Spacer Grid Assembly Research Articles

Thermal-Hydraulic Investigation of Steady-State Single-Phase Natural Circulation of an Integral PWR-Type SMR Test Rig (iPSTR)

The main aim of the natural circulation of the primary coolant in a nuclear reactor is to reject heat from the reactor core to the steam generator without using a circulation pump. In this work, a vertical heater–vertical cooler, high-temperature, high-pressure, nonuniform-diameter, single-phase natural circulation loop is proposed. The rig contains a spacer grid assembly of electrical heaters in a core with a conical section in its upper plenum and a double helical coil steam generator. The proposed loop is analyzed using RELAP5 and various analytical models. First, these models are benchmarked with experimental data from the Facility to Investigate Natural Circulation in SMART or FINCLS. The model results are found to be in good agreement with the experimental data. The same models are then employed to investigate the proposed natural circulation facility, named the Integral PWR-type SMR Test Rig (iPSTR), to investigate the mass flow rate as a function of geometric and process parameters. Core power input was varied from 5 to 82.5 kW at a maximum system pressure of 10 bar and a maximum elevation difference of thermal centers of 3400 mm. Elevation differences of the thermal centers and diameter of core are found to be important parameters that affect thermal-hydraulic performance significantly. However, cone angle, spacer grid, and system pressure are found to have no significant effect on the performance of the iPSTR. Moreover, the proposed iPSTR is found to possess higher Reynolds number compared with the existing facilities.

Development of a spring model to predict the behavior of fuel rods in a spacer grid assembly

This paper presents a newly developed plastic model to represent the nonlinear inelastic behavior of some fuel assembly components, such as leaf springs and dimples, as translational springs. Nuclear fuel rods are bundled into fuel assemblies constrained by spacer grids located at discrete locations, which transfer load or acceleration to the fuel rods through the leaf springs and dimples. In the current state of practice, leaf springs and dimples are modeled as linear, or nonlinear elastic, translational springs, without considering plastic deformations. It is essential to consider the plasticity and deterioration of the spacer grid assembly to better predict the response of the spent nuclear fuel (SNF) rods subjected to the dynamic loads that may take place during storage and transportation. The presented model considers the kinematic strain hardening algorithm to account for the plastic deformation of the spring in compression only, and incorporates some of the peak-oriented hysteretic model rules to account for stiffness deterioration in the plastic region. The material model is written in the C++ language in the open-source finite element (FE) software MASTODON. The material model is validated by applying simple loading protocols to the two-noded spring. The use of the material model in the fuel assembly is demonstrated by using a small segment of the SNF rods. This new model captures the intermediate reversals when the fuel rod is subjected to vibrations. Numerical examples show that the model can accurately predict the plastic deformations of the leaf spring and dimples. The introduction of nonlinear inelastic capabilities in the FuelCompressionSpring resulted in higher deformations of the leaf spring and longer time periods of the displacement response. The developed hysteretic models can efficiently reproduce the elastic and plastic response of leaf springs and dimples, and can be used in large fuel assembly models.

Weld bead formation and corrosion behavior of pulsed laser welded zirconium alloy

Pulsed laser spot welding was developed for the fabrication of Zircaloy spacer grid assembly in pressurized light water reactors. The effects of laser welding parameters on the size of the weld beads were discussed. The influence of oxygen and hydrogen uptakes during the welding process on the corrosion resistance was investigated after the corrosion tests. The results indicate that the optimum weld bead size could be obtained by adjusting the pulsed laser parameters, including laser peak power, number of shots, and pulse width. Pulse width should be taken into consideration to control the weld width. When the moisture and oxygen contents were maintained at a low level, i.e., 200 parts per million (ppm) and 30 ppm, respectively, the oxide thickness of specimens corroded for 120 h was about 0.7 μm compared to about 1.5 μm at high moisture and oxygen contents (1,000 and 280 ppm). The combination of the oxidation and hydriding induced cracking was responsible for the crack of the specimens welded under high moisture and oxygen contents, thereby decreasing the corrosion resistance of the welding specimens.

