Feasibility Study of a New Fabrication Method for the Li4SiO4 Pebbles

In the past decades many R&D efforts have been spent in the development of a suitable Li4SiO4 fabrication method (e.g., melt spray process, graphite bed method, capillary-based microfluidic wet method etc.), nevertheless we are still far from an “industrial standard solution”. The aim of the paper is to develop a new fabrication method capable to produce stable and well-sized pebbles of lithium orthosilicate (Li4SiO4) based on the drip casting. This method is mainly based on the dripping at room temperature, which is novel in the framework of available fabrication processes requiring high temperature: this latter is demonstrated to affect the final product characteristics. It is worthy to remark that the Li4SiO4 is a candidate material for the breeding blanket material of the fusion power reactor. In the paper we will describe the experimental apparatus, designed and built at DICI - University of Pisa with the collaboration of Bitossi Industries, and the procedure adopted in order to produce pebbles of Li4SiO4.

The Characteristics Study of Helium-Xenon Mixture in Closed Brayton Cycle for Space Nuclear Reactor Power

Differing from the adoption of helium as working fluid of closed Brayton cycle (CBC) for terrestrial high temperature gas cooled reactor (HTGR) power plants, helium-xenon mixture with a proper molar weight was recommended as working fluid for space nuclear reactor power with CBC conversion. It is essential to figure out how the component of helium-xenon mixture affects the net system efficiency, in order to provide reference for the selection of appropriate cycle working fluid. After a discussion of the physical properties of different helium-xenon mixtures, the related physical properties are studied to analyze their affection on the key parameters of CBC, including adiabatic coefficient, recuperator effectiveness and normalized pressure loss coefficient. Then the comprehensive thermodynamics of CBC net system efficiency is studied in detail considering different helium-xenon mixtures. The physical properties study reveals that at 0.7 MPa and 400 K, the adiabatic coefficient of helium-xenon mixture increases with increased molar weight, from 0.400 (pure helium) to 0.414 (pure xenon), while recuperator effectiveness firstly increases and then decreases with the increase of molar weight, and the normalized pressure loss coefficient increases monotonically with molar weight increases. The thermodynamic analysis results show that the adiabatic coefficient has less effect on the net system efficiency, while the net system efficiency increases with increased recuperator effectiveness, and the net system efficiency decreases with normalized pressure loss coefficient increases. Finally, the mixture of helium-8.6% xenon was adopted as working fluid, instead of pure helium, for ensuring less turbine mechanicals (turbine and compressor) stages, and resulting maximum recuperator effectiveness. At the given cold / hot side temperature of 400 / 1300 K, the net system efficiency can reach 29.18% theoretically.

Study on Supervision Mode of Floating Nuclear Power Plant With Small Modular Reactor

A floating nuclear power plant (FNPP) with small modular reactor (SMR) is a combination of a civilian nuclear infrastructure and an offshore installation, which is defined as a floating nuclear facility. The article draws the lessons from studying of the engineer combination like Floating Production Storage and Offloading (FPSO) under the regulation of several government departments. It puts forward recommendations for license application and government regulation as follows in consideration with current license application for nuclear power plant and ship survey. A FNPP shall follow the requirements of construction, fueling and operation for civil nuclear installation combined with ship survey. Application is submitted to nuclear safety regulator for construction permit, while the design drawings shall be submitted to department of ship survey which checks the drawings whether meet the requirements of ship survey, considering some nuclear safety needs. The result of ship survey shall be represented in the safety analysis reports. The construction and important devices manufacturing shall be under the supervision of nuclear installation regulators and ship survey departments. In conclusion, National Nuclear Safety Administration (NNSA) and Maritime Safety Administration of the People’s Republic of China (MSA) shall establish united supervisory system for SMR on sea in China. It is suggested that NNSA is in charge of the overall safety of a FNPP, while MSA is responsible of the ship survey. The operator shall undertake obligation of a FNPP and evaluate the ship cooperating with experienced agency. It is suggested that government departments build the mutual recognition agreement of safety review. It is better to solve the vague questions by coordination.

