General Atomics Research Articles

ANALYSIS ON THE PERFORMANCE OF THE BANDUNG CONVERSION FUEL-PLATE TRIGA REACTOR IN STEADY STATE WITH CONSTANT COOLANT FLOW RATE

Bandung TRIGA2000 Reactor, a General Atomic (GA)-made research reactor used for training, research andiIsotope production, has been upgraded to operate at power of 2000 kW using TRIGA fuel rod type. Recently, the TRIGA reactor fuel element producers are going to discontinue the production of TRIGA fuel element. To overcome the unavailability of TRIGA fuel element, BATAN planned to modify TRIGA2000 fuel type from rod-type to U3Si2-Al plate-type fuel with 19.75% enrichment, similar to the domestically fabricated one used in RSG-GAS. The carried out design emphasized on the determination of operation condition limits for setting the reactor protection system in accordance to the reactor safety calculation results. The conceptual design of the innovative fuel plate TRIGA reactor cooling system is expected to remove heat generated by fuels with nominal power of 1 MW up to 2 MW. The design is developed through modelling and safety analysis using COOLOD-N2 validated code. The safety margin is set to its flow instability at transient condition of the fuel plate, which is ≥ 2.38; departure from nucleate boiling ratio ≥1.50; and no onset of nucleate boiling, ΔTONB ≥ 0oC. The primary coolant flow rate accommodating the existing Bandung TRIGA reactor capability is as high as 50 kg/s. The analysis results show that at power of 1 MW, the reactor can safely operate, while at power of 2 MW the safety margin is exceeded. In other words, the plate TRIGA reactor that employs forced convection mode operates safely at 1 MW with excess power 120% of its nominal power.Keywords: 1 MW, Thermalhydraulic design, Steady state condition, TRIGA plate, Constant flowrate

Accelerated Tests on Si and SiC Power Transistors with Thermal, Fastand Ultra-Fast Neutrons.

Neutron test campaigns on silicon (Si) and silicon carbide (SiC) power MOSFETs and IGBTs were conducted at the TRIGA (Training, Research, Isotopes, General Atomics) Mark II (Pavia, Italy) nuclear reactor and ChipIr-ISIS Neutron and Muon Source (Didcot, U.K.) facility. About 2000 power transistors made by STMicroelectronics were tested in all the experiments. Tests with thermal and fast neutrons (up to about 10 MeV) at the TRIGA Mark II reactor showed that single-event burnout (SEB) failures only occurred at voltages close to the rated drain-source voltage. Thermal neutrons did not induce SEB, nor degradation in the electrical parameters of the devices. SEB failures during testing at ChipIr with ultra-fast neutrons (1-800 MeV) were evaluated in terms of failure in time (FIT) versus derating voltage curves according to the JEP151 procedure of the Joint Electron Device Engineering Council (JEDEC). These curves, even if scaled with die size and avalanche voltage, were strongly linked to the technological processes of the devices, although a common trend was observed that highlighted commonalities among the failures of different types of MOSFETs. In both experiments, we observed only SEB failures without single-event gate rupture (SEGR) during the tests. None of the power devices that survived the neutron tests were degraded in their electrical performances. A study of the worst-case bias condition (gate and/or drain) during irradiation was performed.

Determination of βeff for ININ's TRIGA Reactor

Abstract The value of βeff for Training Research Isotopes of General Atomics (TRIGA) Mark III reactor, belonging to the National Institute of Nuclear Researches (ININ), is reported. The TRIGA Mark III reactor core was simulated with MCNP6 to deduce the effective multiplication factor (keff) for critical state and after a small insertion of positive reactivity (∼0.20 $). To perform more realistic simulations, we had to incorporate in the composition of the low-enrichment uranium (LEU) fuel element the produced poisons in a time period of six years, considering the operation time in days, during which the reactor was operating at maximum power (1 MWth). The calculation of the βeff value was obtained with the keff results, calculated with the code, and the reactor periods measured experimentally. We also obtained directly the βeff value, through the card of MCNP6 to calculate keff (KOPTS) card of the MCNP6 code in order to compare both values.

