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.

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.

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.

On Efficient Passive Cooling of Control Rod Drivers

Experimental research of natural convection and the ways of its suppression in an annular vertical channel to simulate the conditions of cooling the control rod drivers of the reactor protection system (RPS) in its so-called wet design, where the drivers are cooled by primary circuit water supplied due to the system that includes branched pipelines, valves, pump, heat exchanger, etc., is reported. Reliability of the drivers depends upon their temperature ensured by operation of an active multi-element cooling system. Its replacement by an available passive cooling system is possible only under significant suppression of natural convection in control rod channel filled with primary coolant. The methods of suppression of natural convection proposed in the work have demonstrated the possibility both of minimization of axial heat transfer and of almost complete elimination of temperature non-uniformity and oscillation inside the channel under the conditions of free travel of moving element (control rod) in it. The obtained results widen the possibilities of substitution of the active systems of cooling the RPS drivers by reliable passive systems, such as high-performance heat-transfer systems of evaporation-condensation type with heat pipes or two-phase thermosyphons as heat-transferring elements.

Measurement of Reaction Rate Distribution and Neutron Spectrum in China Experimental Fast Reactor

China Experimental Fast Reactor (CEFR) has completed physics start-up tests in 2010 and connected the grid on 40%FP in 2011. The reaction rate distribution, neutron spectrum are some important parameters for CEFR neutron field. In order to measure these parameters some low power irradiation tests using foil activation method have been done in CEFR core. Two kinds of special irradiation test subassemblies have been developed and fabricated for irradiation in CEFR core. And a digital high purity Germanium gamma-ray spectrometer system has been established for foil activity measurement. After dozens of low power irradiation tests in CEFR core, the radial and axial distribution of 235U and 238U fission reaction rate have been measured. The distribution of 238U capture reaction rate in CEFR core was also obtained in these tests. The experimental values of reaction rate are according with the calculation values well. Neutron spectrum was measured by means of multifoil activation method. And a neutron spectrum adjusting code was also compiled to determine the neutron spectrum.

A Safety Evaluation of HTTR Core Components Against 2011 Tohoku Earthquake

On March 11th, 2011, the 2011 Tohoku Earthquake which is one of the largest earthquakes in japan occurred and the maximum acceleration in observed seismic wave in the HTTR exceeded the design value in a part of input seismic motions. Therefore, a visual inspection, a seismic analysis and a performance confirmation test of facilities were carried out in order to confirm the integrity of facility after the earthquake. The seismic analysis was carried out for the reactor core structures by using the response magnification factor method. As the results of the evaluation, the generated stress in the graphite blocks in the reactor core at the earthquake were well below the allowable values of safety criteria, and thus the structural integrity of the reactor core was confirmed. The integrity of reactor core was also supported by the visual inspections of facilities and the operation without reactor power in cold conditions of HTTR.