Pulsed laser spot welding of intersection points for Zircaloy-4 spacer grid assembly

A pulsed laser spot welding technology has been developed for joining intersection points of 0.457mm-thick Zircaloy-4 straps. Weld beads size, mechanical properties and microstructure characterization of the weld beads were investigated. The results indicate that peak power of the pulsed laser has significant influence on the penetration depth of weld beads. Pulse width and number of shots should be taken into consideration mainly to control the weld width. The average value of ultimate tensile loads of the intersection points are continuously increasing as the penetration depth and weld width of weld beads increase. Fracture was located at the fusion line and the fracture surface could be characterized as a mixture of ductile and cleavage feature. Column grains and equiaxed grains were observed in fusion zone and heat affected zone, respectively. The fusion zone consists of a mixture of α-Zr and β-Zr phases among which lamellar Widmanstatten structure α-Zr+Zr3Fe was distributed. The Zr(Fe, Cr)2 second phase particles were precipitated inter and intragranular of α-Zr grains.

CFD analysis on a three by three nuclear fuel bundle with five spacer grid assemblies

This study has been conducted to investigate the flow characteristics for SMART 3×3 rod bundle with various turbulence models, namely, κ-ɛ model, κ-ω shear stress transport (SST) model and Baseline (BSL) Reynolds stress model. Commercial code CFX 12.0 was used for the simulation, and pressure drop, second flow intensity, flow velocity and temperature fields were predicted. For the development of an adequate computational model to simulate such a complex geometry, a grid sensitivity study was first examined. In addition, a pressure drop test for SMART rod bundle was shortly introduced, which test was done at Korea Atomic Energy Research Institute and compared with simulation results. From the comparison, with the SST model and BSL model similar pressure drop values were obtained. Maximum surface temperatures in the cross-section increase with the increase in the steps of the span (spacer grid assembly) and the temperature immediately decreases just after the mixing vane and then the temperature slowly rises. Compared with experimental results, the SST model had a similar maximum temperature pattern along rod axis. Also, it was interesting to find that the turbulence model did not affect the secondary flow intensity while affecting the flow velocity values. As a result, the most appropriate turbulence model and the grids were able to be determined.

이동 가능한 연료봉 지지부의 특성 고찰

A spacer grid assembly is one of the most important structural components of the nuclear fuel assembly of a pressurized water reactor (PWR), and it affects the performance of the fuel assembly. The primary design requirement is that the mechanical integrity of the fuel rod should be maintained by the spacer grid assembly during the operation of the reactor. It was known that fretting damage to the fuel rod can be reduced by adjusting the relative moving displacement between the fuel rod and its support. In this study, we used the finite element method to evaluate the characteristics of a sliding support designed to reduce fretting damage of fuel rods.

경수로 핵연료집합체 지지격자체의 횡방향 충격강도 연구

지지격자체는 경수로 핵연료집합체의 가장 중요한 핵심 구조부품 중에 하나이다. 질칼로이 지지격자체 설계시의 우선적으로 고려해야 할 사항은 지지격자체가 원자로심에서 냉각수의 심한 수두손실을 유발하지 않으면서 지진사고를 고려한 설계하중 하에서 충분한 횡방향 충격강도를 갖도록 하는 것이다. 본 연구에서는 시험과 유한요소해석을 통해 지지격자체의 횡방향 충격강도에 영향을 주는 인자들에 대한 분석을 수행하였고, 지지격자체 제조용 질칼로이 원자재 소요량을 획기적으로 줄이면서 지지격자체의 횡방향 충격강도를 개선할 수 있는 방안을 제시하였다.

삽입 및 이동 가능한 연료봉 지지부의 지지격자 형상

지지격자체는 경수로 핵연료집합체의 가장 중요한 핵심 구조부품이다. 지지격자체 설계시의 고려사항은 원자로 운전중에 연료봉의 지지건전성을 유지하도록 하는 것이다. 본 연구에서는 연료봉이 유동기인진동에 의해서 진동할 때 연료봉과 연료봉 지지부 사이에서 상대변위를 완화해 줌으로서 연료봉의 프레팅 마모손상 가능성을 감소시킬 수 있는 이동 가능한 연료봉 지지부로 구성된 새로운 지지격자체 형상을 제안하였다. 아울러 제안된 이동 가능 지지부의 연료봉 지지특성을 유한요소해석을 통해 분석하였다.

