Reactor Vessel Wall Research Articles

Numerical simulation on in-vessel molten corium behavior with external vessel cooling using smoothed particle hydrodynamics

The in-vessel retention through external reactor vessel cooling (IVR-ERVC) strategy is a key management strategy for early termination of a nuclear severe accident that can threaten the integrity of the reactor vessel. To simulate the physical phenomena of the molten corium, the smoothed particle hydrodynamic (SPH) method is utilized in this study. The SPH method is a Lagrangian computational fluid dynamic (CFD) method that can simulate multi-fluid stratification, turbulence, natural circulation, radiative heat transfer, thermal ablation, and crust formation. To address the external vessel cooling, it is coupled with a conventional 1-D nuclear system analysis method. The 1-D system analysis code can calculate the two-phase natural circulation of cooling water and the convective heat transfer on the external reactor vessel wall. These two simulation codes exchange the temperature and heat flux of the reactor vessel outer wall. This study numerically simulated the IVR-ERVC strategy for a Korean high-power reactor and compared it with the traditional lumped parameter method (LPM). Unlike LPM, this study provides localized detailed data about the thermal hydraulic behavior of molten corium and visualization of phenomena in the IVR-ERVC strategy. This enhances our understanding of the phenomena in IVR-ERVC strategy and introduces new perspectives.

Variation of heat load due to in-vessel debris bed in IVR condition

We performed experiments investigating the heat load imposed on the inner reactor vessel wall in the in-vessel retention of core melt condition, varying the debris bed height and configuration. Mass transfer method was adopted to achieve high RaʹH of order ∼1015 with compact test rigs. We tested three types of the debris configurations: cone, flat and skew debris beds. The results showed that Nusselt numbers decreased in both the uppermost and the curved surfaces with the increasing bed height. Due to the friction increased by the debris bed, the natural convective flow in the oxide pool was weakened, which intensified the impairment of the heat transfer. As the rising plume at the center was disturbed by the porous structure of the debris bed, the local average heat transfer of the uppermost surface was weakened at the center. The heat transfer was dramatically decreased at the curved surface where the debris bed was covered.

Coupled Analysis of Heat Transfer in a Molten Pool With Three-Layer Configuration

Understanding a reliable description of the heat transfer in the molten pool with three-layer configuration is inevitable for the design of severe accident mitigation measures in research reactors, such as IVR-ERVC. However, investigations on a thick lower metallic layer are scarce. Thus, further studies on heat transfer in a molten pool with a thick lower metallic layer are highly required, based on previous severe accident studies in light water reactors. This study conducts a numerical simulation of heat transfer in the molten pool with a thick lower metallic layer during a severe accident, combining model development, code verification, and code application. First, the MPCAP code has been developed, based on the coupled simulation of natural convection inside the molten pool, two-dimensional heat conduction in the RPV lower head, and convection outside the reactor vessel wall. Then, a verification case has been conducted using three-layer configuration in AP1000. The AP1000 verification case indicates that the MPCAP code predicts important parameters in the molten pool well.

Development of IVR-ERVC evaluation method and its application to the SMART

An IVR (In-Vessel corium Retention) through ERVC (External Reactor Vessel Cooling) is known to be an effective means for maintaining the reactor vessel integrity during severe accidents in nuclear power plants. The evaluation method of the IVR-ERVC for small reactors was developed and applied to a small integral reactor of the SMART (System-integrated Modular Advanced ReacTor). As a first step, the thermal load analysis from the corium to the external reactor vessel wall is necessary. As a second step, the natural circulation mass flow rate analysis, which is formed in the annular gap between the external reactor vessel wall and insulator, is necessary using the thermal hydraulic computer code. The CHF (Critical Heat Flux), which is the maximum heat removal rate from the external reactor vessel wall depending on the estimated natural circulation mass flow rate, is determined using experimental results. The success criteria of the IVR-ERVC to prevent reactor vessel failure during severe accidents is determined by comparing of the thermal load with the CHF. Finally, a structure analysis on the reactor vessel wall performs to evaluate the structure integrity of the reactor vessel wall in ERVC condition using the structure analysis computer code. Consequently, the IVR-ERVC strategy turned out to be an effective means for maintaining the SMART reactor vessel integrity during severe accidents from the thermal load analysis, CHF analysis, and structure integrity analysis.

