Advanced Pressurized Water Reactor Research Articles

The effects of blade stacking lean angle to 1400 MW canned nuclear coolant pump hydraulic performance

Canned nuclear coolant pump is used in the advanced third generation pressurized water reactor and is key equipment and unique rotating equipment in the first loop of a nuclear power plant. High efficiency, low power consumption, low axial force and low pressure pulsation are the significant hydraulic performance demanding to hydraulic components. Impeller is a key component of canned nuclear coolant pump. The blade stacking lean angle is one of the main geometric parameters in impeller design. Therefore, an investigation into the stacking lean angle to the canned nuclear coolant pump can be useful. In order to understand effects of stacking lean angle to canned nuclear coolant pump, this paper numerically investigated three different stacking lean angles. The validity of numerical simulation was confirmed through a comparison between numerical and experimental results. The performance change of canned nuclear coolant pump with the stacking lean angle was acquired. Hydraulic performance curve, axial force curve, static pressure distribution at impeller outlet and static pressure pulsation were performed to investigate the performance changes caused by the stacking lean angle. The results show that the stacking lean angle has important influence on the performance of canned nuclear coolant pump and should be paid attention to in the impeller design. Positive stacking lean angle is good for efficiency and reducing pressure pulsation in impeller. Negative stacking lean angle is good for increasing the head and reducing the axial force, while reducing pressure pulsation in static components.

Code assessment of ATLAS integral effect test simulating main steam-line break accident of an advanced pressurized water reactor

ABSTRACT KAERI has been operating an integral effect test facility, Advanced Thermal–Hydraulic Test Loop for Accident Simulation (ATLAS), for accident simulations of advanced pressurized water reactors. As an integral effect test database for major design basis accidents has been accumulated, a domestic standard problem (DSP) exercise using ATLAS was proposed in order to transfer the database to domestic nuclear industries and to contribute to improving the safety analysis technology for pressurized water reactors (PWRs). As the third DSP exercise, a double-ended guillotine break of the main steam-line at an 8% power without loss of off-site power was decided as a target scenario. Seventeen domestic organizations joined this DSP exercise. They include universities, government, and nuclear industries. The participants of DSP-03 were classified into three groups and each group has focused on the specific subject related to the enhancement of the code assessment; (1) scaling capability of the ATLAS test data by comparing with the code analysis for a prototype, (2) multi-dimensional thermal–hydraulic phenomena anticipated during the steam-line break transient, (3) effect of various models in the one-dimensional safety analysis codes.

Station black out concurrent with PORV failure using a Generic Pressurized Water Reactor simulator

Station Black Out (SBO) is an accident situation that refers to the total loss of offsite power, along with the unavailability of onsite power, which results from the failure of all Diesel Generators (DG). Probabilistic Safety Assessment (PSA) spans a number of methods that include modeling of event-trees and simulation of accidents scenarios, aimed to quantify risk and ensure safety in nuclear power plants. PSA also deals with prediction of future accidents and calculation of failure probabilities that has been done in this study. A SBO accident was simulated using a Generic Pressurized Water Reactor (GPWR) simulator from KEYMASTER™. The accident scenario consists of two stages; the first stage belongs to normal SBO, in second stage SBO accident with additional failure of Pilot Operated Relief Valve (PORV) opens and it stuck open has been considered for the pressurizer. A comparison of the two stages was made by plotting variables on the same graph. The research has been carried out to analyze the hot and cold leg temperatures, Steam Generator (SG) pressure, SG Narrow Range (NR) level, SG water-level-percentage (PCT), Pressurizer pressure, Fuel Temperature, and containment pressure. Simulation results suggest that failure in closing PORV has negligible impact on hot and cold leg temperatures, results in an overall less pressure in SGs, but higher pressure in the pressurizer. Additionally, containment pressure did not exceed the maximum approved pressure of 8.7kg/cm2, but was approaching the Advanced Pressurized Water Reactor’s (APR-1400) design pressure of 4.218kg/cm2. Finally, nuclear fuel temperature exceeded Probabilistic Risk Assessment (PRA) limit of 726.7°C for both scenarios. The research results will provide help in future for better understanding of accidents in APR-1400 reactors.

