Advanced Power Reactor Research Articles

A practical challenge-response authentication mechanism for a Programmable Logic Controller control system with one-time password in nuclear power plants

Abstract Instrumentation and Control (I&C) systems of nuclear power plants (NPPs) have been continuously digitalized. These systems have a critical role in the operation of nuclear facilities by functioning as the brain of NPPs. In recent years, as cyber security threats to NPP systems have increased, regulatory and policy-related organizations around the world, including the International Atomic Energy Agency (IAEA), Nuclear Regulatory Commission (NRC) and Korea Institute of Nuclear Nonproliferation and Control (KINAC), have emphasized the importance of nuclear cyber security by publishing cyber security guidelines and recommending cyber security requirements for NPP facilities. As described in NRC Regulatory Guide (Reg) 5.71 and KINAC RS015, challenge response authentication should be applied to the critical digital I&C system of NPPs to satisfy the cyber security requirements. There have been no cases in which the most robust response authentication technology like challenge response has been developed and applied to nuclear I&C systems. This paper presents a challenge response authentication mechanism for a Programmable Logic Controller (PLC) system used as a control system in the safety system of the Advanced Power Reactor (APR) 1400 NPP.

Controller tuning constants for the Advanced Power Reactor 1400 Feedwater Control System based on Ms-Constrained Integral Gain Optimization Method

Failure to control the steam generator (SG) water levels causes unexpected shutdowns in the Advanced Power Reactor (APR) 1400 nuclear power plants. This impacts maintenance, operational life, and safety critically. Therefore, it is vital to stabilize the SG water level at different power levels, particularly at low powers, by improving the performance of the proportional integral (PI) controller in the Feedwater Control System (FWCS). The PI controller designed using Ms-Constrained Integral Gain Optimization (MIGO) mitigated this problem significantly. In order to design this PI controller for the FWCS, the SG model was developed as a process model with a system identification algorithm in MATLAB. The algorithm processed data sets obtained from the thermal-hydraulic model by the Safety and Performance Analysis Code (SPACE). The MIGO and the Ziegler-Nichols methods were each used to produce a design for the PI controller, and the controllers were compared. Simulations were run in MATLAB for resulting designs. The PI controller designed by the MIGO is more stable than the controller using the Ziegler-Nichols approach. Finally, the PI MIGO controller was applied to an overall schematic block diagram of the FWCS. The simulations of the diagram in MATLAB Simulink demonstrate the PI MIGO controller offers superior control and enhanced robustness in the FWCS.

Validation of severe accident management guidelines (SAMGs) for advanced power reactor 1400 (APR1400) using colored Petri net (CPN) Tools

Since the Fukushima nuclear accident in 2011, there has been a growing need to strengthen severe accident management guidelines (SAMGs) and find appropriate methods for validating SAMGs. The validation of SAMGs is currently performed in terms of their usability, using a full-scope simulator, a tabletop drill, and a severe accident analysis code. However, given that the use of these rely on human expertise, and the workflow of SAMGs is very complex, a new validation method is needed. In this study, a model of a SAMG workflow has been developed using Colored Petri Nets (CPN) as a new validation method. In addition, time delay, which is required for SAMG actions, has been factored in.In order to verify usable SAMGs instead of existing validations by SAMG documents, the SAMGs were validated using computing technology. The modified CPN models, which are representative of the limited plant states during implementation of SAMGs, were simulated for computer-assisted validation. It has been verified that the novel validation method can complement existing validation methods.

Onsite power system risk assessment for nuclear power plants considering components ageing

In this paper, the influence of ageing on Probabilistic Safety Assessment (PSA) results is estimated for Advanced Power Reactor APR1400 onsite power system. Models for component ageing failure rate including surveillance test and replacement ageing control are presented for utilization. Time-dependent ageing effect and average ageing effect over plant lifetime or replacement intervals are evaluated. Components are prioritized regarding their influence on the change of system unavailability and relative increase of their unavailability due to ageing.

