Experimental study on 1/28 scaled NGNP HTGR reactor building test facility response to depressurization event. (April 2018)

• Record Type: Journal Article
• Authors: Yang, Se Ro; Kappes, Ethan; Nguyen, Thien; Vaghetto, Rodolfo; Hassan, Yassin
• Is Part Of: Annals of nuclear energy. Volume 114(2018)
• Page Start: 154
• Page End: 164
• Publication Date: 2018-04
• Subjects: NGNP -- HTGR -- VLPC -- LOFC -- Depressurization -- Air-ingress -- Reactor building -- Pressure response -- Oxygen concentration; Nuclear energy -- Periodicals; Nuclear engineering -- Periodicals; 621.4805
• DOI: 10.1016/j.anucene.2017.12.023
• Languages: English
• ISSNs: 0306-4549

Abstract

Highlights: A 1/28-scaled HTGR reactor building facility was built to study RB response due to depressurization. Scaling analysis was conducted regarding thermal–hydraulic phenomena expected in air refill phase. Pressure and air-helium mixing inside RBs were studied under D-LOFC medium break in reactor cavity. Measurements covered various flow paths and break locations in steam generator and reactor cavity. Results suggested an increment in the vent path area between reactor and steam generator cavities. Abstract: The analysis and understanding of air ingress events are an important aspect of the design of high-temperature gas-cooled reactor (HTGR) accident scenarios. These include depressurized loss of forced cooling (D-LOFC) events that allow for the possibility of air ingress into the reactor pressure vessel as a result of a break in the helium pressure boundary, which can ultimately result in oxidation of the fuel elements and other nuclear-grade graphite components. To characterize air ingress into the vented low-pressure containment of the next generation nuclear plant HTGR during hypothetical moderate-sized D-LOFC break accidents experimentally, a 1/28 scaled simplified reactor building model was established. A non-dimensional similarity approach was employed for scaling of the experimental facility. Three experiments were designed and conducted to study the dynamic response to the accident scenario. The experimental results suggest an increment in the flow area of the check valve between the reactor cavity (CV1) and steam generator cavity (CV3). Furthermore, qualitative analysis was conducted on the experimental data.