-500SKDI (500 MWe) lightwater integral reactor with natural circulation of the coolant in the vessel with a diameter less than 5 m. The given reactor has a high safety level, simple operability, its specific capital cost and fuel expenditure being lower as compared to a conventional PWR. The development of the B-500SKDI reactor is carried out taking into consideration verified technical decisions of current NPPs on the basis of Russian LWR technology.     

Benchmark analysis of main circulation pump trip events at the Ignalina NPP using RELAP5 code

The state-of-the-art code RELAP5/MOD3 was originally designed for PWRs. Because of unique RBMK designs the application of this code to RBMK-1500 encountered several problems. A successful best estimate RELAP5 model of the Ignalina NPP has been developed. This model includes the reactor main circulation circuit (MCC) and reactor control and protection system required for this kind of transient analysis. Benchmark analysis of all operating main circulation pump (MCP) trip events was performed. During the analysis the characteristics of isolation control valves and MCP throttling regulating valves were established. Comparison of calculated and measured parameters was also used to establish realistic resistances of different MCC components and realistic behaviour of the controllers of the reactor systems. Calculations performed with the RELAP5 model, which includes these modifications, compare favourably with plant data.     

A decade of experiments and recent upgrading at the AMS facility in Bucharest

The Bucharest AMS facility has been in operation since 1998. We shortly present the performed experiments, the major upgrade of the AMS facility at NIPNE – Bucharest and the ongoing progress resulting since. We mounted a new ion source, of NEC 40 sample MC-SNCIS type and we reinforced the vacuum on the injector deck. Computer control on all parameters of the injector deck was implemented through a build-in-house electronic set-up. By converting the Tandem accelerator from a belt-driven charging system to a Pelletron and by introducing a modern GVM we have obtained a reduction of the fluctuations of the terminal voltage by at least two orders of magnitude.     

Development of SIRIUS-N facility with simulated void-reactivity feedback to investigate regional and core-wide stability of natural circulation BWRs

The SIRIUS-N facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of a natural circulation BWR. A real-time simulation was performed in the digital controller for modal point kinetics of reactor neutronics and fuel-rod conduction on the basis of measured void fractions in reactor core sections of the thermal-hydraulic loop. Stability experiments were conducted for a wide range of thermal-hydraulic conditions, power distributions, and fuel rod time constants, including the nominal operating conditions of a typical natural circulation BWR. The results show that there is a sufficiently wide stability margin under nominal operating conditions, even when void-reactivity feedback is taken into account. The stability experiments were extended to include a hypothetical parameter range (double-void reactivity coefficient and inlet core subcooling increased by a factor of 3.6) in order to identify instability phenomena. The regional instability was clearly demonstrated with the SIRIUS-N facility, when the fuel rod time constant matches the oscillation period of density wave oscillations.     

Calculation of the VVER-1000 coolant transient benchmark using the coupled code systems DYN3D/RELAP5 and DYN3D/ATHLET

Plant-measured data provided by the OECD/NEA VVER-1000 coolant transient benchmark programme were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to an experiment that was conducted during the commissioning of the Kozloduy NPP Unit 6 in Bulgaria. In this experiment, the fourth main coolant pump was switched on whilst the remaining three were running normal operating conditions. The experiment was conducted at 27.5% of the nominal level of the reactor power. The transient is characterized by a rapid increase in the primary coolant flow through the core, and as a consequence, a decrease of the space-dependent core inlet temperature. The control rods were kept in their original positions during the entire transient. The coupled simulations performed on both DYN3D/RELAP5 and DYN3D/ATHLET were based on the same reactor model, including identical main coolant pump characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has also been performed to evaluate the respective thermal hydraulic models of the system codes RELAP5 and ATHLET.     

Modeling of a buoyancy-driven flow experiment at the ROCOM test facility using the CFD codes CFX-5 and Trio_U

The influence of density differences on the mixing of the primary loop inventory and the emergency core cooling (ECC) water in the downcomer of a pressurized water reactor (PWR) was analyzed at the ROssendorf COolant Mixing (ROCOM) test facility. ROCOM is a 1:5 scaled model of a German PWR, and has been designed for coolant mixing studies. It is equipped with advanced instrumentation, which delivers high-resolution information for temperature or boron concentration fields.An experiment with 5% of the design flow rate in one loop and 10% density difference between the ECC and loop water was selected for validation of the CFD software packages CFX-5 and Trio_U. Two similar meshes with approximately 2 million control volumes were used for the calculations. The effects of turbulence on the mean flow were modeled with a Reynolds stress turbulence model in CFX-5 and a LES approach in Trio_U. CFX-5 is a commercial code package offered from ANSYS Inc. and Trio_U is a CFD tool which is developed by the CEA-Grenoble, France.The results of the experiment and of the numerical calculations show that mixing is dominated by buoyancy effects: at higher mass flow rates (close to nominal conditions) the injected slug propagates in the circumferential direction around the core barrel. Buoyancy effects reduce this propagation. The ECC water falls in an almost vertical path and reaches the lower downcomer sensor directly below the inlet nozzle. Therefore, density effects play an important role during natural convection with ECC injection in PWRs. Both CFD codes were able to predict well the observed flow patterns and mixing phenomena.     

