Steady-State Thermal-Hydraulic Analysis of a Particle-Bedded Boiling Water Reactor

The current Light Water Reactors both BWR and PWR have extensive nuclear reactor safety systems, which provide safe and economical operation of Nuclear Power Plants. During about forty years of operation history the safety systems of Nuclear Power Plants have been upgraded in an evolutionary manner. The cost of safety systems, including large containments, is really high due to a capital cost and a long construction period. These conditions together with a low efficiency of steam cycle for LWR create problems to build new power plants in the USA and in the Europe. An advanced Boiling Water Reactor concept with micro-fuel elements (MFE) and superheated steam promises a radical enhancement of safety and improvement of economy of Nuclear Power Plants. In this paper, a new type of nuclear reactor is presented that consists of a steel-walled tube filled with millions of TRISO-coated fuel particles (Micro-Fuel Elements, MFE) directly cooled by a light-water coolant-moderator. Water is used as coolant that flows from bottom to top through the tube, thereby fluidizing the particle bed, and the moderator water flows in the reverse direction out of the tube. The fuel consists of spheres of about 2.5 mm diameter of UO2 with several coatings of different carbonaceous materials. The external coating of steam cycle the particles is silicon carbide (SiC), manufactured with chemical vapor deposit (CVD) technology. Steady-State Thermal-Hydraulic Analysis aims at providing heat transport capability which can match with the heat generated by the core, so as to provide a set of thermal hydraulic parameters of the primary loop. So the temperature distribution and the pressure losses along the direction of flow are calculated for equilibrium core in this paper. The calculation not only includes the liquid region, but the two phase region and the superheated steam region. The temperature distribution includes both the temperature parameters of micro-fuel elements and the coolant. The results show that the maximum fuel temperature is much lower than the limitation and the flow distribution can meet the cooling requirement in the reactor core.

Research on Monte Carlo Perturbation Calculation Methods Applied in Reactor Physics

In the reactor design and analysis, there is often a need to calculate the effects caused by perturbations of temperature, components and even structure of reactors on reactivity. And in sensitivity studies, uncertainty analysis of target quantities and unclear data adjustment, perturbation calculations are also widely used. To meet the need of different types of reactors (complex, multidimensional systems), Monte Carlo perturbation methods have been developed. In this paper, several kinds of perturbation methods are investigated. Specially, differential operator sampling method and correlated tracking method are discussed in details. MCNP’s perturbation calculation capability is discussed by calculating certain problems, from which some conclusions are obtained on the capabilities of the differential operator sampling method used in the perturbation calculation model of MCNP. Also, a code using correlated tracking method has been developed to solve certain problems with cross-section changes, and the results generated by this code agree with the results generated by straightforward Monte Carlo techniques.

Axial Development of Local Flow Parameters in Bubbly Two-Phase Flow in Normal and Microgravity Conditions

In relation to the development of the interfacial area transport equation, a precise database of the axial development of void fraction profile, interfacial area concentration and Sauter mean bubble diameter in an adiabatic nitrogen-water bubbly flow in a 9 mm-diameter pipe was constructed for normal and microgravity conditions using stereo image-processing. The flow measurements were performed at four axial locations (axial distance from the inlet normalized by the pipe diameter, z/D = 5, 20, 40 and 60) and with various flows: superficial gas velocity of 0.00840–0.0298 m/s, and superficial liquid velocity of 0.138–0.914 m/s. The effect of gravity on radial distribution of bubbles and the axial development of two-phase flow parameters is discussed in detail based on the obtained database and visual observation.

