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ASME Conference Presenter Attendance Policy and Archival Proceedings

2018;():V001T00A001. doi:10.1115/ICONE26-NS1.
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This online compilation of papers from the 2018 26th International Conference on Nuclear Engineering (ICONE26) represents the archival version of the Conference Proceedings. According to ASME’s conference presenter attendance policy, if a paper is not presented at the Conference by an author of the paper, the paper will not be published in the official archival Proceedings, which are registered with the Library of Congress and are submitted for abstracting and indexing. The paper also will not be published in The ASME Digital Collection and may not be cited as a published paper.

Commentary by Dr. Valentin Fuster

Operations and Maintenance, Engineering, Modifications, Life Extension, Life Cycle, and Balance of Plant

2018;():V001T01A001. doi:10.1115/ICONE26-81003.

Nuclear power plants typically consider a turbine trip and rapid closure of the main turbine stop valves as a normal transient event. As required by ASME Code [1], the piping loads generated by the unbalanced pressures in the system resulting from the rapid valve closure are part of the analyzed spectrum of conditions in the piping and support analysis. The analysis that determines the magnitude and timing of the loads is often referred to as a “steamhammer” analysis.

Currently, there are several computerized analytical techniques to determine the steamhammer piping and support loads [2], but because of compressibility assumptions the equations become more difficult to solve than in the analogous incompressible waterhammer models, which are quite straightforward.

This paper highlights the effect of fluid compressibility by comparing results predicted by both waterhammer (slightly compressible) flow models and compressible (steamhammer) flow models. Guidelines are offered to show how parameters of a piping system (such as pipe length, valve closure time and flow characteristic, steam initial state properties, and velocity) can be interpreted to determine if compressible effects are insignificant or if they play a significant role.

Commentary by Dr. Valentin Fuster
2018;():V001T01A002. doi:10.1115/ICONE26-81183.

SG (steam generator) is one of the most important equipment in fast reactors, the experience in design and operation of fast reactor worldwide show that failures of SG occurred frequently and often caused serious consequences, therefore it’s necessary to conduct reliability analysis on SG in design phase. FMEA (Failure Mode Effect Analysis) is used to identify all potential failure modes and filter out main failure modes. Then, qualitative analysis and quantitative calculation are carried out to evaluate main failure modes. Next, reliability of SG can be obtained by conducting Latin Hypercube Sampling. Analysis results show that the leakage probability of SG in 20 years is 0.130 219, and the most sensitive factor is the quality of weld in the junction of tubes and tube plate, and the SG meet its reliability requirement.

Commentary by Dr. Valentin Fuster
2018;():V001T01A003. doi:10.1115/ICONE26-81212.

In order to guarantee NPP operation safety, it is necessary to ensure the reliability operation of safety related equipment by equipment qualification. In the past, plant equipment qualification function requirement is an envelope requirement based on engineering experience, and it is not beneficial to NPP economical efficiency. These days, representative Gen III NPP (e.g. AP1000 and EPR) adopt improved technology in equipment qualification function requirement design, and more accurate requirement is designed. In this paper, AP1000 and EPR equipment qualification function requirement design methodology is studied and analyzed as the first step. Then, a safety related equipment qualification function requirement design methodology which is applicable for China self intellectual property Gen III NPP is provided. Furthermore, an example of equipment qualification function requirement design is carried out by analyzing nuclear instrument system power range channel sensor.

Commentary by Dr. Valentin Fuster
2018;():V001T01A004. doi:10.1115/ICONE26-81408.

Classification of nuclear power plant waste related buildings and structures is relatively flexible, and research on it is relatively less. In order to better master the classification of waste related buildings and structures and adapt it to the improvement of nuclear power technology and regulations, this paper carries out the function analysis of waste systems and sub-items, and integrates the mature engineering practice experiences and the differences between projects. Finally classification of waste sub-items is analyzed and summarized, such as sectional classification of liquid waste discharge galleries, classification optimization of radioactive waste building. The results of the optimization analysis in this paper can provide sufficient basis and guidance for the classification of nuclear power projects in the future, and improve the economy of nuclear power technology, and shows very good engineering application significance.

Commentary by Dr. Valentin Fuster
2018;():V001T01A005. doi:10.1115/ICONE26-81478.

Nuclear Power Plants (NPP) have multiple levels of defense in depth hierarchy. The NPP accident condition operation strategy belongs to the 3rd level. It is used to supervise the operator to handle the NPP under accident operating condition. NPP accident condition operation strategy is an essential and difficult work in NPP design field, hence only few organizations are able to develop the accident operating strategies independently all over the world. In this paper, a systematic NPP accident operating condition strategy design methodology is raised based on function analysis and task analysis technology. Based on the methodology, a reactivity accident operation strategy is designed and proved to be reasonable through preliminary verification and validation work.

Commentary by Dr. Valentin Fuster
2018;():V001T01A006. doi:10.1115/ICONE26-81480.

For PWR, remote shutdown station (RSS) is a redundant control mean to shut down the reactor when main control room (MCR) inhabitation is challenged (e.g. fire, smoke...). Nowadays, due to nuclear power plants control measures were improved with DCS system, a full function DCS RSS was equipped and more essential equipment could be controlled on RSS.

Under operating conditions that prohibit nuclear power plant operators to stay in the main control room, the operators should move to RSS and shutdown the reactor to ensure plant safety following <Moving to remote shutdown station when main control room is un-inhabitable operating strategy> (RSS strategy for short) to fallback the plant from power operation to cold shutdown. The original operating strategy by nature circulation is no longer the best choice both for operation safety and economy efficiency, and an optimized new strategy should be raised.

Based on the former reason, an optimized operation strategy was raised in this paper. In the optimized strategy, all plant normal standard operation modes were considered as initial conditions, rather than only considering power operation condition in the original one. The fallback mode and fallback strategy for each initial condition was also designed and optimized. To accelerate the depression and heat removal process, a forced circulation operation strategy is adopted when the reactor coolant pumps are available, and less local operation was included by taking advantages of the full function operating measures on RSS. To simplify the whole procedure structure, the operation modules of other general operating procedures are reused.

To validate the effectiveness of the optimized operating strategy, a full scope PWR simulation tool was employed to make thermo hydraulic calculation validation of the reactor response and also the remote control station HMI supporting validation. By simulating the original strategy and the optimized one and related analysis, we found that the optimized strategy is effective, and able to be executed based on the remote control station hardware. By executing the optimized strategy, the unit can fall back to the cold shutdown condition safely and a few hours were saved compared with the original strategy. The optimized strategy had already been implemented on real PWR nuclear power plant.

Commentary by Dr. Valentin Fuster
2018;():V001T01A007. doi:10.1115/ICONE26-81575.

Active magnetic bearing technology is used more and more for its high performance, such as high speed and frictionless operation. But the rotor vibrates sometimes during operation due to the existence of residual unbalanced mass, which may affect the security of the whole system. In order to determine the distribution of residual unbalanced mass, this paper proposes a method based on frequency response, control current analysis, and image data processing. The theoretical and calculated results show the validity of the method.

Topics: Rotors , Feedback
Commentary by Dr. Valentin Fuster
2018;():V001T01A008. doi:10.1115/ICONE26-81584.

In nuclear power plant system, pump is the key equipment to maintain the flow of the primary loop coolant and the secondary loop heat transfer fluid. The main coolant pump and the feed water pump are in long-term operation status. Bearings are the key components to ensure stable operation of the pump, and which could be damaged in abnormal conditions. Once the failure occurred in the bearings, pumps would exhibit periodic vibration, which might cause the flow pulsations of coolant and heat transfer fluid; gradually, these situations could reduce the control accuracy and the stability of pump. Therefore, the detection and diagnosis of pump bearings are significant to improve the safety and stability of reactor system. We proposed an approach combined with signal processing and machine learning to extract the signal features and recognize the signal samples automatically. The proposed approach consists of three main steps: firstly, empirical mode decomposition (EMD) is applied to decompose the signals into several intrinsic mode functions (IMFs) which are corresponding to the different components of the original signals; secondly, calculating the correlation coefficient between each IMF and the original signal, the correlation coefficient sequence imply the components distribution of the signal which can be applied to recognize the signal samples; finally, extracting a part of correlation coefficient sequences to train the support vector machine (SVM), and then an classifier can be obtained and use to recognize the other signal samples automatically. Experimental results show that this method can effectively detect the pump bearing operating conditions and failures, and can provide a reference for the safe and stable operation of reactor pumps.

Commentary by Dr. Valentin Fuster
2018;():V001T01A009. doi:10.1115/ICONE26-81708.

The problem of ensuring the integrity of VVER type reactor equipment integrity is now most significant in connection with justifying the safety of the NPP units and the extension of their service life-time to 60 years and more. This issue primarily first of all concerns long term operated NPP power units with VVER-440s and VVER-1000s. The justification of the VVER equipment integrity depends on the reliability of estimation of the degree of the equipment damage. One of the mandatory requirements [1], providing the reliability of such estimation, and also the evaluation of VVER equipment life-time, is the monitoring of equipment radiation loading parameters. Relative to this requirement there is a problem the challenge of justification of such the normative parameters, used for an estimating of the reactor pressure vessel (RPV) metal embrittlement, as the fluence and fluence rate of fast neutrons with energies above 0,5 MeV. Compliance with these requirements is analyzed during regular monitoring of radiation load parameters, which is performed by SEC NRS for all Russian NPP from the regulatory point of view. As a result of this activity, SEC NRS has recently elaborated one of the new approaches aimed to monitoring the radiation load of all equipment of Russian VVERs. The paper describes these approaches and shows the way of their implementation during monitoring procedures.

Commentary by Dr. Valentin Fuster
2018;():V001T01A010. doi:10.1115/ICONE26-81797.

After a long-term shutdown, the 10MW high temperature gas-cooled test reactor (HTR-10) was restarted, and the operation & safety characteristics of the HTR-10 transition core are tested and verified. A series of the characteristic tests have been implemented, such as the value calibrating test of the control rod and boron absorber ball, the disturbance characteristic of helium circulator, the start-stop characteristic and the stable power operation characteristic, which indicated the characteristics of the reactor transition core meet the design and safety requirements.

Commentary by Dr. Valentin Fuster
2018;():V001T01A011. doi:10.1115/ICONE26-81825.

The paper presents a Relap5 study of the influence of the centrifugal pump characteristics on the dynamic loads on piping system after power failure. Interpolated and experimental pump characteristics are used.

The differences between the interpolated and measured pump curves, the general description of Relap5 model and results of calculations in form of selected time curves for rotational speed, volume flow, pressures and dynamic forces are presented and discussed.

