• Transactions of the Korean Society for Noise and Vibration Engineering
• /
• v.23 no.4
• /
• pp.308-314
• /
• 2013

The vibration and stress analysis program of comprehensive vibration assessment program(CVAP) is to theoretically verify the structural integrity of reactor vessel internals(RVI) and to provide the basis for selecting the locations monitored in measurement and inspection programs. This paper covers the applicability of the vibration and stress analysis method of APR1400 RVI CVAP. The analysis method was developed to use 3-dimensional detail hydraulic and structural models with ANSYS and CFX. To assess the method, the hydraulic loads and structural reponses of OPR1000 were predicted and compared with the measured data in the OPR1000 RVI CVAP. The results predicted with this method were close to the measured values considerably. Therefore, the analysis method was developed properly.

• Nuclear Engineering and Technology
• /
• v.55 no.1
• /
• pp.80-99
• /
• 2023

This paper presents the results of investigating the change in welding residual stresses of the core shroud, which is one of subcomponents in reactor vessel internals, performing finite element analysis. First, the welding residual stresses of the core shroud were calculated by applying the heat conduction based lumped pass technique and finite element elastic-plastic stress analysis. Second, the temperature distribution of the core shroud during the normal operation was calculated by performing finite element temperature analysis considering gamma heating. Third, through the finite element viscoelastic-plastic stress analysis using the calculated temperature distribution and setting the calculated residual stresses as the initial stress state, the variation of the welding residual stresses was derived according to repeating the normal operation. In the viscoelastic-plastic stress analysis, the effects of neutron irradiation on mechanical properties during the cyclic normal operations were considered by using the previously developed user subroutines for the irradiation agings such as irradiation hardening/embrittlement, irradiation-induced creep, and void swelling. Finally, the effect of neutron irradiation on the welding residual stresses was analysed for each irradiation aging. As a result, it is found that as the normal operation is repeated, the welding residual stresses decrease and show insignificant magnitudes after the 10^th refueling cycle. In addition, the irradiation-induced creep/void swelling has significant mitigation effect on the residual stresses whereas the irradiation hardening/embrittlement has no effect on those.

• Nuclear Engineering and Technology
• /
• v.43 no.1
• /
• pp.89-98
• /
• 2011

The construction technology for reactor vessel internals (RVI) modularization is one of the most important factors to be considered in reducing the construction period of nuclear power plants. For RVI modularization, gaps between the reactor vessel (RV) core-stabilizing lug and the core support barrel (CSB) snubber lug must be measured using a remote method from outside the RV. In order to measure RVI gaps remotely at nuclear power plant construction sites, certain core technologies must be developed and verified. These include a remote measurement system to measure the gaps between the RV core-stabilizing lug and the CSB snubber lug, an RVI mockup to perform the gap measurement tests, and a new procedure and schedule for RVI installation. A remote measurement system was developed previously, and a gap measurement test was completed successfully using the RVI mockup. We also developed a new procedure and schedule for RVI installation. This paper presents the new and improved installation procedure and schedule for RVI modularization. These are expected to become core technologies that will allow us to shorten the construction period by a minimum of two months compared to the existing installation procedure and schedule.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2012.10a
• /
• pp.306-311
• /
• 2012

A reactor vessel internals comprehensive vibration assessment program (RVI CVAP) of an advanced power reactor 1400 (APR1400) is being performed as a non-prototype category-2 type of reactor based on the US Nuclear Regulatory Commission Regulatory Guide (NRC RG) 1.20. The aim of this paper is to present the results of structural response analysis and measuring locations of a upper guide structure (UGS) assembly of the APR1400 reactor. The analysis results of the UGS assembly results show that meet the specified integrity levels of the design acceptance criteria. Also, the measuring locations are set by the analysis results of the UGS assembly and selection criteria of measuring locations prior to this study. These analysis results and measuring locations will be used as fundamental materials to design a measurement system for the APR1400 RVI CVAP.

• Computational Structural Engineering
• /
• v.7 no.3
• /
• pp.95-103
• /
• 1994

The safety related components in the nuclear power plant should be designed to withstand the seismic load. Among these components the integrity of reactor internals under earthquake load is important in stand points of safety and economics, because these are classified to Seismic Class I components. So far the modelling methods of reactor internals have been investigated by many authors. In this paper, the dynamic behaviour of reactor internals of Yong Gwang 1&2 nuclear power plants under SSE(Safe Shutdown Earthquake) load is analyzed by using of the simpled Global Beam Model. For this, as a first step, the characteristic analysis of reactor internal components are performed by using of the finite element code ANSYS. And the Global Beam Model for reactor internals which includes beam elements, nonlinear impact springs which have gaps in upper and lower positions, and hydrodynamical couplings which simulate the fluid-filled cylinders of reactor vessel and core barrel structures is established. And for the exciting external force the response spectrum which is applied to reactor support is converted to the time history input. With this excitation and the model the dynamic behaviour of reactor internals is obtained. As the results, the structural integrity of reactor internal components under seismic excitation is verified and the input for the detailed duel assembly series model could be obtained. And the simplicity and effectiveness of Global Beam Model and the economics of the explicit Runge-Kutta-Gills algorithm in impact problem of high frequency interface components are confirmed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.14 no.2
• /
• pp.91-100
• /
• 2016

