Note: New project descriptions are expected in January and added as they are identified.
Identify the project numbers of greatest interest in the “Career Goals” section of your online application.
Project Number: NE-1
NGNP tritium penetration study
As part of engineering workscope of the Next Generation of Nuclear Plant (NGNP), INL investigates tritium penetration from the High Temperature Gas-Cooled Reactor to nuclear hydrogen plants.
The purpose of this study is to analyze tritium transport from the nuclear reactor to hydrogen plants. In this study INL may develop the numerical tool or use a computer code developed by Japan Atomic Energy Agency depending on the bilateral agreement. In the tritium analysis codes, diffusion data and correlations are based on the simple experiments. Therefore, there are some difficulties to apply them directly to a compact heat exchanger (HX) such as printed circuit heat exchanger (PCHE) type HXs, which consist of semi-circular shaped holes. In this work, the diffusion rates of PCHE type HXs will be estimated by computational methods. Computational fluid dynamic (CFD) code or Comsol Multi-physics code will be utilized here. In addition to this, the effect of the HX's pitch arrangement on the tritium penetration rates will be estimated and finally, the best internal configurations of PCHE will be suggested.
This is a multi-year task. In FY-08, we will focus on a number of benchmarking calculations using Peach Bottom data and validation.
A student who has experience in heat transfer and CFD code is preferred and if not, INL will provide necessary training for this work.
Contact: Chang Oh, Send E-mail
Project Number: NE-2
Data Collection and Analysis on Plasma Diagnostic Devices of a Tokamak Magnetic Fusion Experiment
A physics or nuclear engineering student, senior or first year graduate student, is needed to perform data collection and analysis on operating experiences of plasma diagnostics of a US tokamak. The student will learn about various diagnostics, review failure reports, analyze the data to develop device failure rates and repair rates. The student will compare the data analysis results to any other published data on diagnostics. Necessary skills include basic understanding of diagnostic device functions, types, and construction (or willingness to learn such information) and basic undergraduate statistics. The student would prepare a presentation or poster on the summer’s activities and document the analysis in a report.
Contact: Lee Cadwallader, Send E-mail
Project Number: NE-4
ATR National Scientific User Facility
The Advanced Test Reactor (ATR) is one of the world's largest and most capable irradiation testing facilities. The reactor's high flux allows accelerated testing of reactor materials and components, so that a lifetime neutron exposure can be accumulated in a relatively short time. The ATR National Scientific User facility operates to make available these world class capabilities for use by nuclear researchers who agree to publish the results of experiments. The User Facility includes both state-of-the art irradiation testing and post-irradiation examination capabilities. Current major projects include development of an irradiation test vehicles for high temperature instrumented irradiation experiments, development of boiling water capsule experiment hardware, and methodology and design of systems for crack growth rate and fracture toughness testing of irradiated materials.
Contact: Mitchell K. Meyer, Send E-mail
Project Number: NE-5
GFR Vented Fuel Pin Sensitivity Study
Current research on the 2400 MW gas-cooled fast reactor (GFR) concept examines the feasibility of the vented fuel pin design (VFP). The VFP provides a flow circuit to vent fission product gases from the fuel pin, which decreases the cladding hoop stress and the requisite outlet plenum volume. An important concern for the VFP is its performance during off-normal transients. The objective for this intern position is to perform sensitivity analyses that predicts the VFP response during small and large break loss of coolant accidents. Analyses will be performed using CFD (STAR-CD or FLUENT) and the INL RELAP5 system code on a Linux workstation. The preferred candidate is a graduate level nuclear engineering student with a strong background in thermal hydraulics, working knowledge of CFD modeling (meshing and analysis), and basic knowledge of the RELAP5 system code.