The analysis of the results shows that the maximal dynamic force on pipe section calculated with experimental pump curves can be up to 6 % higher than respective calculated using interpolated curves. However, it is not possible to determine to what extent the differences are caused by the interpolation itself or caused by the differences in the design of the centrifugal pumps. The latter since it differs more than 50 years between the pumps whose characteristics are used for interpolation and the pumps with corresponding experimental characteristics.

Commentary by Dr. Valentin Fuster
2018;():V001T01A012. doi:10.1115/ICONE26-81833.

The Transient Reactor Test (TREAT) Facility, located at the Idaho National Laboratory (INL), is a versatile test facility able to subject experimental specimens to various transient nuclear conditions. TREAT was placed in standby after operating from February 1959 through April 1994, resulting in the loss of nearly all transient testing capability in the United States. Recently, the US Department of Energy (DOE) determined this capability was again needed. After DOE completed National Environmental Policy Act actions in February 2014, INL established the Resumption of Transient Testing Program (RTTP). RTTP was a multi-year effort to restart TREAT to reestablish a domestic transient testing capability. After 23 years of standby operations, the RTTP completed restart activities on August 31, 2017, 13 months ahead of schedule and nearly $20 million under budget. RTTP activities included an Environmental Assessment that resulted in a Finding of “No Significant Impact” associated with restarting TREAT, establishment of a compliant Safety Analysis Report (SAR), refurbishment and/or replacement of key reactor systems and components, key system knowledge recovery, reestablishment of configuration management, procedure updates, personnel training and qualification, and demonstration of operational readiness for reactor operations. Several noteworthy factors that contributed to the restart of TREAT include:

• Funding to acquire personnel and material resources provided in a timely fashion.

• Close coordination with the regulator’s (DOE) nuclear safety program during updates, interactive review, and approval of safety documentation provided for timely update of the TREAT SAR and implementing documents.

• Effective management control enabled by utilization of standard outage management techniques with a focus on age-related degradation and updated standards and requirements.

• DOE program management ensured efficient implementation of program management tools. These tools focused on clear high-level milestones and spend plans allowing flexibility for the contractor to respond to evolving facility conditions and information in a near-real time manner and with minimal program overhead. This approach enabled efficient execution of work in an environment where determination of required work scope was dependent on performance of inspection, testing, analysis, and evaluation activities.

• Implementation of the Contractor Assurance System, with frequent internal and externally-led assessments that facilitated process improvements and corrective actions to ensure the operational readiness for required contractor and DOE readiness assessments and safe nuclear operations.

• The RTTP benefited from archived plant documentation and maintenance performed while the plant was in a safe-standby status.

• Unique methods of reactivity control allowed for individual and integrated reactor system functional testing, procedure vetting, and personnel training while maintaining the reactor in a safe state.

Commentary by Dr. Valentin Fuster
2018;():V001T01A013. doi:10.1115/ICONE26-81936.

Gamma camera imaging technology is a non-destructive passive radiation imaging technology, which can quickly find the unknown source location, search the exact number of radioactive sources and relative intensity. Therefore, it is very important and widely used in the fields of effective regulation of radioactive sources, handling of various nuclear emergencies, nuclear arms control and other fields. In the practical application of gamma camera, it often faces the imaging difference caused by the difference of radiation source intensity, detection time and detection distance. It is helpful to study the change of imaging characteristics under different experimental conditions for the practical application of gamma camera under different scenes. In this paper, the structure and imaging principle of gamma camera are analyzed in detail. Using the Cartogam portable gamma camera, a set of comparative experiments are carried out to study the time characteristics, distance characteristics and source intensity characteristics of the gamma camera. The results show that the imaging quality of gamma camera is positively correlated with the time source intensity, negatively correlated with the distance. For a milliCurie source, the gamma camera has very good fast-position resolution at a distance of 1 meter from the radioactive source and can form a more complete hot spot image within 5 minutes. When the distance becomes larger, the radioactive source needs at least 20 minutes to form a more accurate hot spot image. The hot spot is no longer as complete as a concentric circle structure, but can achieve precise positioning. For a strong source of more than ten milliCurie, immediate imaging within two minutes can be basically achieved within two meters. Under multi-source conditions, when the source intensities differ greatly and the distance between sources is relatively close, the detection of weak source can not be achieved by increasing the measurement time. However, by observing the counting images in a short period of time, the possibility of existence of a weak source can be deduced. Therefore, in the practical application of the gamma camera, it is necessary to constantly adjust its imaging conditions to ensure the detection of weak source verification. In this paper, the Monte Carlo model of gamma camera is set up to simulate the imaging. Compared with the actual imaging hot spots, the simulated images can correctly reflect the hot spot graph’s level distribution, which has the value of further research.

Topics: Imaging
Commentary by Dr. Valentin Fuster
2018;():V001T01A014. doi:10.1115/ICONE26-82016.

This paper introduces a method to control the calculation progress of the RELAP5 codes and integrate them with other codes by interacting boundary data (such as general tables, TDVs and TDJs) at each step.

This work is basically finished with the support of the RINSIM simulation platform. The paper gives a brief introduction on RINSIM that how it controls the codes progress, sends the codes control commands, shares the values of different codes’ common blocks or modules.

However, the work can’t be done by just using the RINSIM, it also needs to modify RELAP5 codes. With the codes’ modification of commands responding, data reading/writing interface, data interacting interface, time step control and so on, we can build interface subroutines to integrate codes onto the RINSIM.

At the end, the paper gives out the result of a transient calculation with an advanced PWR model. Compared to some old integration method, the new method has far more strong stability. And the result shows that the integration progress of the code does not obviously affect the calculation accuracy, but definitely extends the application fields because of the multiple functions supplied by the RINSIM.

Topics: Simulation
Commentary by Dr. Valentin Fuster
2018;():V001T01A015. doi:10.1115/ICONE26-82346.

This paper proposes to use robust command shaping methods for reducing the vibrations during remote handling of in-vessel components. The need of deriving efficient vibration control strategies for a safe transportation of large and heavy pay-loads during maintenance procedures in nuclear fusion reactors is the main motivation behind this work. The approach shapes the reference motion command to the component such that the vibratory modes of the system are canceled. We perform the dynamic simulations of a large in-vessel component of the DEMOnstrating fusion power reactor during a remote handling operation. The simulations shows that the method is a possible solution to reduce the vibrations induced by the motion, in both the transient and residual phases. The benefits introduced by command shaping make the method promising towards building control framework for remote handling of in-vessel components in various tokamak devices.

Topics: Tokamaks , Vibration
Commentary by Dr. Valentin Fuster
2018;():V001T01A016. doi:10.1115/ICONE26-82390.

The paper describes the activities of conceptual design of tools and procedures and the virtual simulation of the Remote Handling (RH) tasks provided in the maintenance of the systems present in the Access Cell (AC) of DONES (DEMO Oriented Neutron Source) facility. In particular, the RH maintenance of the Target Assembly (TA) is critical because of its position in the most severe region of neutron irradiation, the Test Cell (TC), where the material specimen are tested to understand the degradation of the materials properties throughout the reactor operational life. The main RH maintenance activity includes the replacement of the entire TA and the cleaning of the surfaces of connection in the TC. The cleaning operation is fundamental because it allows the removal of any lithium solid deposition from the surfaces: any further deposition on the surfaces could compromise the sealing of the TA. The RH is based on the idea of a reconfigurable modular chain of devices connected to the Access Cell Mast Crane (ACMC) located in the AC. To increase the modularity and to reduce the costs of the Remote Handling System (RHS), a telescopic boom is used equipped with a Gripper Change System (GCS) that allows the use of different end effectors. To perform the tasks, a Parallel Kinematic Manipulator (PKM) and a Robotic Arm (RA) are proposed, allowing the tools to move with more degree of freedom in the AC space. The modeling of the devices and the 3D kinematic simulations maintenance operations tasks were simulated and tested in virtual reality environment, aimed at developing and validating the implemented maintenance procedures, in collaboration with the IDEAinVR Laboratory of CREATE/University of Naples Federico II, and the research center at ENEA Brasimone, Italy.

Commentary by Dr. Valentin Fuster
2018;():V001T01A017. doi:10.1115/ICONE26-82422.

During the course of 42 years of irradiated operations in the Hot Fuel Examination Facility at the Idaho National Laboratory (INL), a hot cell window had never been replaced. Recently, slow deterioration of a window seal resulted in mineral oil leaking at a rate of over a liter per month from the window tank unit on through the protective A-slab seal and into the hot cell. A hot cell window consists of both a steel tank unit with five slabs of glass of varying thicknesses with the remaining free space filled with clear mineral oil, and a thinner protective interior A-slab of glass. The repair solution was to remove and replace the A-slab window followed by replacing the window tank unit in two distinct phases.

The facility original A-slab design was a leak tight barrier and a frame that was “L” shaped with a gasket between the glass and the window flange. Problems with the gasket adhering to the glass and the window flange resulted in pulling the glass from the frame during initial installation activities. Due to the adhesion problem, the gasket was changed to a dust seal during commissioning the facility. Over time, the window tank unit mineral oil leak flowed through this dust seal. Replacing the leaking tank unit necessitated the need for a new leak tight boundary as well as provide a method to drain the accumulated oil behind the A-slab until the tank unit could be replaced. These criteria led to a new A-slab design to be installed.

Initially, removal and replacement of the A-slab was performed in the main cell (hot side) to reestablish a leak tight barrier. Transfers of the windows and removal of the bolts/reinstallation of new bolts were all performed with specialized equipment designed for remote operations in a hazardous environment using remote manipulators and cranes. Removal and replacement of the window tank unit via the operating corridor (cold side) was scheduled during a facility outage to accommodate availability of contract service personnel who specialize in hot cell windows.

Due to the complexity of the replacement task, approximately 30% of the personnel on site were involved in the window replacement. Engineering, facility operations and radiation control personnel were primary contributors with electricians, carpenters and the analytical laboratory personnel contributing, as well. The multi-year installation program was safely concluded allowing the facility to resume full operations with the window properly sealed.

Topics: Fuels
Commentary by Dr. Valentin Fuster
2018;():V001T01A018. doi:10.1115/ICONE26-82560.

