Physics and Safety Analysis for the Nist Research Reactor.

Abstract

Detailed reactor physics and safety analyses have been performed for the 20 MW D 2 O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional MCNP Monte Carlo neutron and photon transport calculations were performed to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim arms.The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model including the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP.Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element.

Effect of Shim Arm Depletion in the NBSR

“…MCNP has been previously used to design and analyze the new cold neutron source and to perform neutronics calculations needed for relicensing and safety analysis [7]. In performing the analyses of the changes to the 113 Cd distribution as a function of depletion [4,5], the MCNP model of the NBSR was modified by dividing each shim arm into 20 mesh cells, as shown in Figure 1.…”

Impact of fission neutron energies on reactor antineutrino spectra

“…MCNP5 was used due to extensive benchmarking of all MCNP5-based models used in this study. The list of reactor models investigated for this study are: a typical Westinghouse pressurized water reactor (PWR) [41], the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) [42], and the National Bureau of Standards Reactor (NBSR) at the National Institute of Standards and Technology (NIST) [43]. The Westinghouse PWR operates with LEU fuel, while HFIR and NBSR operate with HEU fuel.…”

Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL

“…While thermal-hydraulic modeling plays a vital role in the design, operation, performance, and safety of a nuclear reactor, the present paper focuses particularly on the safety (or accident analysis) application. Examples of codes used for research reactor thermalhydraulic modeling are RELAP5 [21] for the NIST research reactor [22], the IPR-R1 TRIGA Brazilian reactor [23], and the High Flux reactor (HFR) in Netherlands [24]. The PARET code [25] has been used for the McMaster University research reactor [26], the University of Florida Training Reactor [27], and the NUR Algerian research reactor [28] among several others.…”