Numerical analysis of different break direction effect on hydrogen behavior in containment during a hypothetical LOCA

Large amounts of hydrogen could be generated and released into the containment during the course of a severe accident in a Light Water Reactor nuclear power plant. Getting detail hydrogen distribution in containment is important for analysis of hydrogen risk. In this study, a three dimensional CFD analysis model is established to analyze hydrogen behavior during hypothesized severe accident induced by LOCA in advanced passive PWR containment. Hydrogen distribution in containment simulating with algebraic and k-ε turbulence models is studied, the result shows that when k-ε model is adopted, hydrogen stratification is established in whole upper space and the peak concentration in the dome is 8%. Then the effect of source injection direction with vertical and horizontal injection cases is investigated. In vertical injection case, the local high concentration of hydrogen reaches over 25% in source compartment. In horizontal injection case, it occurs in SG-2 and the channel connecting two steam generator compartment i.e. vertical compartment. In upper space, comparing with horizontal source direction, when source direction is vertically towards upper space, hydrogen flows into the dome driven by momentum and stratification can be established more possible. In long term, with hydrogen distribution tending to be uniform, the effect of direction is not obvious. Due to more steam transporting into local compartments (SG-2, vertical and CMT compartments) when direction is horizontal, hydrogen concentration in these compartment is lower. In addition, source direction change can also lead to the change of location where hydrogen combustion risk probably occurs. When direction is horizontal, combustion risk happens in SG-2, vertical or CMT compartments rather than source compartment.

An experimental study of pool entrainment with side exit

Pool entrainment in upper plenum is an important safety related phenomenon in SBLOCA transient for advanced PWR plants like AP1000. Due to the unique geometric characteristics, the application of standard Kataoka model in the upper plenum/hotleg entrainment phenomena requires significant caution. Based on a careful review of recent pool entrainment studies, this research carried out an experimental study of air-water pool entrainment with and without side exit/outlet to better capture the prototypic geometry of upper plenum/hotleg arrangement. Total 150 sets of test data are obtained at various combinations of gas velocities and water levels. The test data covers several entrainment regions from low gas flux region to near saturation region. The results show that the side exit/outlet will reduce the entrainment rate in high gas flux region and near saturation region. The mechanism is analyzed using visualization and CFD simulation. Based on the experiment data, a new correlation of pool entrainment with side exit/outlet in high gas flux region is proposed.

Controller design and optimization of reactor power control system for ASPWR

The objective of this paper is to design a reactor power control system (RPCS) for the advanced small pressurized water reactor (ASPWR) that adopts constant primary loop average coolant temperature and steam generator secondary side steam pressure programs. In this paper, a two-input and two-output linear nuclear steam supply system (NSSS) model is developed first based on the linearization of the nonlinear point reactor core model with one-group delayed neutron and simplified nonlinear once-through steam generator model. Then, both the conventional trial-and-error design method and the improved version of multi-objective, non-dominated sorting genetic algorithm (NSGA-II) are applied for the RPCS controller design. It has been demonstrated by the numerical simulation results that the RPCS control parameters optimized by the NSGA-II provide better control performance than those tuned by conventional design method. Therefore, the NSGA-II based control parameter optimization method is recommended to tune the RPCS controllers.

Effect of in-core instrumentation mounting location on external reactor vessel cooling

The effect of in-core instrumentation (ICI) mounting location on the application of in-vessel corium retention through external reactor vessel cooling (IVR-ERVC), used to mitigate severe accidents in which the nuclear fuel inside the reactor vessel becomes molten, was investigated. Numerical simulations of the subcooled boiling flow within an advanced pressurized-water reactor (PWR) in IVR-ERVC applications were conducted for the cases of top-mounted ICI (TM-ICI) and bottom-mounted ICI (BM-ICI), using the commercially available computational fluid dynamics (CFD) software ANSYS-CFX. Shear stress transport (SST) and the RPI model were used for turbulence closure and subcooled flow boiling, respectively. To validate the numerical method for IVR applications, numerical simulations of ULPU-V experiments were also conducted. The BM-ICI reactor vessel was modeled using a simplified design of an advanced PWR with BM-ICI; the TM-ICI counterpart was modeled by removing the ICI parts from the original geometry. It was found that TM-ICI was superior to BM-ICI for successful application of IVR-ERVC. For the BM-ICI case, the flow field was complicated because of the existence of ICIs and a significant temperature gradient was observed near the ICI nozzles on the lower part of the reactor vessel, where the ICIs were attached. These observations suggest that the existence of ICI below the reactor vessel hinders reactor vessel cooling.

