Analysis of LOCA Scenarios in the NIST Research Reactor Before and After Fuel Conversion

Conference · Sun Aug 30 00:00:00 EDT 2015

OSTI ID:1183253

Baek, J. S. ^[1]; Cheng, L. Y. ^[1]; Diamond, D. ^[1]

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1. Brookhaven National Lab. (BNL), Upton, NY (United States)

An analysis has been done of hypothetical loss-of-coolant-accidents (LOCAs) in the research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The purpose of the analysis is to determine if the peak clad temperature remains below the Safety Limit, which is the blister temperature for the fuel. The configuration of the NBSR considered in the analysis is that projected for the future when changes will be made so that shutdown pumps do not operate when a LOCA signal is detected. The analysis was done for the present core with high-enriched uranium (HEU) fuel and with the proposed low-enriched uranium (LEU) fuel that would be used when the NBSR is converted from one to the other. The analysis consists of two parts. The first examines how the water would drain from the primary system following a break and the possibility for the loss of coolant from within the fuel element flow channels. This work is performed using the TRACE system thermal-hydraulic code. The second looks at the fuel clad temperature as a function of time given that the water may have drained from many of the flow channels and the water in the vessel is in a quasi-equilibrium state. The temperature behavior is investigated using the three-dimensional heat conduction code HEATING7.3. The results in all scenarios considered for both HEU and LEU fuel show that the peak clad temperature remains below the blister temperature.

• Research Organization: Brookhaven National Lab. (BNL), Upton, NY (United States)
• Sponsoring Organization: USDOE, National Inst. of Standards (NIST), American Nuclear Society (United States)
• DOE Contract Number: SC00112704
• OSTI ID: 1183253
• Report Number(s): BNL-107070-2015-CP; R&D Project: 20773; NN9100000; TRN: US1600154
• Resource Relation: Conference: 16. International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16), Chicago, IL (United States), 30 Aug - 4 Sep 2015
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
HIGHLY ENRICHED URANIUM
LOSS OF COOLANT
NBSR REACTOR
NUCLEAR FUELS
THERMAL HYDRAULICS
THERMAL CONDUCTION
BLISTERS
CONVERSION
THREE-DIMENSIONAL CALCULATIONS
TIME DEPENDENCE
CONFIGURATION
LIMITING VALUES
SHUTDOWN
SLIGHTLY ENRICHED URANIUM
COMPUTERIZED SIMULATION
TEMPERATURE DISTRIBUTION
NBSR
loss-of-coolant accident (LOCA) analysis
water drainage
peak clad temperature