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.

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.

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.

Effect of weld microstructure on brittle fracture initiation in the thermallyaged boiling water reactor pressure vessel head weld metal

Effects of the weld microstructure and inclusions on brittle fracture initiation are investigated in a thermally aged ferritic high-nickel weld of a reactor pressure vessel head from a decommissioned nuclear power plant.As-welded and reheated regions mainly consist of acicular and polygonal ferrite,respectively.Fractographic examination of Charpy V-notch impact toughness specimens reveals large inclusions(0.5-2.5μm)at the brittle fracture primary initiation sites.High impact energies were measured for the specimens in which brittle fracture was initiated from a small inclusion or an inclusion away from the V-notch.The density,geometry,and chemical composition of the primary initiation inclusions were investigated.A brittle fracture crack initiates as a microcrack either within the multiphase oxide inclusions or from the debonded interfaces between the uncracked inclusions and weld metal matrix.Primary fracture sites can be determined in all the specimens tested in the lower part of the transition curve at and below the 41-J reference impact toughness energy but not above the mentioned value because of the changes in the fracture mechanism and resulting changes in the fracture appearance.

Effect of Pre-Deformation Enhanced Thermal Aging on Precipitation and Microhardness of a Reactor Pressure Vessel Steel

Microstructure evolution in neutron irradiated Reactor Pressure Vessel （RPV） steels was experimentally simulated through an improved degradation procedure in this study. The degradation procedure includes austenitizing at 1 150℃ and water quench, deformation 10% and 30% respectively, and then thermal aging at 500℃ for different period of time. The microstructure of the specimens was analyzed in details using transmission electron microscopy （TEM）. The micro-hardness test results showed that all the hardness curves of undeformed, 10% pre-deformed and 30% pre-deformed specimens have two micro-hardness peaks with the first peak value corresponding to different thermal aging time of 1 hour, 5 hours and 10 hours, respectively. It was revealed that the hardness curves were influenced by the precipitation of Cu-rich precipitates （CRPs） and carbides, deposition of martensite and work hardening.

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.

A Peridynamic Approach for the Evaluation of Metal Ablation under High Temperature

In this paper,the evaluations of metal ablation processes under high temperature,i.e.,the Al plate ablated by a laser and a heat carrier and the reactor pressure vessel ablated by a core melt,are studied by a novel peridynamic method.Above all,the peridynamic formulation for the heat conduction problem is obtained by Taylor’s expansion technique.Then,a simple and efficient moving boundary model in the peridynamic framework is proposed to handle the variable geometries,in which the ablated states of material points are described by an additional scalar field.Next,due to the automatic non-interpenetration properties of peridynamic method,a contact algorithm is established to determine the contact relationship between the ablated system and the additional heat carrier.In addition,the corresponding computational procedure is listed in detail.Finally,several numerical examples are carried out and the results verify the validity and accuracy of the present method.

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.

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).

OKMC simulation of vacancy-enhanced Cu solute segregation affected by temperature/irradiation in the Fe–Cu system

The effects of annealing and irradiation on the evolution of Cu clusters in a-Fe are investigated using object kinetic Monte Carlo simulations.In our model,vacancies act as carriers for chemical species via thermally activated diffusion jumps,thus playing an important role in solute diffusion.At the end of the Cu cluster evolution,the simulations of the average radius and number density of the clusters are consistent with the experimental data,which indicates that the proposed simulation model is applicable and effective.For the simulation of the annealing process,it is found that the evolution of the cluster size roughly follows the 1/2 time power law with the increase in radius during the growth phase and the 1/3 time power law during the coarsening phase.In addition,the main difference between neutron and ion irradiation is the growth and evolution process of the copper-vacancy clusters.The aggregation of vacancy clusters under ion irradiation suppresses the migration and coarsening of the clusters,which ultimately leads to a smaller average radius of the copper clusters.Our proposed simulation model can supplement experimental analyses and provide a detailed evolution mechanism of vacancy-enhanced precipitation,thereby providing a foundation for other elemental precipitation research.

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.

Crystal Structure Evolution of the Cu-rich Nano Precipitates from bcc to 9R in Reactor Pressure Vessel Model Steel

The crystal structure evolution of the Cu-rich nano precipitates from bcc to 9R during thermal aging was studied in nuclear reactor pressure vessel （RPV） model steels. The specimens, contained higher copper and nickel contents than commercially available one, were heated at 890 ~C for 0.5 h and then water quenched followed by tempering at 0（50 ~C for I0 h and aging at 400 ~C for 1000 h. It was observed that bcc and 9R orthogonal structure, as well as 9R orthogonal and 9R monoclinic structure, coexist in a single Cu-rich nano precipitate. Further analyses pointed out that Cu-rich nano precipitates of bcc structure were not stable, it may preferentially transform to 9R orthogonal structure and then to 9R monoclinic structure. This results showed that the crystal structure evolution of the Cu-rich nano precipitates was complex.

