Reactor hot spot analysis

The principle methods for performing reactor hot spot analysis are reviewed and examined for potential use in the Applied Physics Division. The semistatistical horizontal method is recommended for future work and is now available as an option in the SE2-ANL core thermal hydraulic code. The semistatistical horizontal method is applied to a small LMR to illustrate the calculation of cladding midwall and fuel centerline hot spot temperatures. The example includes a listing of uncertainties, estimates for their magnitudes, computation of hot spot subfactor values and calculation of two sigma temperatures. A review of the uncertainties that affect liquid metal fast reactors is also presented. It was found that hot spot subfactor magnitudes are strongly dependent on the reactor design and therefore reactor specific details must be carefully studied. 13 refs., 1 fig., 5 tabs

Information technology and data mining for spent fuel treatment

Information technology is being used to provide interactive access to data collected from the electro-metallurgical treatment of spent fuel. The data are results from many hundreds of experiments performed to better characterize the processes by which uranium is separated from the waste products. Web-based display and relational database query capabilities facilitate the identification of trends in the data and the relating of these trends to the underlying electrochemistry. The objectives are to ensure that the process behavior is well understood, to make readily accessible the necessary data for development and validation of models, and to identify unexpected trends in the data as indications of phenomena not yet represented in the models

Dynamic Modeling Efforts for System Interface Studies for Nuclear Hydrogen Production.

System interface studies require not only identifying economically optimal equipment configurations, which involves studying mainly full power steady-state operation, but also assessing the operability of a design during load change and startup and assessing safety-related behavior during upset conditions. This latter task is performed with a dynamic simulation code. This report reviews the requirements of such a code. It considers the types of transients that will need to be simulated, the phenomena that will be present, the models best suited for representing the phenomena, and the type of numerical solution scheme for solving the models to obtain the dynamic response of the combined nuclear-hydrogen plant. Useful insight into plant transient behavior prior to running a dynamics code is obtained by some simple methods that take into account component time constants and energy capacitances. Methods for determining reactor stability, plant startup time, and temperature response during load change, and tripping of the reactor are described. Some preliminary results are presented

Initial Assessment of the Operability of the Vhtr-Htse Nuclear Hydrogen Plant.

The generation of hydrogen from nuclear power will need to compete on three fronts: production, operability, and safety to be viable in the energy marketplace of the future. This work addresses the operability of a coupled nuclear and hydrogen-generating plant while referring to other work for progress on production and safety. Operability is a measure of how well a plant can meet time-varying production demands while remaining within equipment limits. It can be characterized in terms of the physical processes that underlie operation of the plant. In this work these include the storage and transport of energy within components as represented by time constants and energy capacitances, the relationship of reactivity to temperature, and the coordination of heat generation and work production for a near-ideal gas working fluid. Criteria for assessing operability are developed and applied to the Very High Temperature Reactor coupled to the High Temperature Steam Electrolysis process, one of two DOE/INL reference plant concepts for hydrogen production. Results of preliminary plant control and stability studies are described. A combination of inventory control in the VHTR plant and flow control in the HTSE plant proved effective for maintaining hot-side temperatures near constant during quasi-static change in hydrogen production rate. Near constant electrolyzer outlet temperature is achieved by varying electrolyzer cell area to control cell joule heating. It was found that rates of temperature change in the HTSE plant for a step change in hydrogen production rate are largely determined by the thermal characteristics of the electrolyzer. It's comparatively large thermal mass and the presence of recuperative heat exchangers result in a tight thermal coupling of HTSE components to the electrolyzer. It was found that thermal transients arising in the chemical plant are strongly damped at the reactor resulting in a stable combined plant. The large Doppler reactivity component, three times greater than next reactivity component, per unit temperature, is mainly responsible. This is the case even when one of the conditions for out-of-phase oscillations between reactor inlet and outlet temperature, a large time for transport of process heat between the reactor and chemical plant, exists

Scalability of the natural convection shutdown heat removal test facility (NSTF) data to VHTR/NGNP RCCS designs.

Passive safety in the Very High Temperature Reactor (VHTR) is strongly dependent on the thermal performance of the Reactor Cavity Cooling System (RCCS). Scaled experiments performed in the Natural Shutdown Test Facility (NSTF) are to provide data for assessing and/or improving computer code models for RCCS phenomena. Design studies and safety analyses that are to support licensing of the VHTR will rely on these models to achieve a high degree of certainty in predicted design heat removal rate. To guide in the selection and development of an appropriate set of experiments a scaling analysis has been performed for the air-cooled RCCS option. The goals were to (1) determine the phenomena that dominate the behavior of the RCCS, (2) determine the general conditions that must be met so that these phenomena and their relative importance are preserved in the experiments, (3) identify constraints specific to the NSTF that potentially might prevent exact similitude, and (4) then to indicate how the experiments can be scaled to prevent distortions in the phenomena of interest. The phenomena identified as important to RCCS operation were also the subject of a recent PIRT study. That work and the present work collectively indicate that the main phenomena influencing RCCS heat removal capability are (1) radiation heat transport from the vessel to the air ducts, (2) the integral effects of momentum and heat transfer in the air duct, (3) buoyancy at the wall inside the air duct giving rise to mixed convection, and (4) multidimensional effects inside the air duct caused by non-uniform circumferential heat flux and non-circular geometry

Prioritization of VHTR system modeling needs based on phenomena identification, ranking and sensitivity studies.