A Study of the Ageing Effect on Onsite Power System Risk Assessment for Nuclear Power Plants

Ageing is a general process in which properties of a component gradually changes with time or use. Ageing is one of the phenomena that effects the availability of Nuclear Power plant (NPP) electrical systems. Ageing can cause the failure rate of components to increase over time as the component wears out. Operating experience has demonstrated many components failures that have occurred as a result of ageing mechanisms such as corrosion, erosion, embrittlement, fatigue, vibration and wear. In this paper, the influence of ageing on Probabilistic Safety Assessment (PSA) results is estimated for Advanced Power Reactor (APR1400) onsite power system. Models for component ageing failure rate including surveillance test and replacement ageing control are presented for utilization. Time-dependent system unavailability due to ageing is evaluated. The average increase of system unavailability due to the ageing of system components is estimated. Components are prioritized regarding their influence on the change of system unavailability and relative increase of their unavailability due to ageing. The analysis is performed using Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE-8) software. The obtained results show the significant increase in average system unavailability after incorporating ageing effect, and the surveillance test and replacement activities can reduce this effect. Detailed time plot over plant lifetime is obtained showing the progression of ageing impact on system unavailability. The timedependent results show when, in a plant's lifetime, ageing effects become significant from a risk standpoint, and whether components are being replaced frequently enough. System components are ranked in accordance with their risk significance. As the components age, the contribution of some components in system unavailability increases, while it decreases for the others. The method presented is applicable for identification and ranking of the components that should be considered in the ageing management program.

What will the electricity miracle be? [Blueprints for a Miracle

TABLE TENNIS ISN'T meant to be played at Mach 2. At twice the speed of sound, the ping-pong ball punches a hole straight through the paddle. The engineers at TerraPower, a startup that has designed an advanced nuclear power reactor, use a pressurized-air cannon to demonstrate that very point to visitors. The stunt vividly illustrates a key concept in nuclear fission: Small objects traveling at high speed can have a big impact when they hit something seemingly immovable. And perhaps there is a larger point being made here, too-one about a small and fast-moving startup having a big impact on the electric-power industry, which for many years also seemed immovable.

Effect of mineral admixtures in concrete on biological shield wall radiation in APR1400

Although many researchers investigated the effect of radiation on concrete mechanical properties, a little attention was given to the effect of concrete mix design on the radio-activation of biological shield wall, which could be important on decommissioning of nuclear power plants. In this paper, to investigate the effect of concrete mix design, radiation in biological shield wall of the Advanced Power Reactor 1400 MWe (APR1400) was investigated with consideration of several concrete mix designs. As an analytical variable, three different concrete mix designs were considered according to mix proportions of Ordinary Portland Cement (OPC), Fly Ash (FA), and Ground Granulated Blast Furnace Slag (GGBS), which were broadly adopted on construction site. For irradiation analysis of concrete, the compositions of concrete mix were analyzed to obtain the mass of each element in each concrete mix design. Flux distribution along depth of concrete shield wall was evaluated from flux at the outside of the Reactor Pressure Vessel (RPV) which was obtained through regression analysis for flux distribution in the RPV. After 60 years of service life and 10 years of cooling period, the analysis showed that concrete mixed with FA was 9% more radio-activated than concrete in which OPC only or both OPC and GGBS were mixed. The result of this paper will be useful for design of concrete shield wall considering radioactive waste amount generated through decommissioning.

A case study of time-dependent risk informed integrated safety assessment under complex accident sequences

A time-dependent risk informed integrated safety assessment methodology is suggested to evaluate the response of nuclear power plants under complex accident sequences caused by failures of multiple safety features. Event propagation is modeled based on the combination of deterministic and probabilistic safety assessment methods, with results representing changes in plant risk over time. Operators can utilize the time-dependent risk information as a quantitative basis, and results can determine the degree of safety enhancement by the improvement of emergency operator action procedures or by strengthening the design of safety features. The concepts of consequential failure probability and point-estimate failure time are introduced. The consequential failure probability, calculated from central limit theorem, identifies the key safety system and provides a precise risk calculation. The point-estimate failure time, using non-parametric order statistics, justifies the crediting of emergency operator actions. As a case study, risk of large early radioactive release due to pressure-induced multiple steam generator tube rupture following main steam line break accident in Advanced Power Reactor 1400 is considered. Results are used to discuss the credibility of operator mitigation actions, and a remedy to enhance plant safety to satisfy safety goals is suggested.

Application of variable frequency drive on the condensate pump motors of APR1400 nuclear power plants for energy savings

Condensate pumps provide the motive force to deliver condensate from the hotwell to the deaerator storage tanks. Advanced Power Reactor 1400 (APR1400) is comprised of three 50% capacity motor drive...

