Thermal analysis of a storage cask for 24 spent PWR fuel assemblies

Abstract

The purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal, off-normal and accident conditions. The environmental temperature is assumed to be 27°C under the normal condition. The off-normal condition has an environmental temperature of 40°C. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The accident condition is defined as a 100% blockage of air inlet ducts. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of the ventilation system have been carried out for the determination of the optimum duct size and shape. The finite-volume computational fluid dynamics code FLUENT was used for the thermal analysis. From the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal, off-normal and accident conditions.     

On the factors affecting the fretting-wear risk of PWR fuel assemblies

Key factors affecting the rod-to-grid fretting-wear risk of fuel assemblies operated in pressurized water reactors (PWR) are evaluated. The analysis is part of a comprehensive approach to predict fretting-wear risk based on the fuel assembly operating conditions. The assembly wear damage is determined by a non-linear vibration model of the nuclear fuel rod exposed to a turbulent flow. The study evaluates the sensitivity of the wear damage to the grid support forces, fuel rod-to-grid gap size, assembly grids misalignment, rod structural damping and stiffness, assembly bow shape, friction coefficients and turbulence force spectrum. The results of the numerical simulations show that the grid cell clearance and the turbulence forces are key factors in the wear process. Since a good correlation exists between these two parameters and the assembly location in the core, it is recommended to include consideration of the wear risk minimization as an additional criterion for the design of the core loading pattern.     

Assessment of the interchannel mixing model with a subchannel analysis code for BWR and PWR conditions

The influence of the interchannel mixing model employed in a traditional subchannel analysis code was investigated in this study, specifically on the analysis of the enthalpy distribution and critical heat flux (CHF) in rod bundles in BWR and PWR conditions. The equal-volume-exchange turbulent mixing and void drift model (EVVD) was embodied to the COBRA-IV-I code. An optimized model of the void drift coefficient has been devised in this study as the result of the assessment with the two-phase flow distribution data for the general electric (GE) 9-rod and Ispra 16-rod test bundles. The influence of the subchannel analysis model on the analysis of CHF was examined by evaluating the CHF test data in rod bundles representing PWR and BWR conditions. The CHFR margins of typical light water nuclear reactor (LWR) cores were evaluated by considering the influence on the local parameter CHF correlation and the hot channel analysis result. It appeared that the interchannel mixing model has an important effect upon the analysis of CHFR margin for BWR conditions.     

CFD methodology and validation for single-phase flow in PWR fuel assemblies

This paper presents the CFD modeling methodology and validation for steady-state, normal operation in a PWR fuel assembly. This work is part of a program that is developing a CFD methodology for modeling and predicting single-phase and two-phase flow conditions downstream of structural grids that have mixing devices. The purpose of the mixing devices (mixing vanes in this case) is to increase turbulence and improve heat transfer characteristics of the fuel assembly. The detailed CFD modeling methodology for single-phase flow conditions in PWR fuel assemblies was developed using the STAR-CD CFD code. This methodology includes the details of the computational mesh, the turbulence model used, and the boundary conditions applied to the model. The methodology was developed by benchmarking CFD results versus small-scale experiments. The experiments use PIV to measure the lateral flow field downstream of the grid, and thermal testing to determine the heat transfer characteristics of the rods downstream of the grid. The CFD results and experimental data presented in the paper provide validation of the single-phase flow modeling methodology. Two-phase flow CFD models are being developed to investigate two-phase conditions in PWR fuel assemblies, and these can be presented at a future CFD Workshop.     

A non linear model for the PWR fuel assembly seismic analysis

In seismic PWR core analysis, the non linear behavior of fuel assemblies has to be studied taking into account variations of stiffness and damping due to the slippage of rods in spacer grids. Based on the linear CEA’s assembly model, principally composed of two beams representing sets of rods and thimbles, a linear model is described. In the second part of this paper the slippage and loss of contact of rods on a grid relic is analyzed. Based on the Coulomb friction model and plasticity analogy, a grid model constituted of elastoplastic hinge and non linear springs is proposed to simulate the progressive slippage of rods inside the cells. In the third part, a non linear assembly model is designed and validated with experimental data. This model represented with success the increase of damping and decay of frequency when the magnitude of the dynamic loading increases.     

