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The main objective of this project was to identify and characterize the conditions under which abnormal heat transfer phenomena would occur in a Very High Temperature Reactor (VHTR) with a prismatic core. High pressure/high temperature experiments have been conducted to obtain data that could be used for validation of VHTR design and safety analysis codes. The focus of these experiments was on the generation of benchmark data for design and off-design heat transfer for forced, mixed and natural circulation in a VHTR core. In particular, a flow laminarization phenomenon was intensely investigated since it could give rise to hot spots in the VHTR core.

Core bypass flow has been one of key issues in the very high temperature reactor (VHTR) design for securing core thermal margins and achieving target temperatures at the core exit. The bypass flow in a prismatic VHTR core occurs through the control element holes and the radial and axial gaps between the graphite blocks for manufacturing and refueling tolerances. These gaps vary with the core life cycles because of the irradiation swelling/shrinkage characteristic of the graphite blocks such as fuel and reflector blocks, which are main components of a core's structure. Thus, the core bypass flow occurs in a complicated multidimensional way. The accurate prediction of this bypass flow and counter-measures to minimize it are thus of major importance in assuring core thermal margins and securing higher core efficiency. Even with this importance, there has not been much effort in quantifying and accurately modeling the effect of the core bypass flow. The main objectives of this project were to generate experimental data for validating the software to be used to calculate the bypass flow in a prismatic VHTR core, validate thermofluid analysis tools and their model improvements, and identify and assess measures for reducing the bypass flow. To achieve thesemore » objectives, tasks were defined to (1) design and construct experiments to generate validation data for software analysis tools, (2) determine the experimental conditions and define the measurement requirements and techniques, (3) generate and analyze the experimental data, (4) validate and improve the thermofluid analysis tools, and (5) identify measures to control the bypass flow and assess its performance in the experiment.« less

The purpose of the fluid dynamics experiments in the MIR (Matched Index of-Refraction) flow system at Idaho National Laboratory (INL) is to develop benchmark databases for the assessment of Computational Fluid Dynamics (CFD) solutions of the momentum equations, scalar mixing, and turbulence models for the flow ratios between coolant channels and bypass gaps in the interstitial regions of typical prismatic standard fuel element (SFE) or upper reflector block geometries of typical Modular High-temperature Gas-cooled Reactors (MHTGR) in the limiting case of negligible buoyancy and constant fluid properties. The experiments use Particle Image Velocimetry (PIV) to measure the velocity fields that will populate the bypass flow study database.

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope of the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended tomore » be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.« less

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope of the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended tomore » be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.« less

The U.S. Department of Energy is exploring the potential for the VHTR which will be either of a prismatic or a pebble-bed type. One important design consideration for the reactor core of a prismatic VHTR is coolant bypass flow which occurs in the interstitial regions between fuel blocks. Such gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The existence of the gaps induces a flow bias in the fuel blocks and results in unexpected increase of maximum fuel temperature. Traditionally, simplified methods such as flow network calculations employing experimental correlations are used to estimate flow and temperature distributions in the core design. However, the distribution of temperature in the fuel pins and graphite blocks as well as coolant outlet temperatures are strongly coupled with the local heat generation rate within fuel blocks which is not uniformly distributed in the core. Hence, it is crucial to establish mechanistic based methods which can be applied to the reactor core thermal hydraulic design and safety analysis. Computationalmore » Fluid Dynamics (CFD) codes, which have a capability of local physics based simulation, are widely used in various industrial fields. This study investigates core bypass flow phenomena with the assistance of commercial CFD codes and establishes a baseline for evaluation methods. A one-twelfth sector of the hexagonal block surface is modeled and extruded down to whole core length of 10.704m. The computational domain is divided vertically with an upper reflector, a fuel section and a lower reflector. Each side of the sector grid can be set as a symmetry boundary« less

The United States Department of Energy is promoting the resurgence of nuclear power in the U. S. for both electrical power generation and production of process heat required for industrial processes such as the manufacture of hydrogen for use as a fuel in automobiles. The DOE project is called the next generation nuclear plant (NGNP) and is based on a Generation IV reactor concept called the very high temperature reactor (VHTR), which will use helium as the coolant at temperatures ranging from 450 oC to perhaps 1000 oC. While computational fluid dynamics (CFD) has not been used for past safety analysis for nuclear reactors in the U. S., it is being considered for safety analysis for existing and future reactors. It is fully recognized that CFD simulation codes will have to be validated for flow physics reasonably close to actual fluid dynamic conditions expected in normal and accident operational situations. To this end, experimental data have been obtained in a scaled model of a narrow slice of the lower plenum of a prismatic VHTR. The present report presents results of CFD examinations of these data to explore potential issues with the geometry, the initial conditions, the flow dynamics and the data needed to fully specify the inlet and boundary conditions; results for several turbulence models are examined. Issues are addressed and recommendations about the data are made.