In power plants, there are structures made up of thin plates, such as air-conditioning ducts or thin-walled pipes, where corrosion can occur. In this study, we provide a solution to reduce inspection time of the thin plate corrosion measurement and enable monitoring, using a non-contact ultrasonic sensor. The sensor can measure the reduction in thickness of thin plates due to general corrosion without the need to remove or reinstall insulating material that is on the outside of the plate. The proposed sensor is based on the non-contact ultrasonic measurement technique which was originally proposed by Greve et al, further developed and patented by Zhong et al. at the University of Bristol, and commercialized by Inductosense Ltd. In order to ultrasonically measure the thin plate thickness, we use a method based on the group velocity of the guided waves. The proposed method was tested theoretically with numerical simulations and experimentally against our target conditions. The results of the numerical simulations and experiments confirm that the proposed method can be applied to thickness measurements of thin-plates in our target condition. Based on the feasibility test results, we developed a prototype sensor and measurement software. From the results of the performance evaluation tests, we have confirmed that the prototype sensor has sufficient capability to measure the thickness of the thin plates without the removal of the insulator. Even if the offset between the plate and the inspection probe is 100 mm, the prototype sensor still works well.

Commentary by Dr. Valentin Fuster

Instrumentation and Control (I&C) and Influence of Human Factors

2018;():V001T04A001. doi:10.1115/ICONE26-81122.

At present, DCS is widely used as the control system for nuclear power plants both at home and abroad, which prompting many companies to research the technology of DCS debugging. In this paper, taking a certain nuclear power plant within China for reference, the virtual DCS debugging and research platform which based on the full-scope nuclear power plant simulation model is developed. It was developed by first establishing a simulation model on the RINSIM Simulation Platform and ordering a customized set of virtual DCS system, then developing a communication program between the simulation model and the virtual DCS system. Users can observe the actual effects and results if following the pre-designed testing procedures after the configuration of control logics, HMI (Human Machine Interface) and I/O communication interfaces. The virtual DCS platform is aimed at assisting with technology research of DCS project for similar nuclear power plants and also can provide professional training for associated personnel of nuclear power plant.

Commentary by Dr. Valentin Fuster
2018;():V001T04A002. doi:10.1115/ICONE26-81142.

The advanced Mechanical Shim (MSHIM) core control strategy employs two separate and independent control rod banks, namely the MSHIM control banks (M-banks) and axial offset (AO) control bank (AO-bank), for automatic reactivity/temperature and axial power distribution control respectively. The M-banks and AO-bank are independently controlled by two closed-loop controllers called the coolant average temperature (Tavg) controller and AO controller. Since the movement of M-banks and AO-bank can both affect the Tavg and AO, the Tavg controller is coupled with the AO controller. In order to avoid the interference between the two controllers, the MSHIM control system adopts an interlock design between them to avoid the simultaneous movement of the M-banks and AO-bank and make sure the priority of the M-bank movement. This design can enhance the stability of the MSHIM control system. However, the control performance is degraded at the same time. In the present study, the feedforward compensation decoupling method and multimodel approach are used to eliminate the coupling effect between the two controllers in the MSHIM control system during a wide range of power maneuvers. A multiple feedforward compensation system is designed with integration of feedforward compensators for the Tavg and AO controllers at five power levels using the multimodel approach. By implementing it in the MSHIM control system, the interlock between the M-banks and AO-bank can be released to realize the independent and decoupled control between Tavg and AO. The effectiveness of the decoupled MSHIM control system is verified by comparing its control performance with that of the original MSHIM control system during typical load change transients of the AP1000 reactor. The obtained results show that superior and decoupled control of Tavg and AO can be achieved with the proposed decoupled MSHIM control system.

Commentary by Dr. Valentin Fuster
2018;():V001T04A003. doi:10.1115/ICONE26-81156.

Small Pressurized Water Reactors (SPWR) are different from those of the commercial large Pressurized Water Reactors (PWRs). There are no hot legs and cold legs between the reactor core and the steam generators like in the PWR. The coolant inventory is in a large amount. The inertia of the coolant is large and it takes a long time for the primary system to respond to disturbances. Once-through steam generator is adopted and its water inventory is small. It is very sensitive to disturbances. These unique characteristics challenge the control system design of an SPWR. Relap5 is used to model an SPWR. In the reactor power control system, both the reactor power and the coolant average temperature are regulated by the control rod reactivity. In the feedwater flow control system, the coordination between the reactor and the turbine is considered and coolant average temperature is adopted as one measurable disturbance to balance them. The coolant pressure is adjusted based on the heaters and spray in the pressurizer. The water level in the pressurizer is controlled by the charging flow. Transient simulations are carried out to evaluate the control system performance. When the reactor is perturbed, the reactor can be stabilized under the control system.

Commentary by Dr. Valentin Fuster
2018;():V001T04A004. doi:10.1115/ICONE26-81163.

Main control room simulator is widely used in design verification and operator training for nuclear power plant. The simulator needs to implement the arrangement, environment, human machine interface and function of main control room, which should be the same as much as possible. For designer, each type of reactor needs an individual simulator for design verification. As the number of unit increased, the simulator will consume a lot of space and difficult to reuse for other project. In addition, design verification for control room and I&C system need to start at the early stage of a project and is usually an iterative process with the design work. Build a control room facility for simulator needs a lot of time and is difficult to modify once constructed. To make the simulator more flexible and match the project schedule, virtual reality technology can be used to replace or extend traditional control room simulator with approximately the same arrangement, environment, human machine interface and function. In the full scope engineering simulator of HPR1000 unit, virtual reality control room interface has been designed as an extension of real control room implementation. The designer or operator can control and monitor the power plant in virtual reality environment, which just feels like real control room. It also can be used for other type of reactor by connecting to other simulator server and adding corresponding control room model in virtual reality software. With this preliminary application, control room simulator can be implemented in a short time and flexible for modification, which give designer more time and space for design verification and optimization. Once it applied in training simulator of nuclear power plant in future, it may provide a low cost and flexible option for operator training.

Commentary by Dr. Valentin Fuster
2018;():V001T04A005. doi:10.1115/ICONE26-81250.

Discontinuity appears in simulated training instruments of nuclear radiation reconnaissance which adopted ellipse numerical model. In order to solve the problem, an improved ellipse numerical model was proposed, in which the contaminated area was taken into count as a whole. The level of nuclear radiation at any position in the contaminated area can be calculated by the improved ellipse numerical model. On the basis of the improved ellipse numerical model, the architecture of the simulated instrument for training of nuclear radiation reconnaissance was proposed. The results of experiments showed that the improved ellipse numerical model not only had the main characteristics of the contaminated area but also successfully solved the problem of numerical discontinuity. Through adjusting the parameters of the contaminated area, the improved training instrument can adapt to different scope of nuclear radiation reconnaissance without any regional restriction.

Commentary by Dr. Valentin Fuster
2018;():V001T04A006. doi:10.1115/ICONE26-81300.

Digital Reactor Protection System (RPS) is one of the most important systems in instrumentation and control systems of Nuclear Power Plants (NPP). The reliability analysis of RPS plays an important role both in theory and engineering application. Traditional reliability methods, such as fault tree analysis and Markov chain theory, have many limitations in the research of RPS reliability, since the number of system states increases exponentially with the growth of system complexity.

Aiming at the reliability analysis of complex system like RPS, the Monte Carlo method simulates the system behaviors and obtains the reliability calculations through a large number of simulations. This paper takes a preliminary research of RPS reliability based on Monte Carlo Methods, including static reliability analysis based on Monte Carlo simulation of the behavior of every equipment in the RPS, and dynamic characters of the RPS based on the simulation of RPS period tests.

Commentary by Dr. Valentin Fuster
2018;():V001T04A007. doi:10.1115/ICONE26-81301.

A mathematical model is established for the High Temperature Hydrogen Detector (HTHD) used in severe accident conditions of nuclear power plants. The system error caused by the temperature difference of the internal wall between the working thermal conductivity cells and the reference conductivity cells is analyzed. Then the back propagation neural network algorithm is introduced to correct the system error. The test results show that BP neural network can effectively suppress this system error, and it has well generalization performance. At the same time, this method can be extended to correct measurement errors caused by other disruptive factors, such as supply voltage fluctuation, velocity variation due to pressure change, and interfering components (e.g. steam).

Commentary by Dr. Valentin Fuster
2018;():V001T04A008. doi:10.1115/ICONE26-81339.

The active magnetic bearing (AMB) in HTR-PM primary helium circulator (PHC) applies the inductive displacement transducer (IDT) to achieve the closed-loop feedback control. The magnetic anisotropy of the rotor material can be equivalent to structure defect of the rotor and affect the IDT measurement accuracy, leading to internal exciting vibration of the rotor. In this paper, the magnetic field analysis shows that the rotor magnetic anisotropy has effect on the sensor measurement and brings about the displacement measurement error. In the rotor-sensor experiment, the effect of rotor magnetic anisotropy on the IDT is obtained as a curve of magnetic error, which further explains that the rotor magnetic anisotropy will affect the dynamic measurement accuracy of the IDT. With the simulation result, it is observed that the displacement measurement error will lead to the internal exciting force of AMB-rotor system. The force will increase as the rotor speed increases, and bring about high-frequency vibration of the rotor.

Topics: Rotors , Vibration , Helium
Commentary by Dr. Valentin Fuster
2018;():V001T04A009. doi:10.1115/ICONE26-81486.

From the general industrial control system to the nuclear power plant control platform, the threat of information security has its own particularity more than continuity. The original dedicated system in general industrial area is gradually replaced by many common protocol, software and equipment. As a result, the security vulnerabilities are more likely to be used illegally. For a specific nuclear power plant digital control platform-NASPIC, the vulnerability analysis of platform is performed. Mainly two aspects of technology and management are to be carried out. For technical aspects, four categories problems-unauthorized execution, unauthorized write, unauthorized reading and reject service-are analyzed. Management is mainly about the lack of management strategy and strategy vulnerability. By analyzing the fragility of the instrument control platform, the key equipments, key channels and key modules are proposed. The qualitative and quantitative rules are deduced for evaluation of NASPIC information security.

Commentary by Dr. Valentin Fuster
2018;():V001T04A010. doi:10.1115/ICONE26-81570.

Digital instrumentation and control (I&C) systems are widely used in many industrial areas. In the recent years, the digitalization process for nuclear power plants has also been moving on rapidly. Full digital I&C systems are now adopted in almost all new constructed nuclear power plants. The architecture of a digital I&C system plays a pivotal role for the safety, reliability and security of the whole nuclear power plant. Moreover, for the advanced small modular reactors, both the reliability and extensibility of I&C systems are especially required.

Therefore, in this paper we propose a new architecture of the digital I&C systems based on the developed computing performance and communication technology. The control units and the data servers in the new proposed architecture are decentralized and working in a mutually redundant and distributed computing/storage way. Thus the architecture is with a flexible extensibility. Moreover, other control units or data servers can take over the functions of a certain number of failed ones. This characteristic benefits the system’s reliability significantly. The reliability of the new architecture is theoretically evaluated and the results demonstrate that it is much higher than that of the traditional architecture of I&C systems.

