Reactor physics evaluation of the TRIGA LEU fuel in the 20 MW NIST research reactor

This work presents the contribution of the Argonauta research reactor in the education and training of human resources in Nuclear Sciences. Since 1965, the Argonauta reactor, located at Rio de Janeiro, Brazil, has been offering theoretical and experimental classes to undergraduate and graduate students. Nuclear Physics and Reactor Physics are the major areas included in the classes provided by the Argonauta’s staff. Recently, Radiochemical classes were integrated in the program. The Argonauta reactor showed to provide substantial contributions to training and formation in the nuclear domain besides improving its capacity to develop know-how in the areas of Nuclear Science.

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Startup, transition core, and molybdenum-99 production upgrade analyses for low-enriched uranium fuel conversion at the University of Missouri research reactor

10.32469/10355/79479 ◽

2020 ◽

Author(s):

◽

Wilson Cowherd

Keyword(s):

Low Power ◽

Research Reactor ◽

The United States ◽

Fuel Conversion ◽

Reactor Physics ◽

University Of Missouri ◽

Safety Margins ◽

Enriched Uranium ◽

The University ◽

The Impact

Under the direction of the United States Department of Energy (DOE) National Nuclear Security Administration (NNSA) Office of Material Management and Minimization (M3) Reactor Conversion Program, the University of Missouri Research Reactor (MURRÂ®) plans to convert from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Low power physics startup test predictions, transition core planning, and analysis for a proposed fission-based molybdenum-99 production upgrade were done in support of LEU fuel conversion. As a first step to LEU fuel conversion, low-power physics tests will be performed to calculate reactor physics parameters. These parameters include flux distributions, coefficients of reactivity, and critical assembly measurements. To facilitate this test, reactor physics calculations were performed using MCNP5 to predict the values of these parameters. Implications of these predictions and areas of uncertainty in the prediction analysis are also discussed. Once MURR completes the testing of the initial LEU core, MURR will enter into a series of transition cycles until steady-state mixed-burnup operation is reached. A Python program was developed that incorporated the constraints of MURR operation while minimizing the time MURR will have to operate atypically during the transition cycles. The impacts of the transition cycles on experiment performance are reported, as well as the number of fuel elements needed. Finally, preliminary analysis on a proposed molybdenum-99 production device at MURR was performed. This analysis shows the impact on the reactor power distribution with implications to predicted safety margins as a part of the larger scope of the experiment analysis.

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Validation of the MORET 5 Monte Carlo Transport Code on Reactor Physics Experiments

Journal of Nuclear Engineering ◽

10.3390/jne2010007 ◽

2021 ◽

Vol 2 (1) ◽

pp. 65-73

Author(s):

Nicolas Leclaire ◽

Isabelle Duhamel

Keyword(s):

Research Reactor ◽

Reactor Physics ◽

Kinetics Parameters ◽

Continuous Energy ◽

Transport Code ◽

Energy Version ◽

Physics Experiments ◽

Monte Carlo Transport ◽

Instantaneous Increase ◽

Pool Type

The MORET 5 code, which has been developed over more than 50 years at IRSN, has recently evolved, in its continuous energy version, from a criticality oriented code to a code also focused on reactor physics applications. Some developments such as the implementation of kinetics parameters contribute to that evolution. The aim of the paper is to present the validation of the code for the keff multiplication factor used in criticality studies as well as for other parameters commonly used in reactor physics applications. Special attention will be paid on commission tests performed in the CABRI French Reactor (CABRI is a pool-type research reactor operated by CEA and located in the Cadarache site in southern France used to simulate a sudden and instantaneous increase in power, known as a power transient, typical of a reactivity-initiated accident (RIA).) and the IPEN/MB-01 LCT-077 benchmark.

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Calculation and Measurement of Kinetics Parameters of Tehran Research Reactor

Volume 5: Fuel Cycle and High and Low Level Waste Management and Decommissioning; Computational Fluid Dynamics (CFD), Neutronics Methods and Coupled Codes; Instrumentation and Control ◽

10.1115/icone17-75036 ◽

2009 ◽

Author(s):

Seyed Abolfazl Hosseini ◽

Naser Vosoughi ◽

Mortaza Gharib ◽

Mohammad Bagher Ghofrani

Keyword(s):

Research Reactor ◽

Reactor Power ◽

Coolant Temperature ◽

Neutron Generation ◽

Reactor Physics ◽

Kinetics Parameters ◽

Second Stage ◽

Cold State ◽

Tehran Research Reactor ◽

Relative Errors

Effective delayed neutron fraction βeff and neutron generation time Λ are important factors in reactor physics calculation and transient analysis. In first stage of this research, these kinetics parameters have been calculated for two states of Tehran research reactor (TRR), i.e. cold (fuel, clad and coolant temperature 20°C) and hot (fuel, clad and coolant temperature 65, 49 and 44°C, respectively) using MTR_PC code. In second stage, these parameters have been measured with experimental method based on Inhour equation. For cold state, calculated βeff and Λ by MTR_PC are 0.008315 and 30.190 μsec, respectively. Same parameters in hot state are 0.008303 and 33.828 μsec, respectively. The measured βeff and Λ for cold state (reactor power is range of 100–200 Watt) are 0.008088 and 32.001 μsec, respectively. The calculated and measured values are in good agreement. Relative errors are % 2.8 for βeff and % 5.6 for Λ which are smaller than the other reported results.

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VOID COEFFICIENT SENSITIVITY ANALYSIS FOR THE TRIGA MARK II REACTOR AT L.E.N.A. (UNIPV)

EPJ Web of Conferences ◽

10.1051/epjconf/202124715005 ◽

2021 ◽

Vol 247 ◽

pp. 15005

Author(s):

D. Portinari ◽

A. Cammi ◽

S. Lorenzi ◽

M. Aufiero ◽

Y. Calzavara ◽

...

Keyword(s):

Sensitivity Analysis ◽

Cross Sections ◽

Research Reactor ◽

Multiplication Factor ◽

Delayed Neutron ◽

Reactor Physics ◽

Fission Spectrum ◽

Challenging Tasks ◽

Generalized Perturbation Theory ◽

Triga Mark Ii

Sensitivity analysis studies the effect of a change in a given parameter to a response function of the system under investigation. In reactor physics, this usually translates into the study of how cross sections and fission spectrum modifications affect the value of the multiplication factor, the delayed neutron fraction or the void coefficient for example. Generalized Perturbation Theory provides a useful tool for the assessment of adjoint weighed functions such as keff and void coefficient sensitivities. In this work, the capability of SERPENT code to perform sensitivity calculation based on GPT is used to study the TRIGA Mark II research reactor installed at L.E.N.A. of University of Pavia. A general sensitivity analysis to the most important reactor’s cross sections has been performed in order to highlight the biggest reactivity contributions. Two numerically challenging tasks related to GPT calculation have been performed thanks to the relatively quick Monte Carlo approach allowed by this reactor: investigating the linearity of the reactivity injection caused by the flooding of the central channel, and calculating the fuel void coefficient sensitivity to the coolant density.

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