Development and Application of Maintenance Template in Pressurized Water Reactor Nuclear Power Plant

Good practices of maintenance optimization in nuclear power field need to be effectively consolidated and inherited,and maintenance optimization can provide technology support to create a long-term reliable and economic operation for nuclear power plants( NPPs) especially for a large number of nuclear powers under construction. Based on the development and application of maintenance template in developed countries,and combining with reliability-centered maintenance( RCM) analysis results and maintenance experience data over the past ten years in domestic NPPs, the development process of maintenance template was presented for Chinese pressurized water reactor( PWR) NPP,and the application of maintenance template to maintenance program development and maintenance optimization combined with cases were demonstrated. A shortcut was provided for improving the efficiency of maintenance optimization in domestic PWR NPP,and help to realize a safe,reliable,and economic operation for domestic NPPs.

Non-integer Order Control Scheme for Pressurized Water Reactor Core Power

Tracking load changes in a pressurized water reactor(PWR)with the help of an efficient core power control scheme in a nuclear power station is very important.The reason is that it is challenging to maintain a stable core power according to the reference value within an acceptable tolerance for the safety of PWR.To overcome the uncertainties,a non-integer-based fractional order control method is demonstrated to control the core power of PWR.The available dynamic model of the reactor core is used in this analysis.Core power is controlled using a modified state feedback approach with a non-integer integral scheme through two different approximations,CRONE(Commande Robuste d’Ordre Non Entier,meaning Non-integer orderRobust Control)and FOMCON(non-integer order modeling and control).Simulation results are produced using MATLAB■program.Both non-integer results are compared with an integer order PI(Proportional Integral)algorithm to justify the effectiveness of the proposed scheme.Sate-spacemodel Core power control Non-integer control Pressurized water reactor PI controller CRONE FOMCON.

Development of CONTHAC-3D and hydrogen distribution analysis of HPR1000

An in-house code,CONTHAC-3D,was developed to calculate and analyze thermal-hydraulic phenomena in containments during severe accidents.CONTHAC-3D is a three-dimensional computational fluid dynamics code that can be applied to predict gas flow,diffusion,and steam condensation in a containment during a severe hypothetical accident,as well as to obtain an estimate of the local hydrogen concentration in various zones of the containment.CONTHAC-3D was developed using multiple models to simulate the features of the proprietary systems and equipment of HPR1000 and ACP100,such as the passive cooling system,passive autocatalytic recombiners and the passive air cooling system.To validate CONTHAC-3D,a GX6 test was performed at the Battelle Model Containment facility.The hydrogen concentration and temperature monitored by the GX6 test are accurately predicted by CONTHAC-3D.Subsequently,the hydrogen distribution in the HPR1000 containment during a severe accident was studied.The results show that the hydrogen removal rates calculated using CONTHAC-3D for different types of PARs agree well with the theoretical values,with an error of less than 1%.As the accident progresses,the hydrogen concentration in the lower compartment becomes higher than that in the large space,which implies that the lower compartment has a higher hydrogen risk than the dome and large space at a later stage of the accident.The amount of hydrogen removed by the PARs placed on the floor of the compartment is small;therefore,raising the installation height of these recombiners appropriately is recommended.However,we do not recommend installing all autocatalytic recombiners at high positions.The study findings in regard to the hydrogen distribution in the HPR1000 containment indicate that CONTHAC-3D can be applied to the study of hydrogen risk containment.

Transient Behaviors of Thermo-Hydraulic and Thermal Stratification in the Pressurizer Surgeline for the Nuclear Power Plant

In pressurized water reactor(PWR)system,the surgeline plays an important role in bonding the pressurizer and the primary circle.Some considerable problems,including the thermo-hydraulics,the thermal stratification and the accompanying thermal stress under transient conditions,pose risks to the surgeline integrity.Herein,a fully-coupled flow-heat-thermo-elasticity model was developed to investigate the transient behavior of thermo-hydraulic parameters and the thermal stratification phenomenon in PWR.To evaluate the nonuniformity of the stratified flow,a stratification degree indicatorζis introduced.It is found that during the outsurge flow,the increase of temperature variation will enlarge the temperature gradient on the wall,corresponding to a more serious deformation.In the cases of positive temperature variation,the higher temperature variation causes higher stratification degreeζ,and vice versa.The mass flow rate m and the stratification degree are in negative correlation.The local deformation can reach 1.802 cm under a 50 K temperature variation,while its location varies from case to case.More attention should be paid to the regulation between the highest deformation location and the surgeline thermo-hydraulic parameters.

CFD simulation of thermal hydraulic characteristics in a typical upper plenum of RPV

A comparative computational fluid dynamics(CFD)study was conducted on the three different types of pressurized water reactor(PWR)upper plenum,named TYPE 1(support columns(SCs)and control rod guide tubes(CRGTs)with two large windows),TYPE 2(SCs and CRGTs without windows),and TYPE 3(two parallel perforated barrel shells and CRGTs).First,three types of upper plenum geometry information were collected,simplified,and adopted into the BORA facility,which is a 1/5 scale system of the four-loop PWR reactor.Then,the geometry,including the upper half core,upper plenum region,and hot legs,was built using the Salome platform.After that,an unsteady calculation to simulate the reactor balance operation at hot full power scenario was performed.Finally,the differences of flowrate distribution at the core outlet and temperature distribution and transverse velocity inside the hot legs with different upper plenum internals were compared.The results suggest that TYPE 1 upper plenum internals cause the largest flowrate difference at the core outlet while TYPE 3 leads to the most even distributed flowrate.The distribution and evolution pattern of the tangential velocity inside hot legs is highly dependent on the upper plenum internals.Two counter-rotating swirls exist inside the TYPE 1 hot leg and only one swirl revolving around the hog leg axis exist inside the TYPE 2 hot leg.For TYPE 3,two swirls like that of TYPE 1 rotating around the hot leg axis significantly increase the temperature homogenization speed.This research provides meaningful guidelines for the future optimization and design of advanced PWR upper plenum internal structures.