Nuclear engineering program marks 10th anniversary

Ten years ago, Virginians who wanted to study nuclear engineering at the graduate level had to leave the state to do so. But then VCU, with support from Dominion Resources, started a program whose hallmark has been its ability to balance theory and application in its approach to nuclear engineering education

Neutronics, steady-state, and transient analyses for the Poland MARIA reactor for irradiation testing of LEU lead test fuel assemblies from CERCA : ANL independent verification results.

The MARIA reactor at the Institute of Atomic Energy (IAE) in Swierk (30 km SE of Warsaw) in the Republic of Poland is considering conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel assemblies (FA). The FA design in MARIA is rather unique; a suitable LEU FA has never been designed or tested. IAE has contracted with CERCA (the fuel supply portion of AREVA in France) to supply 2 lead test assemblies (LTA). The LTAs will be irradiated in MARIA to burnup level of at least 40% for both LTAs and to 60% for one LTA. IAE may decide to purchase additional LEU FAs for a full core conversion after the test irradiation. The Reactor Safety Committee within IAE and the National Atomic Energy Agency in Poland (PAA) must approve the LTA irradiation process. The approval will be based, in part, on IAE submitting revisions to portions of the Safety Analysis Report (SAR) which are affected by the insertion of the LTAs. (A similar process will be required for the full core conversion to LEU fuel.) The analysis required was established during working meetings between Argonne National Laboratory (ANL) and IAE staff during August 2006, subsequent email correspondence, and subsequent staff visits. The analysis needs to consider the current high-enriched uranium (HEU) core and 4 core configurations containing 1 and 2 LEU LTAs in various core positions. Calculations have been performed at ANL in support of the LTA irradiation. These calculations are summarized in this report and include criticality, burn-up, neutronics parameters, steady-state thermal hydraulics, and postulated transients. These calculations have been performed at the request of the IAE staff, who are performing similar calculations to be used in their SAR amendment submittal to the PAA. The ANL analysis has been performed independently from that being performed by IAE and should only be used as one step in the verification process

Transferring simulation skills from other industries to nuclear

Engineering analysis and simulation has always played a significant role in the nuclear sector and its use continues to increase across all branches of industry. To remain competitive in an increasingly global environment and to ensure the safety and reliability of designs, the nuclear industry must take advantage of the new engineering simulation technologies. Concerns surrounding the inappropriate use of simulation by staff without the appropriate competency persist, as analyses become more advanced, increasingly embracing more complex physical phenomena, often in an effort to model reality more faithfully. Furthermore, the age profile of the skilled staff in the nuclear sector in the UK is such that the skills shortage is likely to increase in future. These trends emphasize the need for life-long learning and continual staff development along with transfer of skills from other industry sectors to the nuclear sector. The nuclear industry has taken some initiatives to address skill shortages through the National Skills Academy for Nuclear and Nuclear Energy Skills Alliance (NESA) but these are mostly focused on manufacturing and R&D skills. The recently completed EU funded EASIT2 project is directly aimed at addressing the engineering analysis and simulation skills. This paper gives a brief overview of the EASIT2 project and its deliverables and points out how it can help the skills issues being faced by the nuclear industry. INTRODUCTIO

Comparison of the PLTEMP code flow instability predictions with measurements made with electrically heated channels for the advanced test reactor.

