A Hierarchically Structured Down-Top Test Equipment Debugging Method for RPS

An accident scenario-based statistical test method for RPS in nuclear power plant

Development of a digital reactor control and protection system

To maintain the cyber security, nuclear digital Instrumentation and Control (I&C) systems must be analyzed for security risks because a single security breach due to a cyber attack can cause system failure, which can have catastrophic consequences on the environment and staff of a Nuclear Power Plant (NPP). Attack trees have been widely used to analyze the cyber security of digital systems due to their ability to capture system specific as well as attacker specific details. Therefore, a methodology based on attack trees has been proposed to analyze the cyber security of the systems. The methodology has been applied for the Cyber Security Analysis (CSA) of a Bistable Processor (BP) of a Reactor Protection System (RPS). Threats have been described according to their source. Attack scenarios have been generated using the attack tree and possible counter measures according to the Security Risk Level (SRL) of each scenario have been suggested. Moreover, cyber Security Requirements (SRs) have been elicited, and suitability of the requirements has been checked.

A reactor protection system (RPS) in a boiling water reactor (BWR) is a unique system since it represents high redundancy compared with other safety-related systems in the BWR. Owning to its high redundancy, NUREG-5500 vol.3 showed that unreliability of the General Electric BWR4-type RPS is dominated by common-cause failures (CCFs) in its highly redundant components. Since the staggered test scheme can reduce the period of latent CCF states of redundant standby components, the scheme is a good candidate to improve the reliability of the RPS. However, the scheme is impracticable for the RPS because of its high redundancy. In this paper, a new concept of a “multi-group staggered test scheme” for a highly redundant system is proposed. In the multi-group staggered test scheme, components are arranged into several groups, and all components in a group are tested simultaneously. This paper presents a preliminary example of the new scheme that is applied to control rod notch tests, in which all withdrawn control rods are moved at least one notch. It is shown that the new scheme can reduce the CCF risk in RPS.

Stability and steady state analysis of control and safety systems of Nuclear Power Plants

The safety of nuclear power plants (NPPs) is achieved through the implementation of a multi-layered protection system and safety functions, tailored to the configuration of the nuclear facility. One of the critical systems in ensuring safety is the reactor protection system, which ensures the automatic transition of the reactor to a controlled safe state during emergency situations and in the event of deviations from normal operating conditions. An effective assessment of the functional safety of reactor protection systems during the design, implementation, and operation phases serves as a vital tool for confirming compliance with safety requirements and ensuring reliable protection of NPPs. In contemporary contexts, the evaluation of functional safety effectiveness is conducted based on a risk-oriented approach, which is implemented in accordance with the regulatory requirements set forth by the IAEA and national regulatory bodies. This article presents a systematic study that investigates the risk-oriented approach to assessing the functional safety of reactor protection systems in NPPs. The research is grounded in a systematic mapping of findings from a sample of 72 articles, selected according to predefined criteria. Consequently, the current state of the field has been defined, and key methodologies, methods, and tools have been identified. Additionally, significant trends have been highlighted, including the integration of modern modeling techniques with traditional approaches. Furthermore, areas for growth and promising directions for future research have been identified, as noted in the reviewed articles.

Communication independence is one of the key criteria of digital safety I&C system design. This paper mainly analyzes the requirements for communication independence in safety regulations and standards, and then introduces the architecture and design features, including communication failure processing measures, of communication networks of ACPR1000 nuclear power plant safety digital protection system based on FirmSys platform developed by CTEC. The communication design meets the regulations requirements and effectively improves the safety and reliability of the system, and it is successfully applied in reactor protection system (RPS) of Yang Jiang nuclear power plant unit 5&6. In addition this design can provide reference for communication designs of other NPPs and industries.

Small modular reactors (SMRs) are factory-built transportable nuclear power plants (NPPs) that can generate up to 300 MWe and support new applications for nuclear energy such as hydrogen production, industrial heat generation, and water desalination. SMRs are poised for near-term deployment in the United States and offer improvements over existing NPPs such as passive safety features, greater plant site flexibility, reduced construction costs, load-following operations, and extended refueling cycles. However, in order to realize the benefits of SMRs, there are several challenges related to instrumentation and control (I&C) system components that must be addressed. The I&C systems of an SMR ensure its safe and efficient operation. I&C sensors measure process parameters such as temperature, pressure, level, flow, and neutron flux and provide input to the Reactor Protection System (RPS) to initiate a rapid shutdown if necessary. The performance of these safety-related I&C sensors and systems must be periodically verified to ensure they meet the plant technical specifications for accuracy and response time. In addition, the associated I&C cabling (which includes low-voltage cable assemblies for sensors, Control Rod Drive Mechanisms (CRDMs), and Rod Position Indication (RPI) Systems) is tested as part of normal plant maintenance activities or in support of aging management programs. Over time, exposure to harsh environmental conditions in the plant can result in degradation and failure of I&C sensors and cables. This is especially true for the I&C sensors and cables in SMRs which will be subjected to elevated temperatures, high radiation, and a vacuum atmosphere within the containment vessel during normal plant operation. These conditions lead to excessive Ohmic heating within the cables which further accelerates damage to the cable insulation material and results in premature failure. Frequent cable replacement due to premature degradation is not practical or economical for SMR plant owners. To combat these environmental stressors, the ampacity of the cable must be derated. However, there is limited experience to date with cable derating in vacuum, and no experience with cable derating in vacuum at high temperature and radiation. As a result, Analysis and Measurement Services Corporation (AMS) is conducting a study on several common NPP I&C cables to determine how various insulation materials perform at conditions that emulate those anticipated in the containment of an SMR. The on-going effort is funded under a research and development (R&D) grant awarded to AMS by the U.S. Department of Energy (DOE). In addition, AMS is partnering with the Oak Ridge National Laboratory (ORNL) to conduct specialized testing in support of this study under a separate initiative called the Gateway for Accelerated Innovation in Nuclear (GAIN). The goal of this R&D is to develop technical guidance and technologies to support initial startup and subsequent operation of SMRs through specialized testing services and condition monitoring technologies. Preliminary results of this R&D are provided in this paper. AMS will continue to investigate challenges associated with I&C systems for SMR applications in support of the timely deployment of the first U.S.-based SMR by the mid-2020s.