<em>Ab initio</em> calculation of the thermal neutron scattering cross sections of uranium mononitride

Nuclear design and neutronic analysis of thermal neutron reactor need high reliable thermal neutron cross sections. Uranium mononitride (UN) is a candidate fuel material for advanced power reactor with its better thermodynamics and accident tolerance. However, in thermal neutron region, reliable thermal neutron scattering cross sections are lacked for UN, which is disadvantageous to reactor physics simulations. The scattering law of the UN fuel material may impact the thermal neutron spectrum and criticality of the reactor systems. Neutron cross sections in thermal range are correlated with energy, temperature, physical and chemical properties of the scattering medium, reflecting the phonon spectra of material itself. In this paper, based on the ab initio method of quantum mechanics, phonon density of states in UN are calculated by VASP/PHONON code, and used for integral to obtain UN heat capacity at a constant volume. Adopting this new phonon density of states, NJOY/LEAPR code is used to generate S (, ) data by thermal neutron scattering theory and NJOY/THERMR utilizes these data to produce thermal scattering matrix in order to investigate thermal kernel effect of UN. Subsequently, thermal neutron scattering cross sections of UN are generated with NJOY code system. Comparison with uranium dioxide (UO2) in the traditional PWR is done. Results indicate that optimized lattice parameter are in good agreement with the database; the optical modes are well separated from the acoustic modes compared with UO2; heat capacity at a constant volume is consistent with experimental value; the inelastic and elastic cross sections of 238U in UN are lower than those of 238U in UO2. N in UN only deals with incoherent part in elastic cross sections. As the temperature increases, elastic cross sections of UN decrease while inelastic ones increase, and cross sections approach to free atom cross section at high energies. Considering the limitations of 14N, the scattering law and inelastic scattering cross sections are also under investigation using 15N in UN compound. This paper's conclusion fulfill the vacancy of thermal neutron scattering cross sections of UN, which laid a foundation for systematic study on the neutronics properties of UN fuel in the light water reactors as well as for the design of new neutron moderators and neutron filer.

Development of wall and interfacial friction models for two-dimensional film flow with local measurement methods

Abstract Based on the advancement of computational capability, high-precision analyses of multi-dimensional thermal hydraulic phenomena in nuclear power plants have been carried out. One of the multi-dimensional phenomena is two-phase flow in the upper downcomer. The quantification of momentum transfer between a downward liquid flow and a transverse gas flow in the reflood phase is significant because it determines the bypass flow rate of the emergency core coolant (ECC). Multi-dimensional modules of thermal-hydraulic analysis codes have been used to simulate this multi-dimensional two-phase flow. However, the model and correlations incorporated in the multi-dimensional modules are ones developed from one-dimensional experimental results. Because the characteristics of multi-dimensional phenomena are different from one-dimensional conditions, the implemented models and correlations might produce inaccurate results. Therefore, new models based on multi-dimensional experimental data should be developed. This study involved the twodimensional, two-phase experiments performed using 1/10 and 1/5 scale facilities. These were simplified unfolded downcomers of the Advanced Power Reactor 1400 (APR1400). The local measurement methods, pitot tubes, ultrasonic thickness gauge, and depth-averaged PIV, were applied to measure local variables: air velocity, liquid film velocity, and liquid film thickness. Local experimental data on the two-dimensional film flow were used to produce the wall and to inform interfacial friction factors from two-dimensional, two-phase momentum conservation equations. New wall and interfacial friction models were developed incorporating newly proposed physical concepts.

Application of CUPID for subchannel-scale thermal–hydraulic analysis of pressurized water reactor core under single-phase conditions

Abstract There have been recent efforts to establish methods for high-fidelity and multi-physics simulation with coupled thermal–hydraulic (T/H) and neutronics codes for the entire core of a light water reactor under accident conditions. Considering the computing power necessary for a pin-by-pin analysis of the entire core, subchannel-scale T/H analysis is considered appropriate to achieve acceptable accuracy in an optimal computational time. In the present study, the applicability of in-house code CUPID of the Korea Atomic Energy Research Institute was extended to the subchannel-scale T/H analysis. CUPID is a component-scale T/H analysis code, which uses three-dimensional two-fluid models with various closure models and incorporates a highly parallelized numerical solver. In this study, key models required for a subchannel-scale T/H analysis were implemented in CUPID. Afterward, the code was validated against four subchannel experiments under unheated and heated single-phase incompressible flow conditions. Thereafter, a subchannel-scale T/H analysis of the entire core for an Advanced Power Reactor 1400 reactor core was carried out. For the high-fidelity simulation, detailed geometrical features and individual rod power distributions were considered in this demonstration. In this study, CUPID shows its capability of reproducing key phenomena in a subchannel and dealing with the subchannel-scale whole core T/H analysis.