Commentary by Dr. Valentin Fuster
2018;():V001T04A011. doi:10.1115/ICONE26-81647.

With the development of control and information technology at NPPs, software reliability is important because software failure is usually considered as one form of common cause failures in Digital I&C Systems (DCS). The reliability analysis of DCS, particularly qualitative and quantitative evaluation on the nuclear safety-critical software reliability belongs to a great challenge. To solve this problem, not only comprehensive evaluation model and stage evaluation models are built in this paper, but also prediction and sensibility analysis are given to the models. It can make basement for evaluating the reliability and safety of DCS.

Commentary by Dr. Valentin Fuster
2018;():V001T04A012. doi:10.1115/ICONE26-81714.

Pressurized Water Reactor (PWR) nuclear power plant sump operator assisted program is applied to monitor unrecognized leaks of reactor coolant. It is very crucial to leak before break (LBB) protection and greatly affects the operational safety of nuclear reactors. In this paper, an algorithm of sump level operator assisted support program is proposed. Compared with the algorithm of traditional PWR, this algorithm adds the identification of working conditions and re-builds the leakage flow calculation method, which eliminates interference factors to the extent practical and improves the accuracy of the calculation results of unrecognized leakage flow.

Commentary by Dr. Valentin Fuster
2018;():V001T04A013. doi:10.1115/ICONE26-81829.

Steam generator (SG) is one of the key equipment of nuclear power units. Because of the large range of its loads changing, the water level control of SG effectively is an essential secure guarantee of nuclear power plants. SG is a complex system, besides imbalance and non-minimum phase characteristic, it also has the properties of nonlinearity, time-varying and with small stability margin. There are many difficulties in water level control of SG. Of which false water level and varying parameters are the most severe problems.

In this paper, first the water level features and the water level control principle of U-tube steam generator (UTSG) are introduced. Then mathematical model mechanism and both the static and dynamic characteristic of it water level are discussed. Finally various control methods are used for comparing the control effect.

Intelligent control is a type of control strategy which imitates human intelligence behavior. It is mainly aimed at the controlled plant with complicate model parameters, or which model structure hard to describe accurately by mathematical method. Cloud Model theory is proposed by Academician Li Deyi based on the idea of artificial intelligence with uncertainty. This theory focus on analyzing the uncertainty of control plant, realizes the uncertain conversion between qualitative concept and quantitative numerical by combining ambiguity and randomness. In the field of control technology, ambiguity and randomness make it difficult to establishing precise mathematical model of control plant, and become a bottleneck during the research of improving stability, accuracy and quickness of control system.

In this context, Cloud Model can be a good conversion between qualitative concept and quantitative numerical due to its ability of showing the uncertainty of qualitative concept which described by natural language. Under the action of external input, system control can be realized by inferencing according to the qualitative concept and uncertainty rules of Cloud Model. In this paper, the researched Cloud Model control system is based on normal distribution, because a large number of random events in nature and society obey or approximately obey normal distribution.

The rate of convergence of Cloud Model control is evidently faster than PID. Moreover, the capability of Cloud Model control in tracking, adapting, anti-interference and overcoming large time lag are apparently superior when comparing with the control effect of PID.

Topics: Nuclear power
Commentary by Dr. Valentin Fuster
2018;():V001T04A014. doi:10.1115/ICONE26-82048.

The complexity of modern safety critical systems is becoming higher with technology level growth. Nowadays the most important and vital systems of automotive, aerospace, nuclear industries count millions of lines of software code and tens of thousands of hardware components and sensors. All of these constituents operate in integrated environment interacting with each other — this leads to enormous calculation task when testing and safety assessment are performed.

There are several formal methods that are used to assess reliability and safety of NPP I&C (Nuclear Power Plant Instrumentation and Control) systems. Most of them require significant involvement of experts and confidence in their experience which vastly affects trustworthiness of assessment results.

The goal of our research is to improve the quality of safety and reliability assessment as result of experts involvement mitigation by process automation. We propose usage of automated FMEDA (Failure Modes, Effects and Diagnostic Analysis) and FIT (Fault Insertion Testing) combination extended whith multiple faults approach as well as special methods for quantitative assessment of experts involvement level and their decisions uncertainty. These methods allow to perform safety and reliability assessment without specifying the degree of confidence in experts.

Traditional FMEDA approach has several bottlenecks like the need of manual processing of huge number of technical documents (system specification, datasheets etc.), manual assignment of failure modes and effects based on personal experience. Human factor is another source of uncertainty. Such things like tiredness, emotional disorders, distraction or lack of experience could be the reasons of under- and over-estimation.

Basing on our research in field of expert-related errors we propose expert involvement degree (EID) metric that indicates the level of technique automation and expert uncertainty degree (EUD) metric which is complex measure of experts decisions uncertainty within assessment. We propose usage of total expert trustworthiness degree (ETD) indicator as function of EID and EUD.

Expert uncertainty assessment and Multi-FIT as FMEDA verification are implemented in AXMEA (Automated X-Modes and Effects Analysis) software tool. Proposed Multi-FIT technique in combination with FMEDA was used during internal activities of SIL3 certification of FPGA-based (Field Programmable Gate Array) RadICS platform for NPP I&C systems. The proposed expert trustworthiness degree calculation is going to be used during production activities of RPC Radiy (Research and Production Corporation).

Our future work is related to research in expert uncertainty field and extension of AXMEA tool with new failure data sources as well as software optimization and further automation.

Commentary by Dr. Valentin Fuster
2018;():V001T04A015. doi:10.1115/ICONE26-82230.

This paper researched the safety functional requirements analysis and the allocation of functions between man and machine for the nuclear power plant. The safety functional requirements are identified from accident handling needs and refined from system configuration consideration. Through the analysis of design conditions, some safety features were extracted to mitigate accidents. Then, components (e.g. pumps, valves, tanks) were determined to implement each of the safety features at the system design stage. At this stage, some implicit safety features, which could not be obtained directly from the accident analysis, were added, according to the specific conditions of system configuration and operation. Finally, after further judgement on possible inconsistency, a complete list of safety functions for the nuclear power plant was formed. As an illustration, this paper provided a list of safety functions related to the safety injection function, and a list of equipment for the safety injection system. Furthermore, these identified safety functions, were appropriately allocated between man and machine, to be performed either by system components automatically, or by operators locally or remotely from the control room, or under the cooperation of operators and system components. Seven factors were considered in the allocation: a) performance requirements; b) the capability or limits of man and machine; c) existing practices; d) operating experience; e) management requirement; f) technical feasibility; g) cost. The allocation of functions for the safety injection system was validated using a simulator.

Commentary by Dr. Valentin Fuster
2018;():V001T04A016. doi:10.1115/ICONE26-82270.

Safety assessment of nuclear power plant instrumentation and control systems (NPP I&Cs) is a complicated and resource consuming process that is required be done so as to ensure the required safety level and comply to normative regulations. A lot of work have been performed in the field of application of different assessment methods and techniques, modifying them and using their combinations so as to provide unified approach in comprehensive safety assessment. Anyway, performed research have shown there are still challenges to overcome, including rationale and choice of the safety assessment method, verification of assessment results, choosing and applying techniques that support safety assessment process, especially in the nuclear field. In our work we present developed framework that aggregates the most appropriate safety assessment methods typically used for NPP I&Cs. Key features that this framework provides are the formal descriptions of all required input information for every safety assessment method, possible data flows between methods, possible output information for every method. Such representation allows to obtain possible paths required to get necessary indicators, analyze the possibility to verify them by application of different methods that provide same indicators etc. During safety assessment of NPP I&Cs it is very important to address software due to its crucial role in I&C safety assurance. Relevant standards like IEC 60880 [1] and IEC 62138 [2] provide requirements for software related activities and supporting processes in the software safety lifecycle of computer-based I&C systems of nuclear power plants performing functions of safety category A, B and C, as defined by IEC 61226 [3]. Requirements and frameworks provided by IEC 60880 and IEC 62138 for the nuclear application sector correspond to IEC 61508, part 3 [4]. These standards define several types of safety related software and specify particular requirements for each software type. So as to verify software and confirm correspondence to required safety level, different techniques are suggested in normative documents. We share our experience obtained during software failure modes and effect analysis (software FMEA) and software fault insertion (software FIT) processes into FPGA-based platform, NPP I&C systems based on that platform, and RPCT, integrated development environment used by RPC Radiy and end users to design user application logic, specify hardware configuration etc. We apply software FIT to outputs of RPCT, considering source code, configuration files and firmware files. Finally, we provide a case study of application the developed safety assessment framework and software FMEA/FIT practices during practical assessment of FPGA-based NPP I&C system.

Commentary by Dr. Valentin Fuster
2018;():V001T04A017. doi:10.1115/ICONE26-82290.

The aim of this work is to show a preliminary investigation on the propagation of electromagnetic fields generated by wireless technologies inside a nuclear facility or power plant. First, a survey of currently proposed wireless technologies for nuclear facilities and plants has been carried out. Then, for selected scenarios, the electromagnetic field propagation has been studied by means of electromagnetic simulation tools, and the presence of the nuclear environment has been simulated by properly modeling environmental parameters and engineered barriers. The choice of the proper simulation techniques and tools is mandatory in order to simulate the effect of the realistic environment on the propagation. Accordingly, the feasibility of wireless technologies application at nuclear facilities can be assessed on the basis of results achieved from simulated scenarios. The goal is to analyze, for selected scenarios, possible issues due to the propagation of an electromagnetic field in presence of simplified barriers mimicking the real nuclear environment. This approach can provide indications on how to deploy potential benefits of wireless technologies in a nuclear environment, evaluating pros and cons of the investigated technologies.

Commentary by Dr. Valentin Fuster
2018;():V001T04A018. doi:10.1115/ICONE26-82377.

Diversity approach is used to decrease risk of common cause failure (CCF) of Nuclear Power Plant (NPP) Instrumentation and Control systems (I&Cs). Application of a multi-diversity, i.e. a few different types of version redundancy allows minimizing CCF risk. On the other side, implementation of diversity increases cost and complicates maintenance of multi-version I&Cs. Hence, it is important to find optimal solution according with criteria “required level of diversity (safety) / minimal cost and maintenance complexity. Modern FPGA technology creates additional possibilities to meet requirements of the standards (such as NUREG/CR-7007, IEEE Std 7-4.3.2-2016, IAEA SSR-2/1:2016, IAEA NP-T-3.17:2016 and others) by developing main and diverse subsystems on the basis of the same FPGA platform. Existing diversity normative base should be enhanced in three directions — scope, depth and rigor to provide more detailed description of possible applied techniques and tools for quantitative assessment.