Investigation of various reactor vessel auxiliary cooling system geometries for a hybrid micro modular reactor

The four different geometries of a reactor vessel auxiliary cooling system (RVACS) are designed to investigate how the heat removal capability changes by the geometries. Each geometry has different heat transfer characteristics regarding airflow and the presence of an air separator and an insulation material. The heat removal performance is evaluated with the reactor vessel (RV) temperature, the RV wall emissivity, and the airflow gap. For the same RVACS volume, 600 °C RV temperature, and 6-cm gap, the heat removal capability varies from 240.8 kW to 308.5 kW, depending on the geometry. The wall emissivity is less effective for Geometry 2, which has a large cavity volume and a small heat transfer area compared to the other geometries. The highest heat removal performance was obtained using Geometry 3 because cold air flows in from the bottom of the RVACS, improving both radiative and convective heat transfer. Reducing the gap size by 3 cm results in only 80.0 kW of heat removal capability for Geometry 1, and the heat removal dramatically decreases to 0.2 kW at an RV temperature of 800 °C. A sufficient RVACS gap size of at least 6 cm with a 0.6-m diameter intake pipe is required to provide adequate natural circulation and eventually enhance the heat removal capability.

Matrix effect impact on measuring hardness of metals bombarded by accelerated ions using laser induced breakdown spectroscopy

For safety reasons, the need of a fast in situ technique for measuring change in surface hardness of nuclear reactor vessel walls due to ion bombardment has great importance. Material hardness measurement using laser induced breakdown spectroscopy has been demonstrated as a new promising technique. The effect of the material matrix is one of the challenges that this technique for hardness measurement may face. In this work, the plasma excitation temperature Te has been used to measure the relative hardness of some pure Ti and Ti alloy samples with different hardnesses. The hardness of these samples was changed by irradiating the samples with 1 MeV carbon (C) ions with different doses. On studying the average crystallite size of these samples using x-ray diffraction and its relation with hardness, it was found that pure Ti samples and Ti alloys show different behaviors. On the other hand, it was found that although there is a good linear relation between Te and the Vickers hardness, the sample matrices have a strong influence on the measurements, especially for samples with small hardness value differences, irrespective of their average crystallite size.

Two Dimensional Performance Analysis of Small HTR Residual Heat Removal System in DLOFC Condition

This research is aimed to investigate a 10 MWth HTR thermal parameter response to long-term Depressurized Loss Of Force Cooling (DLOFC) accidents. By using two dimensions analysis in radial and vertical direction, comparing to one dimension in radial, the results are expected to be closer to real conditions. The method used is firstly develop a Matlab-based temperature distribution analysis program that solves the two-dimensional conduction equation in cylindrical coordinates and also in a function of time. Using the program, DLOFC simulations were conducted over a period of 588 hours after 200 hours of time needed to reach steady state close to normal operating conditions. It can be evaluated how far the temperature of all reactor components starting from the core to the concrete wall will rise and whether they are within their safety limits. The results show that during DLOFC, temperature of all reactor components increase toward the maximum value before decreasing again. The maximum temperature of the reactor core, reactor vessel and concrete wall are respectively 1080°C, 231°C, and 96.1 °C. The maximum value of the core temperature is reach at 11.1 hours after accident. The water temperature in the cooling panel, for 50 m3 of water volume, has reached its boiling point of 100°C at around 100 hours after DLOFC and decreasing again at the time of 360 hours. Evacuation of the residual heat is not only through the cooling water panel, but also through the three directions of concrete wall, ie top surface, peripheral surface and bottom surface. During DLOFC, the maximum heat power evacuated through top, bottom and peripheral surface are respectively 4.57 kW, 4.84 kW and 29.04 kW. The maximum heat power evacuated by water panel is 52.7 kW. It can be concluded that the reactor remains safe in very long time of DLOFC.