Thermal-hydraulic calculations using MARS code applied to low power and shutdown probabilistic safety assessment in a PWR

The methods developed for full-power probabilistic safety assessment, including thermal-hydraulic methods, have been widely applied to low power and shutdown conditions. Experience from current low power and shutdown probabilistic safety assessments, however, indicates that the thermal-hydraulic methods developed for full-power probabilistic safety assessments are not always reliable when applied to low power and shutdown conditions and consequently may yield misleading and inaccurate risk insights. To increase the usefulness of the low power and shutdown risk insights, the current methods and tools used for thermal-hydraulic calculations should be examined to ascertain whether they function effectively for low power and shutdown conditions. In this study, a platform for relatively detailed thermal-hydraulic calculations applied to low power and shutdown conditions in a pressurized water reactor was developed based on the best estimate thermal-hydraulic analysis code, MARS2.1. To confirm the applicability of the MARS platform to low power and shutdown conditions, many thermal-hydraulic analyses were performed for the selected topic, i.e. the loss of shutdown cooling events for various plant operating states at the Korean standard nuclear power plant. The platform developed in this study can deal effectively with low power and shutdown conditions, as well as assist the accident sequence analysis in low power and shutdown probabilistic safety assessments by providing fundamental data. Consequently, the resulting analyses may yield more realistic and accurate low power and shutdown risk insights.     

Analyses of the V1000CT-1 benchmark with the DYN3D/ATHLET and DYN3D/RELAP coupled code systems including a coolant mixing model validated against CFD calculations

Plant-measured data provided within the specification of the OECD/NEA VVER-1000 coolant transient benchmark (V1000CT) were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to the MCP (main coolant pump) switching on experiment conducted in the frame of the plant-commissioning activities at the Kozloduy NPP Unit 6 in Bulgaria. The experiment was started at the beginning of cycle (BOC) with average core expose of 30.7 effective full power days (EFPD), when the reactor power was at 27.5% of the nominal level and three out of four MCPs were operating. The transient is characterized by a rapid increase in the primary coolant flow through the core and, as a consequence, a decrease of the space-dependent core inlet temperature. Both DYN3D/RELAP5 and DYN3D/ATHLET analyses were based on the same reactor model, including identical MCP characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. For an adequate modelling of the redistribution of the coolant flow in the reactor pressure vessel during the transient a simplified mixing model for the DYN3D/ATHLET code was developed and validated against a computational fluid dynamics calculation.

The results of both coupled code calculations are in good agreement with the available experimental data. The discrepancies between experimental data and the results of both coupled code calculations do not exceed the accuracy of the measurement data. This concerns the initial steady-state data as well as the time histories during the transient. In addition to the validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has been performed to evaluate relevant thermal hydraulic models of the system codes RELAP5 and ATHLET and to explain differences between the calculation results.     

An integral loop test and MARS code analysis for a DVI line break LOCA in the APR1400

The transient thermal-hydraulic phenomena of a DVI (Direct Vessel Injection) line break LOCA (Loss-of-Coolant Accident) in pressurized water reactor, APR1400, were investigated. In order to understand the phenomena during the LOCA transient, a reduced-height and reduced-pressure integral loop test facility, the SNUF (Seoul National University Facility), was constructed with scaling down the prototype. For the appropriate test conditions in the experiment with the SNUF, the energy scaling method was suggested with scaling the coolant mass inventory and the thermal power for the reduced-pressure condition. According to the conditions determined by the method, the experimental study was performed with the SNUF. The experimental results showed that the phenomenon of the downcomer seal clearing played a dominant role in the reduction of the system pressure and the recovery of the coolant level in the core. That phenomenon occurred when the steam incoming from cold legs penetrates the coolant in the upper downcomer toward the broken DVI line. The experimental results were compared with the prototype analysis to estimate the energy scaling method, so that the experiment reasonably simulated the phenomena in the prototype. For the analytical investigation, the experiment was simulated with MARS code to validate the calculation capability of the code, especially for the downcomer seal clearing, which showed good agreement with the results of experiment.     

Power-operated relief valve stuck-open accident and recovery scenarios in the Institute of Nuclear Energy Research integral system test facility

Four scaled small break loss-of-coolant accident (LOCA) tests simulating the pressurizer power-operated relief valves (PORVs) stuck-open accidents and the recovery actions in a pressurized water reactor (PWR) were performed at the Institute of Nuclear Energy Research (INER) integral system test (IIST) facility. The objectives of this study are to verify the effectiveness of emergency operating procedure (EOP) and emergency core cooling system (ECCS) on reactor safety. The break sizes were volumetrically scaled down based on one and all three fully-opened PORVs which is equivalent to 0.23% and 0.69% hot leg flow area of the reference plant. The experimental results indicate that in case of high pressure injection (HPI) system failure, the rapid depressurization of the steam generators is proved to be an effective way in the depressurization of the reactor coolant system and the core cooling. In contrast, if only one HPI charging pump operates normally, which injected half (or minimum) flow rate of normal cooling water, the core cooling can be adequately provided without operating the secondary bleeding during PORV stuck-open transient. This paper also presents the scaling methods for the reduced-height, reduced-pressure (RHRP) IIST facility and the test conditions. The validity of the present scaling methodology is confirmed by the results from previous IIST counterpart tests and comparison of the present results with those of the tests performed at the full-height, full-pressure(FHFP) stuck-open tests.