Quantification of uncertainty is a key requirement for the design of a nuclear power plant and the assurance of its safety. Historically the procedure has been to perform the required uncertainty assessment through comparison of the analytical predictions with experimental simulations. The issue with this historical approach has always been that the simulations through experiments could not be at full scale for the practical reasons of cost and scheduling. Invariably, only parts of the system were tested separately or if integral testing was performed for the complete system, the size or scale of the experimental apparatus was significantly smaller than the actual plant configuration

Development of HyPEP, A Hydrogen Production Plant Efficiency Calculation Program

The Department of Energy envisions the next generation very high temperature gas-cooled reactor (VHTR) as a single-purpose or dual-purpose facility that produces hydrogen and electricity. The Ministry of Science and Technology (MOST) of the Republic of Korea also selected VHTR for the Nuclear Hydrogen Development and Demonstration (NHDD) Project. The report will address the evaluation of hydrogen and electricity production cycle efficiencies for such systems as the VHTR and NHDD, and the optimization of system configurations. Optimization of such complex systems as VHTR and NHDD will require a large number of calculations involving a large number of operating parameter variations and many different system configurations. The research will produce (a) the HyPEP which is specifically designed to be an easy-to-use and fast running tool for the hydrogen and electricity production evaluation with flexible system layout, (b) thermal hydraulic calculations using reference design, (c) verification and validation of numerical tools used in this study, (d) transient analyses during start-up operation and off-normal operation. This project will also produce preliminary cost estimates of the major components

Initial VHTR Accident Scenario Classification: Models and Data.

Nuclear systems codes are being prepared for use as computational tools for conducting performance/safety analyses of the Very High Temperature Reactor. The thermal-hydraulic codes are RELAP5/ATHENA for one-dimensional systems modeling and FLUENT and/or Star-CD for three-dimensional modeling. We describe a formal qualification framework, the development of Phenomena Identification and Ranking Tables (PIRTs), the initial filtering of the experiment databases, and a preliminary screening of these codes for use in the performance/safety analyses. In the second year of this project we focused on development of PIRTS. Two events that result in maximum fuel and vessel temperatures, the Pressurized Conduction Cooldown (PCC) event and the Depressurized Conduction Cooldown (DCC) event, were selected for PIRT generation. A third event that may result in significant thermal stresses, the Load Change event, is also selected for PIRT generation. Gas reactor design experience and engineering judgment were used to identify the important phenomena in the primary system for these events. Sensitivity calculations performed with the RELAP5 code were used as an aid to rank the phenomena in order of importance with respect to the approach of plant response to safety limits. The overall code qualification methodology was illustrated by focusing on the Reactor Cavity Cooling System (RCCS). The mixed convection mode of heat transfer and pressure drop is identified as an important phenomenon for Reactor Cavity Cooling System (RCCS) operation. Scaling studies showed that the mixed convection mode is likely to occur in the RCCS air duct during normal operation and during conduction cooldown events. The RELAP5/ATHENA code was found to not adequately treat the mixed convection regime. Readying the code will require adding models for the turbulent mixed convection regime while possibly performing new experiments for the laminar mixed convection regime. Candidate correlations for the turbulent mixed convection regime for circular channel geometry were identified in the literature. We describe the use of computational experiments to obtain correction factors for applying these circular channel results to the specialized channel geometry of the RCCS. The intent is to reduce the number of laboratory experiments required. The FLUENT and Star-CD codes contain models that in principle can handle mixed convection but no data were found to indicate that their empirical models for turbulence have been benchmarked for mixed convection conditions. Separate effects experiments were proposed for gathering the needed data. In future work we will use the PIRTs to guide review of other components and phenomena in a similar manner as was done for the mixed convection mode in the RCCS. This is consistent with the project objective of identifying weaknesses or gaps in the code models for representing thermal-hydraulic phenomena expected to occur in the VHTR both during normal operation and upsets, identifying the models that need to be developed, and identifying the experiments that must be performed to support model development

Gas-Cooled Fast Reactor (GFR) FY04 Annual Report

The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection

Incorporating background frequency improves entropy-based residue conservation measures

BACKGROUND: Several entropy-based methods have been developed for scoring sequence conservation in protein multiple sequence alignments. High scoring amino acid positions may correlate with structurally or functionally important residues. However, amino acid background frequencies are usually not taken into account in these entropy-based scoring schemes. RESULTS: We demonstrate that using a relative entropy measure that incorporates amino acid background frequency results in improved performance in identifying functional sites from protein multiple sequence alignments. CONCLUSION: Our results suggest that the application of appropriate background frequency information may lead to more biologically relevant results in many areas of bioinformatics