200MW低温供热堆盒间流道流动传热数值分析

采用三维CFD软件Phoenics-3.2,计算了200MW低温供热堆燃料组件盒间的流场及温场。研究了旁通流量、控制棒提升等因素的影响。在考虑这些因素之后,得出了最佳旁通入流方案。     

A subchannel analysis of DUPIC fuel bundle for the CANDU reactor

Thermal characteristics of the reference DUPIC fuel has been studied for its feasibility of loading in the CANDU reactor. Half of the DUPIC fuel bundle has been modeled for a subchannel analysis of the ASSERT-IV Code which was developed by AECL. From the calculated mixture enthalpy, equilibrium quality and void fraction distributions in subchannels of the fuel bundle, it is found that the gravity effect may be pronounced in the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. The asymmetric distribution of the coolant in the fuel bundle is known to be undesirable since the minimum critical heat flux ratio can be reduced for a given value of the channel flow rate. On the other hand, the central region of the DUPIC fuel bundle has been found to be cooled more efficiently than that of the standard fuel bundle in the subcooled and the local boiling regimes due to the fuel geometry and the fuel element power changes. Based upon the subchannel modeling used in this study, the location of minimum critical heat flux ratio in the DUPIC fuel bundle turned out to be very similar to that of the standard fuel when the equivalent values of channel power and channel flow rate are used. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the subchannel-wise mixture enthalpy and void fraction peaks are located in the peripheral region of the DUPIC fuel bundle while those are located in the central region of the standard CANDU fuel bundle. Reduced values of the channel flow rates were used to study the effect of channel flow rate variation. The effect of the channel flow reduction on different thermal-hydraulic parameters have been discussed. This study shows that the subchannel analysis for the horizontal flow is very informative in developing new fuel for the CANDU reactor.     

压水堆核电站燃料传输系统抗震性能有限元分析

压水堆核电站燃料传输系统的主要功能是将核燃料组件在反应堆厂房和燃料厂房之间进行相互传送.为保障其结构在不同工况下能够安全运行,需对其在异常工况(OBE)和事故工况(SSE)下进行抗震性能分析.为此,采用有限元分析软件ANSYS 12.0对某百万千瓦级核电站的燃料传输系统进行了结构建模;对传输系统在运行过程中不同构型下的转运通道、承载器、传输小车、反应室倾翻架、燃料室倾翻架、燃料室轨道结构和反应室轨道结构等主要构件在不同工况下进行了应力分析;采用SRSS方法进行了应力组合,并依据RCCM规范对各构件进行不同工况下的强度评定.结果表明,在异常和事故工况下,燃料传输系统在不同构型下各结构均满足强度要求,具有良好的抗震性能.     

Seismic analysis of PWR-RCC fuel assemblies

The present paper investigates the dynamic behaviour of PWR-RCC fuel assemblies under seismic excitation. A simple vibrational model of the fuel assembly is proposed, which leads to natural frequencies whose spacing agree with experimental data. Available experimental results are reviewed. Impact characteristics of Zircaloy spacer grids are also discussed. It is proposed that their soundness criteria be expressed in terms of impact energy rather than in terms of impact force. The computer code CLASH is briefly described; it is utilized to perform a sensitivity analysis. An example of application is also given.     