Status of dynamic water film model development in 3-D CFD code GASFLOW-MPI for analysis of passive containment cooling system

The advanced passive PWR AP-1000 and AP-1400 are pressurized water reactors (PWR) with advanced passive safety systems. Most recognize the passive containment cooling system (PCCS) as being one of the most important concepts for advanced passive safety systems. Following a postulated design-base-accident (DBA), an outside cold liquid-water spray is injected on the apex of the spheroidal dome. The resulting water film drains under gravitational body forces over the outside dome surface and down the cylinder surface. This PCCS removes heat from the containment interior to reduce potentially damaging pressure and temperature loads. A numerical model for simulating the falling water film transport in PCCS has been developed in 3-D CFD code GASFLOW-MPI. The water film model is governed by mass, momentum and energy transport equations with essential source terms for inter-phase exchange. The conjugate heat and mass transfer between the gas mixtures inside the containment, water film, and the gas in the annular space outside of the steel containment can be simulated. The heat conduction in the solid steel containment is calculated as well. In order to facilitate model validation in GASFLOW-MPI, we have developed an exact numerical solution (ENS) for the one-dimensional steady-state liquid film flow over an ellipsoidal-domed cylinder. Although the ENS assumes constant containment wall temperature and outside ambient air temperature, it can incorporates any generalized heat transfer correlation between the containment surface and water film as well as any generalized heat and mass (evaporation) transfer correlations between the film and outside ambient air. ENS results for draining liquid water films clearly present the coupling effects of heat and mass transfer. Excellent comparisons between the ENS and CFD calculation results support the reliability and accuracy of the dynamic water film model in GASFLOW-MPI code which provides an efficient way to analyze PCCS in 3-D full scale advanced light water reactors.

Conventional Controller Design for the Reactor Power Control System of the Advanced Small Pressurized Water Reactor

The objective of this paper is to design a reactor power control system for the advanced small pressurized water reactor that adopts a constant average coolant temperature and a secondary-side steam pressure program. Based on the nonlinear core model with the one-group delayed neutron and simplified nonlinear once-through steam generator model, a two-input and two-output linear nuclear steam supply system (NSSS) model is obtained. Three types of control systems are then proposed and designed on a Bode diagram using analytical methods and second-order approximation. The comparison of the control performance and robustness of the three control systems shows that the double feedback control (DFC) with both power feedback and temperature feedback provides the best performance for reactor power and average coolant temperature with parameter uncertainty due to control rod differential worth variation. The simulations based on the high-order nonlinear NSSS model also show good performance of the DFC system.

Parameter Identification with the Random Perturbation Particle Swarm Optimization Method and Sensitivity Analysis of an Advanced Pressurized Water Reactor Nuclear Power Plant Model for Power Systems

The ability to obtain appropriate parameters for an advanced pressurized water reactor (PWR) unit model is of great significance for power system analysis. The attributes of that ability include the following: nonlinear relationships, long transition time, intercoupled parameters and difficult obtainment from practical test, posed complexity and difficult parameter identification. In this paper, a model and a parameter identification method for the PWR primary loop system were investigated. A parameter identification process was proposed, using a particle swarm optimization (PSO) algorithm that is based on random perturbation (RP-PSO). The identification process included model variable initialization based on the differential equations of each sub-module and program setting method, parameter obtainment through sub-module identification in the Matlab/Simulink Software (Math Works Inc., Natick, MA, USA) as well as adaptation analysis for an integrated model. A lot of parameter identification work was carried out, the results of which verified the effectiveness of the method. It was found that the change of some parameters, like the fuel temperature and coolant temperature feedback coefficients, changed the model gain, of which the trajectory sensitivities were not zero. Thus, obtaining their appropriate values had significant effects on the simulation results. The trajectory sensitivities of some parameters in the core neutron dynamic module were interrelated, causing the parameters to be difficult to identify. The model parameter sensitivity could be different, which would be influenced by the model input conditions, reflecting the parameter identifiability difficulty degree for various input conditions.