Evaluation of Radiation Effects on Residents Living Around the NSRR Under External Hazards

Abstract The nuclear safety research reactor (NSRR) is a research reactor of training research isotopes general atomics—annular core pulse reactor (TRIGA-ACPR) type, located in the Nuclear Science Research Institute (NSRI). The NSRR facility has been utilized for fuel irradiation experiments to study the behaviors of nuclear fuels under reactivity-initiated accident (RIA) conditions. Under the new regulation standards, which was established after the Fukushima Daiichi accident, research reactors are regulated based on the risk of the facilities. The graded approach is introduced in the regulation. To apply the graded approach, the radiation effects on residents living around the NSRR under the external hazards were evaluated, and the level of the risk of the NSRR facility was investigated. This paper summarizes the result of the evaluation in the case where the safety functions are lost due to a tornado, an earthquake followed by a tsunami. There is fuel in the reactor core, fresh fuel storage, and spent fuel storage. As the effects from reactor core, we evaluate the external exposure to radiation and exposure from the release of fission products assuming that loss of function to shut down the reactor, break of cladding tubes, loss of reactor pool water, and collapse of the reactor building. As the effects from fresh fuel storage, we evaluate the internal exposure assuming that the fresh fuel particles released into the air because of breaking into pieces. In addition, we evaluate the critical safety assuming that the critical safety shapes of the fresh fuel storage are lost. As the effects from spent fuel storage, we evaluate the critical safety assuming that the critical safety shapes of the spent fuel storage are lost. All in all, the risk is confirmed to be relatively low, since the effective dose on the residents is found to be below 5 mSv per event due to the loss of the safety functions caused by the tornado, earthquake, and the accompanying tsunami.

Demountable Coaxial Clamped Joint For ITER Central Solenoid Module Final Test Program

The ITER Central Solenoid Modules (CSM) are being fabricated at General Atomics (GA) at their Magnet Technologies Center in Poway, CA. Each of the seven modules will undergo final testing at the GA facility to demonstrate their performance at 48.5 kA, 11 T, and at 4.5 K. In order to perform tests on multiple modules, a demountable coaxial joint using indium wires was developed to connect the modules to the feeders. Two full scale joints were fabricated to establish the assembly technique and preliminarily tested (up to 3 kA) on the CSM qualification coil (Cu conductor). The same joint components were then tested using a superconducting jumper (NbTi) and their resistance measured (2.3 and 4.9 nΩ, at 4.5 K, 40 kA). In parallel to measuring the voltage across the joints, a calorimetric evaluation was also performed and it confirmed the low resistance measurements of both joints. Testing of the same coaxial clamped joints on a heat treated Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn jumper took place in February 2019 and the results are reported in this paper. With these acceptable results, the same joint components and assembly techniques will be utilized to connect the 14 joints on the seven CSM (Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn) for full current testing.

New Fission Product Yield Measurements at Oregon State University

Cumulative fast fission product yields for 235 U and 238 U were measured at Godiva-IV in burst mode. Data was collected starting 45 minutes after the prompt irradiation. Data analysis codes were developed to analyze gammarays and calculate fission product yields with their respective uncertainties. Due to uncertainties in the branching ratios, different γ-rays from the same isotope led to different fission yield. This led to an effort at Oregon State University (OSU) to utilize the Rabbit facility at their 1 MegaWatt Training, Research, Isotopes, General Atomics (TRIGA) nuclear research reactor. The new experimental setup will allow us to measure branching ratio data and short-lived fission product yields starting seconds after a prompt irradiation.

My research experience at General Atomics in the U.S.

My research experience at General Atomics in the U.S.

