Probabilistic Distribution of Parameters Applied to PFM Analysis for Piping in China Experimental Fast Reactor

The safety assessment method based on probabilistic fracture mechanics (PFM) is well applied to pressure vessel and piping. The PFM analysis is more reasonable and reliable than determinate fracture mechanics (DFM) method. In PFM analysis, the uncertainty of main assessment parameters, such as loads, material character parameters, structure dimension and defect sizes are considered to be random, and the probabilistic distribution of these parameters are determined with the theory of probability statistics. Related to the practical engineering of China experimental fast reactor (CEFR), this paper has done some research work on the parameters probabilistic distribution, and a method was given to determine the optimum fitting probabilistic distribution function of parameters applied to PFM analysis for piping in the small sample size. The work of this paper makes the foundation of the further probabilistic safety assessment of CEFR piping.

Enhancement of Plant Availability by Outage Shortening in European Advanced Boiling Water Reactor (EU-ABWR)

European Advanced Boiling Water Reactor (EU-ABWR) is developed by Toshiba. EU-ABWR accommodates an armored reactor building against Airplane Crash, severe accident mitigation systems, the N+2 principle in safety systems, the diversity principle and a large output of 1600 MWe. These features enable EU-ABWR’s design objectives and principles to be consistent with the requirements in the Finnish utility and the safety requirements of Finnish YVL guide. By adopting Scandinavian outage processes, the Plant Availability is aimed to be greater than 95%. ABWRs have an excellent design potential to acheive short outage duration (e.g., shortening of maintenance and inspection duration by applying Fine Motion Control Rod Drive and Reactor Internal Pump). In addition, the EU-ABWR applies following key design improvements to reduce a refueling outage duration; a) Direct Reactor Pressure Vessel (RPV) Head Spray System, b) Self-standing Control Rods and c) Water shielding reactor pool. In this paper, coolability of RPV due to application of the Direct RPV Head Spray System is also verified with numerical evaluations by Computation Fluid Dynamics (CFD) analysis.

A Heat Pipe Cooled Modular Reactor Concept for Manned Lunar Base Application

A lithium heat pipe cooled modular fast reactor (HPCMR) power system concept has been developed for manned lunar base application. The system is designed to use the static thermoelectric conversion module to produce over 100kW electricity for up to ten years. Waste heat is rejected by potassium heat pipe radiator. This system has advantages of low mass, long lifetime, no pumped liquid coolant, and no single point of failure. Main parameters of the system are also given in this paper.

Improvement of Temperature Evaluation Model of Biological Shielding Concrete for HTTR Test Simulating LOFC With VCS Inactive

HTTR is a helium gas cooled graphite-moderated HTGR with the rated power 30 MWt and the maximum reactor outlet coolant temperature 950°C. The vessel cooling system (VCS), which is composed of thermal reflector plates, cooling panel composed of fins connected between adjacent water cooling tubes, removes decay heat from reactor core by heat transfer of thermal radiation, conduction and natural convection in case of loss of forced cooling (LOFC). The metallic supports are embedded in the biological shielding concrete to support the fins of VCS. To verify the inherent safety features of HTGR, the LOFC test is planned by using HTTR with the VCS inactive from an initial reactor power of 9 MWt under the condition of LOFC while the reactor shut-down system disabled. In this test, the temperature distribution in the biological shielding concrete is prospected locally higher around the support because of thermal conduction in the support. A 2-dimensional symmetrical model was improved to simulate the heat transfer to the concrete through the VCS support in addition to the heat transfer thermal radiation and natural convection. The model simulated the water cooling tubes setting horizontally at the same pitch with actual configuration. The numerical results were verified in comparison with the measured data acquired from the test, in which the RPV was heated up to around 110 °C without nuclear heating with the VCS inactive, to show that the temperature is locally high but kept sufficiently low around the support in the concrete due to sufficient thermal conductivity to the cold temperature region.

High Breeding Core of a Supercritical-Pressure Light Water Cooled Fast Reactor

Breeding of plutonium with light water cooling has been studied for many years, but high breeding to meet growing demand for electricity in a developed country has not been accomplished. The purpose of this study is to investigate a high breeding core of Super FBR (supercritical pressure light water cooled fast breeder reactor) with new fuel assemblies consisting of tightly packed fuel rods without gaps, which leads to low coolant to fuel volume fraction. The plant system of a Super FBR is once-through coolant cycle with high head pumps. The coolant flow rate is low due to the high enthalpy rise in the core. It is compatible with the high pressure drop of the new fuel assemblies. Both neutronic and thermal hydraulic design of the core is considered. The challenge of high breeding with light water cooling is to satisfy negative coolant void reactivity, high breeding and low enrichment simultaneously. The core with new assemblies has been designed with the average coolant density of 248 kg/m3. It is achieved by setting 380C inlet and 500C outlet temperature. For satisfying negative void reactivity, a solid moderator layer composed of zirconium hydride (ZrH) rods are adopted in some blanket assemblies. Cross sections of the blanket fuel assemblies with ZrH rods are prepared with assembly-wise calculation, because the pin-wise collision probability calculation overestimates the breeding. MOX fuel is used for seed fuel assemblies. Three types of core layouts with “radially heterogeneous”, “radiating” and “scattered” seed assemblies have been considered, and “radiating” layout shows best breeding characteristics among them. The seed assemblies in a “radiating” layout are not radially separated so that more numbers of blanket assemblies can be placed in high neutron flux region of a core. Fraction of blanket fuel assemblies with ZrH rods is selected for high breeding. Super FBR using the new fuel assemblies achieved both negative void and high plutonium breeding.