Conceptual design of a small nuclear reactor for large-diameter NTD-Si using short PWR fuel assemblies

A design of a small nuclear reactor for a large-diameter NTD-Si using a conventional Pressurized Water Reactors (PWR) full-length assembly was proposed in previous works. The height of the full-length assembly was 400 cm, and the overall size of the reactor and reflector around the core became large. In addition, the irradiation channel became very long, making handling of the Si ingots in the channel more difficult. The use of a short PWR fuel assembly, with a height of 100 cm, was considered in the current work. With the shorter assembly, the design of the reactor became compact and more practical. Gd₂O₃ and control rods were used to suppress excess reactivity. Criticality, neutron transport, and core burn-up calculations were performed using the MVP/GMVP II code and MVP-BURN code. Steady-state single-channel thermal hydraulic analyses were also performed. The calculation results showed that the reactor could be critical over 1200 days, and that heat removal from core was possible under 1 atm operating pressure. Large-diameter ingot up to 20 cm in height could be doped with sufficient uniformity. The reactor semiconductor production rate was estimated, and varied between 48 tons/year and 70 tons/year for the 50 Ω cm target resistivity depending on the position of the control rod.     

Thermal-hydraulics analysis during fuel element failure in an operating PWR

The fuel element failure in an operating pressurized water reactor (PWR), including fuel element breaks, has an effect on the operation safety of PWR. In this paper, the RELAP5 model of the fuel element failure is established for the safety analysis. The RELAP5 time step sensitivity analyses for the element pre-break steady and post-break transient simulation are carried out. And the variations of main thermal-hydraulics parameters related to the fuel element break are quantitatively studied, which include the internal gap pressure and the maximum fuel pellet temperature as well as the releasement of noncondensables in the gap. It is found that (1) the results by the RELAP5 code is very sensitive to the time step in a volume system with the noncondensables, and the time step sensitivity analysis is necessary if the effective time step range is unknown, (2) the larger the break area is, the more quickly the gap pressure increases and the maximum pellet temperature reaches to the stable value, (3) when the gap pressure increases and reaches to the coolant pressure, at the break the liquid inflow from coolant to gap will be turned to the vapor outflow from gap to coolant, (4) during the failure transient, the gap thermal conductivity experiences a sharp decrease in the break instant, which results in the decrease of heat transferred to cladding and the sharp decrease of cladding temperature as well as the sharp increase of minimum departure from nucleate boiling ratio (MDNBR). These conclusions can provide the basic for the operation safety analysis of PWR during the fuel element failure.     

Single-phase subchannel mixing in a simulated nuclear fuel assembly

Single-phase subchannel mixing data were obtained from a 25 rod square array by measuring precisely subchannel exit temperatures over a range of test conditions. A least-squares type statistic operating on exit enthalpy differences was developed to ascertain, in conjunction with the COBRA-II subchannel computer analysis, an optimum value for the coefficient of Rowe and Angles's single-phase mixing correlation β. When the same analytical procedures were applied to data taken under conditions of subcooled nucleate boiling, an approximately linear correlation of β with average exit quality was found. The values obtained were for conditions of natural, or unaided mixing, and for design purposes should be considered as a dependable lower limit. Commercial rod-spacer designs intended to increase mixing will, of course, show substantially higher values of β.     

Coolant mixing and subchannel velocities in an LMFBR fuel assembly

Experimental mixing studies were conducted using water with a 91-pin wire-wrapped fuel assembly to establish a data base for calibrating thermal-hydraulic codes and to supplement and clarify existing mixing data. An electrolytic tracer was employed in conjunction with an isokinetic sampling technique which permitted mixed mean subchannel concentrations and flowrates to be measured without incurring errors resulting from local subchannel gradients. Emphasis was placed upon the behavior of peripheral subchannels.

A complete set of mixing data was obtained which can be used to calibrate thermal-hydraulic codes. Isokinetic velocity measurements revealed that the pitch-average corner, edge, and central subchannel velocities were nearly equal to the bundle average velocity. It was shown that the average peripheral swirl velocity was approximately 1.2 times greater than predicted assuming the fluid follows the wire-wraps. Also the magnitude of the average swirl flow decreases with increasing bundle size. Parametric studies revealed that neither the velocity nor concentration profiles were sensitive to Reynolds number in the range Re = 9000–24000.     