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

This paper performs a neutronics evaluation of the General Atomics UZrHx low enriched uranium fuel – TRIGA fuel – in the National Bureau of Standards Reactor (NBSR) at the National Institute of Standards and Technology (NIST). The objective of this study is to examine the accountability and sustainability of the TRIGA fuel on neutronics aspects when applying it to the NBSR conversion. A feasibility scoping study was previously undertaken with considerations on various fuel dimensions, fuel rod layout configurations, and structure material selections, identifying the best option of deploying the TRIGA fuel to NBSR. Continuing with these efforts, an equilibrium NBSR core using the identified fuel was generated, and a well-round physics assessment was carried out by examining key neutronics performance characteristics of the core. All calculations were completed with MCNP-6, a 3-D Monte Carlo neutron transport code. The same fuel management scheme and fuel cycle length as the existing NBSR was adopted in the equilibrium core generation adopts to retain performance consistencies. The effectiveness of the fuel was examined at four representative burnup states of the fuel cycle. Neutronics performances of the equilibrium core was characterized by the fast and thermal neutron flux level as well as power distribution in the core. Reactor safety related parameters such as kinetics parameters and power peaking factors were also evaluated in the study. All results were compared against the current NBSR fueled with HEU for justifications. The findings in this research prove the viability of the TRIGA fuel for the NBSR conversion, and provide supporting data for future investigations on this subject.

Modeling of fuel-cladding stresses in porous UC/SIC fuel pins

Maximum stress development in nuclear fuel pins during their exploitation in the gas-cooled Gen-IV reactors is an important parameter for analysis of fuel-cladding compatibility, especially in the case of testing of new fuel or cladding materials. This article is devoted to the stress evaluation in the combination of uranium carbide fuel and silicon carbide cladding. Swelling fuel exerts the largest pressure on the cladding when initial gap between fuel and cladding has been already closed. Numerical modeling predicts rapid stress increase in this regime. To alleviate these stresses annular-shaped kernel-based bi-modal porous pellets have been designed in the General Atomics Co. with small pores in the spherical agglomerates (kernels) and comparatively large pores between kernels. A new numerical approach is developed to take into account specific features of the mechanical behavior of this new fuel. The modeling predicts stress decrease in direct proportion to porosity of the fuel, but the most significant stress reduction is expected with the upper bound increase of reactor power accompanied by the temperature increase in the pellet. Because of the temperature non-uniformity, stress relaxation in the pellet is sensitive not only to the mean value of porosity but also to its distribution along the pellet radius. Comparatively cold regions near the cladding must have higher porosity than high-temperature core regions for efficient stress relaxation.

Development and Validation of a Serpent-2 Model for the Former 3 MW TRIGA Core Configuration of the Philippine Research Reactor-1

The 3 MW TRIGA (Training, Research, Isotopes, General Atomics) Philippine Research Reactor-1 (PRR-1) at the DOST-PNRI achieved its first criticality on 08 Mar 1988 after its successful upgrade from a plate-type reactor. However, due to unresolved technical problems discovered weeks after the upgrade, the PRR-1 was considered inoperable and has been in shutdown status since then. The slightly irradiated TRIGA fuel rods of the PRR-1 are currently in an interim storage tank and are planned to be utilized in a subcritical reactor assembly. As part of the project to reuse the fuels, simulation models for both present and proposed configurations are important. In this work, we present the complete model of the former configuration of PRR-1 with 115 TRIGA fuel rods developed with the Serpent Monte Carlo code version 2 for simulation of criticality and neutronic analysis. The model of the TRIGA fuel rods was validated in the fresh fuel configuration through the benchmark analysis described in the 1988 reactor criticality report. The effective multiplication factors from the Serpent-2 simulation (𝒌𝒆𝒇𝒇 = 1.0690 ± 0.0012) and measured value of 1.0661 have been found to agree with a deviation of 259 pcm. Neutron flux and fission power distribution simulations using the same reactor configuration were also presented to serve as reference for future burn up calculations and fuel characterization.

Ionization Chambers to Determine Neutron and Gamma-Ray Kerma in a Research Reactor