New Needs of Fracture Mechanic Analysis at Design and Operation Level: Status of French Nuclear Mechanical Codes

Based on ASME Boilers and Pressure Vessels Code the major fracture mechanic analysis is limited to protection of class 1 components to brittle fracture. All the Operators of future plants have to enlarge the scope of these analyses to different concepts, at design or operation stage: - brittle and ductile analysis of hypothetical large flaw - leak before break approach - break exclusion concept - incredibility of failure of high integrity components - end of fabrication acceptable defect - in-service inspection performance - acceptable standards in operation - Long Term Operation (LTO) All these requirements needs a procedure, an analysis method with material properties and criteria. After a short overview of each topic, the paper will present how RCC-M, RSE-M French Codes and ASME III and XI take care of all these new modern regulatory requirements.

Nonlinear Analysis in Pressure Vessel Design Codes: Recommendations for Codified Rules Improvements

During the past 30 years the main rules to design pressure vessels were based on elastic analyses. Many conservatisms associated to these different elastic approaches are discussed in this paper, like: stress criteria linearization for 3-D components, stress classification in nozzle areas, plastic shake down analysis, fatigue analysis, Ke evaluation, and pipe stress criteria for elastic follow-up due to thermal expansion or seismic loads... This paper will improve existing codified rules in nuclear and non-nuclear Codes that are proposed as alternatives to elastic evaluation for different failure modes and degradation mechanisms: plastic collapse, plastic instability, tri-axial local failure, rupture of cracked component, fatigue and Ke, plastic shakedown. These methods are based on limit loads, monotonic or cyclic elastic-plastic analyses. Concerned components are mainly vessels and piping systems. No existing Code is sufficiently detailed to be easily applied; the needs are stress analysis methods through finite elements, material properties including material constitutive equations and criteria associated to each methods and each failure modes. A first set of recommendation to perform these inelastic analysis will be presented to improve existing codes on an international harmonized way, associated to all material properties and criteria needed to apply these modern methods. An international draft Code Case is in preparation.

AFCEN RCC-F: A New Standard for the Fire Protection Design of New Built Light Water Nuclear Power Plants

Fire hazard is an important issue for the safety of nuclear power plants: the main internal hazard in terms of frequency, and probably one the most significant with regards to the design costs. AFCEN is publishing in 2018 a new code for fire protection of new built PWR nuclear plants, so-called RCC-F. This code is an evolution of the former ETC-F code which has been applied to different EPR plants under construction (Flamanville 3 (FA3, France), Hinkley Point C (HPC, United Kingdom), Taïshan (TSN, China)). The RCC-F code presents significant enhancement and evolutions resulting from eight years of work by the AFCEN dedicated sub-committee, involving a panel of contributors from the nuclear field. It is now opened to any type of PWR (Pressurized Water Reactor) type of nuclear power plants and not any longer limited to EPR (European Pressurized Reactor) plants. It can potentially be adapted to other light water concepts. Its objective is to help engineers design the fire prevention and protection scheme, systems and equipment with regards to the safety case and the defense in depth taking into account the French and European experience in the field. It deals also with the national regulations, with two appendices dedicated to French and British regulations respectively. The presentation gives an overview of the code specifications and focuses on the significant improvements.

Research on Heat Transfer Characteristics of the Thermometric Sphere in HTR-10

During the operation of the High Temperature Gas-cooled Reactor (HTGR), the hot-spot temperature in the reactor core must be lower than the maximum permissible temperature of the fuel elements and the materials of construction, so that the reactor kept safe. However, no fixed temperature-measuring devices can be set in a pebble-bed core. A special spherical temperature-measuring device is adopted to make sure it brings as small impact to the reactor operation as possible. There are several metal wires with different melting points inside. The graphite thermometric balls will be put onto the top of HTR-10 reactor core, and they record and reflect the highest temperature in different positions in the core when flowing in the pebble bed. Before the reactor core temperature-measuring experiment of HTR-10, we must study the heat transfer characteristics of the graphite thermometric sphere to find out the relationship of the melting conditions and the temperature in the reactor core. A 3-D model of the graphite thermometric ball is established, and CFD method is adopted to research and figure out the thermal equilibrium time and temperature difference between the metal wires in the ball and the hot fluid outside the balls. Multiple situations are simulated, and the heat transfer process of the thermometric sphere is comprehensively studied. The heat convection is certified the most important aspect. Thermal equilibrium can be achieved within 19 minutes, far shorter than the period while the spheres flowing through the core. The simulation results can also applied to derive the thermal fluid temperature backward.

