An experimental and analytical investigation of uncovered core heat transfer under high pressure, low heat flux conditions

Experimental and analytical results are reported from two series of high pressure core uncovering experiments. It was determined that the uncovered core is cooled primarily by convection and radiation to dry steam and that droplets are confined to the immediate vicinity of the mixture level. Spacer grids substantially increased heat transfer at and downstream of the grid. A simple heat transfer model is presented which accurately predicts uncovered core heat transfer at modified wall Reynolds numbers greater than 2000. Results are expected to be use in modelling small break loss of coolant accidents.     

压力管式超临界水堆极限事故下“无堆芯熔化”概念评估

在自主研发的事故分析程序SCTRAN的基础上,开发并验证了二维导热模型和辐射换热模型,并将改进后的SCTRAN应用于加拿大压力管式超临界水堆在失水事故（LOCA）叠加丧失紧急堆芯冷却系统（LOECC）事故中的堆芯安全评估,并对燃料棒到慢化剂之间的传热效率以及关键的影响因素进行了评估。计算结果表明,在LOCA叠加LOECC工况下,燃料棒到燃料通道的辐射换热和燃料棒到蒸汽的自然对流换热能够有效导出反应堆的衰变余热,最高功率的燃料组件内、外圈燃料棒的最高包壳温度分别为1278℃和1192℃,均低于不锈钢包壳的熔化温度,因此整个事故过程中不会发生堆芯熔化。      

Estimation of fuel and clad temperature of a research reactor during dry period of de-fuelling operation

Heat transfer and fluid flow studies related to spent fuel bundle of a research reactor in fuelling machine has been carried out. When the fuel is in reactor core, the heat generated in the fuel bundle is removed by heavy water under normal reactor operation. However, during the de-fuelling operation, the fuel bundle is exposed to air for some period called dry period. During this period, the decay heat from fuel bundle has to be removed by air flow. This flow of air is induced by natural convection only. In this period, the temperatures of fuel and clad rise. If clad temperature rises beyond a certain limit, structural failure may occur. This failure can result into release of fission products from fuel rod. Hence the temperature of clad has to be within specified limit under all conditions. The objective of this study is to estimate the clad temperature rise during the dry period.In the CFD simulation, the turbulent natural convection flow over fuel and radiation heat transfer are accounted. Standard k-? model for turbulence, Boussinesq approximation for computing the natural convection flow and IMMERSOL model for radiation are used.The steady state and transient CFD simulation of flow and heat is performed, using the CFD code PHOENICS. The steady state analysis provides the maximum temperature the clad will attain if fuel bundle is left exposed to air for sufficiently long time. For safe operation, the clad temperature should be limited to a specified value. From steady state CFD analysis, it is found that steady state clad temperature for various decay powers is higher than the limiting value. Hence transient analysis is also performed. In the transient analysis, the variation of clad temperature with time is predicted for various decay powers. Safe dry time, i.e. the time required for clad to reach the limiting value, is predicted for various decay powers. Determination of safe dry time helps in deciding the time available to the operator to drop the bundle in light water pool for storage. The analysis is found useful in optimizing the de-fuelling process.     

Natural circulation capability of Pb-Bi cooled fast reactor: PEACER

First-principle calculations were performed to analyze the natural circulation heat removal from the core of a liquid metal reactor (LMR). The lead-bismuth (Pb-Bi) was chosen as the primary coolant for the LMR system. From the single channel analysis the temperature and the pressure drop are calculated along the fuel assembly. The total pressure drop of the core is less than 100kPa due to the large pitch-to-diameter ratio and the small height of the fuel pin. The natural circulation potential is a key characteristics of the LMR design. The steady-state momentum and energy equations are solved along the primary coolant path. The calculations are divided into two parts: an analytical model and a one-dimensional lumped parameter flow loop model. Results of the analytical model indicate that the elevation difference of 4.5m between thermal centers of the core and the steam generators could remove as much as 10% of the nominal operating reactor power. The flow loop model yielded the total pressure drop and the natural circulation heat removal capacity.     