Thermal Design Options Using Uranium Carbide and Uranium Dicarbide in SCWR Uniformly-Heated Fuel Channel

SuperCritical Water-cooled nuclear Reactor (SCWR) options are one of the six reactor options identified in Generation IV International Forum (GIF). In these reactors the light-water coolant is pressurized to supercritical pressures (up to approximately 25 MPa). This allows the coolant to remain as a single-phase fluid even under supercritical temperatures (up to approximately 625°C). SCW Nuclear Power Plants (NPPs) are of such great interest, because their operating conditions allow for a significant increase in thermal efficiency when compared to that of modern conventional water-cooled NPPs. Direct-cycle SCW NPPs do not require the use of steam generators, steam dryers, etc. allowing for a simplified NPP design. This paper shows that new nuclear fuels such as Uranium Carbide (UC) and Uranium Dicarbide (UC2) are viable option for the SCWRs. It is believed they have great potential due to their higher thermal conductivity and corresponding to that lower fuel centerline temperature compared to those of conventional nuclear fuels such as uranium dioxide, thoria and MOX. Two conditions that must be met are: 1) keep the fuel centreline temperature below 1850°C (industry accepted limit), and 2) keep the sheath temperature below 850°C (design limit). These conditions ensure that SCWRs will operate efficiently and safely. It has been determined that Inconel-600 is a viable option for a sheath material. A generic SCWR fuel channel was considered with a 43-element bundle. Therefore, bulk-fluid, sheath and fuel centreline and HTC profiles were calculated along the heated length of a fuel channel.

Heat Exchange Characteristics in Heat Supply and Heat Removal Parts of a Nonisothermal Circuit, Including Temperature and Velocity Fields in a HLMC Flow, for Controlled Impurity Content

Results of the experimental studies of the heat exchange to the lead heat-transfer agent in the annular clearance in the circulation contour with the controlled and operated processes of mass exchange and mass transfer of the oxygen content are presented. And results of experimental research of lead-bismuth heat-carrier stream velocity structure at a varied content of oxygen content are presented.

Defense-in-Depth Approach for Use in a Risk-Informed, Technology-Neutral Design and Licensing Framework for New Nuclear Plants

The regulatory framework for the current generation of operating plants and advanced light water reactors (ALWRs) planned for near term construction has evolved over several decades to permit effective regulation of the light water reactor designs. To address other reactor types, development of a framework that possesses the attributes of being technology neutral, risk-informed and performance-based with corresponding processes (regulations and guidance) is ongoing by several U.S. and international organizations. A key design and operating principle which is applied to existing plants and will continue to be applied to future plants is defense-in-depth. The advent of advanced reactor designs, some of which are not based on light water reactor technology, provides incentive for changes in the regulatory framework in several areas, including defense-in-depth practices. To support development of an integrated framework, the Electric Power Research Institute (EPRI) conducted research to identify and assess specific elements of possible technology neutral, risk-informed, performance based frameworks that had been proposed by others. The intent was to develop a preliminary framework based on the results of this review and evaluation and to provide recommendations in areas where additional development and testing would appear to be most beneficial. “Technical Elements of a Risk-Informed, Technology-Neutral Design and Licensing Framework for New Nuclear Plants”, EPRI Report 1016150 documents this research (Reference [1]). For defense-in-depth (D-in-D) existing viewpoints from various sources were compared and an alternative integrated approach which addresses key issues was developed. These alternative views are contained in publications such as NUREG-1860 [2], Regulatory Guide 1.174 [6], IAEA Safety Standards Series No. NS-R-1 [3], IAEA 75-INSAG-3 Revision 1 [4], INSAG-12 [4], and IAEA INSAG-10 [5]. The results of this research support the ongoing efforts to develop standards and guidance for advanced plants with safety characteristics which differ from existing and advanced LWRs.