The neutron activation inventories in reactor vessel and its internals, and bio-shield of a PWR nuclear power plant were calculated to evaluate the effect of impurity elements contained in the structural materials on the activation inventory. Carbon steel is, in this work, used as the reactor vessel material, stainless steel as the reactor vessel internals, and ordinary concrete as the bio-shield. For stainless steel and carbon steel, one kind of impurity concentration was employed, and for ordinary concrete five kinds were employed in this study using MCNP5 and FISPACT for the calculation of neutron flux and activation inventory, respectively. As the results, specific activities for the cases with impurity elements were calculated to be more than twice than those for the cases without impurity elements in stainless and carbon steel. Especially, the specific activity for the concrete material with impurity elements was calculated to be 30 times higher than that without impurity. Neutron induced reactions and activation inventories in each material were also investigated, and it is noted that major radioactive nuclide in steel material is Co-60 from cobalt impurity element, and, in concrete material, Co-60 and Eu-152 from cobalt and europium impurity elements, respectively. The results of this study can be used for nuclear decommissioning plan during activation inventory assessment and regulation, and it is expected to be used as a reference in the design phase of nuclear power plant, considering the decommissioning of nuclear power plants or nuclear facilities.

• Nuclear Engineering and Technology
• /
• v.44 no.5
• /
• pp.501-506
• /
• 2012

Many innovative design features are employed in the reactor vessel internals of SMART, a small integral-type pressurized water reactor, one of which is the flow skirt, which uniformly distributes flow and horizontally restrains the lower part of the core support barrel. This new design requires a comprehensive investigation of vibration characteristics. Therefore, in this study, modal characteristics of flow skirts are investigated with finite element analysis. Specifically, we investigate how the presence of holes, the presence of three rings attached to the flow skirt, and the thickness of the lowest shell effect vibration characteristics. In addition, the fluid effect is addressed, since the flow skirt is submerged in the fluid.

• Journal of Energy Engineering
• /
• v.24 no.2
• /
• pp.154-166
• /
• 2015

The in-core instrumentation measures core power distribution and coolant temperature in local regions of the core in pressurized water reactors. The installation types are distinguished by the designs of routing paths that exit either through reactor bottom mounted instrument nozzles or through reactor top mounted instrument nozzles. Although each type has unique advantages, it is generally known that top mounted design is more competitive with respect to emphasizing nuclear safety issues and ability to cope with severe accidents. The international nuclear vendors have provided various types of reactors with top mounted design. Nuclear power reactors in Korea, however, only have been designed to be applicable to the use of bottom mounted design, and it has been pointed out that the capabilities of Korean reactors against severe accidents should be further enhanced. The paper deals with technical issues on reactor internal and external design, in-core instrumentation, support assembly, sealing mechanism with nozzles, handling, and analytical issues in order to establish the ways of development.

• Nuclear Engineering and Technology
• /
• v.44 no.7
• /
• pp.817-824
• /
• 2012

The most important component of a nuclear power plant is its nuclear reactor. Studies on the integrity of reactors have become an important part regarding the safety of a nuclear power plant. The US Nuclear Regulatory Commission Regulatory Guide (NRC RG) 1.20 presents a Comprehensive Vibration Assessment Program (CVAP) to be used to verify the structural integrity of the Reactor Vessel Internals (RVI) for flow-induced vibration prior to commercial operation. However, there are few published studies related to the RVI CVAP. We classified the Advanced Power Reactor 1400 (APR1400) RVI CVAP as a non-prototype category-2 reactor as part of an independent validation of its design. The aim of this paper is to present the results of structural response analyses of the Upper Guide Structure (UGS) assembly of the APR1400 reactor. These results show that the UGS and the Inner Barrel Assembly (IBA) meet the specified integrity levels of the design acceptance criteria. The vibration and stress analysis results in this paper will be used as basic information to select measurement locations of the vibration and stress for the APR1400 RVI CVAP.

• Nuclear Engineering and Technology
• /
• v.27 no.5
• /
• pp.710-720
• /
• 1995

The propagation of pump-induced pressure pulsation in a reactor is important because of the potential for vibration and resultant damage of reactor internals. A hydrodynamic model has been developed to obtain the pressure fluctuation due to the operation of pumps in the annulus(between the core support barrel and reactor vessel of a pressurized water reactor) including the coolant inlet pipe. The mathematical analysis is formulated in accordance with the linearized Navier-Stokes equation by assuming a compressible, inviscid flow. Two regions are considered separately and by coupling the solutions of the inlet pipe and the annulus, the inlet nozzle pressure(pressure at pipe and annulus interface) is to be calculated without assumptions. The geometric parameter effect on the pump-induced pressure pulsation is evaluated. Comparison of predicted and measured inlet nozzle pressure values for each forcing frequency shows good order of magnitude agreement.