The NRC recognizes that a more complete technical basis for understanding and regulating advanced digital technologies in commercial nuclear power plants is needed. A concern is that the introduction of digital safety systems may have an impact on risk. There is currently no standard methodology for measuring digital system reliability. A tool currently used to evaluate NPP risk in analog systems is the probabilistic risk assessment (PRA). The use of this tool to evaluate the digital system risk was considered to be a potential methodology for determining the risk. To test this hypothesis, it was decided to perform a limited PRA on a single dominant accident sequence. However, a review of existing human reliability analysis (HRA) methods showed that they were inadequate to analyze systems utilizing digital technology. A four step process was used to adapt existing HRA methodologies to digital environments and to develop new techniques. The HRA methods were then used to analyze an NPP that had undergone a backfit to digital technology in order to determine, as a first step, whether the methods were effective. The very small-break loss of coolant accident sequence was analyzed to determine whether the upgrade to the Eagle-21 process protection system had an effect on risk. The analysis of the very small-break LOCA documented in the Sequoyah PRA was used as the basis of the analysis. The analysis of the results of the HRA showed that the mean human error probabilities for the Eagle-21 PPS were slightly less than those for the analog system it replaced. One important observation from the analysis is that the operators have increased confidence steming from the better level of control provided by the digital system. The analysis of the PRA results, which included the human error component and the Eagle-21 PPS, disclosed that the reactor protection system had a higher failure rate than the analog system, although the difference was not statistically significant.

Validation of a digital computer program to be used in a nuclear power plant protection system must meet quality assurance requirements. Digital systems have not traditionally been used on nuclear reactor protection systems. Licensing of digital system software requires providing assurance that the software performs its intended function. To provide added assurance, the Babcock and Wilcox Company performed software validation on the digital program intended for use on a protection system. Software validation of the Reactor Protection System-II digital program presented a multi-faceted challenge. Quality assurance requirements were imposed on the project. Certain validation ground rules were specified. No known methods existed for proving program correctness for nontrivial software. No precedence had been set to estimate the quality or quantity of testing required as a method of validation. Project schedule constraints were imposed. The need for more documentation than normally furnished was recognized, but how much and what kind was not clear. This paper relates how this challenge was met through a discussion of how the project was performed and the lessons learned through those experiences. A test method was devised within validation ground rules and project schedule constraints to validate that software performed the specified functions. Orderly methods of testing and evaluating were implemented and documented in compliance with a plan to provide auditable, traceable evidence of the validation effort and the digital component program performance.

The Sizewell B control and protection systems represent the most advanced implementations of such systems in an operating nuclear power station. The unique requirements of a nuclear power station have resulted in an approach to the control and protection systems which is both novel but applicable to any application where high integrity is a basic system requirement. The following discusses the implementation of the control and data processing system and reactor protection systems at Sizewell B.

A safety class field programmable gate array based system in nuclear power plant has been developed to improve the diversity. Testbench is necessary to satisfy the technical reference, IEC-62566, for verification and validation of register transfer level code. We use the open verification methodology(OVM) developed by standard body. We show that our testbench can use random input for test. And also we show that reusability of block level testbench for the integration level testbench, which is very efficient for large scale system like nuclear reactor protection system.

Uncertainties in the Power Alignment at the Jordan Research and Training Reactor

The realization method is presented for a logic configuration of the reactor protection system (RPS) for BWRs which enhances their reliability.Higher reliability requires a majority voting logic even when a single logic component has been bypassed. Furthermore the load driver of the RPS on the output side must actuate two separate operation objects. To fulfill the above conditions, a general realization of the 2-out-of-4 logic with a minimum hardware component is proposed. Then, the structure of the microprocessor-configured RPS employing this logic is presented.The reliability of the above RPS is evaluated by computations. The results show that the average frequency of improper operation is less than that with the conventional 1-out-of-2-twice logic under the conditions that the repair rate is 0.1 and the failure detection rate is above 92%. The average probability of failure to operate is reduced by two orders of magnitude under the above conditions.

With the development of modern automatic control technology, digital control technology has advanced technology and inherent advantages compared with traditional analog control technology. Therefore, the use of digital instrument and control system has become an inevitable trend of global nuclear power construction. Nuclear power embedded real-time operating system is an important part of digital instrument and control system in nuclear power plant. It is responsible for managing the hardware resources of the whole system and providing operation environment for application software. Its safety directly affects the safety of instrument and control system. This paper studies the application of embedded information technology in the safety of nuclear power plant, and puts forward a safety related scheme of embedded information system. The scheme fully considers the confidentiality, integrity and availability of information in the digital instrument and control system of nuclear power plant. The application results show that the scheme can meet the information security requirements of nuclear power instrument and control system, and has reference and guiding significance for the research and development of nuclear power embedded real-time operating system.