Effect of Heat on Coating Process of Joint Tube Wall Ascending Tube Fuel Handling System RDE

Fuel forwarding pipeline as part of fuel handling system enters the reactor vessel through a joint tube which is attached in external reactor vessel wall. Joint tube is made of carbon steel and coated inside with stainless steel. The coating has function to protect the carbon steel from rapid corrosion due to high temperature. The present study purposes to investigate the effect of heat on coating process for mechanical properties of material joint tube component. The experiments use carbon plate as base metal. The coating use welding process technique with six variations of current and welding voltage by using Gas Metal Arc Welding (GMAW). There are 18 stainless steel coated specimens and two fresh materials has been prepared for tensile and hardness testing. The obtaining results from the tensile testing indicate that stainless steel coating on carbon steel does not reduce its strength. The hardness of material has increase in the spot point near the metal coating due to the cooling rate process. Thus, the mechanical design of joint tube such as piping stress analysis able to calculate based on the carbon steel thickness and the effect of coating can be neglected.

CHF correlation development under ERVC conditions by using the local liquid velocity from PIV measurements

In-vessel retention through external reactor vessel cooling (IVR-ERVC) is a severe accident management strategy that removes the decay heat from the molten corium in the reactor vessel by flooding the reactor cavity. In ERVC conditions, the critical heat flux (CHF) on the outer wall of the vessel is a crucial factor that indicates a thermal margin of the system. Previous studies used averaged mass flux to predict CHF (Haramura and Katto, 1983; Lee and Mudawwar, 1988, Katto, 1990; Jeong et al., 2005; Park et al., 2013). The averaged value cannot indicate the degree of liquid supply which is the most critical factor in the CHF model. Local liquid velocity near the heated surface is one of the most crucial factors that should be quantified to develop a CHF model. In this study, the local liquid velocity near the surface has been measured. A curved rectangular flow channel was devised to simulate the gap between the external reactor vessel wall and the insulation. The boiling heat transfer and CHF were simulated with air injection at the inner wall of the test section. In the experiment, the flow field under the simulated ERVC conditions was measured using the particle image velocimetry (PIV) technique. The PIV measurement was validated by comparing the measured velocity data and the analytic solution under circular pipe flow. The liquid velocity profile at the middle plane of the curved rectangular test section was obtained in the multiple mass flux and inclination angle conditions. Based on the velocity data, a local velocity correlation was developed. The local velocity correlation was applied to improve the CHF prediction model.

Effect of honeycomb porous plate on critical heat flux in saturated pool boiling of artificial seawater

During a severe nuclear power plant accident, the integrity of the reactor pressure vessel must be assured. In response to a possible fuel meltdown, operators of the current generation of nuclear power plants are likely to inject water into the reactor pressure vessel cavity to cool down the reactor vessel wall, preserving its integrity and avoiding leakage of radioactive material. This study considers the use of seawater to flood a reactor pressure vessel cavity combined with the attachment of a honeycomb porous plate (HPP) on the vessel outer wall as a way to improve the safety margins for in-vessel retention of fuel. In long-duration experiments, saturated pool boiling of artificial seawater was performed with an upward-facing plain copper heated surface 30 mm in diameter. The resulting value for critical heat flux (CHF) was 1.6 MW/m2 at atmospheric pressure, a value significantly higher than the CHF obtained when the working fluid was distilled water (1.0 MW/m2). It was verified that sea-salt deposits could greatly improve surface wettability and capillarity, enhancing the CHF. The combination of artificial seawater and an HPP attached to the heated surface improved the boiling heat transfer coefficient and increased the CHF up to 110% (2.1 MW/m2) as compared to distilled water on a bare surface. After the artificial seawater experiments, most of the wall micropores of the HPP were clogged due to sea-salt aggregation on the HPP top and bottom surfaces. Thus, the CHF enhancement observed in this case was attributed mainly to the separation of liquid and vapor phases provided by the HPP channel structure and improvement of surface wettability and capillarity by sea-salt deposition.

Effects of surface orientation and heater material on heat transfer coefficient and critical heat flux of nucleate boiling