Experiment Study on Collision of Graphite Dowel-Brick Structures

Graphite bricks have important applications in high temperature gas-cooled reactors (HTGRs), the core of HTGR is a pile-up of graphite bricks. So the vibrations and collisions between graphite bricks caused by external excitation have an important influence on structural stability of the core. The locations of bricks are fixed by various kinds of keys and dowels. The collision experiment, with tracks and small railcars as experimental devices and measurement system using optical method, was aimed at studying non-central collisions between two bricks. The passive one of the two bricks was equipped with a key or a dowel. Experiment’s results revealed how the coefficient of restitution and the contact time would change within the range of velocities of the active specimen. It was showed that the contact time would increase with the rise of initial velocity while the coefficient of restitution would rise up firstly and then decrease later in the same process. Besides, qualitative influence caused by different sizes of keys and dowels was briefly discussed, and material properties of graphite was not the dominate factor in the collision of dowel-brick structures experiment, while the velocity of active specimen just before collision and the fact that the collision is non-central have more significant effects on the collision results.

Optimization Design for Air Heat Exchanger of Large Fast Reactor Based on Genetic Algorithm

The passive design for decay heat removal system of future fast reactor will put forward higher requirement for air heat exchanger (AHX), which is directly relevant to the structure and anti-seismic design of stack. Under considering the heat exchanger ability and the structure compactness comprehensively, a strategy for the optimization design of AHX based on genetic algorithm was developed in this paper. The air resistance in shell side of vertical fin tube AHX was chosen as the objective function, and the effect of design parameters including fin pitch, number of tube rows, tube pitch and tube length on the air resistance was discussed. The results of the study show that the method for the optimization design of AHX based on genetic algorithm can effectively optimize the structure of AHX and improve the resistance characteristic of the shell side evidently, which leads to design the fast reactor plant, stack structure and seismic resistance simply.

A Numerical Study of Eccentricity Effect to Heat Transfer Under Single Phase Force Convection Inside a Vertical Annulus

Because annulus channel can be used to develop high efficiency compact heat exchangers, the heat transfer in annulus channel has become great interest to researchers in recent years. Most of the studies focus on the vertical concentric and horizontal eccentric annulus. The investigations about single phase force convection heat transfer inside a vertical eccentric annulus are not enough. In this work, force convection heat transfer is numerically studied to determine the eccentricity effect inside a vertical annulus. For this purpose, full Reynolds-averaged Navier-Stokes equations along with energy equations are solved in a 3-D grid. The discrete method of the equations is based on finite-volume method and the turbulence model is RNG k-ε model. The radius ratio of the annulus is 0.8 in this work. Heat flux of one wall is constant while the other is insulated. Firstly, the feasibility and exactness of the numerical method is proved by comparing the Nusselt number with experiment in concentric annulus. Then the effect of eccentricity is studied in detail.

Research on the Stainless Steel Tubes and Pipes Material Standards of Demonstration Fast Reactor Power Plant

The main characteristics of the sodium pipe system in Demonstration Fast Reactor Power Plant (DFRPP) are high-temperature, thin-wall and big-caliber, which is different from high-pressure and thick-wall of the pressurized water reactor system, and the system is long-term operate in the environment of liquid metal sodium. How to guarantee the reliability of materials in high temperature are most important in material option. Engineering design depend on the criterion. Material standards are different in different countries, and corresponding construction codes are different too. Comparing the stainless steel pipe material standers at home and abroad and analyzing the material standards’ difference according to different construction codes, a stainless steel pipe material criterion system is put forward in this paper which is applicable for the DFRPP.

Nonlinear Crest-Cut-Off Method for Reducing the Stress Concentration of Bolt With Many Threads: Application in the Main Bolt of HTGR’s RPV

For the standard thread, severe stress concentration appears in the root of first several threads which share the major part of bolt load, and this is also an important issue in the design of the main bolt thread in the High Temperature Gas-cooled Reactor’s (HTGR’s) Reactor Pressure Vessel (RPV). The linear crest-cutoff method (LCCOM) linearly cuts down the height of the engaged threads near the bolt head, and it could reduce the stress concentration at the roots of first several threads. However, as revealed in finite element simulation, when there are as many as forty threads in the bolt, even though the axial force shared by the first thread could be decreased by LCCOM, the axial force shared by subsequent threads are still very high, which are often larger than that by the first thread. To settle this problem, a nonlinear crest-cut-off method (NCCOM) is proposed, which employs the curved thread profile of quadratic polynomial function, instead of tapered profile of linear function in LCCOM. The proposed curved thread profile has one additional degree of freedom, and it could also be degenerated to tapered profile. As for the main bolt of forty threads in the HTGR’s RPV, the suitable parameters of curved profile are determined by the intensive numerical simulations. The results show that the proposed NCCOM yields lower axial force of the first several threads, and produces lower stress at the roots of threads in the bolt compared with LCCOM.