Heat transfer in turbulent longitudinal flow through unbaffled assemblies of fuel rods

A numerical analysis of heat transfer in turbulent longitudinal flow through assemblies of unbaffled fuel rods is presented. The solution applies to triangular or rectangular arrays of fuel rods with fully developed velocity and temperature profiles, for fluids with Prandtl number 1 and « 1. In the case of liquid metals, the thermal resistance of the cladding and bond are considered, but the turbulent heat transport component is neglected. For common liquids the circumferential turbulent heat transfer is considered. Results are compared in the range of dimensionless rod spacing of 1.0–1.6. Theoretical predictions and experimental results of other authors dealing with the problem show relatively good agreement.     

Convective and diffusive interchannel transfer in fuel assemblies of a fast-reactor core

Physics and Energy Institute (FÉI), Obninsk. Translated from Atomnaya Énergiya, Vol. 71, No. 2, pp. 114–123, August, 1991.     

我国压水堆装载MOX燃料对核燃料循环情景影响初步分析

为了满足持续增长的国家能源需求,核电将有更大规模的发展。本文对我国未来的核电发展和核燃料循环进行了情景研究,预测了2050年前核电对天然铀资源和燃料制造能力的需求情况,核电站产生的乏燃料量,分离钚产生量。乏燃料后处理能力作为我国核燃料循环体系的重要组成部分,将对我国核燃料循环情景产生重要影响。本文对后处理规模和分离钚的利用进行了假设,研究了两种情景模式下后处理和分离钚利用对我国铀资源需求和核废物产生的影响。     

Dynamic analysis of storage racks for spent fuel assemblies

The dynamic response of storage racks for spent fuel assemblies subjected to base excitation is calculated. While classical methods of linear structural dynamics may be adequate at low levels of excitation, nonlinear effects due to uplift, for example, can no longer be neglected at high excitation levels. Several nonlinear dynamic analyses have been performed for different types of storage racks with uplift capability. With the help of the numerical results, the rocking behavior of storage racks and their structural integrity has been examined.     

The effective thermal conductivity of crud and heat transfer from crud-coated PWR fuel

Water-filled crud on the surface of PWR fuel could offer resistance to the flow of heat, which might be expected to cause higher clad temperatures, and probably more fuel failures, than are actually observed. However, there is some evidence from post-irradiation inspection that the crud is penetrated by pores large enough to permit vapour formation, and it is believed these provide a mechanism for ‘wick boiling’ to occur, which modifies, and indeed can under some circumstances actually improve, heat transfer. This phenomenon is investigated using a two-dimensional coupled multi-physics model, accounting for the flow of water, heat and dissolved species within the crud. The fuel thermal performance is characterized in terms of an effective crud thermal conductivity derived from the use of this model, and the non-linear dependence this effective thermal conductivity has on parameters such as crud thickness and pore density is determined.     

Increasing critical heat flows in VVER-1000 model fuel assemblies by inserting additional spacing lattices

One way to increase the margin to critical heat transfer is investigated – closer lattice spacing on the axial section of a fuel assembly with the minimum margin to critical heat emission. Experiments were performed on the KS facility at the Russian Science Center Kurchatov Institute on two 19-rod VVER-1000 fuel-assembly models with lattice spacings 340, 255, and 170 mm. For coolant mass flow rate 4000 kg/(m²·sec) and relative enthalpy 0–0.1, the critical flow rate increases by 15–20% because the lattice spacing decreases from 340 to 170 mm. As the mass flow velocity and relative enthalpy of the coolant decrease, the critical heat flows do not increase as much. A generalizing relation is obtained for the gain in the critical heat flux as a function of the distance to a spacing lattice and the main regime parameters.