Performance Improvement of Liquid Waste Management System for APR1400

The Liquid Waste Management System (LWMS) is a radioactive waste storage and treatment facility equipped in the nuclear power plants to ensure the safe management of radioactive waste generated in the process of operation. New and improved materials and processes are continuously under consideration and development in various countries to reduce waste generation and to minimize costs. To meet the increasing pressure to reduce the release of radioactive into the environment, this study discusses the possible performance improvements of the existing liquid waste management system of APR1400 without additional equipment installation or design changes of the system. This study proposes three possible items for improvement of the Advanced Power Reactor 1400 (APR1400) liquid waste management system which are chemical additions, increasing ion exchange (I.X.) selectivity and optimal operation conditions to reduce the radioactivity concentration in the effluent and solid waste generation. The results of off-site dose calculated by ENDOS-L shows that maximum individual dose is reduced by 12.7% from original values. And the reduction of solid waste generation from the extended replacement period of the membrane is expected to reduce the annual disposal cost by 16.7% from original cost. Improvements do not involve equipment modification so these can be easily applied to currently operating NPPs. It will be helpful to reduce the offsite public dose and operating cost of Nuclear Power Plants(NPPs).

Integrated Response Time Evaluation Methodology for the Nuclear Safety Instrumentation System

Safety analysis for a nuclear power plant establishes not only an analytical limit (AL) in terms of a measured or calculated variable but also an analytical response time (ART) required to complete protective action after the AL is reached. If the two constraints are met, the safety limit selected to maintain the integrity of physical barriers used for preventing uncontrolled radioactivity release will not be exceeded during anticipated operational occurrences and postulated accidents. Setpoint determination methodologies have actively been developed to ensure that the protective action is initiated before the process conditions reach the AL. However, regarding the ART for a nuclear safety instrumentation system, an integrated evaluation methodology considering the whole design process has not been systematically studied. In order to assure the safety of nuclear power plants, this paper proposes a systematic and integrated response time evaluation methodology that covers safety analyses, system designs, response time analyses, and response time tests. This methodology is applied to safety instrumentation systems for the advanced power reactor 1400 and the optimized power reactor 1000 nuclear power plants in South Korea. The quantitative evaluation results are provided herein. The evaluation results using the proposed methodology demonstrate that the nuclear safety instrumentation systems fully satisfy corresponding requirements of the ART.

Evaluation of mechanistic wall condensation models for horizontal heat exchanger in PAFS (Passive Auxiliary Feedwater System)

PAFS (Passive Auxiliary Feedwater System) is one of the advanced safety features in the design of the APR+ (Advanced Power Reactor) nuclear power plant, in order to cool down the reactor coolant system without any external supply of electricity or water during an accident. The driving force of the PAFS is condensation inside horizontal tubes of PCHX (Passive Condensation Heat Exchanger). To improve the prediction capability for the condensation heat transfer in the PCHX, advanced wall condensation models which mechanistically consider the wall condensation with the different flow regimes in a horizontal channel were implemented into a thermal hydraulic safety analysis code, SPACE (Safety and Performance Analysis Code for Nuclear Power Plant). The models were evaluated by comparing to the experimental data from the PASCAL (PAFS Condensing Heat Removal Assessment Loop) test facility, which is a separate effect test to validate the cooling performance of the PCHX. Calculation results showed that the condensation model package (Ahn et al., 2014) presented the enhanced the prediction capability of the SPACE code in predicting the steam pressure and temperature under quasi-steady state conditions. Since this model package is composed of mechanistic models by considering the film condensation in the steam phase and the convection in the condensate liquid separately, it contributed to enhance the prediction capability of the SPACE code and reduce the conservatism in the prediction of the cooling capability of the PAFS. The SPACE code with improvement of the wall condensation model can be applied in predicting the cooling and operation capability of the PAFS more realistically in the safety analysis of APR+ nuclear power plant.

Optimal design of passive containment cooling system for innovative PWR

Abstract Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC) code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS) of an innovative pressurized water reactor (PWR). A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT) geometry, PCCS heat exchanger (PCCX) location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed.