Contact: Theron Marshall, Send E-mail
Project Number: NE-6
Advanced High-Temperature Oxygen Sensor Design
The large break air ingress transient for high temperature gas cooled reactors poses the challenging safety question of when natural circulation is initiated. Accurate modeling of the transient is dependent upon the oxygen transport rate, which is initially limited to diffusion. Existing oxygen technology does not provide instrumentation for high-temperature operations (T < 450 C) or space-related measurements. The objective for this intern position is to summarize the current state-of-the-art in high-temperature oxygen sensor technology and develop a conceptual design for a high-temperature and time-space cognizant oxygen sensor. The preferred candidate is a graduate level student with a strong background in instrumentation and controls.
Contact: Theron Marshall, Send E-mail
Project Number: NE-7
Data Collection and Analysis for Preventative/Predictive Maintenance on ATR Support Systems
A nuclear engineering student, junior or senior, is desired for performing data collection and analysis for Preventative/Predictive Maintenance on ATR Support systems. The student will collect and compare field data with industry standards for Preventative/Predictive Maintenance data. Results will be presented to engineers for final acceptance and incorporate into established programs. The student would prepare a presentation on the summer's activities and document work in a report.
Contact: Phil Erickson, Send E-mail
Project Number: NE-8
Verify and Validate Nuclear Physics Codes
Each cycle run in the Advanced Test Reactor (ATR) has a different core configuration which must be analyzed for safety. A large suite of codes and scripts are used to perform this safety analysis. These codes are currently run on an older UNIX workstation where they have been verified and validated (V&Vd). The essential codes have been transferred to a new UNIX workstation but the scripts need to be checked and modified to work as expected. Finally, the entire package will need to be verified and validated on the new platform. This work can be performed at an office location in Idaho Falls.
Contact: Paul Roth, Send E-mail
Project Number: NE-9
Nuclear S&T
Evaluate the reactivity worth (in terms of ((delta K)/K)) of the Advanced Test Reactor (ATR) flux traps and all other areas of interest in all stages of voiding. homogeneous densities from 1 to 0 incrementally, and voided by sub region with in the region of interest, (i.e. a half inch cylinder voided down the middle of flux trap probably has more reactivity worth a homogeneous density of .95, but they both have the same mass. Document results.
Contact: Frances Marshall, Send E-mail
Project Number: NE-10
Nuclear S&T
Determine grams of U-235 per 0.1 degree of OSCC rotation at various irradiation positions around the Advanced Test Reactor (ATR) core and in the driver fuel. Do this with all fresh fuel and with burned fuel (as realistically as possible). Document results.
Contact: Frances Marshall, Send E-mail
Project Number: NE-11
Nuclear S&T
Determine the reactivity worth of historic isotope targets loaded into the Advanced Test Reactor (ATR) a-positions, b-positions, h-positions, and Small Irradiation Housing Assembly positions in the South East and Center flux traps. Determine collapsed 1-group cross sections for the reactions of interest (it would also be desirable to evaluate these same isotopes for irradiation in the rabbit). Document results.
Contact: Frances Marshall, Send E-mail
Project Number: NE-12
Nuclear S&T
Evaluate a mixed fuel model for the Advanced Test Reactor (ATR); a combination of 19-plate discrete and 3-region lumped models. Determine the difference in its ability to predict fine detail in areas of interest (peak flux and fission density in driver fuel and other driver fuel limits) compared to the all discrete model. Document results.
Contact: Frances Marshall, Send E-mail
Project Number: NE-13
Experimental graphite characteristic study
As part of an international collaboration program, INL is going to perform experimental work in terms of measuring the internal pore surface area density of nuclear-grade graphite, which we found to be a very important parameter in the early stage of graphite oxidation, and the transient graphite oxidation with oxidation. The results of these measurements will be implemented into an upgraded GAMMA code INL has developed. The following activities will be carried out in this task:
- Measurement of surface area density of nuclear graphite using Brunaur-Emmett-Teller (BET) method
- Measurement of transient graphite oxidation
- Implementation of the advanced graphite material parameters into GAMMA code.
A student who has experimental experience is preferred and will be involved only in the experimental portion of work. INL will provide necessary training for this work.
Contact: Chang Oh, Send E-mail
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Idaho Falls, ID 83415-3790
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