Power up-grading study for the first Egyptian research reactor

In the present work, power up-grading study is performed, for the first Egyptian Research Reactor (ET-RR-1), using the present fuel basket with 4×4 fuel rods, (17.5 mm pitch), and a proposed fuel basket with 5×5 fuel rods, (14.0 mm pitch), without violating the thermal hydraulic safety criteria. These safety criteria are; fuel centerline temperature (fuel melting), clad surface temperature (surface boiling), outlet coolant temperature, and maximum heat flux (critical heat flux ratio). Different thermal reactor powers (2–10 MW) and different core coolant flow rates (450, 900, 1350 m³ h⁻¹) are considered. The thermal hydraulic analysis was performed using the subchannel code COBRA-IIIC for the estimation of temperatures, coolant velocities and critical heat flux. The neutronic calculations were performed using WIMS-D4 code with 5 — group neutron cross section library. These cross sections were adapted to use in the two-dimensional (2-D) diffusion code DIXY for core calculations. The study concluded that ET-RR-1 power can be upgraded safely up to 4 MW with the present 4×4-fuel basket and with the proposed 5×5-fuel basket up to 5 MW with the present coolant flow rate (900 m³ h⁻¹). With the two fuel arrays, the reactor power can be upgraded to 6 MW with coolant flow rate of 1350 m³ h⁻¹ without violating the safety criterion. It is also concluded that, loading the ET-RR-1 core with the proposed fuel basket (5×5) increases the excess reactivity of the reactor core than the present 4×4 fuel matrix with equal U-235 mass load and gave better fuel economy of fuel utilization.     

Gas-cooled fast reactors—Status of CEA preliminary design studies

The gas-cooled fast reactor (GFR) is one of the six reactor concepts selected in the frame of the Generation IV initiative. The most significant GFR option is the use of a helium high temperature primary coolant. The helium option is very attractive (chemical inertness, neutron transparency, etc.) but it leads to very specific thermal-hydraulic issues.As far as the reactor core design is concerned, a ceramic fuel concept with a good thermal conductivity has been chosen. The main requirement is to obtain an average exit core temperature of 850 °C (energy conversion efficiency) with a maximum fuel temperature of about 1200 °C and with a low core pressure drop (in order to ease the decay heat removal). The main core characteristics have been determined for two reactor powers: a medium one (600 MWth) and a large one (2400 MWth). For various reasons, this latter became the CEA reference choice. A consistent set of core parameters has been determined taking into account the different constraints including the thermal-hydraulics. The reference arrangement proposed is based on plate fuel elements.A significant issue for the GFR is the decay heat removal. An innovative approach has been chosen in case of loss of coolant accidents (LOCAs). A “guard containment” enclosing the primary system is used to maintain a medium gas pressure (10 bar) in order to remove the decay heat by low power forced convection systems in the short term and natural convection systems in the long term. This guard containment is not pressurized during normal operations and can be a metallic structure.As far as the energy conversion system is concerned, an indirect-combined cycle has been chosen. The significant advantages of this choice are: a moderate core inlet temperature (400 °C instead of 480 °C for the direct cycle) and an attractive compactness of the primary system (facilitating the guard containment design).Due to the novelty of these options, a significant effort of components pre-sizing and design calculations has been achieved. Following this effort, a CATHARE model of the reactor system has been made and the calculation of the reactor steady-state confirms the consistency of the overall system pre-sizing. This model has been used for a first transient calculation. Other types of transients have to be analyzed, however, it is thought that the proposed GFR design can reach the safety requirements of Generation IV systems.     

The inherently-safe power reactor DYONISOS : Dynamic nuclear inherently-safe reactor operating with spheres

A philosophy of inherent safety is formulated and an inherently-safe thermal power reactor is presented. Solid fuel in the form of spheres a few centimetres in diameter is suspended under the hydrodynamic pressure of molten lead coolant in vertical channels within the graphite moderator. Loss of main pump pressure, or a loss-of-coolant accident (LOCA), results in immediate removal of the fuel to rigid sieves below the core, with consequent subcriticality. Residual and decay heat are carried away by thermal conduction through the coolant or, in the case of a LOCA, by a combination of radiation and natural convection of cover gas or incoming air from the fuel to the reactor vessel and convection of air between the vessel and steel containment wall. All decay heat removal systems are passive, though actively initiated external spray cooling of the containment can be used to reduce wall temperature. This, however, is only necessary in the case of a LOCA and after a period of 24 h.     