Code Acceptance of a New Joining Technology for Storage Containments

One of the activities associated with cleanup throughout the Department of Energy (DOE) complex is packaging radioactive materials into storage containers. Much of this work will be performed in high-radiation environments requiring fully remote operations for which existing, proven systems do not currently exist. These conditions require a process that is capable of producing acceptable (defect-free) welds on a consistent basis; the need to perform weld repair, under fully-remote operations can be extremely costly and time consuming. Current closure-welding technologies (fusion welding) are not well suited for this application and will present risk to cleanup cost and schedule. To address this risk, Fluor and the Pacific Northwest National Laboratory (PNNL) are proposing that a new and emerging joining technology, Friction Stir Welding (FSW), be considered for this work. FSW technology has been demonstrated in other industries (aerospace and marine) to produce near flaw-free welds on a consistent basis. FSW is judged capable of providing the needed performance for fully-remote closure welding of containers for radioactive materials. The performance characteristics of FSW, i.e., high weld quality, simple machine-tool equipment and increased welding efficiency, suggest that this new technology should be considered for radioactive materials packaging campaigns. FSW technology will require some development/adaptation for this application, along with several activities needed for commercialization. One of these activities will be to obtain approval from the governing construction code to use the FSW technology. The American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME B&PVC) will govern this work; however, rules for the use of FSW are not currently addressed. A code case will be required to define appropriate process variables within prescribed limits for submittal to the Code for review/approval and incorporation.

STP Risk Managed Technical Specification Software Design and Implementation

STP uses a custom enterprise software application called RICTCal to calculate risk informed completion times (RICTs). Besides providing the end user interface to the calculation engine, the software also creates electronic regulatory-required reports that are automatically filed in the plant records management system. In addition to regulatory-required information on risk informed completion times and risk managed action times (RMATs), the software provides additional configuration risk information such as risk for reactor trip. The computation methodology and design of the software is described as well as required input data to support the calculation.

Ascertaining Directional Information From Incident Nuclear Radiation

Unprecedented capabilities for the detection of nuclear particles are presented by tensioned metastable fluid states which can be attained via tailored resonant acoustic systems such as the acoustic tensioned metastable fluid detection (ATMFD) systems. Radiation detection in tensioned metastable fluids is accomplished via macro-mechanical manifestations of femto-scale nuclear interactions. Incident nuclear particles interact with the dynamically tensioned metastable fluid wherein the intermolecular bonds are sufficiently weakened such that the recoil of ionized nuclei generates nano-scale vapor cavities which grow to visible scales. Ionized nuclei form preferentially in the direction of incoming radiation, therefore, enabling the capability to ascertain information on directionality of incoming radiation — an unprecedented development in the field of radiation detection. Nuclear particle detection via ATMFD systems has been previously reported, demonstrating the ability to detect a broad range of nuclear particles, to detect neutrons over an energy range of eight orders of magnitude, to operate with intrinsic detection efficiencies beyond 90%, and to ascertain information on directionality of incoming radiation. This paper presents advancements that expand on these accomplishments, thereby increasing the accuracy and precision of ascertaining directionality information utilizing enhanced signal processing-cum-signal analysis, refined computational algorithms, and on demand enlargement of the detector sensitive volume. Advances in the development of ATMFD systems were accomplished utilizing a combination of experimentation and theoretical modeling. Modeling methodologies include Monte-Carlo based nuclear particle transport using MCNP5 and complex multi-physics based assessments accounting for acoustic, structural, and electromagnetic coupling of the ATMFD system via COMSOL’s Multi-physics simulation platform. Benchmarking and qualification studies have been conducted with special nuclear material (SNM), Pu-based neutron-gamma sources, with encouraging results. These results show that the ATMFD system, in its current configuration, is capable of locating the direction of a radioactive source to within 30° with 80% confidence.

Steam-Reheat Option for SCWRs

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30 – 35% to about 45 – 48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs. SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. To achieve higher thermal efficiency a nuclear steam reheat has to be introduced inside a reactor. Currently, all supercritical turbines at thermal power plants have a steam-reheat option. In the 60’s and 70’s, Russia, USA and some other countries have developed and implemented the nuclear steam reheat at subcritical-pressure in experimental reactors. There are some papers, mainly published in the open Russian literature, devoted to this important experience. Pressure-tube or pressure-channel SCW nuclear-reactor concepts are being developed in Canada and Russia for some time. It is obvious that implementation of the nuclear steam reheat at subcritical pressures in pressure-tube reactors is easier task than that in pressure-vessel reactors. Some design features related to the nuclear steam reheat are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors with the nuclear steam reheat is feasible and significant benefits can be expected over other thermal-energy systems.