Ionization chambers were designed and constructed to determine the kerma rates in various materials within several centimeters of a Training, Research, Isotopes, General Atomics (TRIGA) reactor core operating at 1 MW. The primary aim of this article was to compare kerma measurements with the advanced Monte Carlo code calculations of nuclear heating. Wall thickness, collection gap, and fill gas pressure were chosen to satisfy Bragg–Gray criteria, so that the measured ionization current was related to the kerma rate in the wall material. Chamber wall materials composed of low mass number elements, including hydrogen-rich C552 air-equivalent plastic and beryllium, were selected to measure the kerma due to fast neutron elastic scattering. By operating these neutron sensitive chambers coincidentally with relatively neutron insensitive chambers composed of aluminum and Zircaloy-4, we were able to measure the total heating due to fast neutrons and gamma rays in a material and to differentiate these heating components. A chamber composed of borated stainless steel was used in a similar fashion to measure the thermal neutron flux. The total kerma rate was also measured in various materials typically found in a reactor core. Chamber collection volumes were initially determined using ambient air fill gas and National Institute of Standards and Technology (NIST)-traceable air-kerma rates from a 60Co source. All chambers were sealed with argon gas to provide thermal and compositional stability. Chamber properties, including stability, saturation, and gas-phase mass subject to charge collection, were determined using the 60Co source. Chambers were operated for approximately 30 min adjacent to the reactor core, and the integrity of gas seals was subsequently verified by repeating the measurement with the 60Co source.

Benchmarking COMSOL Multiphysics Single-Subchannel Thermal-Hydraulic Analysis of a TRIGA Reactor with RELAP5 Results and Experimental Data

COMSOL Multiphysics has been used to conduct thermal-hydraulic analysis in multiple nuclear applications. The aim of this study is to benchmark the prediction accuracy of COMSOL Multiphysics in performing thermal-hydraulic analysis of TRIGA (Training, Research, Isotopes, General Atomics) reactors such as the Geological Survey TRIGA Reactor (GSTR) by comparing its predictions with RELAP5 (a widely used code in nuclear thermal-hydraulic analysis) results and experimental data. The GSTR type is Mark I with a full thermal power of 1 MW, and it resides at the Denver Federal Center (DFC) in Colorado. The numerical investigation of the present work is carried out by developing single-subchannel thermal-hydraulic models of the GSTR utilizing RELAP5 and COMSOL codes. The models estimate the temperatures (fuel, outer clad, and coolant) and water flow patterns in the core as well as fuel element powers at which void starts to form within the coolant subchannels. Then, these models’ predictions are quantitatively evaluated and compared with the measured data.

Multiple Analytical Approach to Isotopic Transport Analysis in Magnetic Fusion Devices

High-Z impurities released from plasma-material interactions have been shown to limit the performance of fusion plasmas, and understanding these impurity transport mechanisms throughout the plasma scrape-off layer is a major challenge. Presented herein is a study of tungsten (W) erosion and transport by uniquely measuring absolute quantities of isotopic W in order to determine the source of natural and enriched 182W isotopes that have traveled throughout the tokamak discharges on the DIII-D National Fusion Facility at General Atomics. Two primary analysis methods have been implemented to characterize this W on graphite collector probes that were inserted into DIII-D’s outboard midplane. Results from experiments using Rutherford backscattering spectrometry (RBS) have measured W particle areal densities down the centerline of the probes as high as 6E14 atoms/cm2 with a detection limit of 1E12 atoms/cm2. Laser ablation inductively coupled plasma mass spectrometry (LAMS) has confirmed the elemental trends found with RBS and has provided additional insight into collector probe surface profiles. Two-dimensional elemental and isotopic maps from LAMS are used to reveal new collector probe features and further refine the source of collected W. Variations in isotopic profiles and total W content are coupled to (a) the face of the probe being analyzed, (b) the dimensions of the probe, and (c) the plasma pulse parameters that were used during probe exposure. These results provide one-of-a-kind empirical evidence that is now being utilized for validation of tokamak impurity transport through theoretical models and in codes such as 3D-LIM and OEDGE.

C-ring testing of nuclear grade silicon carbide composites at temperatures up to 1900 °C

Silicon carbide matrix reinforced with silicon carbide fiber (SiC-SiC) composite is an advantageous material for many high temperature applications due to retention of room temperature strength above 1100 °C, which is the current operational limit for many high temperature metals. For the nuclear industry, this property, along with excellent radiation tolerance, has made SiC-SiC an attractive cladding material for the accident tolerant fuel program. The mechanical strength at high temperatures of these composites needs to be evaluated to address improvements in the third generation of Nicalon fibers, fabrication processes, as well as tested in a form-factor relevant to fuel cladding. In this work, mechanical strength of SiGA™ composite tubes, a SiC-SiC composite fabricated at General Atomics using the current generation of Hi-Nicalon Type S SiC fiber and chemical vapor infiltration processing (CVI) of the matrix, was evaluated at temperatures up to 1100 °C in ambient air and ∼1900 °C in an inert gas environment with nominal 30–60 min dwell time. The results of this study show full retention of room temperature strength at both test regimes. Some degradation of material toughness was observed at temperatures above ∼1760 °C. During this study, effects of interphase, fiber, and matrix morphological changes as a function of temperature were investigated.