The goals of the paper which overviews practical issues of diversity application are the following:

- present extended classification of diversity considering additional types of version redundancy for FPGA platform based I&Cs (logical processing equipment, life cycle, logic/algorithm etc.) in comparing to NUREG7007;

- describe the modified technique of diversity assessment taking into account three and more levels of diversity classification;

- illustrate and discuss variants of assurance of the required degree of diversity by use of the RadICS FPGA platform to develop main and diverse subsystems.

The classification is specified considering diversity of hardware and FPGA designs. In particular, diversity of hard logic and soft processors, interfaces and buses, self-diagnostics means and others are described and embedded into NUREG/CR-7007 classification.

The NUREG7007-based diversity assessment techniques supporting all stage of analyzing options are discussed, and algorithms for versions choice are described. This technique takes into account more detailed specification of diversity classification (for types, subtypes and sub-subtypes of diversity for logic diversity, logic processing equipment diversity and others) and options to evaluate weight coefficients.

Case study is based on description of two options of RadICS FPGA platform application to develop two-version NPP I&C, which meets standard requirements to diversity.

Commentary by Dr. Valentin Fuster
2018;():V001T04A019. doi:10.1115/ICONE26-82483.

Nuclear safety is one of the key issues for a nuclear power plant (NPP). The alarm system plays a critical role for the safe and efficient operation of an NPP which affects the correctness and efficiency of the operators in dealing with the accidents. It is even more important for the alarm system of a multi-modular NPP which has more than one reactor modules in a single unit because all the modules are usually monitored in the same main control room. The alarm generation and display mechanism of a typical multi-modular NPP, the High Temperature Gas-Cooled Reactor - Pebble bed Module (HTR-PM), is analyzed in this paper which has two reactor modules coupled to one steam turbine. Three operators are responsible for the operation of two nuclear islands and a conventional island, respectively. The alarm generation and display processes will be discussed in this paper. Firstly, the architecture of the RPS and the alarm system related to the red and yellow alarms is introduced. Then the generation and display mechanism of the red and yellow alarms is proposed. A protection variable of a design basis accident is given as an example for the alarm signal handling. The characteristics of the alarm system are then discussed. More optimization directions on the alarm design for multi-modular NPPs are proposed in the end.

Topics: Accidents
Commentary by Dr. Valentin Fuster
2018;():V001T04A020. doi:10.1115/ICONE26-82498.

There was a common use of instrument and control (I&C) system based on analog technology in design and construction of nuclear power plant built more than ten years ago. With the development and update of automation technology, digital control system has almost completely replaced the older generation technology in many areas of industry. For the nuclear power plant still using the analog I&C system, it caused the reduction of the production lines of related components and the lack of specified technical engineer. Along with the aging and obsolescence of analog technology equipment, the unsustainability of spares prompted the owners of nuclear power plant to implement modernization using digital technologies. Different from the design of digital control system in new nuclear power plant, the modernization project design is limited by the setting of the original system. Therefore, most owners adopt the function alternative strategy to implement the upgrading project. This strategy can effectively solve the problem of spares shortage, but it is difficult to fully elaborate the advantages of digital control system. Based on the technical characteristics of digital control system and the design experience derived from new nuclear power project construction, this paper puts forward the optimized design measures, under the limitation of the old power plant, to enhance the safety and economy of the nuclear power plant after final digital upgrading.

Commentary by Dr. Valentin Fuster
2018;():V001T04A021. doi:10.1115/ICONE26-82556.

The 200 MWth nuclear heat reactor II (NHR200-II) is a typical integral pressurized water reactor (iPWR) being developed by the institute of nuclear and new energy technology (INET) in Tsinghua university. The NHR200-II, which has inherent safety features such as full-range natural circulation, passive residual heat removal, self-pressurization and control rod hydraulically driving, can be adopted as a clean base-load energy source for a sea-water desalination plant having the process of multi-effect desalination with thermal vapor compression (MED-TVC). Dynamic modelling of the sea-water desalination plant coupled by the NHR200-II and MED-TVC is necessary for the design of its plant control strategy, which is important for the stable and efficient operation. In this paper, a lumped parameter dynamic model of NHR200II-based sea-water desalination plant with the process of MED-TVC is proposed based upon the conservation laws of mass, momentum and energy. The modeling verification in both the steady-state and open-loop dynamic-state are given, which show the suitability of applying this model for control system design. Finally, the closed-loop responses in the case of power-level maneuver from 100% to 50% full power is given.

Commentary by Dr. Valentin Fuster
2018;():V001T04A022. doi:10.1115/ICONE26-82558.

Passive residual heat removal system (PRHRS) is of great significance for reactor shutdown safety. The PRHRS of a small modular reactor, such as the integral pressurized water reactor (iPWR) and the modular high temperature gas-cooled reactor (MHTRG), is composed of the primary loop (PL), intermediate loop (IL) and air-cooling loop (AL). The AL is a density-difference-driven natural circulation caused by the difference of air temperature at the inlet and outlet of the air-cooling tower. Thus, it is possible to adopt the air flow in AL to generate electricity for post-shutdown reactor monitoring. In this paper, a novel residual heat electricity generation system (RHEGS), which is composed of the PRHRS and a vertical wind generator installed in the air-cooling tower, is proposed for the power supply of post-shutdown monitoring instruments. To verify the feasibility of practical implementation, the dynamical model of this newly designed RHEGS including the dynamics of PRHRS, windmill, rotor as well as doubly-fed induction generator (DFIG) are all given. Then, both steady-state and transient verification for the RHEGS of a nuclear heating reactor NHR200-II plant with a rated thermal power of 200 MWth is carried out, which shows that the output active power of RHEGS can be 20∼30kW which is about 1% the residual heat flux and can fully meet the power requirements of post-shutdown monitoring instruments.

Commentary by Dr. Valentin Fuster
2018;():V001T04A023. doi:10.1115/ICONE26-82561.

The nuclear power plant with two modular high-temperature gas-cooled reactors (HTR-PM) is under construction now. The control room of HTR-PM is designed. This paper introduces the alarm displays in the control room, and describes some verification and validation (V&V) activities of the alarm system, especially verification for some new human factor issues of the alarm system in the two modular design. In HTR-PM, besides the regular V&V similar to other NPPs, the interference effect of the alarm rings of the two reactor modules at the same time, and the potential discomfort of the two reactor operators after shift between them are focused. Verifications at early stage of the two issues are carried on the verification platform of the control room before the integrated system validation (ISV), and all the human machine interfaces (HMIs) in the control room, including the alarm system are validated in ISV. The test results on the verification platform show that the alarm displays and rings can support the operators understand the alarm information without confusion of the two reactors, and the shift between the two reactor operators have no adverse impact on operation. The results in ISV also show that the alarm system can support the operators well.

Commentary by Dr. Valentin Fuster

Innovative Nuclear Power Plant Design and SMRs

2018;():V001T13A001. doi:10.1115/ICONE26-81002.

A novel combined small modular reactor (SMR) and gas turbine cycle is presented. This SMR-GT cycle is modeled using fundamental thermodynamic relationships and compared to existing state-of-the-art power generation cycles. The SMR-GT cycle includes an 82 MWe SMR cycle that is combined with a 54 MWe gas turbine cycle. A heat exchanger is used to extract energy from the gas turbine exhaust to create superheated main steam and provide reheat downstream of the LP turbine. This results in a 32 MWe increase in the SMR cycle for total unit output of 136 MWe.

Comparisons of thermal efficiency, heat rate, CO2 emissions, and net generation are made between a stand-alone SMR, a typical combined cycle gas turbine (CCGT), standalone gas turbine and the combined SMR-GT cycles. Several advantages of the SMR-GT cycle are discussed.

In addition, the rapid deployment of a gas turbine allows for a power station to deliver power and earn revenue prior to completion of the more complex SMR portion of the plant. The SMR portion of the cycle also reduces the overall fuel cost volatility associated with gas turbine based power station.

Commentary by Dr. Valentin Fuster
2018;():V001T13A002. doi:10.1115/ICONE26-81159.

Compared with the land-based nuclear power plant, the operating conditions of offshore nuclear power plant (ONPP) are much more complicated. For example, the barge-mounted platform malfunction, which is as important as the natural events and human events, should be considered in the plant safety analysis,. As a result, a two dimension operating condition coupled with barge and reactor status should be considered in the development of relevant power plant operating procedures. On the other hand, the beyond design basis hazards induced by the combination of unique and unanticipated external events of ONPP may lead to a blind area to both traditional and two dimension procedures mentioned above. Due to the insufficiency of existing operating condition and relevant procedures to tackle with the above events mentioned, an expanded operation strategy, namely the beyond design basis hazards and the extended ultimate response measures, is developed, Injecting sea water into reactor pressure vessel directly after primary system depressurized and venting the containment when necessary, formed the basis of ultimate response measure, which was proposed by Taiwan Power Company after Fukushima Accident. Considering the offshore and barge-mounted features, the ultimate response measure can be extended to include sea water injection into steam generator indirectly through secondary side passive residual heat removal lines and reactor cabin flooding by sea water through Kingston valves, to rebuild a newly, hierarchical one. Finally, the extended ultimate response measures, provided mainly for the plant command staff and operators, are analyzed utilizing thermal-hydraulic integral computer code preliminarily, to prove the effectiveness of the system configuration and operating strategy. It is concluded that injecting sea water into steam generator can remove the decay heat effectively, and the sensitivity study shows that operator intervention is good enough in accident mitigation.

Commentary by Dr. Valentin Fuster
2018;():V001T13A003. doi:10.1115/ICONE26-81188.

The United Kingdom (UK) Small Modular Reactor (SMR) is being developed by a Rolls-Royce led consortium to provide a market driven, affordable, low carbon energy, generation capability. The UK SMR is a Pressurised Water Reactor (PWR) design based on proven technology with a high level of safety achieved through multiple active and passive systems. This paper presents the approach that has been taken in the early design phases of the pressure vessels for the UK SMR. It considers the key design principles e.g. standardisation, simplification and design for manufacture, inspection and assembly which are being applied to enable the cost and lead-time reductions which are necessary for the UK SMR to be a viable alternative to larger conventional nuclear plants. The Reactor Pressure Vessel (RPV) is used as an example to illustrate some of the key design requirements which need to be addressed. Nuclear components are required to be designed and constructed to standards which are commensurate with the significance of the safety functions which they perform. This paper covers the practice established in the UK of designing to Incredibility of Failure for those components with catastrophic failure modes such as the RPV. It describes the additional features including more stringent materials specification and testing, additional defect tolerance studies and the qualification of manufacturing inspections which need to be addressed in the design to satisfy the high reliability claim.