Laser Welding Unit for Intersection Line Welding of Spacer Grid Inner Straps and its Application

A spacer grid assembly is one of the most important structural components in a Pressurized Wa-ter Reactor fuel assembly. The spacer grid assembly, which supports nuclear fuel rods laterally and vertically with a friction grip, is an interconnected array of slotted grid straps welded at the intersec-tions to form an egg-crate structure. From a structural point of view, the spacer grid assembly is re-quired to have enough crush strength under lateral loads so that the nuclear fuel rods are maintained in a coolable geometry, and that control rods are able to be inserted. The capacity of a spacer grid assembly to resist lateral loads is usually characterized in terms of its crush strength. Zircaloy is prevailing as the material of a spacer grid because of its low neutron absorption characteristic and its extensive successful in-reactor use. One of the primary considerations is to provide a Zircaloy spacer grid with crush strength sufficient enough to resist design basis loads, without significantly increasing the pressure drop across the reactor core. Based on the fact that the crush strength of a spacer grid assembly is known to be proportional to a weld penetration at the intersections of the slotted grid straps, a laser welding unit which is able to weld along the intersection line of spacer grid inner straps was proposed, and spacer grid specimens were fabricated using the proposed laser welding unit in this study. Also, a crush strength test and analysis of the spacer grid specimens were carried out to confirm the effect of the resultant crush strength enhancement.

Impact of Nuclear Fuel Assembly Design on Grid-to-Rod Fretting Wear

The fluid-induced vibration at the nuclear fuel assembly may cause grid-to-rod fretting wear, and subsequently, fuel rod perforation. The fluid-induced grid-to-rod fretting wear occurs with a certain fuel assembly design and/or in certain nuclear power plants, which may be strongly correlated with external and internal vibration causes acting on the fuel assembly. The external vibration causes may include reactor coolant flow velocity, nonuniform radial flow profiles caused by reactor internals, and interfuel assembly gap and fuel assembly-shroud gap, whereas the internal vibration causes may include asymmetric mixing vane pattern across the spacer grid assembly and inadequate spacer grid design. In this study, the external and internal vibration causes are described including three internal vibration mechanisms acting on the fuel assembly. The impact of internal vibration on the grid-to-rod fretting wear was investigated, based on out-of-pile vibration and grid-to-rod fretting wear test results for two kinds of fuel assembly designs as well as their operating experiences in commercial reactors. In addition, fuel assembly design optimization guidelines are proposed to eliminate the grid-to-rod fretting wear-induced failure.

Impact of Nuclear Fuel Assembly Design on Grid-to-Rod Fretting Wear

The fluid-induced vibration at the nuclear fuel assembly may cause grid-to-rod fretting wear, and subsequently, fuel rod perforation. The fluid-induced grid-to-rod fretting wear occurs with a certain fuel assembly design and/or in certain nuclear power plants, which may be strongly correlated with external and internal vibration causes acting on the fuel assembly. The external vibration causes may include reactor coolant flow velocity, nonuniform radial flow profiles caused by reactor internals, and interfuel assembly gap and fuel assembly-shroud gap, whereas the internal vibration causes may include asymmetric mixing vane pattern across the spacer grid assembly and inadequate spacer grid design. In this study, the external and internal vibration causes are described including three internal vibration mechanisms acting on the fuel assembly. The impact of internal vibration on the grid-to-rod fretting wear was investigated, based on out-of-pile vibration and grid-to-rod fretting wear test results for two kinds of fuel assembly designs as well as their operating experiences in commercial reactors. In addition, fuel assembly design optimization guidelines are proposed to eliminate the grid-to-rod fretting wear-induced failure.

Parameter study for a dimple location in a space grid under the critical impact load

The spacer grid assembly, an interconnected array of slotted grid straps embossed with dimples and springs, is one of the main structural components of a pressurized light-water reactor (PWR). It takes the role of supporting the nuclear fuel rods which experience a severe expansion and contraction caused by harsh operational conditions such as an earthquake. The external load by an earthquake can be mainly represented as a lateral load, and the resistance to it is evaluated in terms of dynamic crush strengths. It has been reported that a dimple location in a space grid has an effect on this strength. In this paper, based on this fact, the effect of a dimple location in a 3×3 support grid on impact strength has been investigated as a preliminary parameter study for a full sized support grid. The optimal location of the dimple, about 3.5 mm from the tip of the strap, has been found and some design guidelines for a support grid such as reducing the spring length and the dimple gap have been provided.

