Standard Design Approval Research Articles

Unique safety features and licensing requirements of the NuScale small modular reactor

Small modular reactors (SMR) offer a novel approach to the construction and operation of nuclear power plants. The NuScale VOYGR™ plant uses a simplified SMR design that is based on proven light-water reactor technology with substantial improvements in nuclear safety. It consists of a 250 MWt reactor core housed with other primary system components in an integral reactor pressure vessel surrounded by a steel containment vessel, all of which is immersed in a large pool of water that also serves as the ultimate heat sink. At the core of the NuScale safety case are three primary safety systems: the decay heat removal system, the emergency core cooling system, and the containment. The ability of the NuScale Power Module (NPM) passive safety systems to remove core decay heat for an unlimited duration is demonstrated through analysis of a beyond-design-basis extended loss of AC power with no replenishment of water to the ultimate heat sink or operator actions. The NuScale methodology to evaluate an indefinite loss of AC power uses the proprietary NRELAP5 systems analysis computer code. Analysis results show that the reactor coolant system liquid level above the core is maintained and that containment pressure remains below the vessel design pressure. Once full passive air cooling is established, containment pressure and temperature will decrease over time with decreasing core decay heat. NuScale received standard design approval in September 2020 and design certification in January 2023 for its 50 MWe NPM configured as a 12 module plant. NuScale is currently seeking standard design approval to increase its core power to 250 MWt, nominally 77 MWe per module, in a 6-module plant (VOYGR™-6) configuration. The high-level safety of NuScale’s SMR technology is foundational to a new standard of nuclear power plant resilience.

Long term cooling safe shutdown performance analysis for SMART with passive safety system using MARS-KS

Passive safety systems have been adopted in reactor design to enhance safety especially in SMRs after the TMI-2 accident in 1979. The passive system has high operational reliability with natural driving forces such as gravity, natural circulation and stored energy. In recent years, development of innovative SMRs has emerged as a global R&D initiative for sustainable energy to challenge the carbon neutrality of climate change. The passive safety system makes reactor design simple and thus enhances the reliability of the reactor. As advanced reactors adopt passive safety systems, regulatory requirements and guides are required. Thus, development of regulatory guide and detailed regulatory review methodology for the passive safety systems is needed. In this paper, long term cooling safe shutdown acceptance criteria and performance analysis methodology are proposed for the regulatory standard design approval for the SMART100 integral reactor with fully passive system which has been jointly developed by KAERI in Korea and K.A.CARE in Saudi Arabia. The long term cooling safe shutdown performance analyses are performed until 72 hours to evaluate the methodology, and sensitivity analyses are also performed to provide a technical basis for the regulatory independent review for the long term cooling performance of the SMART100 passive safety systems.

Researches on thermal hydraulics and fuel performance of ATFs during extreme steam generator tube failure without ECCS and DHRS in NuScale

The small modular reactor NuScale has completed Standard Design Approval of Nuclear Regulatory Commission review by good safety performance and it makes history with the first SMR to receive NRC design approval. In this paper, to ensure the safety of the fuel rod in steam generator tube failure, accident tolerant fuels are selected as candidates for the traditional fuel rod. The primary and second circuit of NuScale were modeled, and compared with standard data to ensure accuracy. Thermal hydraulics and fuel performance of ATFs at NuScale during SGTF are analyzed. The results show that the fuel centerline temperature lower by loading with Th0.923U0.077O2-SiC ATF system in the steady-state. Moreover, it can prolong the time for radiotoxicity to leak during SGTF. In short, the safety of NuScale is improved with the replacement of ATF fuel system.

Risk reduction design alternatives for low-power and shutdown operation of a 1500-MW advanced power reactor with a passive auxiliary feedwater system

The Korean nuclear power regulatory body issued the standard design approval for a 1500-MW advanced power reactor in 2014. The reactor was designed based on a four-train independent safety concept with a passive auxiliary feedwater system. Its full power risk, or core damage frequency (CDF), was determined to be significantly lower than that of the Advanced Power Reactor 1400 (APR1400) used in Units 3 and 4 of the Shin-Kori Nuclear Power Plant. However, because the ratio of the CDF contribution to the internal full power to that of the low-power and shutdown (LPSD) is 17:83, the risk during LPSD operation requires further assessment and reduction. In this study, we investigated design alternatives for reducing LPSD operation risks and analyzed the associated sensitivities. We propose various design alternatives, and their combinations, which are analyzed based on the reduction effect they have on the CDF for plant operating states and on initiating events during LPSD operation. The proposed alternatives, and their combinations in particular, exhibited significant reductions in LPSD CDF, indicating high potential applicability.

Recent achievement and future prospects of the ATLAS program

As a critical thermal-hydraulic test facility, the ATLAS facility has been widely utilized since its first commission test in 2006 to resolve urgent safety concerns raised by nuclear industries and regulatory body, to verify the performance of new safety systems, to validate safety analysis codes, and to provide a cooperation network for developing safety analysis capabilities. In the first phase of the ATLAS program, major R&D efforts were put into producing code validation data to support the standard design approval of the APR1400 when design basis accident (DBA) scenarios were the main interest of the program. In the following phase, the ATLAS program’s focus shifted to beyond DBA scenarios, and the OECD-ATLAS international joint project started reflecting the safety concerns of many NEA countries following the Fukushima Daiichi disaster in 2011. In this paper, the current status of the ATLAS program, including the OECD-ATLAS Phase-1 and 2 projects, are briefly described together with their major outcomes. The next R&D program based on the ATLAS platform is also briefly discussed.

