Determination of Excess Reactivity and Control Worth for the Core Test Reactor

This report addresses the determination of excess reactivity and control worth for the core test reactor

Nuclear characteristics of a fissioning uranium plasma test reactor with light-water cooling

An analytical study was performed to determine a design configuration for a cavity test reactor. Test section criteria were that an average flux of 10 to the 15th power neutrons/sq cm/sec (E less than or equal to 0.12 eV) be supplied to a 61-cm-diameter spherical cavity at 200-atm pressure. Design objectives were to minimize required driver power, to use existing fuel-element technology, and to obtain fuel-element life of 10 to 100 full-power hours. Parameter calculations were made on moderator region size and material, driver fuel arrangement, control system, and structure in order to determine a feasible configuration. Although not optimized, a configuration was selected which would meet design criteria. The driver fuel region was a cylindrical annular region, one element thick, of 33 MTR-type H2O-cooled elements (Al-U fuel plate configuration), each 101 cm long. The region between the spherical test cavity and the cylindrical driver fuel region was Be (10 vol. % H2O coolant) with a midplane dimension of 8 cm. Exterior to the driver fuel, the 25-cm-thick cylindrical and axial reflectors were also Be with 10 vol. % H2O coolant. The entire reactor was contained in a 10-cm-thick steel pressure vessel, and the 200-atm cavity pressure was equalized throughout the driver reactor. Fuel-element life was 50 hr at the required driver power of 200 MW. Reactor control would be achieved with rotating poison drums located in the cylindrical reflector region. A control range of about 18 percent delta k/k was required for reactor operation

Neutronics Studies on the NIST Reactor Using the GA LEU fuel

The National Bureau of Standards Reactor (NBSR) located on the National Institute of Standards and Technology (NIST) Gaithersburg campus, is currently underway of fuel conversion from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. One particular challenging part of the conversion of the NBSR is the high average flux level (2.5×1014 n/cm2-s) required to maintain experimental testing capabilities of the reactor, without significant changes to the external structures of the reactor. Recently the General Atomics (GA) Training Research Isotopes General Atomics (TRIGA) fuel has shown some promising features as a LEU candidate for the high performance research reactors such as the NBSR. The GA fuel has a long history of success in conversion of research reactors since it was developed in 1980s. The UZrH compound in the GA fuel has seen success in long term TRIGA reactors, and is a proven safe LEU alternative. This study performs a neutronics evaluation of the TRIGA fuel under the schema of the NBSR’s heavy conversion requirements in order to examine whether the TRIGA fuel is a viable option for conversion of the NBSR. To determine the most optimal path of conversion, we performed a feasibility study with particular regard to the fuel dimensions, fuel rod configurations, cladding, as well as fuel structure selection. Based on the outcome of the feasibility study, an equilibrium core is then generated following the NBSR’s current fuel management schema. Key neutronics performance characteristics including flux distribution, power distribution, control rod (i.e., shim arms) worth, as well as kinetics parameters of the equilibrium core are calculated and evaluated. MCNP6, a Monte Carlo based computational modeling software was intensively used to aid in these calculations. The results of this study will provide important insight on the effectiveness of conversion, as well as determine the viability of the conversion from HEU to LEU using the GA fuel

Design and performance of a nuclear reactor simulator for nonnuclear testing of space power systems

Desigg and performance of SNAP-8 nuclear reactor simulator for nonnuclear testing of space power system

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

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 (MURR®) 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.Includes bibliographical reference

Fuel burnup simulation and analysis of the Missouri S&T Reactor

The purpose of this work is to simulate the fuel burnup of the Missouri S&T Reactor. This work was accomplished using the Monte Carlo software MCNP. The primary core configurations of MSTR were modeled and the power history was used to determine the input parameters for the burnup simulation. These simulations were run to determine the burnup for each fuel element used in the core of MSTR. With these simulations, the new predicted isotopic compositions were added into the model. New core configurations were determined, and the burnup corrected model was used to predict the excess reactivity and control rod worth of the three shim rods in the new configurations labeled 125 and 126. The reactor core was arranged to these configurations in order to determine excess reactivity and control rod worth for the shim rods. When core 125 was tested, the reactor could not attain criticality without external source, which was predicted by the simulations. Core 126 had sufficient excess criticality to support measurement. Those measurements had inconsistent results, and indicated methodology errors. Based on this work, it is recommended to revise the code for temperature gradients and Doppler effects. The experimental methodology for the rod drop tests should also be revised to ensure the methods are applicable to core parameters. These corrections allow the model and the reactor itself to better reflect each other so that the predictions by MCNP will better reflect the measurements at MSTR --Abstract, page iii

Master of Science

thesisThe objectives of this thesis are twofold: to determine the highest achievable power levels of the current University of Utah TRIGA Reactor (UUTR) core configuration with the existing three control rods, and to design the core for higher reactor power by optimizing the control rod worth. For the current core configuration, the maximum reactor power, eigenvalue keS, shutdown margin, and excess reactivity have been measured and calculated. These calculated estimates resulted from thermal power calibrations, and the control rod worth measurements at various power levels. The results were then used as a benchmark to verify the MCNP5 core simulations for the current core and then to design a core for higher reactor power. This study showed that the maximum achievable power with the current core configuration and control rod system is 150kW, which is 50kW higher than the licensed power of the UUTR. The maximum achievable UUTR core power with the existing fuel is determined by optimizing the core configuration and control rod worth, showing that a power upgrade of 500 kW is achievable. However, it requires a new control rod system consisting of a total of four control rods. The cost of such an upgrade is $115,000