Requirement Analysis of Computer-Based Instrumentation and Control System for Reaktor Daya Eksperimental

Currently, in few countries, there have been many projects related to the development of Field Programmable Gate Array (FPGA)-based Instrumentation and control (I&C) system designs for their nuclear power plant (NPP), whether for modernization or the new development of NPPs. Indonesia already develops a study related to High Temperature Gas Cooled Reactor as the generation IV for NPPs type for these few years with a specific type which is High Temperature Reactor–10 MW (HTR-10). Since the FPGA-based technology is new in the field, and both the international standards and guidance, from national and international has not defined provisions this issue, therefore verification and validation (V&V) activities for this HTR-10 are mandatory to be included in the design life cycle of I&C system development to get license approval from the national regulatory body. The FPGA-based system needs concentration both in hardware and software development, but commonly only software-related standards that usually included in the design and V&V in the previous computer-based I&C system. Therefore, in this study the V&V plan needs to be developed to state the methods to be required regarding the correctness of the applications of FPGA based I&C system. By arranging this V&V method, we will see whether they are easily applicable to actual systems, had the result to reach the requirement to have reliability, availability, maintainability and safety aspects, and has shortened and easier V&V step than computer-based I&C system. From the results of this study, it can be seen that the results of the method used for V&V between Programmable Logic Controller (PLC)-based and FPGA are almost the same because both have almost the same requirements, especially in software design. Some testing is not mandatory for FPGA-based because of the reduced system complexity. The use of FPGA can shorten the V&V process which can also reduce costs.

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Requirement analysis for human factor engineering on SMR-NuScale design

Maerani

Suryono

Santoso

2022

J. Phys.: Conf. Ser.

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Human Factors Engineering (HFE) is an important element of control room facility design and Man Machine Interface (MMI) design whose principles must be applied systematically to ensure the safe operation of power reactors. In the design of the NuScale Small Modular Reactor (SMR), the United States-Nuclear Regulatory Commission (US-NRC) as an international regulatory body has published Design-Specific Review Standards (DSRS) for Nuclear SMR design which are specific for NuScale. There is no specific discussion related to HFE in the document. However, in order to develop the NuScale design in In Indonesia, we can learn from Oregon State University, the first Nuscale developer, regarding the requirements when those are compared to the large nuclear reactors. Therefore, reviewing the NuScale SMR design needs to be carried out starting with the conducting a requirement traceability analysis by tracing the guidelines from the regulatory body and knowing the recommended guideline in developing the HFE design, especially for the specific SMR/NuScale. It is hoped that by understanding the process of verifying these requirements documents, Indonesia will be able to apply the HFE for the NuScale reactor by meeting the criteria required by BAPETEN and be able to prepare the validation method so that the design can obtain a license.

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Requirement Analysis of Computer-Based Instrumentation and Control System for Reaktor Daya Eksperimental

Cited by 3 publications

References 5 publications

Reliability program plan for field programmable gate array-based I&C system of nuclear power plant

Reliability program plan for field programmable gate array-based I&C system of nuclear power plant

V&V Methods Used in the Development of FPGA-based I&C Systems for HTR-10

Requirement analysis for human factor engineering on SMR-NuScale design

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