An accident, thermal fluids, and reactor physics phenomena identification and ranking process was conducted by a panel of experts on the next generation nuclear plant (NGNP) design (consideration given to both pebble-bed and prismatic gas-cooled reactor configurations). Safety-relevant phenomena, importance, and knowledge base were assessed for the following event classes: (1) normal operation (including some reactor physics aspects), (2) general loss of forced circulation (G-LOFC), (3) pressurized loss-of-forced circulation (P-LOFC), (4) depressurized loss-of-forced circulation (D-LOFC), (5) air ingress (following D-LOFC), (6) reactivity transients - including anticipated transients without scram (ATWS), (7) processes coupled via intermediate heat exchanger (IHX) (IHX failure with molten salt), and (8) steam/water ingress. The panel's judgment of the importance ranking of a given phenomenon (or process) was based on the effect it had on one or more figures of merit or evaluation criteria. These included public and worker dose, fuel failure, and primary (and other safety) system integrity. The major phenomena of concern that were identified and categorized as high importance combined with medium to low knowledge follow: (1) core coolant bypass flows (normal operation), (2) power/flux profiles (normal operation), (3) outlet plenum flows (normal operation), (4) reactivity-temperature feedback coefficients for high-plutonium-content cores (normal operation andmore » accidents), (5) fission product release related to the transport of silver (normal operation), (6)emissivity aspects for the vessel and reactor cavity cooling system (G-LOFC), (7) reactor vessel cavity air circulation and heat transfer (G-LOFC), and (8)convection/radiation heating of upper vessel area (P-LOFC).« less

This project characterizes typical two-phase stratified flow conditions in advanced water reactor horizontal pipe sections, following activation of passive cooling systems. It provides (1) a means to educate nuclear engineering students regarding the importance of two-phase stratified flow in passive cooling systems to the safety of advanced reactor systems and (2) describes the experimental apparatus and process to measure key parameters essential to consider when designing passive emergency core cooling flow paths that may encounter this flow regime. Based on data collected, the state of analysis capabilities can be determined regarding stratified flow in advanced reactor systems and the best paths forward can be identified to ensure that the nuclear industry can properly characterize two-phase stratified flow in passive emergency core cooling systems.

The Very-High-Temperature Reactor (VHTR) is one of six reactor technologies chosen for further development by the Generation IV International Forum. In addition this system is the leading candidate for the Next Generation Nuclear Power (NGNP) Project in the U.S which has the goal of demonstrating the production of emissions free electricity and hydrogen by 2015. In preparation for the thermal-hydraulics and safety analyses that will be required to confirm the performance of the NGNP, work has begun on readying the computational tools that will be needed to predict the thermal-hydraulics conditions and safety margins of the reactor design. Meaningful feasibility studies for VHTR designs will require accurate, reliable predictions of material temperatures which depend upon the thermal convection in the coolant channels of the core and other components. Unfortunately, one-dimensional system codes for gas-cooled reactors typically underpredict these temperatures, particularly for reduced power operations and hypothesized accident scenarios. Likewise, most turbulence models in general-purpose CFD codes also underpredict these temperatures. Matched-Index-of-Refraction (MIR) fluid dynamics experiments have been designed and built to develop benchmark databases for the assessment of CFD solutions of the momentum equations, scalar mixing and turbulence models for typical VHTR plenum geometries in the limiting case of negligiblemore » buoyancy and constant fluid properties.« less