Topics: Design
Commentary by Dr. Valentin Fuster
2018;():V001T13A004. doi:10.1115/ICONE26-81222.

A FLiNaK high temperature test loop, which was designed to support the Thorium Molten Salt Reactor (TMSR) program, was constructed in 2012 and is the largest engineering-scale fluoride loop in the world. The loop is built of Hastelloy C276 and is capable of operating at the flow rate up to 25m3/h and at the temperature up to 650°C. It consists of an overhung impeller sump-type centrifugal pump, an electric heater, a heat exchanger, a freeze valve and a mechanical one, a storage tank, etc. Salt purification was conducted in batch mode before it was transferred to and then stored in the storage tank. The facility was upgraded in three ways last year, with aims of testing a 30kW electric heater and supporting the heat transfer experiment in heat exchanger. Firstly, an original 100kW electric heater was replaced with a 335kW one to compensate the overlarge heat loss in the radiator. A pressure transmitter was subsequently installed in the inlet pipe of this updated heater. Finally, a new 30kW electric heater was installed between the pump and radiator, the purpose of which was to verify the core’s convective heat transfer behavior of a simulator design of TMSR. Immediately after these above works, shakedown test of the loop was carried out step by step. At first the storage tank was gradually preheated to 500°C so as to melt the frozen salt. Afterwards, in order to make the operation of transferring salt from storage tank to loop achievable, the loop system was also preheated to a relatively higher temperature 530°C. Since the nickel-base alloy can be severely corroded by the FLiNaK salt once the moisture and oxygen concentration is high, vacuum pumping and argon purging of the entire system were alternatively performed throughout the preheating process, with the effect of controlling them to be lower than 100ppm. Once the salt was transferred into the loop, the pump was immediately put into service. At the very beginning of operation process, it was found that flow rate in the main piping could not be precisely measured by the ultrasonic flow meter. Ten days later, the pump’s dry running gas seal was out of order. As a result, the loop had to be closed down to resolve these issues.

Commentary by Dr. Valentin Fuster
2018;():V001T13A005. doi:10.1115/ICONE26-81271.

SPICRI (State Power Investment Central Research Institute) is developing a new conceptual design of heating-reactor, named Heating-reactor of Advanced low-Pressurized and Passive safetY system (HAPPY), which is targeted for the district heating, desalination of seawater, and other heat applications. It is a 200MWth two-loop low-pressurized water reactor with low thermal parameters. The whole reactor vessel is deployed inside a shielding and cooling pool with thermal insulation measure. The conceptual design of HAPPY is described in this paper, including the design criteria, safety features, main parameters and main components.

A preliminary safety analysis is carried out to provide a reference for the design and optimization of HAPPY. In this paper, four different LOCA analyses are described and compared. The results show that the current design can deal well with all the selected LOCA scenarios and the effectiveness of the safety systems is proved.

Topics: Safety , Heating
Commentary by Dr. Valentin Fuster
2018;():V001T13A006. doi:10.1115/ICONE26-81309.

Studies on the suppression of the reactivity of sodium itself have been performed on the basis of the concept of suspended nanoparticles in liquid sodium (sodium nanofluid). According to the theoretical and experimental results of studies for sodium nanofluid, velocity and heat of sodium nanofluid reaction with water (sodium nanofluid/water reaction) are lower than those of the pure sodium/water reaction. The analytical model for the peak temperature of a sodium nanofluid/water reaction jet has been developed by the authors in consideration of these suppression effects. In this paper, the prediction method for mitigation effects on damage of adjacent tubes in steam generator tube rupture (SGTR) accidents is developed by applying this analytical model for the peak temperature of the reaction jet. On the assumption that the sodium nanofluid is used for the secondary coolant of sodium-cooled fast reactor (SFR), mitigation effects under the design basis accident (DBA) condition and the design extension condition (DEC) of SGTR are estimated by using this method. The results indicate a clear possibility to reduce the number of damaged tubes and to suppress the pressure generated in SGTR accidents by using sodium nanofluid as the secondary coolant.

Commentary by Dr. Valentin Fuster
2018;():V001T13A007. doi:10.1115/ICONE26-81311.

Rolls-Royce and a UK Consortium are progressing the design and development of a Small Modular Reactor (SMR) Power Station. The SMR programme is a phased design cycle, progressing through the Rolls-Royce gated review process. The project aims to deploy the first of a kind SMR in the UK by the end of the next decade. In this paper, the development methodology for the reactor core design is discussed, along with a selection of the key technical challenges that have been addressed during the concept design phase. Lessons learned from past projects have been identified, to help improve the design efficiency for the SMR.

The concept design has been developed in an iterative fashion, with different analysis disciplines carefully integrated around a common set of objectives. Key economic requirements for an SMR core include maximising fuel economy, cycle length and thermal power while remaining small enough to enable a modular build approach. Top-level safety requirements include control of reactivity, control of core temperature and control of release of radioactivity/radioactive material.

A set of surrogate design limits has been used alongside the true safety limits to avoid the need for detailed transient subchannel or fuel performance analysis in this phase. This has allowed the design to mature and be characterised very quickly, while also maintaining high confidence that all performance and safety requirements will be met when detailed analyses are undertaken.

This paper describes the different analyses that have been undertaken to date, including a variety of reactor physics and thermal hydraulics calculations. The paper discusses the limits used, how they have been used to optimise the design solution and why they provide high confidence in the core design’s performance.

Commentary by Dr. Valentin Fuster
2018;():V001T13A008. doi:10.1115/ICONE26-81318.

A new multi-purpose modular small pressurized water reactor with once-through steam generators is being designed in China. Its key parameters are different from traditional large pressurized water reactor. There are sixteen once-through steam generators divided into two groups inside of the pressure vessel. The four coolant pumps are located on the periphery of the pressure vessel. The coolant is heated by the core and transported the heat to the secondary loop by once-through steam generators. The superheated steam is generated, and its dynamics are different from those of U-tube steam generators. The relationship between the reactor and turbine is also complicated and needs to investigate. The control strategies of traditional large pressurized water reactor cannot be applied directly to the small reactor with once-through steam generators. Therefore, it is necessary to investigate suitable control strategies of the multi-purpose modular small reactor with once-through steam generators.

Three control strategies are proposed and investigated in this study: turbine-leading, reactor-leading and feedwater-leading. With the reactor-leading strategy, the reactor power is adjusted by moving the control rod. The coolant temperature follows the change of the reactor power. Feedwater flow is applied to regulate the steam pressure. The steam flow rate follows the change of the feedwater flow rate to satisfy the demand power. With the turbine-leading strategy, the steam valve is adjusted which will influence the steam flow to satisfy the demand power. The feedwater-leading control strategy is adjusting the feed water flow rate corresponding to the demand power which has been measured. And reactor power and turbine load vary with feedwater flow rate. Input-output pairings of the control systems are determined based on the different strategies and proportion-integral-derivative (PID) controllers are tuned to meet the control requirements.

To evaluate the performance of control strategies, power maneuvering events including a 10%FP (Full Power) step change and a ramp change with a rate of 5%FP/min are simulated. The processes of important control parameters varying with time are compared and evaluated to obtain the suitable one. Conclusions can be drawn from the simulation analyses of the control performance. The reactor-leading control strategy is best for the base-load operation. The turbine-leading control strategy is more suitable for load-following operation. The feedwater leading control strategy can be applied to load-following operation with smooth load adjustment.

Commentary by Dr. Valentin Fuster
2018;():V001T13A009. doi:10.1115/ICONE26-81331.

In sodium-cooled fast reactors, the core is not arranged in its most reactive configuration. In this case, when the fuel melts to form a molten pool, the recriticality may occur by positive reactivity insertion due to core compaction. To prevent such recriticality, special devices of the fuel subassembly structure for discharging the molten fuel from the core region, have been investigated by the Japan Atomic Energy Agency (JAEA). On the other hand, the inherent feature of core geometry and the neutron characteristics may provide the similar effect to prevent such recriticality. The purpose of this study is to design the core specification its deformation in CDA causes negative feedback to subcritical condition, without any fuel discharge device.

The convex shaped core has the longer fuel length in the inner-core region and the shorter fuel in the outer-core region. Therefore, the core geometry as intact status has a lower neutron leakage effect. When the fuel melts in CDA, the core height is compacted and negative reactivity insertion is expected during molten pool formation. The convex shaped core is based on the large-scale cylindrical homogeneous core (3,600 MWth, 4.95m in core diameter, and 0.75m in core height). The calculation showed that the compaction of cylindrical core leads to a reactivity gain, whereas the convex shaped core results in negative reactivity effect.

In this geometry, both inner-core and outer-core are divided into two regions. Furthermore, we introduced the smaller diameter pin for inner-core and keep uniform Pu enrichment for all regions. The smaller diameter pins in high importance region are effective for flat-distribution. Through pin diameter survey, we confirmed the advantages of smaller diameter pin, such as reducing pressure loss of core coolant and decreasing the height of molten pool.

Commentary by Dr. Valentin Fuster
2018;():V001T13A010. doi:10.1115/ICONE26-81550.

Light-water cooled Small Modular Reactors (SMRs) are a potential game-changing technology for energy supply. The potential benefits of SMRs are however conditional on solving the key standardisation and construction issues that have troubled large reactor (LR) projects, which have in turn led to high build costs and long project durations.

Initiatives to determine the build schedule of SMRs are hindered by a lack of SMR construction experience and related data. The methodology used in this paper, to deal with the lack of SMR-specific data, draws conclusions about SMRs based on data from actual large pressurised water reactor (PWR) construction experience.

It is expected that SMR build schedules can be greatly reduced because of the smaller physical size of structures, fewer components, and other size-related features. However, the construction work space will be more constrained, which could negatively impact build durations. As a result, simple geometric scaling and reduction arguments cannot necessarily be applied to SMR schedules. This paper defines the key areas in which SMR construction differs from LRs, such as smaller geometries as well as modularised and standardised build processes, and describes how these differences might be expected to impact build duration quantitatively.

The model developed in this paper presents an approach to determining SMR build schedule durations for a range of reactor sizes. It starts with an LR build schedule based on real data from the UK’s only PWR, Sizewell B. The available data are used to establish a reference point for a non-modular, stick-built SMR schedule. This scheduling approach assumes that, for each major element, part of the time spent on fabrication and installation tasks will vary with reactor size while the remaining fraction will remain constant regardless of reactor size (e.g. due to quality assurance and commissioning tasks). The accuracy of the model generated here is assessed against available construction data and models from a range of actual reactor build projects.

