Fuel Rod Performance Uncertainty Analysis During Overpressurization Transient for Kuosheng Nuclear Power Plant with TRACE/ FRAPTRAN/ DAKOTA Codes in SNAP Interface (NUREG/IA-0465)

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Publication Information

Manuscript Completed: August 2015
Date Published: March 2016

Prepared by:
Chunkuan Shih*, Hao-Chun Chang*, Jong-Rong Wang*, Shao-Wen Chen*, Hao-Tzu Lin, Show-Chyuan Chiang**, Tzu-Yao Yu**

Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.
1000, Wenhua Rd., Chiaan Village, Lungtan, Taoyuan, 325, Taiwan

*Institute of Nuclear Engineering and Science, National Tsing Hua University; Nuclear and New Energy Education and Research Foundation
101 Section 2, Kuang Fu Rd., HsinChu, Taiwan

**Department of Nuclear Safety, Taiwan Power Company
242, Section 3, Roosevelt Rd., Zhongzheng District, Taipei, Taiwan

K. Tien, NRC Project Manager

Division of Systems Analysis
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

Prepared as part of:
The technical exchange and cooperation between the U.S. Nuclear Regulatory Commission and the Task Force for
Safety Critical Software in the Field of Nuclear Safety Resear

Published by:
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

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Abstract

After the MUR (measurement uncertainty recapture) power uprates, Kuosheng Nuclear Power Plant has uprated the power to 2943 MWt. Recently, Taiwan Power Company is concerned in SPU (stretch power uprated) plan and uprates the power to 3030 MWt, which is 104.7% of the designed power. Before the stretch power uprates, several transient analysis should be done for ensuring that the power plant could maintain stability in high power operating conditions. The advanced thermal hydraulic analysis code TRACE which is conducted by U.S. NRC was applied for overpressurization transient including main steam line isolation valves closure, turbine stop valves closure and turbine control valves closure. This closure of valves increased the dome pressure; as a result, the void fraction inside the reactor core decreased. The decline of the void fraction will increase the reactivity feedback; hence, the power increased rapidly until the reactor scram. Further, with safety relief valves open, the dome pressure would not exceed the criteria regulated by ASME. In addition, to cover the insufficiency of thermal hydraulic code, fuel rod transient analysis code FRAPTRAN was applied. To perform fuel rod transient analysis, thermal information from TRACE code would be entered into FRAPTRAN with power history. In addition, DAKOTA code was applied to concern the influence of fuel rod manufacturing tolerance. With uncertainty bands from DAKOTA analysis, the criteria could be judged with more confidence. This research successfully established a procedure of thermal hydraulic and fuel rod property analysis. Further, with SNAP interface, the FRAPTRAN and DAKOTA were combined successfully to perform the uncertainty analysis.