Traditionally, nuclear reactor safety analysis has been performed using systems analysis codes such as RELAP5, which was developed at the INL. However, goals established by the Generation IV program, especially the desire to increase efficiency, has lead to an increase in operating temperatures for the reactors. This increase pushes reactor materials to operate towards their upper temperature limits relative to structural integrity. Because there will be some finite variation of the power density in the reactor core, there will be a potential for local hot spots to occur in the reactor vessel. Hence, it has become apparent that detailed analysis will be required to ensure that local ‘hot spots’ do not exceed safety limits. It is generally accepted that computational fluid dynamics (CFD) codes are intrinsically capable of simulating fluid dynamics and heat transport locally because they are based on ‘first principles.’ Indeed, CFD analysis has reached a fairly mature level of development, including the commercial level. However, CFD experts are aware that even though commercial codes are capable of simulating local fluid and thermal physics, great care must be taken in their application to avoid errors caused by such things as inappropriate grid meshing, low-order discretization schemes, lack ofmore » iterative convergence and inaccurate time-stepping. Just as important is the choice of a turbulence model for turbulent flow simulation. Turbulence models model the effects of turbulent transport of mass, momentum and energy, but are not necessarily applicable for wide ranges of flow types. Therefore, there is a well-recognized need to establish practices and procedures for the proper application of CFD to simulate flow physics accurately and establish the level of uncertainty of such computations. The present document represents contributions of CFD experts on what the basic practices, procedures and guidelines should be to aid CFD analysts to obtain accurate estimates of the flow and energy transport as applied to nuclear reactor safety. However, it is expected that these practices and procedures will require updating from time to time as research and development affect them or replace them with better procedures. The practices and procedures are categorized into five groups. These are: 1.Code Verification 2.Code and Calculation Documentation 3.Reduction of Numerical Error 4.Quantification of Numerical Uncertainty (Calculation Verification) 5.Calculation Validation. These five categories have been identified from procedures currently required of CFD simulations such as those required for publication of a paper in the ASME Journal of Fluids Engineering and from the literature such as Roache [1998]. Code verification refers to the demonstration that the equations of fluid and energy transport have been correctly coded in the CFD code. Code and calculation documentation simply means that the equations and their discretizations, etc., and boundary and initial conditions used to pose the fluid flow problem are fully described in available documentation. Reduction of numerical error refers to practices and procedures to lower numerical errors to negligible or very low levels as is reasonably possible (such as avoiding use of first-order discretizations). The quantification of numerical uncertainty is also known as calculation verification. This means that estimates are made of numerical error to allow the characterization of the numerical« less

The power density in the core of the block next generation nuclear power plant (NGNP) will not be uniform due to geometry, core layout and fuel pin design. Recent calculations performed to optimize the core design indicate that the maximum radial variation will be 1.25 times the average. This significant radial variation in the local power density will create a variation in the temperature of the helium coolant as it cools the core. The coolant channel with the highest outlet temperature is referred to as the ‘hot’ channel. The concern is that the high temperature channels, which exit into the lower plenum as jets, called ‘hot streaking,’ will adversely affect materials in the lower plenum, the exit duct and the turbine, as well as affect the performance of the turbine. The objective of the present study is to determine or bound the maximum exit temperature of the ‘hot’ channel. The maximum hot channel temperature depends on the total coolant flow rate, which has not yet been fixed. Experiments need to be designed to capture the complex physics of the lower-plenum flow to allow assessment and validation of numerical simulations. While preliminary CFD simulations are not yet validated, they can bemore » of use in the planning of the experiments, particularly in estimating where there are regions of high and low turbulence intensity. Mixing of the coolant is related to the turbulence intensity as well as to the overall nature of the mean flow. The purpose of the present task is to provide preliminary flow calculations of the coolant in the lower plenum to examine flow patterns and turbulence intensity.« less

Summaries of RELAP5/MOD3 code assessments, a listing of the assessment matrix, and a chronology of the various versions of the code are given. Results from these code assessments have been used to formulate a compilation of some of the strengths and weaknesses of the code. These results are documented in the report. Volume 7 was designed to be updated periodically and to include the results of the latest code assessments as they become available. Consequently, users of Volume 7 should ensure that they have the latest revision available.

The NRC confirmatory phase I testing for the AP600 safety systems has been completed in the modified ROSA (Rig of Safety Assessment) test facility located at the Japan Atomic Energy Research Institute (JAERI) campus in Tokai, Japan. The test matrix included a variety of accident scenarios covering both design and beyond-design basis accidents. The test results indicate the AP600 safety systems as reflected in ROSA appear to perform as designed and there is no danger of core heatup for the accident scenarios investigated. In addition, no detrimental system interactions nor adverse effects of non-safety systems on the safety system functions were identified. However, three phenomena of interest have been identified for further examination to determine whether they are relevant to the AP600 plant. Those three phenomena are: (1) a potential for water hammer caused by rapid condensation which may occur following the actuation of the automatic depressurization system (ADS), (2) a large thermal gradient in the cold leg pipe where cooled water returns from the passive residual heat removal system and forms a thermally stratified layer, and (3) system-wide oscillations initiating following the ADS stage 4 actuation and persisting until the liquid in the pressurizer drains and steam generation inmore » the core becomes insignificant.« less