Elsevier

Progress in Nuclear Energy

Volume 85, November 2015, Pages 617-623
Progress in Nuclear Energy

Strength assessment of SMART design against anticipated transient without scram

https://doi.org/10.1016/j.pnucene.2015.08.005Get rights and content

Highlights

  • Acronym of SMART is a system-integrated modular advanced reactor.

  • SMART adopts a passive system to enhance its safety.

  • SMART design minimizes a potential for an occurrence of transients.

  • Intrinsic SMART design can maintain a stable condition for ATWSs without DPS.

Abstract

SMART is a small sized integral type pressurized water reactor and its design minimizes the potential for an occurrence of transients and accidents. To achieve these purposes, engineered safety features enhance the accident resistance. A large volume of semi-passive pressurizer can accommodate a wide range of pressure transients. Low core power density increases the thermal margin, and negative fuel and moderator temperature coefficients yield beneficial effects on a core self-stabilization, and limit the reactor power during accidents. Considering the enhanced engineered safety features and design characteristics, anticipated transient without scram scenarios are adopted to evaluate the strength of the SMART design. The peak pressure does not exceed 116% of the nominal pressure and the fuel can maintain its integrity for the transient. The intrinsic SMART design characteristics having passive safety features and a large primary coolant inventory can maintain a stable condition safely although the reactor protection and diverse protection systems are not available for the anticipated transient without scram transients.

Introduction

Most countries operate small sized power plants for their electricity supply (Kim et al., 2014). These countries have recently taken an interest in small and medium sized nuclear power plants. These plants are considered to be a suitable option for nuclear system deployment in developing countries and non-electrical applications for various facilities (IAEA, 2005, IAEA, 2014). SMART (system-integrated modular advanced reactor) for exporting to countries having small electric grids was developed by KAERI (Korea atomic energy research institute). SMART is a small sized integral type pressurized water reactor (PWR) with a rated thermal power of 330 MW, which adopts a hybrid technology of new innovative design features and proven technologies aimed at achieving highly enhanced safety and improved economics (Kim et al., 2014). The design features are basically inherent safety improvement and passive safety features. The SMART design minimizes the potential for an occurrence of transients and accidents. To achieve these purposes, the engineered safety features enhance the accident resistance. The integral arrangement of the primary system removes large size pipe connections between major components such as a steam generator (SG) or a reactor coolant pump (RCP). It fundamentally eliminates the possibility of large break loss-of-coolant-accident (LOCA). A large volume of the semi-passive pressurizer (PZR) can accommodate a wide range of pressure transients during pressure increase transients and accidents. Low core power density lowers the fuel element temperature rise under accident conditions and increases the thermal margin. Negative fuel and moderator temperature coefficients yield beneficial effects on a core self-stabilization, and limit the reactor power during accidents. The passive residual heat removal system (PRHRS) can remove a residual heat passively for the reactor shutdown conditions.

An anticipated transient without scram (ATWS) transient is one of the important accidents to find the design stiffness and safety characteristics. Performance evaluation of an integral reactor against an anticipated transient without scram carried out by considering a heat removal decrease to the secondary system, a loss of offsite power and an inadvertent control rod withdrawal event as an initiating event (Yang et al., 2006). An ATWS analysis of a control rod withdrawal in addition to bypass valve failure was performed to check the integrity due to the changed operation parameters (Lang and Dong, 2014). It showed that the plant was safe during the ATWS transient of a control rod withdrawal. Also, Chen et al. showed an overall ATWS transient for Maanshan PWR using a system analysis code (Chen et al., 2014). A simulation of a station black-out ATWS was performed by applying response surface methodology (Cacciabue et al., 1985). The proposed technique was an effective tool for selecting the important accident variables and that the body of information gained was significant with respect to the number of observations performed. Containment venting was studied as a mitigation strategy for preventing or delaying severe fuel damage following an ATWS accident initiated by a main steam isolation valve closure (Harrington, 1988). It was proposed that two alternative strategies to delay or prevent severe fuel damage. An ATWS transient was analyzed using a coupled code with DYN3D and ATHLET codes for a PWR (Kliem et al., 2009). A variation of the MOX-fuel assembly portion in the core had an effect on the reactivity coefficients of the fuel temperature and the moderator density. In addition, Ninokata et al. (2002) evaluated the self-controllability against ATWS. They described a simple and general method to evaluate the inherent and passive safety characteristics that used self-controllability limit lines. The previous works studied for the conventional loop type nuclear plants not considering characteristics of the integral nuclear power plants and a passive residual heat removal system. To evaluate the strength of the SMART design, which is an integral type pressurized water reactor, considering a passive residual heat removal, a large volume of the pressurizer, and a low core power density compared with conventional nuclear power plants, anticipated transient without scram transients are performed at nominal operation conditions.

Section snippets

Inherent characteristics of SMART and engineered safety features

The SMART plant can produce 100 MW of electricity, or 90 MW of electricity and 40,000 ton per day of desalinated water concurrently, which is sufficient for 100,000 residents. The reactor assembly of SMART contains fuel assemblies, steam generators, a pressurizer, and reactor coolant pumps in a single reactor pressure vessel (RPV). The integrated arrangement of these components enables the removal of the large size pipe connections between major primary components at the design stage and thus,

System analysis code, TASS/SMR and system modeling

TASS/SMR code is used to analysis the ATWS transient of the SMART plant. The code has been developed for an analysis of the design based transients and accidents in an integral type reactor reflecting the characteristics of the SMART design (Chung et al., 2012). Verification of the code was performed to assess the software correctness and numerical accuracy of the solution. Then, computational models based on comparisons between computational simulations and experimental data were assessed.

Analysis of ATWS transient without diverse protection system

To analyze ATWS transients, an uncontrolled control element assembly withdrawal (CEAW), a loss of feedwater flow (LOFA), and a total loss of reactor coolant flow (TLCF) are selected. The initial and boundary conditions, and specific assumptions are as follows; the initial conditions for the analysis are nominal conditions and realistic values as shown in Table 1. The core power and primary system pressure are 330 MWt and 15.0 MPa, respectively. The primary system is completely filled with

Conclusions

Anticipated transient without scram (ATWS) transients including an uncontrolled control element assembly withdrawal, a loss of feedwater flow, and a total loss of reactor coolant flow were assessed to evaluate the strength of the SMART design considering enhanced engineered safety features and design characteristics. The system thermal hydraulic behaviors were analyzed by TASS/SMR code under realistic initial and boundary conditions.

The peak pressure did not exceed 116% of the initial pressure

Acknowledgments

This work was supported by the National Research Foundation of Korea (NRF) grant funded by the Korea government (MSIP) (No. 2012M2A8A4025980).

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