Numerical Study on Hydrogen Flow Behavior in Two Compartments with Different Connecting Pipes

Hydrogen accumulation in the containment compartments under severe accidents would result in high concentration, which could lead to hydrogen deflagration or detonation. Therefore, getting detailed hydrogen flow and distribution is a key issue to arrange hydrogen removal equipment in the containment compartments. In this study, hydrogen flow behavior in local compartments has been investigated in two horizontal compartments. The analysis model is built by 3-dimensional CFD code in Cartesian coordinates based on the connection structure of the Advanced Pressurized Water Reactor (PWR) compartments. It consists of two cylindrical vessels, representing the Steam Generator compartment (SG) and Core Makeup Tank compartment (CMT). With standard k-ε turbulence model, the effects of the connecting pipe size and location on hydrogen concentration distribution are investigated. Results show that increasing the diameter of connection pipe (IP) which is located at 800 mm from 150 mm to 300 mm facilitates hydrogen flow between compartments. Decreasing the length of IP which is located at 800 mm from 1000 mm to 500 mm can also facilitate hydrogen flow between compartments. Lower IP is in favor of hydrogen mixing with air in non-source compartment. Higher IP is helpful for hydrogen flow to the non-source term compartment from source term compartment.

Feasibility to convert an advanced PWR from UO2 to a mixed U/ThO2 core – Part I: Parametric studies

This work presents the neutronics and thermal hydraulics feasibility to convert the UO2 core of the Westinghouse AP1000 in a (U-Th)O2 core by performing a parametric study varying the type of geometry of the pins in fuel elements, using the heterogeneous seed blanket concept and the homogeneous concept. In the parametric study, all geometry and materials for the burnable poison were kept the same as the AP 1000, and the only variable was the fuel pin material, in which we use several mass proportion of uranium and thorium but keeping the enrichment in 235U, as LEU (20w/o). The neutronics calculations were made by SERPENT code, and to validate the thermal limits we used a homemade code. The optimization criteria were to maximize the 233U, and conversion factor, and minimize the plutonium production. The results obtained showed that the homogeneous concept with three different mass proportion zones, the first containing (32% UO2-68%ThO2); the second with (24% UO2-76% ThO2), and the third with (20% UO2-80% ThO2), using 235U LEU (20w/o), and corresponding with the 3 enrichment zones of the AP 1000 (4.45w/o; 3.40w/o; 2.35w/o), satisfies the optimization criteria as well as attending all thermal constrain. The concept showed advantages compared with the original UO2 core, such a lower power density, and keeping the same 18months of cycle a reduction of B-10 concentration at the soluble poison as well as eliminating in the integral boron poison coated (IFBA).

Screening the design space for optimized plutonium incineration performance in the thorium-based I2S-LWR

This paper presents an optimization of a thorium-plutonium (Th-Pu) fuel cycle by screening various design options for the Integral Inherently Safe Light Water Reactor (I2S-LWR). The I2S-LWR is an advanced 2850MWt integral pressurized water reactor with enhanced safety beyond that of Gen-III+ reactors. The features of this reactor, such as material choice, make it attractive for alternative fuel cycles including the use of thorium. Recently, the feasibility of the Th-Pu cycle was studied and the benefits associated with it were demonstrated. More specifically, the Pu incineration performance was enhanced by adopting multi-batch (i.e. more than 3-batch) schemes and extended burnup (above 100MWd/kg). The optimized design with the most favorable loading pattern was obtained by applying the Simulated Annealing optimization technique. This paper demonstrates further plausible modifications to the Th-Pu cycle design that may enhance its performance considerably. The paper seeks to identify the contributory factors, such as cladding types, plutonium vectors and initial plutonium loadings, with major impact on the incineration performance. The post-irradiation characteristics are also analyzed and suggest that such a cycle may simplify the design and operation of the waste repository.