The objective of this work is to consider how modularisation can reduce build schedule of SMRs of varying size, by employing modular design and construction principles to both remove tasks that are of long duration from the critical path and to improve construction productivity. Mechanisms by which modularisation reduces build schedule are investigated. Build reduction scenarios are presented based on analysis and subsequent modularisation of the SMR critical path and are compared with other related analyses.

Commentary by Dr. Valentin Fuster
2018;():V001T13A011. doi:10.1115/ICONE26-81561.

High-temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM), which is designed by Tsinghua university of China, is under construction in Shidao Bay of China. It will be the world’s first pebble-bed type modular HTGR commercial demonstration plant. In HTR-PM project, steam-Rankine cycle has used in the power conversion system because it represents current state-of-the-art technology. Meanwhile, helium turbine for HTGR has been investigated for many years in Tsinghua University. Mock-up machine for HTR-10GT has been built. Helium turbine for 250MW HTGR, which is based on HTR-PM, has completed conceptual designed. However, supercritical carbon dioxide (S-CO2) Brayton cycle has shown to have great potentials for future HTGR technology in recent years because of its critical properties. Helium turbine cycle and S-CO2 Brayton cycle are both candidates for future HTGR. Therefore, comparative study is conducted in this paper. Comparison is focused on achievable efficiencies for each cycle mentioned above and on cycle layout with respect to simplicity and compactness, which primarily determines capital cost. Firstly, the physical model for helium turbine cycle with recuperator and intercooler is built and cycle performance is analyzed based on the parameters of HTR-PM. Then the model for S-CO2 Brayton cycle with recompression is also built and the cycle efficiency is analyzed with the same parameters of HTR-PM. Secondly, comparison between helium turbine cycle and S-CO2 Brayton cycle is made from the view of thermodynamics. Moreover, parameters optimization of both cycles based on HTR-PM is carried out. At last, advantage and drawback of both cycles are discussed from the engineering point. In conclusion, cycle simplicity and technology maturity of helium turbine cycle are better than S-CO2 Brayton cycle. But on the other side, smaller size equipment and less compression work of S-CO2 Brayton Cycle are more competitive than helium turbine cycle. Helium turbine with higher temperature and S-CO2 Brayton Cycle with higher pressure can achieve higher efficiency than steam Rankine cycle.

Commentary by Dr. Valentin Fuster
2018;():V001T13A012. doi:10.1115/ICONE26-81604.

The key characteristics of small modular reactors (SMRs), as their name emphasized, are their size and modularity. Since SMRs are a family of novel reactor designs, there is a gap of empirical knowledge about the cost/benefit analysis of modularization. Conversely, in other sectors (e.g. Oil & Gas) the empirical experience on modularization is much greater. This paper provides a structured knowledge transfer from the general literature (i.e. other major infrastructure) and the Oil & Gas sector to the nuclear power plant construction world. Indeed, in the project management literature, a number of references discuss the costs and benefits determined by the transition from the stick-built construction to modularization, and the main benefits presented in the literature are the reduction of the construction cost and the schedule compression. Additional costs might arise from an increased management hurdle and higher transportation expenses. The paper firstly provides a structured literature review of the benefits and costs of modularization divided into qualitative and quantitative references. In the second part, the paper presents the results of series of interviews with Oil & Gas project managers about the value of modularization in this sector.

Commentary by Dr. Valentin Fuster
2018;():V001T13A013. doi:10.1115/ICONE26-81651.

Small Modular Reactors (SMRs) are economically competitive nuclear power systems aimed to provide sustainable clean safe and reliable nuclear energy free from the risk of fissile material proliferation. They are smaller versions of the present-day large nuclear power reactors with additional design simplifications, improved and reliable passive safety systems incorporating innovative concepts. With the intrinsic advantage of high power density and carbon-free emissions, SMRs and especially their innovative features are the signals for a nuclear comeback, or in Dr Alvin Weinberg’s words “the second nuclear era” in many ways. According to some estimates, there could be up to 96 SMRs by 2030. This paper addresses three vital areas to the understanding of the SMR’s in emerging global environments: (i) design, (ii) production of plutonium during operation, and (iii) their scope of applications. A representative, though very small SMR, Toshiba’s innovative 4S design is used for presenting estimates of plutonium production which are applicable to other SMRs as well. To better understand the viability of SMRs, this work considers the emerging developers, exporters and markets where SMRs can make significant improvements to the overall socio-economic development of societies challenged with formidable barriers.

Commentary by Dr. Valentin Fuster
2018;():V001T13A014. doi:10.1115/ICONE26-81705.

With the development of small modular reactors, the hydrogen risk reducing technology cannot be ignored. Special safety facilities of small modular reactor (SMR) are investigated and studied, and a serious accident analysis program model for SMR is established. The combination of Pre-inerting and hydrogen recombination was used to control the hydrogen risk. The effectiveness of the hydrogen control system is analyzed by using the GASFLOW program. The results show that the volume fraction of hydrogen in the containment dome is higher than that in the other parts of the containment during the calculation. Because of the small size and tight internal structure, hydrogen accumulates in the narrow channel, which increases the hydrogen concentration in the local channel. Inerting reduces the concentration of oxygen in the containment and effectively controls the possibility of flame acceleration and blasting transition in high hydrogen concentration regions.

Commentary by Dr. Valentin Fuster
2018;():V001T13A015. doi:10.1115/ICONE26-81718.

A new conceptual design of intermediate heat exchanger (IHX) is proposed for application to the gas turbine high temperature reactor system (GTHTR300C) which is being developed by Japan Atomic Energy Agency (JAEA). The GTHTR300C cogenerates hydrogen using the iodine-sulfur (IS) hydrogen production process and electric power using gas turbine. The IHX is used to transport high temperature heat from the nuclear reactor to the hydrogen plant. The IHX proposed in this paper is a horizontal design as opposed to conventional vertical design. Therefore, JAEA investigated the advantage of the horizontal IHX and the economic evaluation when scaling up from conceptual design of high temperature engineering test reactor (HTTR) / IHX to GTHTR300C. To meet the performance requirement, both thermal and structural designs were performed to select heat transfer tube length, tube bundle diameter, insulation thickness, and the length of shell support in a horizontal pressure vessel. It is found that the length of the heat exchanger tube can be shortened and the superalloy-made center pipe structure can be eliminated, which results in reducing the quantity of construction steel by about 30%. Furthermore, the maximum stress concentration in the tubes is found to be significantly reduced such that the creep strength to withstand continuous operation is extended to 40 years, equaling the nuclear reactor life time, without replacement.

Commentary by Dr. Valentin Fuster
2018;():V001T13A016. doi:10.1115/ICONE26-81985.

Small Modular Reactor (SMR) is getting more and more attention due to its safety and multi-purpose application. License structure is an important issue for SMR licensing. Modular design, construction and operation, shared or common structure, system and components (SSC) challenge existing large light water reactor license structure. Existing nuclear power plant license structure, characteristics of SMR and its effect on license structure, and research progress of U.S Nuclear Regulatory Commission (NRC) are analyzed, SMR license structure in China are proposed, which can be used as a reference for SMR R&D, design and regulation.

Commentary by Dr. Valentin Fuster
2018;():V001T13A017. doi:10.1115/ICONE26-82079.

High breeding with light water cooling has been studied for decades, though is not easy to be achieved. The main obstacle is the moderating effect of light water, which softens the neutron spectrum. To harden the neutron spectrum and thereby to enhance the fuel utilization or even to achieve breeding with light water cooling, the tight-lattice assembly was proposed and applied to High Conversion LWRs. Nonetheless, none of them achieved high breeding. Until recently, the tightly packed fuel assembly (TPFA) is designed for the purpose of high breeding. The ratio of hydrogen atoms to heavy metal atoms (H/HM) in this assembly is significantly reduced to be less than 0.1. Super Fast Breeding Reactor (Super FBR) adopts TPFA and achieves breeding performance with compound system doubling time (CSDT) of 43 years. In this study, the breeder BWR core also applies TPFA and achieves CSDT of 50 years. BWR is one type of the most extensively built reactors in the world, with abundant operation experience and mature technologies. Breeder BWR is considered to be capable of being incorporated into the current BWR plants with a handful of modifications, thus obtaining optimal economy.

Commentary by Dr. Valentin Fuster
2018;():V001T13A018. doi:10.1115/ICONE26-82231.

The gas/liquid metal magnetohydrodynamic generator (G/LM-MHD) with the mixture of gas and liquid metal as working fluids shows a promising future due to recent development of liquid metal cooled nuclear reactors. Previous efforts on the G/LM-MHD energy conversion systems have predicted a higher efficiency than traditional thermodynamics cycle. However, most of the earlier studies focus on the conception designs, feasibility analysis and preliminary experiments, while less attention paid on some specific problems such as the bubble phenomenon in the two-phase flow. Therefore, this paper deals with numerical study on the performance characteristics of the gas/liquid metal two-phase flow in an ideal Faraday-type MHD channel, of which the geometry structure is 30 × 30 × 80 mm cuboid segmentary style. The conductive mixture fluid is composed of nitrogen as the gas phase and gallium as the liquid phase (N2/Ga). The temperature at the channel inlet is about 600 K considering the heat transfer after the mixing chamber, while the inlet velocity is around 10 m/s and gas volumetric void fraction is 50%. The external magnetic field is assumed as 4 Tesla adopting the superconducting technology, which seems essential for MHD industrial practice. Then the simulation is accomplished using a modified two-phase mixture model considering the electromagnetic influence. The simulation results show that the distribution of temperature changes much weaker than pressure and velocity, which agrees with earlier one-dimension analysis. On the other hand, the results characterizes clearly the increase of the void fraction close to the electrodes, which can explain intuitively the decrease of the power-generating capacity. Besides, the power output is predicted to reach maximum 22.5 kW while the voltage is 1.2 V and the power density can be 312.5 MW/m3 which is far beyond traditional steam turbines. This study shows a promising future of the gas/liquid metal MHD generator for the small nuclear plants and power systems.

Commentary by Dr. Valentin Fuster
2018;():V001T13A019. doi:10.1115/ICONE26-82239.