Fully ceramic microencapsulated fuel in prismatic high-temperature gas-cooled reactors: Design basis accidents and fuel cycle cost

Fully ceramic microencapsulated (FCM) fuel may enhance the already strong inherent reactor safety characteristics of a conventional High-Temperature Gas-Cooled Reactor (HTGR). FCM fuel uses an SiC matrix that exhibits higher stability under irradiation with limited swelling relative to a conventional graphite matrix. Additionally, the SiC matrix is expected to exhibit improved mechanical performance relative to graphite. The feasibility of FCM-fueled HTGRs has been explored in a previous study (Lu et al., 2018), and three FCM fuel concepts were identified to be able to maintain the cycle length of the reference conventional General Atomics 350 MWt prismatic modular HTGR.In this work, the previous investigation of the reactor core safety characteristics of HTGRs fueled by these three FCM fuels was significantly extended to detailed thermal hydraulics and neutronics analysis of normal operation and design basis accident scenarios. We found that these FCM-fueled cores would have a maximum fuel temperature ∼20–40 K higher than that of the reference HTGR core at the beginning of cycle (BOC) and ∼30–60 K higher at the end of cycle (EOC) during normal operating conditions. Pressurized (P-) and depressurized (D-) loss of forced cooling (LOFC) accidents as well as the control rod withdrawal accident were explored. The FCM-fueled cores would at most have a maximum fuel temperature ∼30–50 K higher than that of the reference core in a P-LOFC accident, ∼10–30 K in a D-LOFC accident and ∼30–50 K in a rod withdrawal accident. The conclusion of the study is that FCM fuel has a small impact on the reactor performance and safety characteristics during normal operation and design basis accident conditions within the existing and generic 350 MWt prismatic modular HTGR design we considered.Additionally, a fuel cycle cost comparison was also performed in the current study. We found that the FCM fuels may be up to 40%–74% more expensive than the conventional fuel. Further analysis of the economics of FCM fuel are needed, because the uncertainties in the assumptions for this study are high.

Online Heat Load Test and Analysis for the Whole 5-K GA Feeder System

Heat conduction load test is a main and tough issue in the complex and huge general atom (GA) feeder system cold test. The GA feeder system is a superconducting system connects from the room temperature power supply system to the 4.5-K busbar of the magnet, supply 52 kA current and cryogenic provisions to the central solenoid magnet (CSM) coil, used for International Thermonuclear Experimental Reactor CSM magnet test at General Atomics Company, San Diego, CA, USA. This system have been developed and constructed at the Institute of Plasma Physics, Chinese Academy of Sciences, and completed the full cold test before transferred to the USA. In all the cold test items, the requirement of overall heat load 68 W is a big challenge, for its many leakage heat sources and cooling circuits: high temperature superconducting current lead, vacuum barrier, busbars, pipes, supports, etc. In order to get the steady results, many efforts were took to achieve the required value, such as reduce the inlet pressure of the 5-K loop as much as possible, minimize the flow rate of different 5-K cooling circuits, and at the same time, a dedicated program was developed to online calculate the density and the heat load while monitoring and analyzing the heat load. This paper will give a detail description of the 5-K overall heat load measurement.