Aerobic granules cultivated and operated in continuous-flow bioreactor under particle-size selective pressure

A novel method based on the selective pressure of particle size （particle-size cultivation method, PSCM） was developed for the cultivation and operation of aerobic granular sludge in a continuous-flow reactor, and compared with the conventional method based on the selective pressure of settling velocity （settling-velocity cultivation method, SVCM）. Results indicated that aerobic granules could be cultivated in continuous operation mode by this developed method within 14 days. Although in the granulation process, under particle-size selective pressure, mixed liquor suspended solids （MLSS） in the reactor fluctuated greatly and filamentous bacteria dominated the sludge system during the initial operation days, no obvious difference in profile was found between the aerobic granules cultivated by PSCM and SVCM. Moreover, aerobic granules cultivated by PSCM presented larger diameter, lower water content and higher specific rates of nitrification, denitrifieation and phosphorus removal, but lower settling velocity. Under long term operation of more than 30 days, aerobic granules in the continuous-flow reactor could remain stable and obtain good chemical oxygen demand （COD）, NH4^＋-N, total nitrogen （TN） and total phosphorus （TP） removal. The results indicate that PSCM was dependent on the cultivation and maintenance of the stability of aerobic granules in continuous-flow bioreactors.

Static recrystallization behavior of SA508Gr.4N reactor pressure vessel steel during hot compressive deformation

The two-pass isothermal hot compression method was used to study the effect of different thermal deformation conditions on static recrystallization behavior in Ni-Cr-Mo series SA508Gr.4N low alloy steel with interval holding time ranging from 1 to 300 s,temperature ranging from 950 to 1150℃,strain rate ranging from 0.01 to 1 s^(-1),true strains ranging from 0.1 to 0.2,and initial austenite grain size ranging from 175 to 552μm.It can be concluded that the static recrystallization volume fraction gradually increases with the increase in the deformation temperature,strain rate,strain and pass interval,and the decrease in the initial grain size,which is mainly due to the increase in the deformation energy storage and dislocations.Moreover,strain-induced grain boundary migration is the nucleation mechanism for static recrystallization of SA508Gr.4N low alloy steel.Based on the stress-strain curve,the predicted value obtained from the established static recrystallization kinetics model is in good consistence with the experimental value,and the static recrystallization thermal activation energy of SA508Gr.4N steel was calculated as 264,225.99 J/mol.

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.

Mechanical and fatigue properties of SA508-Ⅳ steel used for nuclear reactor pressure vessels

The mechanical and fatigue properties of SA508-Ⅳ steel with martensite and granular bainite, respectively, were studied. The mechanical tests results showed that the ultimate tensile strength and impact toughness of the specimen with martensite were 830 MPa and 158 J, respectively, and those of the specimen with granular bainite were 811 MPa and 115 J, respectively. The former had higher tensile strength and impact toughness than the latter. The impact tests results showed that the former belonged to typical dimple fracture, while the latter belonged to brittle fracture. The fatigue tests results showed that the fatigue life of the specimen with martensite was 2717 cycles, and that of the specimen with granular bainite was 1545 cycles under the strain amplitude of ± 0.45%. The specimen with martensite had fewer crack initiation points, narrower fatigue striations separation, and larger volume fraction of high-angle grain boundaries than the latter. The fewer crack initiation points meant fewer fatigue cracks, the narrower fatigue striations separation meant slower crack propagation rate, and the larger volume fraction of high-angle grain boundaries could more effectively hinder fatigue crack propagation. Based on these facts, the fatigue life of the specimen with martensite was higher than that of the specimen with granular bainite.

A modified theta projection model for creep behavior of RPV steel 16MND5

During a hypothetical severe accident of light water reactors,the reactor pressure vessel(RPV) could fail due to its creep under the influence of high-temperature corium.Hence,modelling of creep behavior of the RPV is paramount to reactor safety analysis since it predicts the transition point of accident progression from in-vessel to ex-vessel phase.In the present study we proposed a new creep model for the classical French RPV steel 16 MND5,which is adapted from the "theta-projection model" and contains all three stages of a creep process.Creep curves are expressed as a function of time with five model parameters θ_i(i=1-4 and m).A model parameter dataset was constructed by fitting experimental creep curves into this function.To correlate the creep curves for different temperatures and stress loads,we directly interpolate the model’s parameters θ_i(i=1-4 and m) from this dataset,in contrast to the conventional "theta-projection model" which employs an extra single correlation for each θ_i(i=1-4 andm),to better accommodate all experimental curves over the wide ranges of temperature and stress loads.We also put a constraint on the trend of the creep strain that it would monotonically increase with temperature and stress load.A good agreement was achieved between each experimental creep curve and corresponding model’s prediction.The widely used time-hardening and strain-hardening models were performing reasonably well in the new method.

Improvement in irradiation resistance of FeCu alloy by pre-deformation through introduction of dense point defect sinks

The irradiation resistance of pre-deformed FeCu alloy was studied using a 3 MeV Fe ion irradiation experiment at room temperature in comparison with that of the as-received sample.Nanoindentation and atom probe tomography(APT) were used to characterize the mechanical properties and solute distribution.The stress-strain curve obtained by nanoindentation shows that the yield strength(σ0.2) of the pre-deformed sample is unexpectedly reduced with an increase in the irradiation dose to five displacements per atom(dpa).We suggest that it results both from the decrease in the dislocation density and the suppression of defects during irradiation.APT shows that the nucleation of the Cu cluster is suppressed;however,its growth is promoted in the pre-deformed sample,resulting in the formation of sparse and coarse clusters at 1 dpa irradiation.These coarse Cu clusters were then unexpectedly refined to finer grains with an increase in the irradiation dose to 5 dpa.Theoretically,the improvement in the resistance to irradiation in the pre-deformed sample is attributed to the dense point-defect sinks,that is,the dislocations and grain boundaries introduced by pre-deformation.In addition,the contributions of the dislocations and grain boundaries to the sink strength are estimated for both the as-received and pre-deformed samples.The results indicate that dislocations,rather than grain boundaries,play a major role after deformation.