https://hdl.handle.net/1721.1/67478 Thu, 09 Apr 2026 10:13:57 GMT 2026-04-09T10:13:57Z https://hdl.handle.net/1721.1/75090 Conceptual Design of Molten Salt Loop Experiment for MIT Research Reactor Bean, Malcolm K.; Dewitt, Gregory Lee; Cabeche, Dion T.; Gerrity, Thomas P.; Haratyk, Geoffrey; Kersting, Alyssa R.; Lee, Youho; Virgen, Matthew M.; Lenci, Giancarlo; Lin, Christie; Metzler, Florian; Ochoukov, Roman Igorevitch; Reed, Mark; Sobes, Vladimir; Sugrue, Rosemary M.; Shwageraus, Eugene; Wisniowska, Agata Elzbieta; Youchak, Paul M. Molten salt is a promising coolant candidate for Advanced High Temperature Reactor (AHTR) Gen-IV designs. The low neutron absorption, high thermal capacity, chemical inertness, and high boiling point at low pressure of molten salt coolants could potentially lead to enhanced safety and lower cost of AHTR designs as compared with conventional Light Water Reactors. Improved economics are expected to be a result of the higher possible operating temperatures (700oC), improving thermal efficiency, availability of process heat for industrial applications, and reduced containment costs. Improved safety margins arise from the use of highly robust TRISO particles fuel in either pebble or graphite block form, greater thermal inertia, low pressure and high boiling point of molten salts relative to water cooled reactor designs. Currently, one of the main challenges associated with further advancement of AHTR design is predicting reactor core materials’ interactions with molten salt coolant over long time scales in a radiation environment. In the Fall of 2010, the Nuclear Engineering Design Project Course (22.033/33) undertook the challenge to design a molten salt test loop to be installed in the MIT Research Reactor (MITR) that would recreate anticipated AHTR operating conditions and fill the knowledge gap in understanding of materials behavior in such environment. In addition to simulating neutronic, thermal and chemical conditions similar to those of AHTR, the test loop must also meet the safety and operating requirements of the MITR. During the course, a preliminary design was developed that features an annular in core molten salt flow channel to maximize the volume available for testing materials’ samples and maintaining the salt temperature at 700oC and flow velocity at 6 m/s, while avoiding boiling at the outside surface of the loop, as prescribed by MITR safety requirements. A number of additional requirements were addressed by the students including reactivity insertion, power peaking, tritium production, shielding, and others. The design tasks were subdivided into four key areas of neutronics, thermal hydraulics, chemistry and materials, and instrumentation and control. The molten salt chosen was LiF-BeF[subscript 2] (FLiBe) with lithium enriched in [superscript 7]Li isotope up to 99.995% because this salt is the leading coolant candidate for AHTR. Hastelloy-N was chosen as the material in contact with the molten salt due to its high resistance to corrosion, good material properties at high temperature and extensive use in previous experiments. The presence of corrosion products, free fluorine and production of tritium in the molten salt were found to be important phenomena challenging the loop design. Therefore, various methods for the salt chemistry control and tritium release were evaluated and resulted in a design of multi-component system for monitoring the salt conditions, maintaining redox potential and removing the impurities and tritium from the salt. Another challenge was managing the loop operation given the relatively high freezing point of the salt at about 460oC. Procedures were developed for start-up, steady-state, shutdown and transient operation of the loop. The thermal hydraulic analyses indicate that 1.8 kW of strap heating along the loop outside the core section and a 1.5 to 2 kW pump were required, depending on final design choices. In addition, preliminary cost estimates of constructing the loop experiment at MITR were performed. The main constraints on the choice of the loop’s individual components and diagnostics were: 1) the ability to function at the designed operating temperature, pressure, and flow rates; 2) the ability to function in a nuclear radiation environment; and 3) the necessity to meet MITR safety requirements. A database of vendors for the loop’s components, instrumentation, and diagnostics was compiled. To support further work on the molten salt test loop an electronic library of references was compiled as well. Finally, a number of potential accident scenarios were examined and their effects on the safety and operation of MITR were evaluated and found to represent no danger to the public or interfere with normal operations. Minor leakages of either the reactor water or the molten