From Chooz to the Ling'ao NPP:The Technology Transfer of Pressurized Water Reactor Technology from France to China

The transfer of pressurized water reactor(PWR)technology from France to China is an important event in the history of Sino-French scientific and technological relations.China has gradually achieved self-reliance in the field of PWR technology through the introduction and subsequent absorption of France's 900 MW reactors.Compared with the process of introducing and absorbing similar technology from the United States by France,China's experience has been more complicated.This circumstance reflects the differences in the nuclear power technology systems between the two countries.France's industrial strength and early acquisition of nuclear power technology laid a solid foundation for mastering PWR technology.On the other hand,although China established a weak foundation through the implementation of the"728 Project,"and tried hard to negotiate with France,the substantive content of the technology transfer was very limited.By way of the policy transition from"unhooking of technology and trade"to"integration of technology and trade,"China ultimately accomplished the absorption and innovation of PWR technology through the Ling'ao NPP.

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.

A Novel Computerized Water Level Control System of PWR Steam Generator of Nuclear Power Plant

This paper presents a novel method to solve old problem of water level control system of pressurized water reactor (PWR) steam generator (SG) of nuclear power plant (NPP) .The level control system of SG plays an important role which effects the reliablity,safty,cost of SG and its mathematical models have been solved.A model of the conventional controller is presented and the existing problems are discussed. A novel rule based realtime control technique is designed with a computerized water level control (CWLC) system for SG of PWR NPP.The performance of this is evaluated for full power reactor operating conditions by applying different transient conditions of SG′s data of Qinshan Nuclear Power Plant (QNPP).

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.

Response characteristics of PWR primary circuit under SBLOCAs considering steam bypass discharging

Small-break superposed station blackout(SBO)accidents are the basic design accidents of nuclear power plants.Under the condition of a small break in the cold leg,SBO further increases the severity of the accident,and the steam bypass discharg-ing system(GCT)in the second circuit can play an important role in guaranteeing core safety.To explore the influence of the GCT on the thermal-hydraulic characteristics of the primary circuit,RELAP5 software was used to establish a numerical model based on a typical pressurized water reactor nuclear power plant.Five different small breaks in the cold-leg super-posed SBO were selected,and the impact of the GCT operation on the transient response characteristics of the primary and secondary circuit systems was analyzed.The results show that the GCT plays an indispensable role in core heat removal during an accident;otherwise,core safety cannot be guaranteed.The GCT was used in conjunction with the primary safety injection system during the placement process.When the break diameter was greater than a certain critical value,the core cooling rate could not be guaranteed to be less than 100 K/h;however,the core remained in a safe state.

The General Design and Technology Innovations of CAP1400

A historical review of in-vessel melt retention （IVR） is given, which is a severe accident mitigation mea- sure extensively applied in Generation III pressurized water reactors （PWRs）. The idea of IVR actually originated from the back-fitting of the Generation 11 reactor Loviisa WER-440 in order to cope with the core-melt risk. It was then employed in the new deigns such as Westinghouse APIO00, the Korean APR1400 as well as Chinese advanced PWR designs HPRIO00 and CAP1400. The most influential phe- nomena on the IVR strategy are in-vessel core melt evolution, the heat fluxes imposed on the vessel by the molten core, and the external cooling of the reactor pressure vessel （RPV）. For in-vessel melt evolution, past focus has only been placed on the melt pool convection in the lower plenum of the RPV; however, through our review and analysis, we believe that other in-vessel phenomena, including core degradation and relocation, debris formation, and coolability and melt pool formation, may all contrib- ute to the final state of the melt pool and its thermal loads on the lower head. By looking into previous research on relevant topics, we aim to identify the missing pieces in the picture. Based on the state of the art, we conclude by proposing future research needs.

Combined Heat and Power Design Considerations for the APR1400

To date, nuclear cogeneration applications have been limited, primarily to district heating in Eastern Europe and heavy water production in Canada. With the current global price for oil and energy, this technology is not economically viable for most countries. However, oil and fossil fuel prices are known to be highly volatile, and the Paris Agreement calls for a reduction in fossil fuel use. Under these circumstances, heat supplied by nuclear power may abruptly return to favor. To prepare for such a scenario, this study will investigate design considerations for a prototypical modem nuclear power plant, the Korean APR1400 （advanced power reactor 1400） （e.g., Shin Kori Units 3, 4, Shin Hanul 1, 2, Barakah Units 1, 2, 3, 4）. Nuclear cogeneration can impact balance of plant system and component design for the condensate, feedwater, extraction steam, and heater drain systems. The APR1400 turbine cycle will be reviewed for a parametric range of pressures and flow rates of the steam exported for cogeneration to identify major design challenges.

Evaluation of Nonlinear Finite Element Module for the Simulation of Fuel Behavior

Because zirconium alloy cladding is the first containment barrier for fission products, its mechanical integrity is the most important concern. In view of the mechanical integrity, stress and strain are the main factors that affect the cladding performance during normal or off-normal operation, which induces force interaction between the pellet and cladding. In the case of a normal operation period, to estimate the cladding stress and strain, various models and codes have been developed using a simplified 1D （one-dimensional） assumption. However, in the case of a slow ramp during start-up and shut-down and a fast transient such as an AOO （anticipated operational occurrence）, it is difficult for a 1D model to simulate the cladding stress and strain accurately due to its modeling limitation. To model a large deformation along the radial and axial directions such as a ＂＇ballooning＂ phenomenon, FE （finite element） modeling, which can simulate a higher degree of freedom, is an indispensable requirement. In this work, an axisymmetric two-dimensional FE module, which will be integrated into the transient fuel performance code, has been developed. To solve the mechanical equilibrium of the pellet-cladding system, taking into account the geometrical and material non-linearities, the FE module employs an ESF （effective-stress-function） algorithm. Verifications of the FE module for the cases of thermal and elastic analyes were performed using the results of ANSYS 13.0.

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.

Influence evaluation of loading conditions during pressurized thermal shock transients based on thermal-hydraulics and structural analyses

The thermal hydraulic （TH） behavior of coo- lant water is a key factor in the structural integrity assessments on reactor pressure vessels （RPVs） of pressurized water reactors （PWRs） under pressurized thermal shock （PTS） events, because the TH behavior may affect the loading conditions in the assessment. From the viewpoint of TH behavior, configuration of plant equipment and their dimensions, and operator action time considerably influence various parameters, such as the temperature and flow rate of coolant water and inner pressure. In this study, to investigate the influence of the operator action time on TH behavior during a PTS event, we developed an analysis model for a typical Japanese PWR plant, including the RPV and the main components of both primary and secondary systems, and performed TH analyses by using a system analysis code called RELAP5. We applied two different operator action times based on the Japanese and the United States （US） rules： Operators may act after 10 min （Japanese rules） and 30 min （the US rules） after the occurrence of PTS events. Based on the results of TH analysis with different operator action times, we also performed structural analyses for evaluating thermal-stress distributions in the RPV during PTS events as loading conditions in the structural integrity assessment. From the analysis results, it was clarified that differences in operator action times significantly affect TH behavior and loading conditions, as the Japanese rule may lead to lower stresses than that under the US rule because an earlier operator action caused lower pressure in the RPV.

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.