External reactor vessel cooling (ERVC) is the key technology for In-Vessel Retention (IVR) to ensure the safety of a nuclear power plant (NPP) under severe accident conditions. The thermal margin of nucleate boiling heat transfer on the reactor pressure vessel (RPV) lower head is important for ERVC and of wide concern to researchers. In such boiling heat transfer processes, the reactor vessel wall inclination effect on the heat transfer coefficient (HTC) and critical heat flux (CHF) should be considered. In this study, experiments were performed to investigate the effects of surface orientation and heater material on the HTC and CHF of nucleate boiling. Copper (Cu), stainless steel (SS) as well as prototype material (SA508) of the reactor pressure vessel in the nuclear power plant were used to perform boiling tests under atmosphere pressure, respectively. The orientation angle of all boiling surfaces were varied between 0° (upward) and 180° (downward). The experimental results show that the HTC is enhanced for the downward heater surfaces (φ>90°) under the low heat flux conditions and the enhancement ratio decreases with the increase of the heat flux. In addition, the relationship of measured CHF value with the orientation angle was obtained and it shows that the CHF value changes little as the inclination angle increases from 0° to 120°, but it decreases rapidly as the orientation angle increases towards 180° for all boiling surfaces. The material effect, which indicated by thermal effusivity ρcpk, on CHF is also investigated that the surface with larger thermal effusivity has the higher CHF value. In addition, the SA508 surface has the highest CHF value comparing with those of Cu surface and SS surface due to surface corrosion in the boiling process. Based on the experimental data, a correlation for CHF prediction is developed which includes the effects of surface orientation, thermal effusivity and corrosion.

Reactor nozzle inspection based on the laser-guided mobile robot

Most inspection and repair works are performed with robots in a nuclear power plant, as a human operator cannot access the facility due to radiation. In order to ensure the integrity of the reactor vessel, the reactor nozzle welds should be inspected periodically. The inspections have been performed using a conventional inspection machine with a large sturdy column; however, these machines are so large and heavy that their handling and maintenance requires much effort. In order to resolve these kinds of problems, the authors have developed a mobile robotic manipulator, which is very small and light and easy to handle. The underwater mobile robot with magnetic wheels is guided by a laser beam for its accurate positioning. Even though the inspection system is light and simple, its accurate positioning for reactor nozzle weld inspection is much more difficult and complex. This article explains how to position our laser-guided mobile robot manipulator at the reactor vessel wall and determines the pan and tilt angle of the laser pan-tilt device to guide the robot along a specified path around the nozzle. We derive the path planning of the mobile robotic manipulator when inspecting nozzle welds on the upper shell of the vessel, and we experimented the whole robotic system to identify position accuracy of the robot movements through the inspection scanning of reactor nozzle.

Transition from molecular diffusion to natural circulation mode air-ingress in high temperature helium loop

High Temperature Gas-cooled Reactors (HTGRs) have significantly robust passive heat removal capabilities owing largely to conduction and radiation heat transfer through graphite-fuel matrix and reactor vessel wall respectively. However, these capabilities may be significantly impacted if air was to replace Helium in the high temperature graphite under accident scenarios. In case of break in the coolant system high pressure Helium will escape into the reactor cavity leading to depressurization of the reactor. This will allow air from reactor cavity to enter the plenum or core of HTGRs via different mechanisms – diffusion, gravity currents, or natural circulation. Experimental studies were conducted in a geometrically scaled set-up with a shape of small English letter ‘h’, to depict the role of upper plenum in this process. Main focus of these studies was to understand and observe the transition time from diffusion to natural circulation. Thermal camera is used to identify this transition and flow transducer is used to measure the flow rates. The experimental results on incipience time of natural convection after a chamber pre-filled with Helium is opened to atmosphere are qualitatively similar to observations reported in literature using ‘inverse U’ shaped experimental setup, but also show distinct quantitative effect of extended leg on the transition times. Prior to onset of natural circulation (ONC), molecular diffusion plays the significant role in air-ingress. However, at higher temperatures convection currents may be influencing air ingress and in-turn ONC times.

EXPERIMENTAL STUDY OF STAINLESS STEEL 304L INTERACTION WITH ZIRCALOY AT HIGH TEMPERATURES

One of the critical factors in the analysis of the in-vessel retention of corium during a postulated severe accident in a nuclear power plant is the interaction of corium constituents with the reactor vessel wall material. In a CANDU© reactor, after fuel channel disassembly, corium constituents would come into contact with the calandria vessel wall. Ablation of the wall due to physicochemical interaction with corium at high temperature could cause vessel failure, increasing the likelihood of radioactive release to the environment. Therefore, the interaction of the calandria vessel wall material (stainless steel 304L) with each of the main corium constituents (Zircaloy-4, Zr-2.5%Nb, Zircaloy-2, and UO2) was studied experimentally. Arrhenius equations that can be used for analysis of in-vessel corium retention were derived from experimental results. For alloys of Zr, measured rates of interaction were comparable with those reported for similar light-water reactor materials. For UO2, the rate of interaction ...