salt coolant inside the loop were found to be self-sealing with little to no effects on the safety and operation of MITR. A complete failure of the loop’s heating and pumping systems was found to lead to FLiBe’s cooling and freezing inside the loop, with the freezing time ranging from several minutes to ~1 hour depending on the choice of the loop thermal insulation material. Mon, 01 Aug 2011 00:00:00 GMT https://hdl.handle.net/1721.1/75090 2011-08-01T00:00:00Z https://hdl.handle.net/1721.1/75089 Design and Optimization of a High Thermal Flux Research Reactor Via Kriging-Based Algorithm Kempf, Stephanie A.; Hu, Lin-Wen; Forget, Benoit In response to increasing demands for the services of research reactors, a 5 MW LEUfueled research reactor core is developed and optimized to provide high thermal flux within specified limits upon thermal hydraulic performance, cycle length, irradiation utilization, and manufacturability. A novel fuel assembly concept which makes use of integral flux traps is postulated to meet these requirements. Each assembly can be rotated into one of three different configurations to produce flux traps of different size, shape, and neutron energy spectrum within the core. A method for predicting and guiding the search for the optimum geometry was sought. Kriging has been chosen to predict the values of eigenvalue and thermal flux at untested geometric parameters. Because kriging treats all measurements as the sum of a global deterministic function and a stochastic departure from that function, predictions come with a measurement of uncertainty. As a result, the analyst can search the design space for likely improvement, or probe areas of high uncertainty for improvements that might have been missed using other methods. The technique is used in an algorithm for constrained optimization of the design, and a set of best practices for use of this are described. The optimized design produces a peak thermal flux of 1.56 x 10[superscript 14] n/cm[superscript 2]s. Safety is demonstrated by presentation of reactivity feedback coefficients and the results of loss of flow and reactivity insertion transient analysis. A single fission target can be used to produce 96 6-day Ci of [superscript 99]Mo per week. When the reactor is oriented to take advantage of high fast flux, steels can be subjected to damage rates of 5.76 dpa per year. Silicon carbide can be damaged at a rate of 2.79 dpa/y. The concept is a safe, versatile, proliferation-resistant means of supplying current and future irradiation needs. Wed, 01 Jun 2011 00:00:00 GMT https://hdl.handle.net/1721.1/75089 2011-06-01T00:00:00Z https://hdl.handle.net/1721.1/75088 Estimate of Radiation Release from MIT Reactor with Low Enriched Uranium (LEU) Core During Maximum Hypothetical Accident Plumer, Kevin E.; Newton, Thomas H., Jr.; Forget, Benoit In accordance with a 1986 NRC ruling, the MIT Research Reactor (MITR) is planning on converting from the use of highly enriched uranium (HEU) to low enriched uranium (LEU) for fuel. A component of the conversion analysis includes calculating the maximum hypothetical accident (MHA) dose implications for the two types of fuel. In this work, the dose levels at the site exclusion area boundary were calculated for the MITR MHA using both the HEU and LEU models of the MITR core. The core inventories from the reactor were calculated using the ORIGEN-S pointdepletion code linked to the MITR spectrum. The MITR spectrum was used from an MCODE simulation of the equilibrium LEU and HEU versions of the core. Release fractions from the melted fuel to containment were established using melt test data from plate-type fuel as well as modified release fractions from NRC Regulatory Guides. The dose paths considered were the same paths used in the previous work, consisting of atmospheric release through potential containment leakage as well as direct and scattered gamma dose from the containment source term. The Total Effective Dose Equivalent (TEDE) values were calculated in addition to the whole body and thyroid doses. For dose comparison the LEU thermal power was 17% higher than the HEU thermal power in order to provide equivalent total flux levels to the experimental ports. The results showed that the LEU core operating at 7 MW will yield TEDE levels 22% higher than the HEU core operating at 6 MW for equivalent release fractions. The two-hour dose at the exclusion area boundary from the LEU core operating at 7 MW using the plate-type fuel release fractions was 0.440 rem at 21 m and 0.344 rem at 8 m, while the dose from the HEU core operating at 6 MW was 0.361 rem at 21 m and 0.281 rem at 8 m. These doses are within the public dose NRC regulatory limit of 0.500 rem TEDE. Wed, 01 Jun 2011 00:00:00 GMT https://hdl.handle.net/1721.1/75088 2011-06-01T00:00:00Z https://hdl.handle.net/1721.1/75087 Developing Fuel Management Capabilities Based On Coupled Monte Carlo Depletion in Support of the MIT Research Reactor Conversion Romano, Paul Kollath; Newton, Thomas H., Jr.; Forget, Benoit Pursuant to a 1986 NRC ruling, the MIT Reactor (MITR) is planning on converting from the use of highly enriched uranium (HEU) to low enriched uranium (LEU) for fuel. Prior studies have shown that the MITR will be able to operate using monolithic U-Mo LEU fuel while achieving neutron fluxes close to that of an HEU core. However, to date, detailed studies on fuel management and burnup while using LEU fuel have not been performed. In this work, a code package is developed for performing detailed fuel management studies at the MITR that is easy to use and is based on state-of-the-art computational methodologies. A wrapper was written that enables fuel management operations to be modeled using MCODE, a code developed at MIT that couples MCNP to the point-depletion code ORIGEN. To explicitly model the movement of the control blades in the MITR as the core is being depleted, a criticality search algorithm was implemented to determine the critical position of the control blades at each depletion timestep. Additionally, a graphical user interface (GUI) was developed to automate the creation of model input files. The fuel management wrapper and GUI were developed in Python, with the PyQt4 extension being used for GUI-specific features. The MCODE fuel management wrapper has been shown to perform reliably based on a number of studies. An LEU equilibrium core was modeled and burned for 640 days with the fuel being moved in the same pattern every 80 days. The control blade movement and nuclide concentrations were shown to be in agreement with what one would intuitively predict. The fuel management capabilities of REBUS-PC and the MCODE fuel management wrapper were compared by modeling the same refueling scheme using an HEU core. The element power peaking factors for the two models showed remarkable agreement. Together, the fuel management wrapper and graphical user interface will help the staff at the MITR perform in-core fuel management calculations quickly and with a higher level of detail than that previously possible. Mon, 01 Jun 2009 00:00:00 GMT https://hdl.handle.net/1721.1/75087 2009-06-01T00:00:00Z https://hdl.handle.net/1721.1/75086 Design of a Low Enrichment, Enhanced Fast Flux Core for the MIT Research Reactor Ellis, T.S.; Forget, Benoit; Kazimi, Mujid S.; Newton, T.; Pilat, Edward E. Worldwide, there is limited test reactor capacity to perform the required irradiation experiments on advanced fast reactor materials and fuel designs. This is particularly true in the U.S., which no longer has an operating fast reactor but depends upon two aging thermal reactors for testing the behavior of various materials in an irradiation environment. The MIT Research Reactor is planning for a new core to end the need for highly-enriched uranium and operate the reactor with uranium enrichments under 20%. This study explores the use of the central region in the newly proposed MIT reactor core to boost the production of fast neutrons, thus making the new core more beneficial for materials testing. The Fast Flux Trap introduces a region of fissile material surrounding a central irradiation facility which is cooled by liquid lead-bismuth eutectic. This arrangement maximizes the fast neutron production by avoiding neutron moderation in the center. The fissile material, arranged in a tight hexagonal pin array, can be uranium enriched in either [superscript 235]U or [superscript 233]U, to the limit allowable for non-proliferation. Insertion of the Fast Flux Trap in the proposed low enriched uranium core operated at a 10 MW power level, can provide a 252-271% higher fast neutron flux than the previously proposed designs with low enriched fuel for the MIT research reactor and a 235%-253% higher fast neutron flux than the existing highly enriched uranium MITR-II core at 5 MW. This new core fast flux capability is within a factor of 2 to 4 of the much larger national test reactors, the Advanced Test Reactor and the High Flux Isotope Reactor, and hence can allow the MIT research reactor to be more useful for fast irradiation. The work covered both steady state and transient events involving the Fast Flux Trap, using the Monte Carlo N-Particle (MCNP) transport code. It was shown that the power distribution within the Fast Flux Trap pins as well as the plates in the rest of the core will be satisfactory; in other words, no excessive power peaking will develop. The limits of the Fast Flux Trap lifetime were found to exceed the expected licensing time of the new core. Furthermore, the reactivity implications of metallic coolant leaks, water flooding of the Fast Flux Trap and various possible test materials were all found to be acceptable. The loss of flow following a pump trip event was analyzed using the RELAP5-3D code, and found not to result in excessive temperatures with regard to materials strength and corrosion resistance. While the specific design developed in this dissertation is particular to the MIT research reactor core, the Fast Flux Trap design concept can potentially be applied in other reactor cores so that other thermal spectrum research and test reactor facilities can benefit from this enhanced capability. Sun, 01 Feb 2009 00:00:00 GMT https://hdl.handle.net/1721.1/75086 2009-02-01T00:00:00Z https://hdl.handle.net/1721.1/75085 Evaluation of the Thermal-Hydraulic Operating Limits of HEU-LEU Transition Cores for the MIT Research Reactor Wan, Yunzhi; Hu, Lin-Wen The MIT Research Reactor (MITR) is in the process of conducting a design study to convert from High Enrichment Uranium (HEU) fuel to Low Enrichment Uranium (LEU) fuel. The currently selected LEU fuel design contains 18 plates per element, compared to the existing HEU design of 15 plates per element. A transitional conversion strategy, which consists of replacing three HEU elements with fresh LEU fuel elements in each fuel cycle, is proposed. The objective of this thesis is to analyze the thermo-hydraulic safety margins and to determine the operating power limits of the MITR for each mixed core configuration. The analysis was performed using PLTEMP/ANL ver 3.5, a program developed for thermo-hydraulic calculations of research reactors. Two correlations were used to model the friction pressure drop and enhanced heat transfer of the finned fuel plates: the Carnavos correlation for friction factor and heat transfer, and the Wong Correlation for friction factor with a constant heat transfer enhancement factor of 1.9. With these correlations, the minimum onset of nucleate boiling (ONB) margins of the hottest fuel plates were evaluated in nine different core configurations, the HEU core, the LEU core and seven mixed cores that consist of both HEU and LEU elements. The maximum radial power peaking factors were assumed at 2.0 for HEU and 1.76 for LEU in all the analyzed core configurations. The calculated results indicate that the HEU fuel elements yielded lower ONB margins than LEU fuel elements in all mixed core configurations. In addition to full coolant channels, side channels next to the support plates that form side coolant channels were analyzed and found to be more limiting due to higher flow resistance. The maximum operating powers during the HEU to LEU transition were determined by maintaining the minimum ONB margin corresponding to the homogeneous HEU core at 6 MW. The recommended steady-state power is 5.8 MW for all transitional cores if the maximum radial peaking is adjacent to a full coolant channel and 4.9 MW if the maximum radial peaking is adjacent to a side coolant channel. Fri, 01 May 2009 00:00:00 GMT https://hdl.handle.net/1721.1/75085 2009-05-01T00:00:00Z https://hdl.handle.net/1721.1/75084 Pressure Drop Measurements and Flow Distribution Analysis for MIT Research Reactor with Low-Enriched Fuel Yuen-Ting Wong, S.; Hu, Lin-Wen; Kazimi, Mujid S. The MIT Nuclear Research Reactor (MITR) is the only research reactor in the United States that utilizes plate-type fuel elements with longitudinal fins to augment heat transfer. Recent studies on the conversion to low-enriched uranium (LEU) fuel at the MITR, together with the supporting thermal hydraulic analyses, propose different fuel element designs for optimization of thermal hydraulic performance of the LEU core. Since proposed fuel design has a smaller coolant channel height than the existing HEU fuel, the friction pressure drop is required to be verified experimentally. The objectives of this study are to measure the friction coefficient in both laminar and turbulent flow regions, and to develop empirical correlations for the finned rectangular coolant channels for the safety analysis of the MITR. A friction pressure drop experiment is set-up at the MIT Nuclear Reactor Laboratory, where static differential pressure is measured for both flat and finned coolant channels of various channel heights. Experiment data show that the Darcy friction factors for laminar flow in finned rectangular channels are in good agreement with the existing correlation if a pseudo-smooth equivalent hydraulic diameter is considered; whereas a new friction factor correlation is proposed for the friction factors for turbulent flow. Additionally, a model is developed to calculate the primary flow distribution in the reactor core for transitional core configuration with various combinations of HEU and LEU fuel elements. Mon, 01 Sep 2008 00:00:00 GMT https://hdl.handle.net/1721.1/75084 2008-09-01T00:00:00Z