Mechanical design of the high-powered helicon antenna and stripline feed in the DIII-D Tokamak

A high-powered “comb-line” helicon antenna for use within the DIII-D Tokamak is currently in design and fabrication at General Atomics. The antenna will drive current in high beta discharges using electromagnetic helicon waves. The high-powered helicon antenna (HPHA) is expected to couple up to one MW of power into DIII-D plasmas at a frequency of 476 MHz. The antenna design includes 30 individual, inductively coupled modules fastened to six internally water-cooled Inconel back-plates. The end modules have special design features to provide RF stripline feed attachment points. Each back-plate is mounted to a pedestal, which is secured to the vessel wall via a combination of studs and welded brackets. The mechanical design includes features to minimize and survive disruption forces, thermal loading from RF losses, vacuum vessel bake (350C), and installation considerations. Thermal stress design challenges include plasma heat loads, thermal ratcheting, and bake-out cycles that result in thermal growth differences between the stripline and the vessel. Splitting the back-plate into six sections each with a central pedestal support significantly reduces the disruption loads. A single stripline folded in half, fed near the fold provides a 180° phase shift between the stripline module connections for driving the two module straps. This looped mono RF coaxial stripline design was chosen to enable using one feed-through, to optimize space limitations, ease of installation, structural rigidity, and RF tuning ability. The striplines are supported by a quarter-wavelength “stub”, mid span, for support during thermal cycling and plasma disruption events. Multiphysics FEA analyses are performed to optimize the geometric shapes to meet the aforementioned design challenges. A description of the mechanical design, fabrication, installation, and analyses of the HPHA and stripline feeds will be presented

Remaining useful life estimation of critical DIII-D subsystems

DIII-D plays a vital role in the development of the physics basis for fusion energy and the ITER design. Designed and built in the early 1980’s, the system started operations in 1986 and has provided a reliable platform for fusion experiments for over 30 years. A hallmark of DIII-D operations has been its ability to adapt to the changing needs of the fusion research community and support a wide range of tokamak experiments for various stakeholders. An important factor in ensuring continued reliable operations is gaining a thorough understanding of the remaining useful life (RUL) of critical DIII-D systems by means of quantitative assessments of RUL. General Atomics has conducted a pilot RUL assessment project to develop an RUL estimation methodology applicable to various DIII-D systems and demonstrate its feasibility on a critical subsystem (i.e. the F8 field shaping coil). The pilot F8-coil RUL assessment project showed that under typical operating conditions, the F8-coil may be expected to achieve an overall life of >50 years and/or >106 shots and demonstrated that employing reliability engineering techniques coupled with modern analytical tools, quantitative estimates for RUL can be developed with reasonable investment.

From manufacture to assembly of the ITER central solenoid

The Central Solenoid (CS), a key component of the ITER Magnet system, using a 45 kA Nb3Sn conductor, includes six identical coils, called modules, to form a solenoid, enclosed inside a structure providing vertical pre-compression and mechanical support. Procurement of the components of the ITER CS is the responsibility of US ITER, the US Domestic Agency (USDA), while the assembly of these components will be carried out by the ITER Organization (IO). Procurement of all the coil modules was awarded in 2011 to General Atomics, while procurement of the structure is split among several manufacturers, using existing equipment, sometimes among the largest ones in the world. Assembly of the ITER CS will require a dedicated area in the ITER Assembly Hall, conventional tooling and special tooling. US ITER is in charge of the procurement of special tooling, while IO is responsible for the procurement of the conventional ones. A detailed assembly procedure is under development at US ITER, in close collaboration with IO and with the support of CEA. Procurement of the special Assembly Tooling is carried out by US ITER and the main part of the first item, the Assembly Platform, was delivered to IO in 2017.

Design and Performance of Microwave Components for ECH and ECE Applications at General Atomics

This paper provides an overview of high power components for the application of Electron Cyclotron Heating transmission lines, and broadband devices for Electron Cyclotron Emission detection systems. The unique fabrication and assembly challenges are discussed, particularly in the context of ITER. The ITER ECH system will require robust, vacuum-compatible components such as polarizers, dummy loads, and switches that are sufficiently cooled to withstand 1 MW for 3,600 seconds. These elements, along with overmoded corrugated waveguide, are necessary to form transmission lines with efficiencies of 90%, and 90% transmitted HE11mode purity. Recent high power test results are summarized and scaled from the 63.5 mm internal diameter design to the 50 mm diameter version that will be used for ITER. Elements designed for Electron Cyclotron Emission detection and reflectometry systems are discussed, such as frequency filters and polarization rotators. The large frequency operating range of corrugated waveguide is exploited for such applications. The application of additive manufacturing technology towards both low and high power components is considered as a promising new area of development.