Parametric study on rewetting velocities obtained with a two-dimensional heat conduction code

A comprehensive parametric study has been performed to quantify the effect of different variables on the rewetting velocity in a light water reactor following a loss-of-coolant accident. To this purpose, a numerical solution of the general two-dimensional (axial and radial) heat conduction equation in cylindrical geometry has been obtained. The method used is the alternating-direction implicit procedure developed by Peaceman and Rachford. The model accounts for decay heat generation in the fuel, coolant subcooling, different wall temperatures and different heat transfer coefficients across the gap and at the clad surface. The two-dimensional model can be reduced to a one-dimensional model by setting the heat conduction in either the radial or axial direction to zero. Results with the new model agree with previous models and with experimental data.The variables studied were: axial and/or radial heat conduction, clad temperature, quench temperature, coolant temperature, temperature for the onset of nucleate boiling, heat transfer coefficients, stored and decay heats, clad material and clad thickness. The critical thickness (clad thickness for which the calculated rewetting velocity remains constant) was also determined and found to be larger than the clad thickness of light water reactor fuel pins under usual reflood conditions. According to these calculations, the stored and decay heats affect the rewetting velocity significantly.     

基于开放栅格结构的空间堆三维热工水力特性研究

高温气冷堆结合磁流体发电是一种高效的空间电源系统,可以满足空间任务对于大功率、高效率的需求,具有广阔的应用前景。本文参考美国普罗米修斯计划中的开放栅格方案,结合磁流体发电需满足的设计条件,提出了一种由三角形布置、217根燃料棒构成的堆芯方案。在通过试验数据确定流动模型后,对该空间堆进行了三维建模,并在考虑气隙结构、燃料棒功率分布及堆内辐射的基础上研究其热工水力特性,重点针对环境温度及外壁面发射率展开了热工参数敏感性分析。计算结果表明,该堆芯热工设计满足材料温度、压降限值等指标要求。冷却剂在燃料区横向流动不明显,不存在复杂涡结构,流动现象相对较为简单。稳态热工计算结果对环境温度的改变并不敏感,但发射率的改变影响相对较大。     

Reactivity insertion limits in a typical pool-type research reactor cooled by natural circulation

The course of reactivity insertion in a pool type research reactor, with scram disabled under natural circulation condition is numerically investigated. The analyses were performed by a coupled kinetic–thermal–hydraulic computer code developed specifically for this task. The 10-MW IAEA MTR research reactor was subjected to unprotected reactivity insertion (step and ramp) for both low and high-enriched fuel with continuous reactivity feedback due to coolant and fuel temperature effects. In general, it was found that the power, core mass flow rate and clad temperature under fully established natural circulation are higher for high-enriched fuel than for low enriched fuel. This is unlike the case of decay heat removal, where equal clad temperatures are reported for both fuels. The analysis of reactivity represented by the maximum insertion of positive reactivity ($0.73) demonstrated the high inherent safety features of MTR-type research reactor. Even in the case of total excess reactivity without scram, the high reactivity feedbacks of fuel and moderator temperatures limit the power excursion and avoid consequently escalation of clad temperature to the level of onset of nucleate boiling and sub-cooled void formation. The code can also be modified to provide an accurate capability for the analyses of research reactor transients under forced convection.     