Performance evaluation of passive containment cooling system of an advanced PWR using coupled RELAP5/GOTHIC simulation

Motivated to investigate the thermal–hydraulic characteristics and performance of a passive containment cooling system (PCS) for a Generation III pressurized water reactor (PWR), a coupled RELAP5/GOTHIC model was developed, which was then employed to simultaneously simulate the transient responses of the PCS and the containment during a large break loss of coolant accident of the reactor. The results show that the PCS is capable of lowering the containment pressure to an acceptable level for a long period (up to 3days). In a separate-effect study, it was found that the height of the PCS loop plays an important role in determining the flow characteristics and heat removal performance of the PCS. Within the range of the considered loop heights, phase change occurs in the riser of the loop after the height exceeds a specific value (between 13m and 15m), below which only single-phase flow takes place. With increasing height of the loop, the heat removal capability increases monotonically at first; however, it is no longer sensitive to the height after two-phase flow appears. Finally, a feed-and-bleed operation for the cooling tank of the PCS was proposed as an enhancement measure of the heat removal capacity, and the simulation results show it further mitigates the accident. Moreover, a simplified analytical model is developed to predict the impact of the feed-and-bleed flowrate on the PCS performance, which can be used in engineering design.

Study on containment pressure response with passive containment cooling system performance under design basis accident

The passive containment cooling system which utilizes natural phenomena to remove the energy released from reactor will prevent steel containment from overpressure during the postulated accident and provide long-term cooling to decrease the pressure and temperature of containment. The peak pressure in containment under the design basis accidents is important to the design and evaluation on the passive containment cooling system. In this paper, a dedicated code PCCSAP-3D for the evaluation on the performance of the passive containment cooling system is used to analyze the thermal-hydraulic transients in the containment of the advanced pressurized water reactor AP1000. Some factors are compared to study their e on the peak pressure of containment under the typical loss-of-coolant and main steam-line break accident. The results show that the amount of released coolant, the free volume and shell thickness of containment are dominant for containment pressure during the postulated accident, in comparison with the initial conditions. And the containment pressure response is more sensitive to internal initial conditions than to ambient conditions. The peak pressure under the postulated accidents is mainly contributed from rapid coolant release and it is not sensitive to the containment cooling performed by the passive containment cooling system.

Simulation of the in-vessel retention device heat-removal capability of AP-1000 during a core meltdown accident

Westinghouse Electric Company’s design for the Advanced Pressurized Water Reactor (AP-1000) incorporates a special in-vessel retention (IVR) device to enhance the capability of removing heat through the outer wall of the reactor vessel during a core meltdown. External Reactor Vessel Cooling (ERVC) is established by flooding the reactor cavity during a severe accident. The amount of heat removed is limited by the critical heat flux (CHF) at the outer surface of the vessel wall. In this paper, the CHF correlations developed specifically for this purpose are reviewed. The RELAP5-3D thermal-hydraulic code is used to simulate the natural convection flow within the water channel of the IVR device. The results of RELAP5-3D simulations are substituted into ULPU, SULTAN, SBLB, and KAIST critical heat flux correlations to determine the safety margin of the AP-1000 IVR design. Among these four correlations, the CHFs predicted by the ULPU and SBLB correlations are insensitive to the thermal-hydraulic conditions of the coolant channel. The CHF predicted by SULTAN is the lowest among the four correlations. The results demonstrate that the predicted CHF increases with the angle. Nevertheless, the ratio of CHF to heat flux is lowest around the upper quarter of the vessel. The heat load to the lower head is based on two bounding melt configurations. Configuration I involves a stratified light metallic layer on top of a molten ceramic pool, and Configuration II represents the condition that an additional heavy metal layer forms below the ceramic pool. The impact of corium pool configuration on the safety performance of IVR is very significant. Under Configuration II, as predicted by the SULTAN and KAIST correlations, the performance of the current AP-1000 IVR design is marginal. The results of the RELAP5-3D IVR simulation show that the total mass flow rate in the coolant channel increases sharply as the widths of the annular channels vary from 0.0762 to 0.762m. The mass-flow rate decreases gradually when the gap width increases beyond 0.762m. The KAIST CHF correlation predicts that CHF is maximal when the gap width is around 0.4572m and decreases gradually when the gap width increases.