Determination of Optimum LASER Welding Parameters and the Development of a LASER Welding Technique to Enhance the Lateral Impact Strength of a Spacer Grid Assembly for a PWR Fuel Assembly

A spacer grid assembly, which is an interconnected array of slotted grid straps welded at the intersections to form an egg crate structure, is one of the core structural components of the nuclear fuel assemblies of a Pressurized light Water Reactor (PWR). The spacer grid assembly is structurally required to have enough buckling strength under lateral loads so that the nuclear fuel rods are maintained in a coolable geometry, and that control rods are able to be inserted. The ability of a spacer grid assembly to resist lateral loads is usually characterized in terms of its dynamic and static crush strengths. Since the crush strengths of a spacer grid assembly are known to depend on the weld qualities at the intersections of the slotted grid straps, high-tech welding methods, such as a TIG welding, LASER beam welding or Electron beam welding method, have been used recently in the nuclear fuel manufacturing fields. In this study, to meet the above requirements, two kinds of researches were carried out. First, by adjusting the LASER beam welding parameters, an optimum welding combination of the LASER beam welding parameters was obtained for welding a spacer grid assembly. Second, a new LASER beam welding technique was proposed to obtain a longer weld line and a smaller weld bead size by tilting the LASER beam. The buckling strength of the spacer grid welded by the new LASER beam welding technique was enhanced by up to 30 % when compared to that by the conventional LASER beam welding method.

Development of a LASER Welding Apparatus and a Method for an Inner-Strap Welding of a Spacer Grid Assembly for a PWR Fuel Assembly

A spacer grid assembly, which is an interconnected array of slotted grid straps welded at the intersections to form an egg crate structure, is one of the core structural components for the nuclear fuel assemblies of a Pressurized light Water Reactor (PWR). The commercial spacer grid assembly is spot-welded at the crossing points of the intersections by a TIG welding, LASER beam welding or Electron beam welding method. In this study, a LASER beam welding apparatus and a method for an inner strap welding has been proposed to obtain a longer and finer weld line and a smaller weld bead size for a spacer grid assembly for a PWR fuel assembly. Also a rapid welding and excellent weld quality have been achieved by the proposed welding apparatus and method.

IMPACT ANALYSIS AND TEST FOR THE SPACER GRID ASSEMBLY OF A NUCLEAR FUEL ASSEMBLY

A spacer grid assembly is one of the main structural components of the nuclear fuel assembly for a Pressurized light Water Reactor (PWR). The spacer grid assembly supports and aligns the fuel rods, guides the fuel assemblies past each other during a handling and, if needed, sustains lateral seismic loads. The ability of a spacer grid assembly to resist these lateral loads is usually characterized in terms of its dynamic and static crush strengths, which are acquired from tests. In this study, a finite element analysis on the dynamic crush strength of spacer grid assembly specimens is carried out. Comparisons show that the analysis results are in good agreement with the test results to within about a 30 % difference range. Therefore, we could predict the crush strength of a spacer grid assembly in advance, before performing a dynamic crush test. And also a parametric study on the crush strength of a spacer grid assembly is carried out by adjusting the weld penetration depth for a sub-sized spacer grid, which also shows a good agreement between the test and analysis results.

Finite element analysis of the optimized H type grid spring by using gap elements

The optimized H type grid spring of a new design for the Zirconium alloy spacer grids for an advanced LWR fuel assembly is developed. Characterization tests are conducted for this spring and the load vs. displacement curve of the spring characteristics is obtained. Also, four FEM models, which consisted of two 5 × 5 cell models, one 6 × 9 cell model and one 8 × 9 cell model, were developed by the IDEAS code and they were analyzed by the ABAQUS code in order to simulate the newly designed spring characteristics. The results of the FE analysis are compared with those of the spring characterization test and the following conclusions were obtained. First, since a model which includes a lot of cells below the location of the applied load has many gaps due to more slits, the spring characteristics of the model are more flexible. Second, among the four FEM models, the result of the 5 × 5 cell model applied to the displacement load at the top of the model was in good agreement with the test result. Therefore, when a newly designed spring for a spacer grid assembly is developed, its spring characteristics can be estimated by an analysis with an appropriate FEM model without the spring characterization test.

