The Simulation Test to Start up the PIUS-Type Reactor from Isothermal Fluid Condition

A feedback control system of a primary pump is proposed for a PIUS-type reactor based on the temperature distribution in the lower density lock. This control system maintains the fluid temperature at the axial center of the lower density lock at the average of the fluid temperatures below and above the lower density lock in order to prevent the poison water from penetrating into the core during normal operation. This control system was examined in a series of startup tests under different conditions and the test results demonstrated the effectiveness of the control system to startup the reactor from a small initial temperature difference between primary system and poison water system.

The startup of a PIUS-type reactor from an isothermal condition, preventing the inflow of the poison water into the core, is the most difficult transient among normal operations. We installed a preheater system at the exit of the primary pump for the startup test from isothermal condition with the purpose of making a initial temperature difference between primary system and poison water pool. It was confirmed from the results of the tests that the preheater and the present pump feedback control system is quite effective to start up the reactor from an isothermal fluid condition.     

Experimental determination of the boron concentration distribution in the primary circuit of a PWR after a postulated cold leg small break loss-of-coolant-accident with cold leg safety injection

It is known that under-borated coolant can accumulate in the loops and that it can be transported towards the reactor core during a loss-of-coolant-accident. Therefore, the mixing of weakly borated water inside the reactor pressure vessel was investigated using the ROCOM test facility. Wire-mesh sensors based on electrical conductivity measurement are used to measure in detail the spreading of a tracer solution in the facility. The mixing in the downcomer was observed with a measuring grid of 64 azimuthal and 32 vertical positions. The resulting distribution of the boron concentration at the core inlet was measured with a sensor integrated into the lower core support plate providing one measurement position at the entry into each fuel assembly.

The boundary conditions for this mixing experiment are taken from an experiment at the thermal hydraulic test facility PKL operated by AREVA Germany. The slugs, which have a lower density, accumulate in the upper part of the downcomer after entering the vessel. The ECC water injected into the reactor pressure vessel falls almost straight down through this weakly borated water layer and accelerates as it drops over the height of the downcomer. On the outer sides of the ECC streak, lower borated coolant admixes and flows together with the ECC water downwards. This has been found to be the only mechanism of transporting the lower borated water into the lower plenum. In the core inlet plane, a reduced boron concentration is detected only in the outer reaches of the core inlet. The minimum instantaneous boron concentration that was measured at a single fuel element inlet was found to be 66.3% of the initial 2500 ppm.     

Dynamic Response of Hot/Cold Liquid Interfaces to Pump Speed Perturbations in a Thermal-Hydraulic Loop Simulating a PIUS-type Reactor

The operation of a PIUS-type reactor requires controlling the reactor pump speed to keep stationary the hot/cold liquid interfaces between the reactor coolant and cold borated water. The dynamic response of the interface location to pump speed perturbations is analyzed for an experimental loop simulating a PIUS-type reactor. The transfer function between the pump speed and the interface location is obtained by perturbing and Laplace-transforming the one-dimensional fluid momentum equations. The analytical results agree well with experimental data taken from the same facility. It is shown that the magnitude of the phase lag in the response of the interface location, which needs to be considered in designing a pump speed controller, primarily depends on the fluid inertia in the loop, the density lock flow area, and the density difference between the simulated reactor coolant and borated water.     

PIUS - status and perspectives

Today's environmental concerns show that nuclear energy is an important option for meeting future increases in global energy demand. Significant nuclear expansion will probably require new reactor designs in which safety is ensured by simple, convincing means. PIUS represents such a reactor design. It is a re-configured 600 MWe PWR, in which the primary safety goal, protection of the reactor core integrity, is entrusted to built-in, self-protective, passive features, without reliance on any monitoring, detection or actuation system, nor operator action. Its basic design features a core that is openly connected, in a natural circulation loop, to a large pool of borated water. The pool is enclosed in a prestressed concrete pressure vessel provided with redundant leakage barriers. The reactor coolant pumps are operated in such a way that there is hydraulic balance in the openings between the primary coolant loop and the pool. Thereby, the hot, low boron content primary loop water is kept separated from the pool water, in spite of the always open natural circulation loop. In severe transients this balance is disturbed, and pool water ingress occurs, shutting down the reactor, or restricting the power to a safe level. The decay heat is transferred to the pool by the natural circulation loop, and a passive pool cooling system, utilizing natural circulation and natural draft cooling towers, prevents boiling of the pool water, even in a station blackout situation. Transient analyses have shown that this passive long-term RHR function will be available in all accident situations, even after double-ended cold leg breaks. Such breaks result in a temporary pressurization of the reactor containment, but the releases of radioactivity will be extremely small and the doses at the fence boundary very low. Cost estimates indicate that PIUS will be quite competitive, and evaluation studies are now under way in several countries.     