When the University of Missouri Research Reactor (MURR) was designed in the 1960s the potential for fuel element burnout by a phenomenon referred to at that time as 'autocatalytic vapor binding' was of serious concern. This type of burnout was observed to occur at power levels considerably lower than those that were known to cause critical heat flux. The conversion of the MURR from HEU fuel to LEU fuel will probably require significant design changes, such as changes in coolant channel thicknesses, that could affect the thermal-hydraulic behavior of the reactor core. Therefore, the redesign of the MURR to accommodate an LEU core must address the same issues of fuel element burnout that were of concern in the 1960s. The Advanced Test Reactor (ATR) was designed at about the same time as the MURR and had similar concerns with regard to fuel element burnout. These concerns were addressed in the ATR by two groups of thermal-hydraulic tests that employed electrically heated simulated fuel channels. The Croft (1964), Reference 1, tests were performed at ANL. The Waters (1966), Reference 2, tests were performed at Hanford Laboratories in Richland Washington. Since fuel element surface temperatures rise rapidly as burnout conditions are approached, channel surface temperatures were carefully monitored in these experiments. For self-protection, the experimental facilities were designed to cut off the electric power when rapidly increasing surface temperatures were detected. In both the ATR reactor and in the tests with electrically heated channels, the heated length of the fuel plate was 48 inches, which is about twice that of the MURR. Whittle and Forgan (1967) independently conducted tests with electrically heated rectangular channels that were similar to the tests by Croft and by Walters. In the Whittle and Forgan tests the heated length of the channel varied among the tests and was between 16 and 24 inches. Both Waters and Whittle and Forgan show that the cause of the fuel element burnout is due to a form of flow instability. Whittle and Forgan provide a formula that predicts when this flow instability will occur. This formula is included in the PLTEMP/ANL code.Error! Reference source not found. Olson has shown that the PLTEMP/ANL code accurately predicts the powers at which flow instability occurs in the Whittle and Forgan experiments. He also considered the electrically heated tests performed in the ANS Thermal-Hydraulic Test Loop at ORNL and report by M. Siman-Tov et al. The purpose of this memorandum is to demonstrate that the PLTEMP/ANL code accurately predicts the Croft and the Waters tests. This demonstration should provide sufficient confidence that the PLTEMP/ANL code can adequately predict the onset of flow instability for the converted MURR. The MURR core uses light water as a coolant, has a 24-inch active fuel length, downward flow in the core, and an average core velocity of about 7 m/s. The inlet temperature is about 50 C and the peak outlet is about 20 C higher than the inlet for reactor operation at 10 MW. The core pressures range from about 4 to about 5 bar. The peak heat flux is about 110 W/cm{sup 2}. Section 2 describes the mechanism that causes flow instability. Section 3 describes the Whittle and Forgan formula for flow instability. Section 4 briefly describes both the Croft and the Waters experiments. Section 5 describes the PLTEMP/ANL models. Section 6 compares the PLTEMP/ANL predictions based on the Whittle and Forgan formula with the Croft measurements. Section 7 does the same for the Waters measurements. Section 8 provides the range of parameters for the Whittle and Forgan tests. Section 9 discusses the results and provides conclusions. In conclusion, although there is no single test that by itself closely matches the limiting conditions in the MURR, the preponderance of measured data and the ability of the Whittle and Forgan correlation, as implemented in PLTEMP/ANL, to predict the onset of flow instability for these tests leads one to the conclusion that the same method should be able to predict the onset of flow instability in the MURR reasonably well

Applications of neutron activation spectroscopy

Since the discovery in 1932, neutrons became a basis of many methods used not only in research, but also in industry and engineering. Among others, the exceptional role in the modern nuclear engineering is played by the neutron activation spectroscopy, based on the interaction of neutron flux with atomic nuclei. In this article we shortly describe application of this method in medicine and detection of hazardous substances.Comment: Presented at Symposium on applied nuclear physics and innovative technologies, Cracow, 03-06 June 201

Systems Engineering of a Nuclear-Electric Spacecraft

Studies have shown that nuclear-electric propulsion systems will provide superior payload capability and unique advantages over chemical systems for high-energy deep-space missions. Conceptual design studies of unmanned spacecraft employing nuclear-electric propulsion systems have been undertaken to determine some of the major integration problems. Early recognition of these problems will help to stimulate the development effort that will be required to bring these systems into fruitful utilization. Typical designs under consideration for interplanetary missions for the next decade employ a nuclear reactor providing thermal energy to a turbogeneration system which, in turn, supplies electrical power to an ion engine for primary propulsion and additional utility power for guidance and control, powered-flight radio transmission, instrumentation, et cetera. The major systems and components which form a complete spacecraft are listed in this Report, and a review of the significant physical and operational characteristics of these various systems and components which affect spacecraft integration is made. Conceptual.configurations and detailed weight studies for a 60-kilowatts-electric Venus-capture spacecraft and a 1-megawattelectric Jupiter-capture spacecraft are shown to illustrate typical physical arrangements based on the various hardware constraints. From these configurations, the major development goals are ascertained and summarized