The sensitivity analysis for APR1400 nodalization under Large Break LOCA condition based on mars code

The phenomena of loss of coolant accident have been investigated for long time and the result of experiment shows that the flow condition in the downcomer during the end-of-blowdown were highly multi-dimensional at full-scale. However, the downcomer nodalization of input deck for large break loss of coolant accident used in advanced power reactor 1400 analyses are made up with 1-D model and improperly designed to describe realistic coolant phenomena during loss of coolant accident analysis. In this paper, the authors modified the nodalization of MARS code LBLOCA input deck and performed LBLOCA analysis with new input deck. From original LBLOCA input deck file, the nodalization of downcomer and junction connections with 4 cold legs and direct vessel injection lines are modified for reflecting the realistic cross-flow effect and real downcomer structure. The analysis results show that the peak cladding temperature of new input deck decreases more rapidly than previous result and that the drop of peak cladding temperature was advanced by application of momentum flux term in cross-flow. Additionally, the authors developed a new input deck with multi-dimensional downcomer model and ran MARS code with multi-dimensional input deck as well. By using the modified input deck, the Emergency core cooling system by-pass flow phenomena is better characterized and found to be consistent with both experimental report and regulatory guide.

연성회로기판 기반 수평전열관 표면의 비등기포거동 가시화 실험 연구

The Passive Auxiliary Feedwater System(PAFS) is one of the advanced safety concepts adopted in the Advanced Power Reactor Plus(APR+). To validate the operational performance of the PAFS, detailed understanding of a boiling heat transfer on horizontal tube outside is of great importance. Especially, in the mechanistic boiling heat transfer model, it is important to visualize the phenomena but there are some limitations with conventional experimental approaches. In the present study, we devised a heater based on the Flexible Printed Circuit Board (FPCB) for a more comprehensive visualization and subsequently, a digital image processing technique for the bubble motion measurement was established. Using the measurement technique, important parameters of the nucleate boiling are analyzed.

Preliminary study for a PWR steam generator with CUPID/MARS multi-scale thermal-hydraulics simulation

For the thermal hydraulics analysis of nuclear reactor components of LWRs (Light Water Reactors) such as reactor vessel, steam generator, containment, a multi-dimensional two-phase flow code, named CUPID, has been being developed. The CUPID code pursues a capability of multi-physics and multi-scale thermal hydraulics analysis. In the present study, multi-scale simulation was performed by coupling with system-scale code, MARS. The coupled code was assessed to visualize the flow behavior of the steam generator of the Advanced Power Reactor (APR1400). The primary side of the steam generator and remaining Reactor Coolant System (RCS) is modeled by MARS and secondary side is by CUPID. For the secondary side simulation by the CUPID part, a porous media approach was adopted to two-fluid model and conductor model to simplify the complicated geometry of the steam generator. In order to obtain a porosity of a given computing cell, a special algorithm was employed to directly calculate volume ratio by mapping the 3D CAD file onto the grid system. Besides, the proper constitutive relationships for U-tubes are considered further. To treat the complex thermo-hydraulic phenomena on the shell side of a steam generator, a set of constitutive models available in the literature for a two-phase flow map, interfacial heat and mass transfer, interfacial drag, wall friction, wall heating, and heat partitioning in flows over tube bundles were applied to close the numerical model. This paper presents the description of the coupling method, porous media approach to simplify the steam generator, and the simulation results using the coupled code.

Detailed evaluation of two phase natural circulation flow in the cooling channel of the ex-vessel core catcher for EU-APR1400

The ex-vessel core catcher of the PECS (Passive Ex-vessel corium retaining and Cooling System) is installed to retain and cool down the corium in the reactor cavity of the EU (European Union)-APR (Advanced Power Reactor) 1400. A verification experiment on the cooling capability of the PECS has been conducted in the CE (Cooling Experiment)-PECS. Simulations of a two-phase natural circulation flow using the RELAP5/MOD3 computer code in the CE-PECS and PECS have been conducted to predict the two-phase flow characteristics, to determine the natural circulation mass flow rate in the cooling channel, and to evaluate the scaling in the experimental design of the CE-PECS. Particularly from a comparative study of the prototype PECS and the scaled test facility of the CE-PECS, the orifice loss coefficient in the CE-PECS was found to be 6 to maintain the coolant circulation mass flux, which is approximately 273.1kg/m2s. The RELAP5 results on the coolant circulation mass flow rate are very similar to the CE-PECS experimental results. An increase in the coolant injection temperature and the heat flux lead to an increase in the coolant circulation mass flow rate. In the base case simulation, a lot of vapor was injected into the downcomer, which leads to an instability of the two-phase natural circulation flow. A super-step design at a downcomer inlet is suitable to prevent vapor injection into the downcomer piping.