Numerical analysis and visualization experiment on behavior of borated water during MSLB with RCP running mode in an advanced reactor

The core bypass phenomenon of borated water injected through direct vessel injection (DVI) nozzles in APR1400 (Advanced Power Reactor 1400MWe) during main steam line break (MSLB) accidents with a reactor coolant pump (RCP) running mode has been simulated using a two-channel and one-dimensional system analysis model code (MARS), and a three-dimensional computational fluid dynamics (CFD) code (FLUENT). A visualization experiment has also been performed using a scaled-down model of the APR1400. The MARS analysis has predicted a serious core bypass phenomenon of borated water, while the CFD analysis has shown results opposite to the MARS results. The CFD analysis has shown that the flow pattern in the downcomer is fully three-dimensional and that vortex flow structures are formed near the cold legs so that the borated water might pass without difficulty into the high flow region of the cold legs and flow well into the lower downcomer. The visualization experiment has shown that the borated water flows well to the lower plenum, as in the CFD analysis. Both the CFD analysis and visualization experiment have proved that a serious core bypass phenomenon of borated water might not happen in the APR1400. These results are quite different from those predicted by MARS.     

Boron dilution transients during natural circulation flow in PWR—Experiments and CFD simulations

Coolant mixing in the cold leg, downcomer and the lower plenum of pressurized water reactors is an important phenomenon mitigating the reactivity insertion into the core. Therefore, mixing of the de-borated slugs with the ambient coolant in the reactor pressure vessel was investigated at the four loops 1:5 scaled Rossendorf coolant mixing model (ROCOM) mixing test facility. In particular thermal hydraulics analyses have shown, that weakly borated condensate can accumulate in the pump loop seal of those loops, which do not receive a safety injection. After refilling of the primary circuit, natural circulation in the stagnant loops can re-establish simultaneously and the de-borated slugs are shifted towards the reactor pressure vessel (RPV).In the ROCOM experiments, the length of the flow ramp and the initial density difference between the slugs and the ambient coolant was varied. From the test matrix experiments with 0 resp. 2% density difference between the de-borated slugs and the ambient coolant were used to validate the CFD software ANSYS CFX. To model the effects of turbulence on the mean flow a higher order Reynolds stress turbulence model was employed and a mesh consisting of 6.4 million hybrid elements was utilized. Only the experiments and CFD calculations with modeled density differences show stratification in the downcomer. Depending on the degree of density differences the less dense slugs flow around the core barrel at the top of the downcomer. At the opposite side, the lower borated coolant is entrained by the colder safety injection water and transported to the core. The validation proves that ANSYS CFX is able to simulate appropriately the flow field and mixing effects of coolant with different densities.     

Atmospheric-Pressure Small-Scale Thermal-Hydraulic Experiment of a PIUS-Type Reactor

An experimental small-scale low-pressure setup of a PIUS (Process Inherent Ultimate Safety)-type reactor was used for the examination of the stability during normal operation such as startup and load following operation and of the safety during accidents such as loss-of-feed- water and pump runaway. Automatic feedback pump control system based on differential pressure at lower honeycomb density lock was quite effective to maintain the stratified interface between primary and pool water in the honeycomb density lock during normal operation. The process inherent ultimate safety characteristics of the PlUS-type reactor was confirmed with pump-trip scram at the pump speed limit for the various simulated accidents such as a loss-of- feedwater and pump runaway.     

5MW低温供热反应堆微沸腾运行启动可行性研究

文章用RETRAN-02程序,对清华大学核能技术研究所设计和建造的5MW低温供热堆的微沸运行启动方式进行了较为系统的研究;分析了控制反应性引入速率、主回路蒸汽冷凝量大小及主回路对外总传热量的大小对启动稳定性的影响。结果表明,在一种新颖的启动方式下,只要对反应堆的某些特定参数作适当的实时控制,反应堆就能从单相向两相微沸运行方式稳定过渡。     

Thermal and Stability Considerations of Super LWR during Sliding Pressure Startup