From the view of practical engineering application, a compacter nuclear power plant is expected. The weight and the volume of a nuclear power plant can be reduced by optimal selection of the operational parameters. In this work, a thermal-hydraulic model of the reactor, mathematical models of the reactor vessel, the main pipe, the pressurizer, the steam generator, the turbine and the condenser were established for the Qinshan-I nuclear power plant based on the related technical materials. The responses of the optimal targets to the changes of the design variables were studied by the sensitivity analyses. The non-dominated solution front of the nuclear power plant was obtained by means of the immune memory clone constrained multi-objective optimization algorithm. The study shows that the component mathematical models are reliable for the optimization process, the distribution of the non-dominated solution is decided by the steam generator secondary pressure. The volume and the weight of the system could be at least reduced by 23.0% and 9.5%, respectively.

Commentary by Dr. Valentin Fuster
2018;():V001T13A020. doi:10.1115/ICONE26-82292.

Supercritical carbon dioxide Brayton cycle power converters can benefit advanced nuclear reactors, as well as small modular reactors, by reducing the plant cost and increasing plant electrical output. The sCO2 cycles can also be designed for operation under direct dry air cooling. This paper presents the results of the coupled control analysis of a sCO2 cycle for a 100 MWe sodium-cooled fast reactor. The plant control mechanisms were investigated and optimized for load following operation.

Commentary by Dr. Valentin Fuster
2018;():V001T13A021. doi:10.1115/ICONE26-82295.

Supercritical carbon dioxide Brayton cycle power converters can benefit advanced nuclear reactors, as well as small modular reactors, by reducing the plant cost and increasing plant electrical output. The sCO2 cycles can also be designed for operation under direct dry air cooling. The paper presents the results of the coupled control analysis of a sCO2 cycle for a 100 MWe sodium-cooled fast reactor under changing ambient air temperatures. The optimum plant operation modes are identified.

Commentary by Dr. Valentin Fuster
2018;():V001T13A022. doi:10.1115/ICONE26-82343.

The purpose of the European Utility Requirements (EUR) Organisation is to actively develop and promote harmonised requirements for new mid- and large-size LWR NPPs that are proposed for construction in Europe. The harmonisation which is sought by the fourteen member utilities of the EUR Organisation, aims at delivering the safest and most competitive designs based on common requirements shared across Europe.

The harmonised requirements are presented in the EUR document. This consists of an extensive set of requirements covering all aspects (safety, performance, competitiveness) and all parts of a NPP (Nuclear and Conventional Islands). It can be used by Utilities (e.g. for assessment of the GEN3 designs proposed by vendors, technical reference for call for bids) and by Vendors (e.g. understanding of customer’s expectation of GEN3 NPPs, facilitating the licensing process).

The presentation will describe the main outcomes of the last 3 years of EUR Organisation activities (roadmap 2016–2018) and the challenges for the coming near future, in the following three fields.

First, completion of Revision E of the EUR Document was achieved in December 2016 and issued in July 2017, followed in October 2017 by a training course that was attended by 90 participants. The presentation will describe the most significant updates including revision of the safety requirements to align to the most recent European and International safety standards issued by WENRA and IAEA, lessons learned from the Fukushima accident, including re-evaluated Seismic and External Natural Hazards approach and updated international standards (e.g. for I&C and for European Grid code). Revision E also includes feedback from previous design assessments of NPPs. Future possible development of the EUR Document (and of the assessment process) will be considered within the EUR Organisation ‘Roadmap’.

Assessment of new designs is the second main technical activity. The assessment of the KHNP EU-APR (European version of APR1400) was completed in 2017 and an assessment of the Russian AEP’s VVER-TOI is planned to complete in 2018. These assessments are against EUR Revision D. The first assessment against Revision E (of the CGN HPR1000 “Hualong” design) is planned between 2018 and 2020. The presentation will recall the EUR design assessment objectives and process and the outcomes and progress of the different assessments.

The third topic is the interaction between the EUR and other stakeholders, in particular other international organisations (ENISS, WNA/CORDEL, WENRA, EC, IAEA, EPRI/URD) with the aim of promoting Industry Requirements and influencing prospective regulation where appropriate. The presentation will describe how the EUR Organisation is connected to these stakeholders and how it presents Utility requirements to the wider nuclear industry.

Commentary by Dr. Valentin Fuster
2018;():V001T13A023. doi:10.1115/ICONE26-82428.

iB1350 stands for an innovative, intelligent and inexpensive boiling water reactor 1350. It is the only Generation III.7 reactor incorporating Fukushima lessons learned and complying with Western European Nuclear Regulation Association (WENRA) safety objectives. It is about twice safer than any existing Gen III.5 reactors. It has 7-day grace period for SBO and SA without containment venting. It enables no evacuation and no long-term relocation in SA. It, however, is based on the well-established proven ABWR. The NSSS and TI are exactly the same as those of the existing ABWR. The iB1350 only enhanced the ABWR safety by adding an outer well (OW) as additional PCV volume, built-in passive safety systems (BiPSS) for SA, DEC systems and an APC shield dome over the containment. The BiPSS include an isolation condenser (IC), an innovative passive containment cooling system (iPCCS), in-containment filtered venting system (IFVS), and innovative core catcher (iCC). All the BiPSS are embedded and protected in the containment building against APC. No specialized safety features remote from the R/B are necessary, which reduces plant cost. The primary system has only one integrated RPV. There are no SGs, no pressurizer, no core makeup tanks, no accumulators, no hot legs, and no cold legs. The iB1350 consists of only one integrated RPV and passive safety systems inside the containment building. This configuration is simpler than the simplest large PWR and as simple as SMR. While SMR have rather small outputs, the iB1350 has 1350 MWe output. It is simple, large and economic. As for the safety design it has an in-depth hybrid safety system (IDHS). The IDHS consists of 4 division active safety systems for DBA, 1 or 2 division active safety systems for DEC and the built-in passive safety systems (BiPSS) for SA. The IDHS is originally based on the four levels of safety that have provided an explicit fourth defense level against devastating external events even before 3.11. It also can be explained along with WENRA Defense in Depth (DiD). It is said that independence between DiD levels are important. However, there are many exceptions for independence between DiD levels. For example, SCRAM is used in DiD2, DiD3a and DiD3b. Any DiD that allows exceptions of independence of DiD levels is fake. The iB1350 is rather based on the three levels of safety proposed by Clifford Beck (AEC, 1967). There is complete independence between level 2 (core systems) and level 3 (containment systems) without any exceptions of independence. DiD without exceptions of independence is a real DiD. Only passive safety reactors can meet the real DiD.

Commentary by Dr. Valentin Fuster
2018;():V001T13A024. doi:10.1115/ICONE26-82445.

Nuclear heating reactor is a new type of power plant that uses nuclear energy as heat source. Low temperature nuclear heating reactor should be the forerunner and main force for developing nuclear heating plant in China. Due to the lower water temperature required by the heating system, this dedicated, non-power generating nuclear reactor works at low temperatures and pressures with inherent safety features. The design, construction and operation of the nuclear heating reactors in various countries in the world were reviewed in this paper, and China’s new demonstration nuclear heating project and NHR-200 low-temperature heating reactor which would be used was discussed in the paper. We put forward the developing route and suggestion for the development of low-temperature heating reactor in China.

Commentary by Dr. Valentin Fuster
2018;():V001T13A025. doi:10.1115/ICONE26-82552.

iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident, and has incorporated both the lessons learned from the Fukushima Daiichi accident and the WENRA safety objectives. It has a double cylinder RCCV (Mark W containment) and an in-depth hybrid safety system (IDHS). The IDHS currently consists of 4 division active safety systems for a DBA, and 2 division active safety systems as well as built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and an innovative passive containment cooling system (iPCCS) for a Severe Accident (SA), which brings the total to 6 division active safety systems. Taking into account of excellent feature of the BiPSS, the IDHS has potential to optimize its 6 division active safety systems. The iPCCS that composes the BiPSS has been enhanced and has greater capability to remove decay heat than the conventional PCCS. While the conventional PCCS never can cool the S/P, the iPCCS can automatically cool the S/P directly with benefits from the structure of the Mark W containment. That makes it possible for the iB1350 to cool the core using only core inject systems and the iPCCS without RHR system: conventional active decay heat removal system. To make the most of this excellent feature of the iPCCS, it is under consideration to take credit for the iPCCS as safety systems for a DBA to optimize configuration of the IDHS. Currently, there are several proposed configurations of the IDHS that are expected to achieve cost reduction and enhance its reliability resulting from passive feature of the iPCCS. To compare those configurations of the IDHS, Level 1 Internal Events Probabilistic Risk Assessment (PRA) and sensitivity analyses considering external hazards have been performed for each configuration to provide measure of plant safety.

Topics: Safety , Optimization
Commentary by Dr. Valentin Fuster
2018;():V001T13A026. doi:10.1115/ICONE26-82579.

The nuclear heating reactor (NHR) is a typical integral pressurized water reactor (iPWR) developed by the institute of nuclear and new energy technology (INET) of Tsinghua University, which has the safety advanced features such as the primary circuit integral arrangement, full-range natural circulation, self-pressurization. Power-level control is crucial for the operational stability and efficiency of the NHR, and the dynamic modeling is a basis for control system design and verification. From the conservation laws of mass, energy and momentum, a lumped-parameter dynamical model is proposed for the nuclear steam supply system (NSSS) based on the 200MWth nuclear heating reactor II (NHR200-II). The steady-state model validation is given by the comparing the parameter values of this model and that for plant design. Then, both the open-loop responses under the disturbances of reactivity and coolant flowrates as well as the closed-loop responses under the case of power ramp are given, where the rationality of the responses are analyzed from the viewpoint of plant physics and thermal-hydraulics. This model can be utilized for not only the control system design but also the development of a real-time simulator for the hardware-in-loop control system verification.

Commentary by Dr. Valentin Fuster
2018;():V001T13A027. doi:10.1115/ICONE26-82620.

Over the course of the last several years the Canadian Nuclear Safety Commission (CNSC) has engaged with numerous vendors and potential licenses of small modular reactors (SMR) technology. This paper describes why Canada, and the CNSC, is of such interest to the international SMR community for prelicensing engagement and potential licensing of SMRs. It discusses what an SMR is and what potentially differentiates them from standard nuclear power plants (NPP). Readiness activities for the potential licensing of SMRs are described as well as modifications being made to the CNSC’s existing regulatory framework to facilitate the same, without reducing safety. The role of the CNSC’s discussion paper (DIS-16-04, Small Modular Reactors: Regulatory Strategy, Approaches and Challenges) and how feedback received on it helped confirm the CNSC’s modifications to be undertaken to the regulatory framework, as well as areas requiring further clarity, are highlighted. Finally, The CNSC Vendor Design Review (VDR) process is described as well as its part in ensuring a state of readiness to evaluate a licence application.

Commentary by Dr. Valentin Fuster

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