Low-enrichment and long-life Scalable LIquid Metal cooled small Modular (SLIMM-1.2) reactor

The Scalable LIquid Metal cooled small Modular (SLIMM-1.0) reactor with uranium nitride fuel enrichment of 17.65% had been developed for generating 10–100MWth continuously, without refueling for ∼66 and 5.9 full power years, respectively. Natural circulation of in-vessel liquid sodium (Na) cools the core of this fast energy spectrum reactor during nominal operation and after shutdown, with the aid of a tall chimney and an annular Na/Na heat exchanger (HEX) of concentric helically coiled tubes. The HEX at the top of the downcomer maximizes the static pressure head for natural circulation. In addition to the independent emergency shutdown (RSS) and reactor control (RC), the core negative temperature reactivity feedback safely decreases the reactor thermal power, following modest increases in the temperatures of UN fuel and in-vessel liquid sodium. The decay heat is removed from the core by natural circulation of in-vessel liquid sodium, with aid of the liquid metal heat pipes laid along the reactor vessel wall, and the passive backup cooling system (BCS) using natural circulation of ambient air along the outer surface of the guard vessel wall. This paper investigates modifying the SLIMM-1.0 reactor design to lower the UN fuel enrichment. To arrive at a final reactor design (SLIMM-1.2), the performed neutronics and reactivity depletion analyses examined the effects of various design and material choices on both the cold-clean and the hot-clean excess reactivity, the reactivity shutdown margin, the full power operation life at 100MWth, the fissile production and depletion, the Beginning-Of-Life (BOL) spatial distributions of the fission power in the core at 100MWth, and the negative temperature reactivity feedback effects. Besides decreasing the UN fuel enrichment, other parameters examined are the materials of the followers to the B4C rods for RSS and RC, the rods in the radial blanket assemblies, and the thickness of the scalloped BeO walls of the UN fuel and radial blanket assemblies. Despite ∼13% lower fuel enrichment (15.35%) and ∼22% lower hot-clean excess reactivity ($6.29 versus $8.06 for SLIMM-1.0 at 100MWth), the operation life of the SLIMM-1.2 is ∼6.8% longer (6.3 versus 5.9 FPY for SLIMM-1.0), and the total negative temperature reactivity feedback is slightly smaller. At BOL, the radial blanket assemblies in ring 4, and the six UN fuel assembles in ring 1 of the SLIMM-1.2 core generate 2% and 23% of the total reactor thermal power, respectively. The twelve and eighteen UN fuel assemblies in rings 2 and 3 of the SLIMM-1.2 core generate 38% and 37% of the reactor thermal power, respectively. At EOL, the thermal power generation by the UN assemblies in rings 1 and 2 of the SLIMM-1.2 core decreases to 21.5% and 35.9%, respectively, while that by the assemblies in rings 3 and 4 increases to 37.6% and 5%, respectively.

Comparison of Analysis Results Between 2D/1D Synthesis and RAPTOR-M3G in the Korea Standard Nuclear Plant (KSNP)

The 2D/1D synthesis methodology has been used to calculate the fast neutron (E > 1.0 MeV) exposure to the beltline region of the reactor pressure vessel. This method uses the DORT 3.1 discrete ordinates code and the BUGLE-96 cross-section library based on ENDF/B-VI. RAPTOR-M3G (RApid Parallel Transport Of Radiation-Multiple 3D Geometries) which performs full 3D calculations was developed and is based on domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architecture. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor. Both methods are applied to surveillance test results for the Korea Standard Nuclear Plant (KSNP)-OPR (Optimized Power Reactor) 1000 MW. The objective of this paper is to compare the results of the KSNP surveillance program between 2D/1D synthesis and RAPTOR-M3G. Each operating KSNP has a reactor vessel surveillance program consisting of six surveillance capsules located between the core and the reactor vessel in the downcomer region near the reactor vessel wall. In addition to the In-Vessel surveillance program, an Ex-Vessel Neutron Dosimetry (EVND) program has been implemented. In order to estimate surveillance test results, cycle-specific forward transport calculations were performed by 2D/1D synthesis and by RAPTOR-M3G. The ratio between measured and calculated (M/C) reaction rates will be discussed. The current plan is to install an EVND system in all of the Korea PWRs including the new reactor type, APR (Advanced Power Reactor) 1400 MW. This work will play an important role in establishing a KSNP-specific database of surveillance test results and will employ RAPTOR-M3G for surveillance dosimetry location as well as positions in the KSNP reactor vessel.