The feasibility of the sliding pressure startup of a high-temperature supercritical-pressure light water reactor (super LWR, SCLWR-H) is assessed from both thermal and stability considerations. In the sliding pressure startup, nuclear heating starts at subcritical pressure and the reactor is pressurized to supercritical pressure at a low power and high enough flow rate. The reactor power and flow rate are then raised gradually to the rated normal values at constant supercritical operating pressure. During startup, the maximum cladding surface temperature must not exceed 620°C. For two-phase flow at subcritical pressures, the homogeneous equilibrium model is used. The thermal-hydraulic and coupled neutronic thermal-hydraulic stabilities during pressurization and power-raising are investigated by a frequency-domain linear analysis for both supercritical-pressure and subcritical-pressure operating conditions. The same stability criteria as those of BWRs are used. From the analysis results, a sliding pressure startup procedure is proposed for super LWR. The thermal criteria are satisfied by keeping the core power between the maximum allowable limit and minimum limit required for turbine startup and operation. The thermal-hydraulic stability and coupled neutronic thermal-hydraulic stability can be maintained by applying an orifice pressure drop coefficient at the inlet of fuel assembly and by controlling the power and flow rate during startup.     

Experimental investigation of natural circulation instability in a BWR-type small modular reactor

The Purdue NMR (Novel Modular Reactor) represents a BWR-type small modular reactor with a significantly reduced reactor pressure vessel (RPV). Specifically, the NMR is one third the height and area of a conventional BWR RPV with an electrical output of 50 MWe. Experiments are performed in a well-scaled test facility to investigate the thermal hydraulic flow instabilities during the startup transients for the NMR. The scaling analysis for the design of natural circulation test facility uses a three-level scaling methodology. Scaling criteria are derived from non-dimensional field and constitutive equations. Important thermal hydraulic parameters, e.g. system pressure, inlet coolant flow velocity and local void fraction, are analyzed for slow and fast normal startup transients. Flashing instability and density wave oscillation are the main flow instabilities observed when system pressure is below 0.5 MPa. And the flashing instability and density wave oscillation show different type of oscillations in void fraction profile. Finally, the pressurized startup procedure is recommended and tested in current research to effectively eliminate the flow instabilities during the NMR startup transients.     

Startup transient simulation for natural circulation boiling water reactors in PUMA facility

In view of the importance of instabilities that may occur at low-pressure and -flow conditions during the startup of natural circulation boiling water reactors, startup simulation experiments were performed in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility. The simulations used pressure scaling and followed the startup procedure of a typical natural circulation boiling water reactor. Two simulation experiments were performed for the reactor dome pressures ranging from 55 kPa to 1 MPa, where the instabilities may occur. The experimental results show the signature of condensation-induced oscillations during the single-phase-to-two-phase natural circulation transition. The results also suggest that a rational startup procedure is needed to overcome the startup instabilities in natural circulation boiling water reactor designs.     

DynMo-TE: Dynamic simulation model of space reactor power system with thermoelectric converters

A space reactor power system (SRPS) has been developed for avoidance of single point failures in reactor cooling and energy conversion. The sectored compact reactor (SCoRe) in this system is lithium-cooled and the reactor core is divided into six equal sectors with liquid metal heat pipes dividers. These reactor sectors are neutronically, but not thermal-hydraulically, coupled. Each sector has its own primary and secondary circulating lithium loops, which are thermally coupled both in a SiGe thermoelectric (TE) power conversion assembly (PCA) and a thermoelectric conversion assembly (TAC) that powers the electromagnetic pumps in the primary and secondary loops. Each secondary loop also has a separate, segmented radiator panel that is optimized for low specific mass and low liquid lithium inventory. The primary loops transport the thermal power generated in the reactor to six PCAs that nominally supply a total of 111.5 kW_e to the load at 450 V DC. Each of the 12 primary and secondary loops has its own bellows-type accumulator that is designed to regulate the lithium pressures in the loops. A dynamic simulation model of this thermoelectric SRPS (DynMo-TE) has been developed and used to investigate the transient operation of the system during a startup from a fully-thawed condition at 600 K, to nominal steady-state operation at which the lithium coolant exits the reactor at only 1179 K. Also investigated is the load-following characteristic of the SCoRe-TE SRPS, following a change in the electrical load demand.     

Petri-net based modelling approach for ALFRED reactor operation and control system design

In this work, the Petri-net modelling approach applied to the control system design of the Advanced Lead Fast Reactor European Demonstrator (ALFRED) is presented, paying particular attention to the startup procedure. The reactor startup is the operational transient in which all the systems of the plant are brought from the cold shutdown condition to the full power mode, close to load-frequency control. In this phase, the several control actions to be taken need to be properly coordinated. To this end, the operational sequence which constitutes the reactor startup procedure has been described by adopting the Petri-nets approach, i.e., a useful formalism for the modelling and the analysis of Discrete Event Systems. Thanks to this quantitative representation, it is possible to easily derive the corresponding control scheme. In addition, the Petri-nets approach has been also exploited for the two-level control system architecture, namely a master system coordinates the operation of the plant by sending suitable signals to the slave system, in which feedback controllers are implemented. As a major outcome of this work, the procedure for the reactor startup and the transition to the full power mode has been simulated in order to assess the control system performance.     