CFX simulation of high temperature thermal–chemical experiment: CS28-2

A computational fluid dynamics (CFD) model of a post-blowdown fuel channel analysis for aged CANDU reactors with crept pressure tube has been developed, and validated against a high temperature thermal–chemical experiment: CS28-2. The CS28-2 experiment is one of three series of experiments to simulate the thermal–chemical behavior of a 28-element fuel channel at a high temperature and a low steam flow rate which may occur in severe accident conditions such as a LBLOCA (large break loss of coolant accident) of CANDU reactors. Pursuant to the objective of this study, the current study has focused on understanding the involved phenomena such as the thermal radiation and convection heat transfer, and the high temperature zirconium-steam reaction in a multi-ring geometry. Therefore, a zirconium-steam oxidation model based on a parabolic rate law was implemented into the CFX-10 code, which is a commercial CFD code offered from ANSYS Inc., and other heat transfer mechanisms in the 28-element fuel channel were modeled by the original CFX-10 heat transfer packages. To assess the capability of the CFX-10 code to model the thermal–chemical behavior of the 28-element fuel channel, the measured temperatures of the fuel element simulators (FES) of three fuel rings in the test bundle and the pressure tube, and the hydrogen production in the CS28-2 experiment were compared with the CFX-10 predictions.     

European supercritical water cooled reactor

The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 °C average core outlet temperature. It is designed and analyzed by a European consortium of 10 partners and 3 active supporters from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small fuel assemblies in boxes with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. Besides the scientific and technical progress, the HPLWR project turned out to be most successful in training the young generation of nuclear engineers in the technologies of light water reactors. More than 20 bachelor or master theses and more than 10 doctoral theses on HPLWR technologies have been submitted at partner organizations of this consortium since the start of this project.     

Two-dimensional simulations of natural convection/radiation heat transfer for BWR assembly within isothermal enclosure

Abstract

The current scoping study identifies the significant heat transfer effects for a 7 × 7 boiling water reactor (BWR) assembly within an isothermal basket opening inside a transport cask. A two-dimensional finite volume mesh is constructed that models the assembly components and cover gas. Computational fluid dynamics (CFD) simulations calculate the buoyancy induced gas motion, conduction and radiation within the components. Simulations use different basket surface temperatures, fuel heat generation rates and cladding surface emissivities, for both nitrogen and helium cover gases at atmospheric pressure. An analytical conduction/radiation model is developed for the thermal resistance between the channel and basket. Results using buoyancy induced gas motion compared to stagnant gas simulations show that natural convection is significant only at low basket temperatures, with nitrogen gas. Helium and high basket temperature simulations exhibit no significant temperature reduction from natural convection. Simulations with varying cladding emissivity ? show that a 10% increase in ? causes a 7˙2% decrease in the interior temperature difference for nitrogen and a 5˙3% decrease for helium.     

Analysis of radiant heat transfer in a BWR fuel assembly

A computer code ‘CIDER’ was developed which analyzes radiant heat transfer in a BWR fuel rod bundle under loss of coolant conditions. In the code, (1) a channel box and fuel rods are considered to be gray bodies, (2) reflection and absorption of radiation beams in the atmosphere is neglected, (3) a fuel rod is approximated by a regular polygonal rod, and (4) radiant heat flux is calculated considering circumferential temperature distribution on each fuel rod surface, which is determined from radial and circumferential heat conduction calculations in a fuel rod. It was found that the conventional model with uniform cladding temperature overestimated heat flux about 30% in a typical situation, or correspondingly underestimated the temperature rises.     

Scaling study of in-core boil-off and heating process

A simple analytical method, which describes uncovery and heatup in the core under accident conditions, is derived, tested against experimental data, and used for generating the scaling criteria. Void fraction and core uncovery levels are analytically derived integrating mass and energy equations under the assumption of quasi-steady state. The coolant energy equation in the uncovered region is integrated to convert the partial differential equation for the fuel temperature into an ordinary differential equation through the assumption of the same axial distribution of the amount of energy loss from the fuel to the coolant as that of the decay heat generation rate. The ordinary differential equation for the fuel temperature, combined with the governing equation for cladding oxidation, is analytically solved assuming a linear variation of fuel temperature and oxidation thickness with time over a period.The present analytical model is tested against the Power Bursting Facility Scoping Test (PBF-ST) and SCDAP calculation. The model produces the estimation of inlet flow rates and its results which are in good agreement with the measured levels. There is an overprediction of the fuel temperatures and an underprediction of the rate of increase of the fuel temperatures by the model, presumed to be mainly caused by no consideration of reflux condensation and the higher prediction of radiation energy loss to the shroud through the treatment of one radial region of the bundle.The PBF-ST is examined with the scaling criteria generated by the present model. It is found out that the linear heat generation rate in the PBF should be by four times larger than that in the prototype system and the radiation number is highly distorted in the PBF.     