Impact Analysis for the Egg-Crate Structure of a Nuclear Fuel Assembly by Using FEM

A spacer grid assembly is one of the main structural components of the nuclear fuel assembly of a Pressurized light Water Reactor (PWR). The spacer grid assembly supports and aligns the fuel rods, guides the fuel assemblies past each other during a handling and, if needed, sustains lateral seismic loads. The ability of a spacer grid assembly to resist these lateral loads is usually characterized in terms of its dynamic and static crush strengths, which are acquired from tests. In this study, a finite element analysis on the dynamic crush strength of spacer grid assembly specimens is carried out. Comparisons show that the analysis results are in good agreement with the test results within an 8 % difference range. Therefore, we could predict the crush strength of a spacer grid assembly in advance, before performing the dynamic buckling test. And also a parametric study on the crush strength of a spacer grid assembly is carried out by adjusting the weld penetration depth for a sub-sized spacer grid, which also shows a good agreement between the test and analysis results.

Determination of the Optimum Welding Parameters for a Laser Welded Spacer Grid Assembly for PWRs

A spacer grid assembly, which is an interconnected array of slotted grid straps and welded at the intersections to form an egg crate structure, is one of the core structural components of the nuclear fuel assemblies of a Pressurized light Water Reactor (PWR). The spacer grid assembly is structur-ally required to have enough buckling strength under lateral loads due to lateral seismic accelera-tions, lateral Loss Of Coolant Accident (LOCA) blowdown forces, and shipping and handling loads so that the nuclear fuel rods are maintained in a coolable geometry, and that control rods are able to be inserted. The ability of a spacer grid assembly to resist lateral loads is usually characterized in terms of its dynamic and static crush strengths, which are acquired from tests. The crush strengths and spacer grid dynamic stiffness of a spacer grid assembly are required for the fuel assembly seis-mic and LOCA blowdown analyses to verify that a coolable grid geometry is maintained. Since the crush strengths of the spacer grid assembly are known to depend on the weld qualities at the inter-sections of the slotted grid straps, high-tech welding methods, such as the TIG welding, LASER beam welding or Electron beam welding method, have been used recently in the nuclear fuel manu-facturing fields.

Impact Analysis and Test of the Spacer Grid Assembly for PWRs

The spacer grid assembly, which is an interconnected array of slotted grid straps and welded at the intersections to form an egg crate structure, is one of the main structural components of the nuclear fuel assembly of a Pressurized light Water Reactor (PWR). The spacer grid assembly supports and aligns the fuel rods, guides the fuel assemblies past each other during handling and, if needed, sustains lateral seismic loads. The ability of the spacer grid assembly to resist the lateral loads is usually characterized in terms of its dynamic and static crush strengths, which are acquired from tests. In this study, dynamic buckling tests and finite element analyses on spacer grid assembly specimens are carried out. Comparisons show that the analysis results are in good agreement with the test results to within an 8 % difference range. Therefore, we could predict the crush strength of a spacer grid assembly in advance before performing the dynamic buckling test.

Finite Element Analysis Model for Estimating the Dynamic Crush Strength of the Spacer Grid Assembly for PWRs

The spacer grid assembly, which is an interconnected array of slotted grid straps and welded at the intersections to form an egg crate structure, is one of the main structural components of the nuclear fuel assemblies of a Pressurized light Water Reactor (PWR). The spacer grid assembly is structurally required to have enough crush strength under lateral loads due to lateral seismic accelerations, lateral Loss Of Coolant Accident (LOCA) blowdown forces, and shipping and handling loads so that the fuel rods are maintained within a coolable geometry, and that the control rods are able to be inserted. The ability of the spacer grid assembly to resist the lateral loads is usually characterized in terms of its dynamic and static crush strengths, which are acquired from the relevant tests. In this study, dynamic buckling tests and finite element analyses on spacer grid assembly specimens are carried out. As a result of the comparisons, the analysis results are in good agreement with the test results to within an 8 % difference range. Therefore, we could predict the crush strength of a spacer grid assembly in advance before performing the dynamic buckling test.