Improvement of TASS/SMR code using a simple correlation for subcooled boiling flow prediction

SMART, which was an advanced integral type small modular PWR, was developed by KAERI (Kim et al., 2014). To analyze the thermal hydraulic phenomena including behaviors at the SMART specific components, the TASS/SMR code which can predict a heat transfer for various thermal hydraulic conditions, has been developed. Information of the void distribution in a subcooled boiling flow is important in predicting the inception of a two-phase flow and an onset of the critical heat flux condition. The TASS/SMR code adopts an energy partitioning method and a critical enthalpy correlation determining a point of net vapor generation for subcooling conditions. A range of the subcooling degree investigated is 1.5–50.6K to validate the method for a subcooled boiling flow prediction. The TASS/SMR code predicts well the void distribution along the height for the subcooled boiling flow conditions compared with the experimental data. The predicted location of the onset of void generation is simulated well at most investigated conditions and delayed slightly at the very high subcooling condition.

Comparative economic analysis of the Integral Molten Salt Reactor and an advanced PWR using the G4-ECONS methodology

The assessment of economic viability of a new reactor concept is crucial particularly during the early stages of its concept development. The G4-ECONS methodology provides a standardized top-down estimate of electricity cost and parametric sensitivities, not specifically targeted toward an accurate prediction of the final cost when deployed, but rather seeking an approximation of cost variations relative to other systems. This study presents an analysis of the Integral Molten Salt Reactor (IMSR) concept in comparison with a consistent analysis of an advanced PWR reactor (represented by AP1000). Estimation of levelized unit electricity costs, as well as sensitivity analyses to the discount rate and uranium or SWU prices, are presented using this methodology.

Effects of staggered blades on the hydraulic characteristics of a 1400-MW canned nuclear coolant pump

A canned nuclear coolant pump is used in an advanced third-generation pressurized water reactor. Impeller is a key component of a canned nuclear coolant pump. Usually, the blade is installed between the hub and the shroud as an entire part. The blade is divided into two parts and is staggered in the circumferential direction is an approach of blade design. To understand the effects of staggered blades on a canned nuclear coolant pump, this article numerically investigated different types of staggering. The validity of the numerical simulation was confirmed by comparing the numerical and experimental results. The performance change of a canned nuclear coolant pump with staggered blades was acquired. Hydraulic performance curves, axial force curves, static pressure distributions at the impeller outlet, and static pressure pulsations were performed to investigate the performance changes caused by the staggered blades. The results show that the staggered blade has an important influence on the performance of canned nuclear coolant pumps. A staggered blade does not improve hydraulic performance but does improve the axial force and pressure pulsation. Specifically, the staggered blades can significantly reduce the pressure pulsation amplitude on the impeller pass frequency.

Enhancing plutonium incineration in the thorium-based I2S-LWR design with loading pattern optimization

This paper presents an optimization of a thorium–plutonium fuel cycle, through a multi-batch reloading scheme, developed for the Integral Inherently Safe Light Water Reactor (I2S-LWR). The I2S-LWR is an advanced 2850MWt integral pressurized water reactor with enhanced safety beyond that of Gen-III+ reactors. Its baseline fuel and cladding materials are U3Si2 and advanced FeCrAl steel, respectively. The advanced steel cladding can withstand longer exposure periods with significantly lower degradation rates compared to traditional Zr-based alloys. In principle, increasing the number of batches allows higher discharge burnups and thus deeper Pu and transuranic elements incineration to be achieved. Therefore, various refuelling strategies were considered in this study, namely 3-, 5- and 7.56-batch schemes. The Simulated Annealing optimization technique was applied for the different batch schemes to obtain the most favourable loading pattern with respect to cycle length performance. The results confirm that increasing the number of batches allows the discharge burnup to be increased by 20% (above 100MWd/kg), which improves the Pu incineration performance. In addition, the increased number of batches improves the reactivity coefficients without violating the power peaking factors limits.