Detailed evaluation of natural circulation mass flow rate in the annular gap between the outer reactor vessel wall and insulation under IVR-ERVC

IVR (In-Vessel corium Retention) through ERVC (External Reactor Vessel Cooling) is known to be an effective means for maintaining the reactor vessel integrity during a severe accident in a nuclear power plant. Detailed simulations of a two phase natural circulation flow in the reactor cavity of an APR (Advanced Power Reactor) 1400 and advanced OPR (Optimized Power Reactor) 1000 under the IVR-ERVC have been conducted to determine the coolant circulation mass flow rate at the annular gap between the outer reactor vessel wall and the insulation of the reactor vessel using the RELAP5/MOD3 computer code. This coolant mass flow rate is very effective on the CHF (Critical Heat Flux) on the outer reactor vessel wall, which is a key factor to the success of the IVR-ERVC. The RELAP5 results have shown that the coolant circulation mass flow rate of a high power reactor of the APR1400 is higher than that of a low power reactor of the advanced OPR1000. Increases in the coolant injection temperature and the water level in the reactor cavity lead to an increase in the coolant circulation mass flow rate, which leads to an increase in the CHF on the outer reactor vessel wall.

Laser cutting of thick steel plates and simulated steel components using a 30 kW fiber laser

Laser cutting of thick steel plates and simulated steel components using a 30 kW fiber laser was studied for application to nuclear decommissioning. Successful cutting of carbon steel and stainless steel plates up to 300 mm in thickness was demonstrated, as was that of thick steel components such as simulated reactor vessel walls, a large pipe, and a gate valve. The results indicate that laser cutting applied to nuclear decommissioning is a promising technology.

Evaluation of heat-flux distribution at the inner and outer reactor vessel walls under the in-vessel retention through external reactor vessel cooling condition

Background: A numerical simulation was carried out to investigate the difference between internal and external heat-flux distributions at the reactor vessel wall under in-vessel retention through external reactor vessel cooling (IVR-ERVC). Methods: Total loss of feed water, station blackout, and large break loss of coolant accidents were selected as the severe accident scenarios, and a transient analysis using the elementbirth-and-death technique was conducted to reflect the vessel erosion (vessel wall thickness change) effect. Results: It was found that the maximum heat flux at the focusing region was decreased at least 10% when considering the two-dimensional heat conduction at the reactor vessel wall. Conclusion: The results show that a higher thermal margin for the IVR-ERVC strategy can be achieved in the focusing region. In addition, sensitivity studies revealed that the heat flux and reactor vessel thickness are dominantly affected by the molten corium pool formation according to the accident scenario.

LAROB: Laser-Guided Underwater Mobile Robot for Reactor Vessel Inspection

In a nuclear power plant, there are several cylindrical vessels such as a reactor vessel and pressurizer. The vessels are usually constructed by welding large rolled plates or nozzle pipes together. To ensure the integrity of the vessel, their welds should be periodically inspected using sensors such as ultrasonic transducers or visual cameras. To inspect these welds effectively, we developed an underwater mobile robot, which is guided by a laser pointer. The robotic system was devised to reduce the inspection time and schedule during mandatory code inspections compared to a conventional inspection machine with a large structure. The system mainly consists of an underwater mobile robot, a laser positioning unit, and a main control station. The underwater mobile robot is guided by a laser positioning unit with a precise resolution of 0.05°. The mobile robot moves on the reactor vessel wall with four magnetic wheels. This paper presents the design and implementation of the underwater mobile robot. The laser guidance control of the mobile robot is also described along with the experimental results. The system was integrated with the main control station, and tested in the reactor vessel of a Korean nuclear power plant. After many improvements in its design and engineering, the system is expected to dramatically reduce the critical path of the reactor vessel inspection, if the system is used practically.