Assessment of an integral small modular reactor during rod ejection accident by using DRAGON/PARCS codes

Small modular reactors (SMRs) have benefits like lower initial costs and investment risks, higher safety features, longer refueling cycle and also beside than electricity production, capability to be used as sea water desalination, district heating and process heat for industries and small isolated grids, which make these reactor an alternative option for the future of nuclear power industries. Different SMRs all over the world are in the different stages of design and construction and different aspects of these reactors are under evaluation by researchers and engineers. The Korean System Integrated Modular Advanced Reactor (SMART) categorized as SMR that has received its standard design approval. In this study Rod ejection accident (REA) that categorized as a design basis accident (DBA) has been evaluated for SMART core, according to the standard safety analysis report (SSAR) of SMART. The DRAGON code has been used for cell calculation and also for spatial kinetic and thermal hydraulic calculations, the feedback, transient and TH blocks of PARCS code as a thermo-neutronic code has been used. Finally the SMART core response to the REA has been evaluated by comparing the attained results with SMART SSAR that shows proper match. Also the capability of DRAGON/PARCS codes for predicting SMR behavior during REA has been approved.

Contribution of thermal–hydraulic validation tests to the standard design approval of SMART

Abstract Many thermal–hydraulic tests have been conducted at the Korea Atomic Energy Research Institute for verification of the SMART (System-integrated Modular Advanced ReacTor) design, the standard design approval of which was issued by the Korean regulatory body. In this paper, the contributions of these tests to the standard design approval of SMART are discussed. First, an integral effect test facility named VISTA-ITL (Experimental Verification by Integral Simulation of Transients and Accidents-Integral Test Loop) has been utilized to assess the TASS/SMR-S (Transient and Set-point Simulation/Small and Medium) safety analysis code and confirm its conservatism, to support standard design approval, and to construct a database for the SMART design optimization. In addition, many separate effect tests have been performed. The reactor internal flow test has been conducted using the SCOP (SMART COre flow distribution and Pressure drop test) facility to evaluate the reactor internal flow and pressure distributions. An ECC (Emergency Core Coolant) performance test has been carried out using the SWAT (SMART ECC Water Asymmetric Two-phase choking test) facility to evaluate the safety injection performance and to validate the thermal–hydraulic model used in the safety analysis code. The Freon CHF (Critical Heat Flux) test has been performed using the FTHEL (Freon Thermal Hydraulic Experimental Loop) facility to construct a database from the 5 × 5 rod bundle Freon CHF tests and to evaluate the DNBR (Departure from Nucleate Boiling Ratio) model in the safety analysis and core design codes. These test results were used for standard design approval of SMART to verify its design bases, design tools, and analysis methodology.

Integrated Testing Environment of Instructor Station for SMART Simulator

SMART is the reactor that has been researched for many years by KAERI in order to provide the small-mid scale of power for typically seawater desalination. Now Korea Atomic Energy Research Institute (KAERI) has issued Standard Safety Analysis Report (SSAR) and acquired Standard Design Approval (SDA) for SMART. In order to conduct the design verification and validation for license, the integrated simulation test environment that is composed of 1) the system specific simulation codes formerly developed in the name of Nuclear Plant Analyzer (NPA) including NSSS and BOP simulation, 2) Instructor Station (IS), 3) Supervisory Control and Data Acquisition (SCADA), 4) operator and instructor Human Machine Interface (HMI), and 5) soft-controller has been considered as an important area for operator training and system validation. These sub-components has been designed and implemented for verifying and validating the SMART design and training of operators and for generating the backup data for licensing. This paper introduces the structure of integrated simulation test environment for SMART, explains the efforts to assist system-specific simulation code interface, and also addresses the effort for implementing and optimizing the test environment by maintaining its own simulation functionality and performance in order to review the simulation results efficiently.

Thermo-hydraulic characteristics of the helically coiled tube and the condensate heat exchanger for SMART

SMART is a 330MWt advanced integral PWR. An experimental validation for the SMART-specific thermal hydraulic phenomena is performed to generate background information required for SMART standard design approval. Series of fundamental experiments are carried out. The two-phase heat transfer at the helically coiled tube is affected linearly by a heat flux and system pressure. However, it has nearly the same value at the high quality region regardless of the pressure change. The condensate heat transfer depends on the tube diameter and the mass flux at the PRHRS heat exchanger. The condensate heat transfer increases at a high steam flow, low pressure and large diameter. The calculation results of MARS and TASS/SMR codes are predicted to have similar values because the two codes incorporate the same heat transfer correlations for a helically coiled tube and condensate heat exchanger.

DESIGN SCOPE AND LEVEL FOR STANDARD DESIGN CERTIFICATION UNDER A TWO STEP LICENSING PROCESS

A small integral reactor SMART (System Integrated Modular Advanced ReacTor), being developed in Korea since late 1990s and targeted to obtaining a standard design approval by the end of 2011, is introduced. The design scope and level for design certification (DC) is well described in the U.S. NRC SECY documents published the early 1990s. However, the documents are valid for a one-step licensing process called a combined operating license (COL) by the U.S. NRC, while Korea still uses a two-step licensing process. Thus, referencing the concept of the SECY documents, we have established the design scope and level for the SMART DC using the contexts of the standard review plan (SRP). Some examples of the results and issues raised during our review are briefly presented in this paper. The same methodology will be applied to other types of reactor under development in Korea, such as future VHTR reactors.