Effect of Rolling Motion on Flow Instability of Parallel Rectangular Channels of Natural Circulation

In order to study the effect of rolling motion on flow instability of parallel rectangular channels of natural circulation, the natural circulation reactor simulation system is used for physical prototype. And theory analysis model of parallel rectangular channels of natural circulation system under rolling motion is established and coded by Fortran. The results of the program are verified to the experiments, and the results are in good agreement. The flow instability boundaries of different pressure under static and rolling motion are calculated respectively. The results show that: 1) under static condition, with the increase of the pressure, the instability boundary line changes, and the system becomes more stable; 2) under rolling conditions, the heating power of instability boundary decreases comparing to the stable conditions. The instability occurs earlier; 3) the stability of the system decreases with the increasing of rolling amplitude and frequency.

Example of Graded and Lifecycle Phase-Specific Security Controls for Nuclear I&C and EPS Use Cases

Working Group WGA9 of IEC SC45A (Nuclear I&C and ES), has recently completed a further working draft (WD) of the new IEC 63096 (unpublished) standard, aptly entitled Nuclear Power Plants – Instrumentation, Control and Electrical Systems – Security Controls. IEC 63096 specifically focuses on the selection and application of computer security controls for computer-based I&C and ES systems. This standard follows the commonly accepted ISO/IEC 27000 series security objectives of confidentiality, integrity and availability, and borrows and expands the objectives and implementation guidance from ISO/IEC 27002, while considering recommendations on sector-specific standards by ISO/IEC 27009. In addition, this guidance introduces a security grading, as well as lifecycle phase-specific controls. The grading aligns with the stringency of security controls, starting with Baseline Requirements (BR), Security Degree S3 and up to S1 (from lowest to highest degree). The lifecycle phase concerns the I&C development (D), project engineering (E) and operation and maintenance phases (O). This paper applies a sub-clause of IEC 63096 clause 15 (Supplier Relationships), to a programmable logic controller (PLC) that is typically used in power plants, to show the intended use of this standard and how it complements highest safety requirements in power plants. The Supplier Relationship clause concerns topics related to supply chain security, and is used to develop a use case example for the PLC. This example demonstrates how the controls and security degrees fits the implementation guidance from ISO/IEC 27002 and how they can be methodically applied to an I&C system.

“Begging the Question” in Licensing Basis Accident Analyses

“Begging the question” describes a situation in which the statement under examination is assumed to be true (i.e., the statement is used to support itself). Examples of this can be found in analysis reports that were prepared by analysts who are not mindful (or maybe uninformed) of the analysis criteria they’re required to fulfill. This is generally seen in analyses of anticipated operational occurrences (AOOs). AOOs are defined in Appendix A of 10 CFR §50 [1], and in ANS-N18.2-1973 [2], where they’re also known as American Nuclear Society (ANS) Condition II events. This standard [2] also defines more serious, Condition III and IV events. Analyses of AOOs, or ANS Condition II events are required to show that: (1) reactor coolant system (RCS) pressure will not exceed its safety limit, and (2) no fuel damage will be incurred, and (3) a more serious accident will not develop, unless there is a simultaneous occurrence of another, independent fault. The three requirements are often demonstrated by three different analyses, each of which is designed to yield conservative results with respect to one of the requirements. Accident analyses that are performed to demonstrate compliance with the first two requirements are relatively straightforward. They rely mostly upon the design of safety valves and the timing of reactor trips. “Begging the question” is seen in analyses that are designed to demonstrate compliance with the third requirement. This paper will describe how this logical fallacy has been applied in licensees’ accident analyses, and accepted by the NRC staff.