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Title: LOCA Analysis for the NIST Research Reactor

Authors:
; ;
Publication Date:
Research Org.:
Brookhaven National Laboratory (BNL), Upton, NY (United States)
Sponsoring Org.:
American Nuclear Society
OSTI Identifier:
1341675
Report Number(s):
BNL-113403-2017-CP
R&D Project: 21798; DN3001010
DOE Contract Number:
SC00112704
Resource Type:
Conference
Resource Relation:
Conference: 2017 ANS Annual Meeting; San Francisco, CA; 20170611 through 20170615
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; NBSR; National Institute of Standards and Technology (NIST); Loss-of-coolant accident (LOCA)

Citation Formats

Baek J. S., Diamond D., and Cheng, L-Y. LOCA Analysis for the NIST Research Reactor. United States: N. p., 2017. Web.
Baek J. S., Diamond D., & Cheng, L-Y. LOCA Analysis for the NIST Research Reactor. United States.
Baek J. S., Diamond D., and Cheng, L-Y. 2017. "LOCA Analysis for the NIST Research Reactor". United States. doi:. https://www.osti.gov/servlets/purl/1341675.
@article{osti_1341675,
title = {LOCA Analysis for the NIST Research Reactor},
author = {Baek J. S. and Diamond D. and Cheng, L-Y},
abstractNote = {},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2017,
month = 6
}

Conference:
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  • 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-enrichedmore » 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.« less
  • Detailed safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The time-dependent analysis of the primary system is determined with a RELAP5 transient analysis model that includes the reactor vessel, the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. A post-processing of the simulation results has been conducted to evaluate minimum critical heat flux ratio (CHFR) using the Sudo-Kaminaga correlation. Evaluations are performed for the following accidents: (1) the control rod withdrawal startup accidentmore » and (2) the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur because of sufficient coolant flow through the fuel channels and the negative scram reactivity insertion.« less
  • The University of Missouri Research Reactor (MURR) has been licensed to operate at 10 MW power since 1974. A preliminary study to increase its present power to a much higher level (/approx/30 MW) shows that this is readily achievable for steady-state operating conditions. The methods used to approach the goal of power upgrade operation include the flattening of the radial power distribution by varying the fuel loading of the plates, and the changing of operating conditions so as to provide a somewhat higher flow rate and greater heat exchanger capability. These changes are easily accomplished without major alterations to thatmore » portion of the primary system within the pool portion of the loop. However, one of the principal considerations in the power upgrade study for licensing is the loss-of-coolant accident (LOCA), which is addressed in this analysis.« less
  • The National Institute of Standards and Technology (NIST) operates a 20-MW research reactor. A 540-mm-diameter cryogenic beam port houses a liquid hydrogen cold neutron source, and a network of 7 neutron guides transports cold neutrons to 13 instruments in the guide hall. The available and planned neutron-scattering instruments are briefly described, along with an advanced cold neutron source that will nearly double the previous flux of cold neutrons.
  • A methodology for calculating inventories for the NBSR has been developed using the MCNPX computer code with the BURN option. A major advantage of the present methodology over the previous methodology, where MONTEBURNS and MCNP5 were used, is that more materials can be included in the model. The NBSR has 30 fuel elements each with a 17.8 cm (7 in) gap in the middle of the fuel. In the startup position, the shim control arms are partially inserted in the top half of the core. During the 38.5 day cycle, the shim arms are slowly removed to their withdrawn (horizontal)more » positions. This movement of shim arms causes asymmetries between the burnup of the fuel in the upper and lower halves and across the line of symmetry for the fuel loading. With the MONTEBURNS analyses there was a limitation to the number of materials that could be analyzed so 15 materials in the top half of the core and 15 materials in the bottom half of the core were used, and a half-core (east-west) symmetry was assumed. Since MCNPX allows more materials, this east-west symmetry was not necessary and the core was represented with 60 different materials. The methodology for developing the inventories is presented along with comparisons of neutronic parameters calculated with the previous and present sets of inventories.« less