超临界水堆全系统启动特性研究

为研究超临界水堆（SCWR）全系统启动特性,以SCTRAN程序为计算工具,基于中国超临界水堆（CSR1000）堆芯参数、高性能轻水反应堆（HPLWR）热力循环回路和日本SCWR再循环启动回路,建立了SCWR完整再循环启动系统模型。通过与HPLWR热力循环回路的稳态参数对比,验证了完整回路模型的正确性。分析在控制系统控制下的CSR1000再循环启动过程,得到了启动过程中堆芯、汽鼓、汽轮机、各级抽汽、再热器、各级回热器的瞬态响应曲线。计算结果表明,启动序列和启动过程各热工参数的变化符合预期,系统稳定启动;堆芯始终处于单相状态;汽轮机入口为超临界蒸汽;经过高压和低压回热器后堆芯入口温度能够达到280℃;高压缸入口压力维持恒定;在启动的过程中最大燃料包壳表面温度低于限值温度650℃,整个启动过程安全可靠。      

CPR1000核电机组堆芯水位监测系统及其调试

CPR1000核电机组反应堆堆芯水位监测系统是反应堆发生LOCA事故后监测堆芯淹没状态的重要系统,由其测量的水位直接用于反应堆事故规程的导向。本文对该系统的测量原理、系统构成进行了详细的介绍,通过对CPR1000核电机组首台机组的调试,实现了该系统的首次自主化调试的目标。     

Qualification of the CFD code Trio_U for full scale reactor applications

The article presents a procedure to qualify the Trio_U code for the prediction of the boron concentration at the core inlet of a French 900 MWe pressurized water reactor under accidental conditions (inherent dilution problem).¹ The objective of this procedure is to ensure that the validation calculations are performed with the same modelling hypotheses as the full scale reactor analysis, for which usually no experimental data are available. A density driven ROCOM experiment as well as an UPTF Tram-C3 experiment have been used for the qualification of the Trio_U code. Both experiments present similar thermal hydraulic conditions as the reactor case. The predicted boron concentration at the core inlet of the reactor shows that the potential return to criticality might not be excluded in the case of a small break LOCA. Further neutronic calculations are necessary to confirm this result.     

Experimental studies and CFD calculations for buoyancy driven mixing phenomena

In nuclear reactor safety the mixing of borated and deborated water is a critical issue that needs investigation, assessment and prediction. Such mixing is buoyancy driven and numerical codes must correctly model momentum transfer between fluids of different density. To assess and develop CFD models for buoyancy driven mixing we set up a simple vertical mixing test facility (VeMix) and equipped it with a newly developed planar electrical imaging sensor. This imaging sensor acquires conductivity images of the liquid at the rear channel wall with a speed of 2,500 frames/s. By adding NaCl tracer to the denser fluid we were able to visualize the mixing process in high spatial and temporal detail. Furthermore, an image processing algorithm based on the optical flow concept was implemented and tested which allows the measurement of flow pattern velocities. Selected experiments at different Richardson numbers were run with two components of different density (pure water and glucose-water mixture) simulating borated and deborated water in a light water reactor scenario. These experiments were compared to CFD calculations using standard turbulence models. Good agreement between experimental data and CFD simulations was found.     

不对称工况对池式钠冷快堆堆芯入口温度的影响

对于池式钠冷快堆,堆芯入口温度是重要的热工参数之一,电厂设计过程中堆芯入口温度的确定受多种因素制约,其中包括不同电厂工况的影响。不对称工况是一种典型的电厂工况,本文以600 MW两环路设计的池式钠冷快堆为研究对象,采用钠冷快堆系统分析程序分析不对称工况对堆芯入口温度的影响。研究结果表明,在所分析的不对称工况下,冷池温度会出现明显的不对称现象,且其中1个环路的冷池温度明显上升。通过分析可知,作为电厂的重要热工参数,在不对称工况下,堆芯入口温度变化的影响主要体现在对冷池内设备的影响上,对电厂整体功能和性能有所影响但不构成该工况下影响电厂功能和性能的关键因素。     

Thermal power calibrations of the IPR-R1 TRIGA reactor by the calorimetric and the heat balance methods

Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculated as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor.     

Thermal power calibrations of the IPR-R1 TRIGA reactor by the calorimetric and the heat balance methods

Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculated as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor.