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Mitigating the release of tritium produced from neutron irradiation of molten salts containing lithium or beryllium is a technical challenge for several advanced reactor designs. In a pebble bed Fl...

Advanced reactor applications that use a molten fluoride salt coolant and graphite moderator are under consideration as next-generation energy technologies. For molten salts with lithium or berylli...

The updated safety basis of the Advanced Test Reactor (ATR) has specific criteria for experiments that are cooled by the primary coolant. It requires that during a Condition 2 transient (where flow coastdown is typically the most limiting), the Critical Heat Flux ratio (CHFR) is greater than 2.0 after consideration of several power multipliers. This safety criteria is often overly conservative for fueled experiments and restricts the acceptance of experiments. In the updated safety basis, a provision for a stochastic approach is included. In this study, coupled Dakota-RELAP5 and RAVEN-RELAP5 are used to propagate uncertainties. The objective of this study is to provide a quantitative basis for utilizing the provisions in the ATR experiment safety criteria. Three different fueled experiments, MP-1, FSP-1, and AFIP-6 are considered. RELAP5 models for each experiment are used to simulate the flow coastdown transient. Several thermal-hydraulic correlations are assessed to determine the Onset of Significant Voiding (OSV), Onset of Flow Instability (OFI), and Critical Heat Flux (CHF). The OSV and OFI correlations considered indicate that there is a large safety margin. The CHF correlations also indicate that there is a moderate to large safety margin. The high mass flux of the experiments exceed the Look-up Table (LUT) correlation range. This results in an artificially restricted prediction of CHF value. A high mass flux modifier was implemented to improve LUT applicability. Due to the specifications of the experiments considered, the correlations must be applicable to higher mass flux ranges. It was also found that, often, the conventional CHFR >2.0 criterion is not met. However, the statistical CHFR margin ( μ − 3 σ > 1.2 ) is met. This result suggests that new criteria, while maintaining adequate safety margins, can be used to expand the operating envelope of high power fueled experiments at ATR. The Pearson ranking indicates that the power level and mass flux have the largest impact on the CHFR and OFI predictions. If a reduction of uncertainty in predictions is sought, the factors affecting fuel experiment fission power should be prioritized for scrutiny, followed by the mass flow rate.

The first Solid Fueled Thorium Molten Salt Reactor (TMSR-SF1), a Fluoride-salt-cooled High Temperature Reactor (FHR) designed by the TMSR group of Chinese Academy of Science (CAS), is expected to be constructed in the near years. The combination of TRistructural-ISOtropic (TRISO) coated-fuel particles, pebble bed core, and Flibe coolant leads to various neutronics challenges, such as the double heterogeneity, distribution of TRISO particles, high density packed pebble bed, neutronics characteristics of Flibe salt, and etc. For the purpose of evaluating candidate neutronics codes, benchmark study will be conducted for the infinite fuel pebble lattice with explicitly described fuel particle distribution. For investigating key physics phenomena of the randomly packed pebble bed core, two coordinate generation methods will be explored in this study: Discrete Element Method (DEM) that accounts for inter-particle forces and a mathematical model (MM) that fills arbitrary domains through simple geometric rules on the addition of particles. While the former generates packing distributions closer to real-world scenarios, the latter is more computationally effective. The pebble loading effect on eigenvalues and fission power distribution will be studied. Emphasis is placed on the Flibe salt density reactivity effect -- the most unique feature of the FHR comparing to the conventional High Temperature Gas-cooled Reactors (HTGRs). Neutron balance based reactivity decomposition method will be employed for describing the underlying physics.

A machine learning based system performance prediction model is currently created to support the development of autonomous control for small reactors, such as the Transportable Fluoride-salt-cooled High-temperature Reactor (TFHR), which is a 20 MWth compact core proposed by MIT for remote site applications. The prediction model consists of a reactor physics model and a thermal-hydraulic model. It is presently constructed using support vector regression (SVR) with training data generated by multiple cases of a one-dimensional reactor system model. A particle filtering framework is utilized to estimate and update model parameters with noisy instrument measurements. Verifications of the prediction and filtering models have been carried out using the TFHR reactivity insertion events. Satisfactory performance in predicting the core behavior and in recognizing transient parameters such as reactivity insertion timing and rate is concluded.

The development of the Fluoride-salt-cooled High-temperature Reactor (FHR) is ongoing. The molten salt coolant provides several inherent safety characteristics that traditional water-cooled nuclear reactors lack. The first test reactor to be built in China is a pebble-bed type. Research efforts are currently focused on determining the heat transfer characteristics of such a reactor. In the past, the pebble-bed geometry has been studied for the development of the High Temperature Gas-cooled Reactor (HTGR). The previous work has indicated that the profile of the Nusselt number depends on both the lattice structure and pebble contact. In this work, a new series of simulations is sought to characterize heat transfer in an FHR pebble-bed. Due to the physical properties of molten salt and the FHR design, the Reynolds number is roughly three orders of magnitude lower than the HTGR while the Prandtl number is roughly an order of magnitude higher. Thus, the flow in an FHR lattice is expected to be unlike the HTGR counterpart. Three different lattices (Simple Cubic (SC), Body Centered Cubic (BCC), and Face Centered Cubic (FCC)) are considered. The study found that the different lattice geometries have a significant impact on the heat transfer characteristics. The SC lattice has a simplified flow that minimizes pressure-drop and Nusselt number. The BCC and FCC lattice have a complicated flow structure resulting in an order of magnitude higher pressure-drop and almost double the Nusselt number. It was concluded that heat transfer characteristics in a random pebble-bed would be highly sensitive to the mixing, and thus turbulent heat transfer, along the flow path. It was also found that while an interstitial gap allows heat transfer characteristics to be replicated, the pressure loss is significantly lower, while the computational costs increase. Correlations available in the literature (developed primarily for HTGRs) have a mixed performance in prediction of the Nusselt number. The result suggests that new experimental efforts are necessary to develop correlations that pertain to FHR conditions.

Fluoride salts are crucial to achieving the benefits of Fluoride-salt-cooled High-temperature Reactors (FHRs). Intensive studies and modeling are being performed in different countries aimed at FHR technology development. Better understanding of liquid fluoride salt coolant properties and their uncertainties are needed for design and analysis of nuclear facilities. The main objective of the present study is to perform a literature survey of the experimental data, numerical studies, reports, and other review compilations for the main thermophysical properties of liquid fluoride salts. The review recommends density, heat capacity, thermal conductivity, and viscosity properties for potential primary coolants LiF-BeF₂ (flibe) and NaF-ZrF₄ (nafzirf) and secondary coolant LiF-NaF-KF (flinak). It is found that there is a dearth of experimental data. Thus recommended property uncertainties are included, which range from 2 to 20%, complete with a discussion on the recommended definition of the uncertainties. The recommended properties and their uncertainties provide a reference point for incorporating uncertainty in modeling to understand its impacts and for code benchmarking and validation. Keywords: FHR; Fluoride salt; Thermophysical properties; Flibe; Nafzirf; Flinak

Summary The current study explores an innovative option for demonstrating the Fluoride-salt–cooled High- temperature Reactor (FHR) technologies with a reactor-driven subcritical facility. The FHR uses clean salt coolants, carbon-matrix coated-particle fuel similar to that used in High-temperature Gas-cooled Reactors and can be coupled to a nuclear air-Brayton combined cycle. Recent assessments indicate favorable economics and safety characteristics, but no FHR has been built. The question is what experimental facilities should be constructed to reduce technical uncertainties before a decision to build a test or demonstration reactor? The MIT Reactor design and license would allow the construction and operation of a subcritical facility with 700°C salt circulating through multiple full-width partial-height fuel assemblies operating with a power density up to 60% of a commercial FHR. This option would allow hot systems testing as a major step toward building the test or demonstration reactor. Preliminary system design, power control options, testing capabilities, and key nuclear characteristics of such a reactor-driven subcritical facility are described. A method of deriving subcritical multiplicity using surface source has been proposed and verified in this study. Finally, the neutronic impacts on the driver facility, ie, the MIT Reactor, have been evaluated.

The conversion of the Massachusetts Institute of Technology Reactor (MITR) from the use of highly-enriched uranium (HEU) fuel-plate assemblies to low-enriched uranium (LEU) by replacing the HEU fuel plates with specially designed General Atomics (GA) uranium zirconium hydride (UZrH) LEU fuel rods is evaluated. The margin to critical heat flux (CHF) in the core, which is cooled by light water at low pressure, is evaluated analytically for steady-state operation. A form of the Groeneveld CHF lookup table method is used and described in detail. A CHF ratio of 1.41 was found in the present analysis at 10 MW with engineering hot channel factors included. Thus, the nominal reactor core power, and neutron flux performance, would need to be reduced by at least 25% in order to meet the regulatory requirement of a minimum CHF ratio of 2.0.

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. At 6 MW, it delivers neutron flux and energy spectrum comparable to light water reactor (LWR) power reactors in a compact core using highly enriched uranium (HEU) fuel. In the framework of non-proliferation policy, the international community aims to minimize the use of HEU in civilian facilities. Within this context, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel based on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MITR. The current study focuses on the impacts of MITR Maximum Hypothetical Accident (MHA), which is also the Design Basis Accident (DBA), with LEU fuel. The MHA for the MITR is postulated to be a coolant flow blockage in the fuel element that contains the hottest fuel plate. It is assumed that the entire active portion of five fuel plates melts. The analysis shows that, within a 2-h period and by considering all the possible radiation sources and dose pathways, the overall off-site dose is 302.1 mrem (1 rem = 0.01 Sv) Total Effective Dose Equivalent (TEDE) at 8 m exclusion area boundary (EAB) and a higher dose of 392.8 mrem TEDE is found at 21 m EAB. In all cases the dose remains below the 500 mrem total TEDE limit goal based on NUREG-1537 guidelines.

The MIT Nuclear Reactor Laboratory (NRL) has irradiated lithium-beryllium fluoride (flibe) salt as part of an on-going U.S. Department of Energy-funded Integrated Research Project to develop a Fluoride Salt High-Temperature Reactor (FHR). As part of this project, the NRL has carried out two irradiations of FHR materials in static flibe at 700°C in the MIT Research Reactor. These irradiations marked the start of a program evaluating the tritium production and release from the fluoride salt system at high temperature; in particular, there is interest in the evolution of tritium from the salt into solid materials and cover gasses. This paper describes the experience gained from the irradiation of flibe with respect to the detection of tritium. It covers the development of techniques for monitoring the evolution of tritium from the salt during irradiation and the factors particular to the FHR system that influence this process, including the radiolytic production and release of volatile fluorine and f...

The technology for the 20-MW(thermal) Transportable Fluoride Salt–Cooled High-Temperature Reactor (TFHR) is proposed by Massachusetts Institute of Technology for off-grid applications such as Antarctic bases and remote mining sites. The preliminary thermal-hydraulic analyses and improvements based on a 1/12th full-core model were performed using three-dimensional computational fluid dynamics (CFD). A benchmark study was conducted by comparing the CFD results against empirical correlations and experimental data obtained by Cooke, Silverman, and Grele. In the 1/12th full-core analysis, three practical considerations that may challenge the TFHR temperature limits are evaluated as bounding analysis. These include (1) helium gap between fuel compact and graphite block, (2) thermal conductivity degradations of graphite matrix due to neutron irradiation, and (3) full-core scale power distribution obtained from neutronic calculations. These design considerations lead to insufficient margin between the nor...

As part of an effort to evaluate water-based nanofluids for nuclear applications, preparation and characterization has been performed for nanofluids being considered for MIT’s nanofluid heat transfer experiments. Three methods of generating these nanofluids are available: creating them from chemical precipitation, purchasing the nanoparticles in powder form and mixing them with the base fluid, and direct purchase of prepared nanofluids. Characterization of nanofluids includes colloidal stability, size distribution, concentration, and elemental composition. Quality control of the nanofluids to be used for heat transfer testing is crucial; an exact knowledge of the fluid constituents is a key to uncovering mechanisms responsible for heat transport enhancement.

Further understanding of tritium transport mechanisms in the combined molten fluoride salt and graphite environment is necessary for the design and licensing of a Fluoride-Salt-Cooled High-Temperature Reactor (FHR). The three in-core fluoride salt irradiations completed at the Massachusetts Institute of Technology Reactor (MITR) are a useful parallel for studying transport phenomena expected in a FHR environment. During the irradiations, evolution of tritium from the flibe salt was monitored and compared to the calculated total generation rate. A difference of 22 ± 10% between the integrated calculated tritium generation rate and the total release was measured for the third MITR irradiation (FS-3). The fraction of tritium which was not released from the salt could be explained by tritium retention in graphite. For post irradiation examination, a thermal desorption furnace was used to heat nuclear graphite samples in order to release and measure retained tritium. The desorption analysis in this work utilized seven subsections of graphite from the second salt irradiation (FS-2); three from a disc of IG-110U and four from ARB matrix graphite. Observed desorption versus temperature as well as total tritium content in the samples after irradiation indicate that the graphites were not volumetrically saturated with tritium, but rather tritium retention was likely limited to the near-surface region. Measurements of the samples resulted in 2.90 ± 0.29 μCi/mm2 of tritium retained by IG-110U and 1.83 ± 0.31 μCi/mm2 for ARB during the 300 h FS-2 in-core irradiation. Based on the desorption measurements, the estimated total tritium retention in graphite from the FS-2 samples is consistent with the tritium release measurements from the FS-3 experiment.

The Fluoride salt-cooled High-temperature Reactor (FHR) is a subset of next-generation Nuclear Reactors that are under active development. The proposed pebble-bed FHR designs use a high Prandtl number molten salt coolant, 2LiF–BeF2, with spherical fuel elements. In order to determine convective heat transfer coefficients, packed-bed correlations available in the literature are used. However, past correlations were developed for either gas or water as the coolant, which have a low Prandtl number ( N Pr ≈ 1 ). This study assesses the applicability of such correlations using numerical simulations. A range of Reynolds numbers ( 10 2 ⩽ N Re ≤ 2.25 · 10 3 ) is evaluated that encompass the proposed FHR designs ( N Pr ≈ 16 ). The study finds that most conventional correlations miss the qualitative and quantitative trend predicted by both k e and LES turbulence models. The Meng correlation (Meng et al., 2012) was found to provide a good agreement with numerical results. Both turbulence models have good agreement up to N Re 10 3 for surface-averaged parameters. At higher flowrates, the secondary flows are more prominent for the LES solutions – however, the impact on the convective heat transfer coefficient is minimal.

The Massachusetts Institute of Technology Reactor (MITR) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. In the framework of non-proliferation policy, research and test reactors have started a program to convert highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. A new type of LEU fuel based on a high-density alloy of uranium and molybdenum (U-10Mo) is expected to allow conversion. A transition core plan, from 22 fresh LEU fuel elements to 24 in an equilibrium state, is proposed. This study evaluates the thermal-hydraulic safety margins of each transition cycle and state (i.e. beginning, middle and end). The STAT7 and RELAP5 codes are used in the study. STAT7 provides an integrated platform for statistical propagation of uncertainties in determining steady-state thermal-hydraulic margins. A RELAP5 model was created to verify MITR's STAT7 model, and provide transient analyses. The steady-state analyses determined that, during the entire transition from a fresh to equilibrium core, the Limiting System Safety Settings (LSSS) has a 10.2% margin to the minimum power at which onset of nucleate boiling (ONB) is precluded at 3 σ confidence. The loss of flow (LOF) transient, which is the most limiting anticipated transient for the MITR, was simulated. With a conservative initial state of the reactor (operating at μ + 3 σ power level and LSSS mass flow rate), the maximum calculated centerline fuel temperature was 115 °C, which is a significantly lower value than the blistering limit of 350 °C. The cladding wall temperatures did not exceed the ONB temperature throughout the transient. Therefore, during LOF, no nucleate boiling is expected, precluding any critical heat flux trigger. This study indicates that the proposed LEU transition core specifications have significant margin to thermal-hydraulic limits, during steady-state operation and the LOF transient.

The Massachusetts Institute of Technology Reactor (MITR) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. At 6 MW, it delivers neutron flux and energy spectrum comparable to power light water reactors (LWRs) in a compact core using highly enriched uranium (HEU) fuel. In the framework of non-proliferation policy, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel based on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow conversion of U.S. high performance research reactors (USHPRRs) like the MITR. The Preliminary Safety Analysis Report (PSAR) for the MITR LEU conversion has been submitted and currently under review by the U.S. Nuclear Regulatory Commission (NRC). A transition core plan, from 22 fresh LEU fuel elements (i.e., beginning-of-life) gradually to 24 of them arranged in an equilibrium configuration, is expected to serve as an appendix chapter in the final SAR. The current study presents the fuel cycle development, which eventually leads to the transition core plan. The results confirm the equilibrium state, where both the shim bank movement (i.e., core reactivity) as well as the content of fissile materials stabilize, can be achieved by fixed-pattern fuel management. Fission density has been evaluated for a number of fully discharged LEU fuel elements, using a best-estimate approach. The fuel cycle calculations also generate power profiles at each core state. A steady-state thermal-hydraulic safety analysis has thus been performed, where onset of nucleate boiling (ONB) is considered as the safety criterion. The results confirm significant margin to ONB at all analyzed transition and equilibrium fuel cycle states.

This study experimentally investigates the reduction efficiency of tritium permeation through 316 stainless steel tubing coated with alumina as a tritium permeation barrier (TPB) in support of the development of molten salt nuclear reactors, particularly for fluoride salt-cooled high-temperature nuclear reactors (FHRs). The TPB coatings composed ofan intermediate bond layer of NiCr, a transition layer of NiCr + alumina, and a pure alumina layer were successively added onto the outer surface of commercial 316 stainless steel tubing via plasma thermal spray. In order to generate a continuous gaseous tritium source, 35 g of purified natural-lithium FLiBe salt was irradiated by thermal neutron flux at 620 °C in the Massachusetts Institute of Technology Research Reactor (MITR). The preliminary results suggest that the TPB coatings on tube surfaces significantly reduced the tritium permeation rate at 700 °C. To get a better understanding of the TPB, the microstructure of the coated tubes was characterized with various techniques.

International audience; This paper reports the testing results of radiation resistant fiber Bragg grating (FBG) in random air-line (RAL) fibers in comparison with FBGs in other radiation-hardened fibers. FBGs in RAL fibers were fabricated by 80 fs ultrafast laser pulse using a phase mask approach. The fiber Bragg gratings tests were carried out in the core region of a 6 MW MIT research reactor (MITR) at a steady temperature above 600°C and an average fast neutron (>1 MeV) flux >1.2 × 10$^{14}$ n/cm$^2$/s. Fifty five-day tests of FBG sensors showed less than 5 dB reduction in FBG peak strength after over 1 × 10$^{20}$ n/cm$^2$ of accumulated fast neutron dose. The radiation-induced compaction of FBG sensors produced less than 5.5 nm FBG wavelength shift toward shorter wavelength. To test temporal responses of FBG sensors, a number of reactor anomaly events were artificially created to abruptly change reactor power, temperature, and neutron flux over short periods of time. The thermal sensitivity and temporal responses of FBGs were determined at different accumulated doses of neutron flux. Results presented in this paper reveal that temperature-stable Type-II FBGs fabricated in radiation-hardened fibers can survive harsh in-pile conditions. Despite large parameter drift induced by strong nuclear radiation, further engineering and innovation on both optical fibers and fiber devices could lead to useful fiber sensors for various in-pile measurements to improve safety and efficiency of existing and next generation nuclear reactors.

This paper reports testing results of radiation resilient fiber Bragg grating (FBG) in radiation resistant fibers in the nuclear reactor core at MIT Research Reactor Lab. FBGs were fabricated by 140-fs ultrafast laser pulse using a phase mask approach. In-core test of fiber Bragg gratings was carried out in the core region of a 6-MW research reactor at temperature > 600°C and an average fast neutron (>1 MeV) flux >1×1014 n/s/cm2. First 100-day tests of FBG sensors shows less than 5 dB reduction in FBG peak strength after over 1×1020 n/cm2 of accumulated fast neutron dosage. To test temporal responses of FBG sensors, a number of reactor anomaly events were artificially created to abruptly change reactor power, temperature, and neutron flux over short periods of time. The thermal optical coefficients and temporal responses of FBG sensors are determined at different accumulated dosages of neutron flux. Results presented in this paper reveals that temperature-stable Type-II FBGs fabricated in radiation-hardened fibers could be used as sensors to perform in-pile measurements to improve safety and efficiency of existing and next generation nuclear reactors., International Conference on Optical Instruments and Technology 2017: Advanced Optical Sensor and Applications, October 28-30, 2017, Beijing, China, Series: Proceedings of SPIE; no. 10618

The Massachusetts Institute of Technology Reactor (MITR) is a 6 MW tank-type reactor with a compact core design. It is moderated and cooled by light-water and has a heavy-water reflector. The MITR is a unique test bed for a wide range of nuclear material and fuel research. In one of the recent fuel cycles, detectors found elevated neutron levels on the reactor top lid during the operation with the Water-loop facility installed. A preliminary investigation theorized that the neutrons were streaming out via a thin CO2 gas gap, which is used as a thermal isolation space for the in-core experiment. The current study employs neutron transport simulations to investigate the physical interpretation of the observed phenomenon. The state-of-the-art general purpose Monte Carlo code MCNP5 is adopted as the computational tool. The calculation results identify the typical characteristics of the streaming effect by quantifying the fraction of upward-streaming neutrons. Efforts have also been made to explore feasible engineering options for mitigating this neutron streaming from the Water-loop facility. It is concluded that a borated guide tube is not sufficient to reduce the neutron level on the reactor top lid; however, a stepped thimble design is effective in reducing the neutron source contribution.

Fluoride-salt-cooled, high-temperature reactor (FHR) technology combines the robust coated-particle fuel of high-temperature, gas-cooled reactors with the single phase, high volumetric heat capacity coolant of molten salt reactors and the low-pressure pool-type reactor configuration of sodium fast reactors. FHRs have the capacity to deliver heat at high average temperature, and thus to achieve higher thermal efficiency than light water reactors. Licensing of the passive safety systems used in FHRs can use the same framework applied successfully to passive advanced light water reactors, and earlier work by the NGNP and PBMR projects provide an appropriate framework to guide the design of safety-relevant FHR systems. This paper provides a historical review of the development of FHR technology, describes ongoing development efforts, and presents design and licensing strategies for FHRs. A companion review article describes the phenomenology, methods and experimental program in support of FHR.

Due to their combination of high-temperature coated-particle fuel, molten salt coolant and related materials requirements, fluoride-salt-cooled, high-temperature reactors (FHRs) exhibit different thermal hydraulic, neutronic and structural mechanics phenomena compared to conventional and more extensively studied other advanced nuclear reactor concepts. This paper highlights key phenomena unique to FHRs, and reviews general issues for developing, verifying, and validating evaluation models for FHR technology that may apply to other advanced reactors. System response codes that are appropriate to predict the behavior of FHRs under steady-state operation and licensing basis events are identified, along with experimental data needs to validate these codes. FHR materials requirements are highlighted, and the missions and licensing program for an FHR test reactor, providing ultimate validation data and proof of concept before a commercial prototype is built, are presented. This review draws upon information compiled in a series of four white papers based on FHR experts workshops held in 2012 in the U.S.

Hydride metallic fuel is known to have several advantages compared to uranium oxide fuel, including enhanced thermal conductivity, large negative temperature reactivity coefficient, and reduced requirement for water moderator. In order to investigate the fuel performances under the irradiation condition, the hydride fuel irradiation (HYFI) experiment was designed and tested in the MIT Research Reactor-II (MITR-II) core. The HYFI facility consists of three fuel rod capsules with cover gas system, and instrumentation and data acquisition system. Each fuel rod capsule includes a fuel rod housing installed with two K-type thermocouples (TCs) in the fuel center and zircaloy cladding outer surface. The gap between the fuel pellets and cladding is filled with lead-bismuth eutectic (LBE) to reduce the thermal resistance and thereby the peak fuel temperature. This fuel rod is encapsulated in a titanium outer capsule. The assembly was designed to accommodate up to three fuel capsules and was irradiated in a center position of the MITR-II core. Neutronic and thermal analyses were performed iteratively for the final design of the HYFI facility, and installation and operation are governed by the MITR-II Technical Specifications. MCNP code was used for the neutronics analysis and FLUENT 6.3 code was used for thermal–hydraulic calculation. Initial irradiation results suggest that neutronics and thermal predictions are in good agreement with experimental results.

The fluoride salt-cooled high-temperature reactor (FHR) is an advanced reactor concept, which uses high-temperature TRISO fuel with a low-pressure liquid salt coolant. The design of a fluoride salt-cooled high-temperature test reactor (FHTR) is a key step in the development of the FHR technology and is currently in progress in both China and the United States. An FHTR based on a pebble bed core design with coolant temperature 600 degrees C to 700 degrees C is being planned for construction by the Chinese Academy of Sciences' Thorium Molten Salt Reactor Research Center, Shanghai Institute of Applied Physics (SINAP). This paper provides a preliminary thermal-hydraulic licensing analysis of an FHTR using SINAP' s pebble core design as a reference case. The operation limits based on criteria outlined in U.S. regulatory guidelines are evaluated. Limiting safety system settings (LSSSs) considering uncertainties for forced convection and natural convection are obtained. The LSSS power and coolant outlet temperature, respectively, are 24.83 MW and 720 degrees C for forced convection and 1.19 MW and 720 degrees C for natural convection. The maximum temperature for the structural materials of 730 degrees C is the most limiting constraint of the FHTR design.

The Massachusetts Institute of Technology Research Reactor (MITR), with thermal power of 6 MW, is one of the five high-performance research reactors operated in the U.S. An in-house program MCODE-FM (MCNP-ORIGEN Coupled Depletion Program – Fuel Management) as the reference code supporting routine operation is currently developed by the MITR staff. Adopting Monte Carlo methods enables exact modeling of the MITR core geometry with use of continuous-energy nuclear data. A criticality search algorithm to track control blade movement is implemented in MCODE-FM. The code also features automation of input file generation, data manipulation, and post-processing of output data for fuel cycle analysis. Some verification and validation runs have been carried out by comparisons with the multiplication factor and the reactivity worth induced by control blades. In this study, further validation runs are performed, with two sets of measured data: 1) reactivity effects of fission product poisoning and fuel depletion and 2) reaction rate based thermal and fast neutron flux. Fission product poisoning effects calculated by MCODE-FM are in good agreement with the measured data obtained during the reactor start-up process (within 10%). The trend of peak xenon reactivity after the reactor shutdown is predicted accurately as well. As compared to the neutron activation experiment, MCODE-FM is found to slightly over-estimate the thermal flux with less than 10% discrepancy and the fast neutron flux with less than 20% discrepancy.

The effect of an external magnetic field on the convective heat transfer and pressure drop of magnetite nanofluids under laminar flow regime conditions (Re

Due to growing concerns over anthropogenic effects on the climate, there is increasing need to replace engineered fluids of high global warming potentials (GWPs), such as hydrofluorocarbons (HFCs) and perfluorocarbons (PFCs), with more environmentally friendly alternatives in thermal management systems. This article presents experimental data and compares various correlations for predicting the pool boiling heat transfer coefficient of a new low-GWP fluid, FK-649. Using a pressurized boiling facility with a smooth aluminum heater, the critical heat flux (CHF) and heat transfer coefficient were measured for the pool boiling of FK-649 at various saturation conditions. The commonly used refrigerant tetrafluoroethane (R-134a) is tested in the same pressurized facility to act as a benchmark for the new fluid. While R-134a exhibited a higher heat transfer coefficient and CHF, this behavior is expected from the fluid properties. Two-phase heat transfer performance of FK-649 is expected to be similar to that of t...

The Massachusetts Institute of Technology Research Reactor (MITR) is evaluating a transitional core conversion strategy for converting from high-enrichment uranium (HEU) to low-enrichment uranium (...

The fluoride-salt-cooled high-temperature reactor (FHR) is a new reactor concept, which combines low-pressure liquid salt coolant and high-temperature tristructural isotropic (TRISO) particle fuel. The refractory TRISO particle coating system and the dispersion in graphite matrix enhance safeguards (nuclear proliferation resistance) and security. Compared to the conventional high-temperature reactor (HTR) cooled by helium gas, the liquid salt system features significantly lower pressure, larger volumetric heat capacity, and higher thermal conductivity. The salt coolant enables coupling to a nuclear air-Brayton combined cycle (NACC) that provides base-load and peak-power capabilities. Added peak power is produced using jet fuel or locally produced hydrogen. The FHR is, therefore, considered as an ideal candidate for the transportable reactor concept to provide power to remote sites. In this context, a 20-MW (thermal power) compact core aiming at an 18-month once-through fuel cycle is currently under design at Massachusetts Institute of Technology (MIT). One of the key challenges of the core design is to minimize the reactivity swing induced by fuel depletion, since excessive reactivity will increase the complexity in control rod design and also result in criticality risk during the transportation process. In this study, burnable poison particles (BPPs) made of B4C with natural boron (i.e., 20% B10 content) are adopted as the key measure for fuel cycle optimization. It was found that the overall inventory and the individual size of BPPs are the two most important parameters that determine the evolution path of the multiplication factor over time. The packing fraction (PF) in the fuel compact and the height of active zone are of secondary importance. The neutronic effect of Li6 depletion was also quantified. The 18-month once-through fuel cycle is optimized, and the depletion reactivity swing is reduced to 1 beta. The reactivity control system, which consists of six control rods and 12 safety rods, has been implemented in the proposed FHR core configuration. It fully satisfies the design goal of limiting the maximum reactivity worth for single control rod ejection within 0.8 beta and ensuring shutdown margin with the most valuable safety rod fully withdrawn. The core power distribution including the control rod’s effect is also demonstrated in this paper.

The onset of nucleate boiling (ONB) serves as the thermal-hydraulic operating limit for many research and test reactors. However, boiling incipience under forced convection has not been well-characterized in narrow channel geometries or for oxidized surface conditions. This study presents experimental data for the ONB in vertical upflow of deionized (DI) water in a simulated materials test reactor (MTR) coolant channel. The channel gap thickness and aspect ratio were 1.96 mm and 29∶1, respectively. Boiling surface conditions were carefully controlled and characterized, with both heavily oxidized and native oxide surfaces tested. Measurements were performed for mass fluxes ranging from 750 to 3000 kg/m2 s and for subcoolings ranging from 10 to 45°C. ONB was identified using a combination of high-speed visual observation, surface temperature measurements, and channel pressure drop measurements. Surface temperature measurements were found to be most reliable in identifying the ONB. For the nominal (native oxide) surface, results indicate that the correlation of Bergles and Rohsenow, when paired with the appropriate single-phase heat transfer correlation, adequately predicts the ONB heat flux. Incipience on the oxidized surface occurred at a higher heat flux and superheat than on the plain surface.

Kriging, a geospatial interpolation technique, has been used in the present work to drive a search-and-optimization algorithm which produces the optimum geometric parameters for a 5 MW research reactor design. The technique has been demonstrated to produce an optimal neutronic solution after a relatively small number of core calculations. It has additionally been successful in producing a design which significantly improves thermal neutron fluxes by 30% over existing reactors of the same power rating. Best practices for use of this algorithm in reactor design were identified and indicated the importance of selecting proper correlation functions.