ATWS analysis for PWR using the coupled code system DYN3D/ATHLET

The ATWS transient “Loss of main feed water supply” in a generic four-loop PWR at the nominal power of 3750 MW was analyzed using the coupled code system DYN3D/ATHLET. A variation of the MOX-fuel-assembly portion in the core has an effect on the reactivity coefficients of the fuel temperature and the moderator density. These two parameters mainly influence the behaviour of the coolant pressure, which is safety-relevant. It has been demonstrated that the pressure maximum decreases with an increasing portion of MOX. For all core loadings considered, both primary-circuit mechanical integrity and sufficient core cooling are guaranteed.     

Numerical simulation of AP1000 LBLOCA with SCDAP/RELAP 4.0 code

The risk of large-break loss of coolant accident (LBLOCA) is that core will be exposed once the accident occurs, and may cause core damages. New phenomena may occur in LBLOCA due to passive safety injection adopted by AP1000. This paper used SCDAP/RELAP5 4.0 to build the numerical model of AP1000 and double-end guillotine of cold leg is simulated. Reactor coolant system and passive core cooling system were modeled by RELAP5 modular. HEAT STRUCTURE component of RELAP5 was used to simulate the fuel rod. The reflood option in RELAP5 was chosen to be activated or not to study the effect of axial heat conduction. Results show that the axial heat conduction plays an important role in the reflooding phase and can effectively shorten reflood process. An alternative core model is built by SCDAP modular. It is found that the SCDAP model predicts higher maximum peak cladding temperature and longer reflood process than RELAP5 model. Analysis shows that clad oxidation heat plays a key role in the reflood. From the simulation results, it can be concluded that the cladding will keep intact and fission product will not be released from fuel to coolant in LBLOCA.     

Thermodynamic analysis of a solid nuclear fuel element surrounded by flow of coolant through a concentric annular channel

A continuous quest for efficient utilization of energy resources has motivated the researchers to search for optimal design and operating conditions during various energy conversion techniques. These conditions for such systems are often proposed by minimizing the destroyed exergy potential in course of the process. In the present paper a second law analysis is done for a nuclear fuel element inside a concentric annular coolant passage. The entropy generation analysis has been carried out through a conjugate approach, with steady state temperature profiles within the fuel element and a thermodynamic approach within fluid. The effect of solid core heat generation and the temperature gradients inside solid core, fuel-clad gap and cladding are considered as well along with the irreversibilities arising out of fluid flow under turbulent condition. The effect of Reynolds number, duty parameter, diameter ratio, Biot number, dimensionless heat flux and thermal conductivity ratios on overall entropy generation characteristics have been investigated and interpreted physically. The validation of the present calculations was confirmed by best-estimate thermal-hydraulic code RELAP. The new thermodynamic design methodology presented in this paper adheres to the safety limits in temperature. The present analysis can be extended for complex fuel pellet arrangements in subchannel structures by an “equivalent annulus model”.     

Innovations in the ENHS reactor design and fuel cycle

A number of approaches were explored for improving characteristics of the encapsulated nuclear heat source (ENHS) reactor and its fuel cycle, including: increasing the ENHS module power, power density and the specific power, making the core design insensitive to the actinides composition variation with number of fuel recycling and reducing the positive void coefficient of reactivity. Design innovations examined for power increase include intermediate heat exchanger (IHX) design optimization, riser diameter optimization, introducing a flow partition inside the riser, increasing the cooling time of the LWR discharged TRU, increasing the minor actinides' concentration in the loaded fuel and split-enrichment for power flattening. Another design innovation described utilizes a unique synergism between the use of MA and the design of reduced power ENHS cores.

Also described is a radically different ENHS reactor concept that has a solid core from which heat pipes transport the fission power to a coolant circulating around the reflector. Promising features of this design concept include enhanced decay heat removal capability; no positive void reactivity coefficient; no direct contact between the fuel clad and the coolant; a core that is more robust for transportation; higher coolant temperature potentially offering higher energy conversion efficiency and hydrogen production capability.