Heat Transport System Research Articles

Estimation of Flow-Accelerated Corrosion Rate in Nuclear Piping System

Abstract Flow-accelerated corrosion (FAC) is a life-limiting factor for the piping network of the primary heat transport system (PHTS) in CANDU® reactors. The pipe wall thinning caused by FAC is monitored by carrying out periodic in-service inspections (ISI) to ensure the fitness-for-service of the piping system. Accurate prediction of the lifetime of various components in the PHTS piping network requires estimation of FAC thinning rate. The traditional Bayesian inference techniques commonly employed for parameter estimation are computationally costly. This paper presents an inexpensive and intuitive simulation-based Bayesian approach to FAC rate estimation, called approximate Bayesian computation using Markov chain Monte Carlo (ABC-MCMC). ABC-MCMC is a likelihood-free Bayesian computation scheme that generates samples directly from an approximate posterior distribution by simulating data sets from a forward model. The efficiency of ABC-MCMC is demonstrated by presenting a comparison with a likelihood-based Bayesian computation scheme, Metropolis-Hastings (MH) algorithm, using a practical data-based example. Furthermore, an innovative step has been proposed for reducing the Markov chain burn-in time in the proposed scheme. To indicate the need of a Bayesian approach in quantifying the uncertainties related to the FAC model parameters, results from the linear regression method, a common industrial approach, are also presented in this study. The numerical results show a notable reduction in computational time, suggesting that ABC-MCMC is an efficient alternative to the traditional Bayesian inference methods, specifically for handling noisy degradation data.

Enhanced sorption heat transportation cycles with large concentration glide

Sorption heat transportation is promising in low-grade thermal energy transportation over long distance due to its small energy transportation loss. However, to compete with the sensible heat transportation from economic aspect, energy transportation density and efficiency of sorption heat transportation still need to be further improved. In this paper, two novel sorption heat transportation cycles with double-stage input or output are proposed. Compared with the conventional cycle, larger concentration glide is achieved by the novel cycles for cycle improvement. Performance of the conventional cycle and novel cycles are calculated and compared using mathematical models developed in software Engineering Equation Solver. Results show that higher energy transportation density could be achieved by the enlargement of concentration glide. The improvement in energy density is stronger with lower heat source temperature, higher heat output temperature, or lower cooling temperature. In addition, the performance of novel cycles is more stable under the same temperature range. The improved sorption heat transportation system could be used as an effective way to transport the low-grade industrial excess heat, which is an efficient connection between the heat source and end user.

Analysis of Loss of Heat Sink for ITER Divertor Cooling System Using Modified RELAP/SCDAPSIM/MOD 4.0

The present work includes thermal hydraulic modeling and analysis of loss of heat sink (LOHS) accident for the ITER divertor cooling system. The analysis is done for the new design of full tungsten divertor. The new design is also analyzed for different local heat loads ranging from 10 MW/m2 to 20 MW/m2 (while maintaining the total heat load 200 MW) under the steady-state fluid conditions. The LOHS event is selected since divertor is the most sensitive component to loss or reduction in coolability of divertor primary heat transport system (DV-PHTS) loop as it receives large heat flux from plasma. The main objective of this analysis is to find margins to unwanted conditions like overstress temperatures of structure and elevated water level in the pressurizer. The analysis is done by modified thermal hydraulic code RELAP/SCDAPSIM/MOD 4.0. The results obtained are compared with the results of old divertor design which uses carbon fiber composite (CFC) layer to show that how the new design of divertor behaves compared to the older design under the accident scenario. A detailed model of DV-PHTS loop and its ancillary system is presented. The model includes promotional integral differential (PID) controller for controlling the pressurizer heater and spray system. A detailed pump model is also included in the present analysis which was previously used as a time-dependent junction. The analysis shows that under the accident scenario, (a) the divertor structure temperature at the critical sites (inner vertical target (IVT) and outer vertical target (OVT)) is always within the design limit and does not affect the structural integrity of the divertor. (b) The water level in the pressurizer increases moderately and finely controlled by the PID controller, and pressurizer safety valve does not open.

Significance of Velocity Slip in Convective Flow of Carbon Nanotubes

The present article inspects velocity slip impacts in three-dimensional flow of water based carbon nanotubes because of a stretchable rotating disk. Nanoparticles like single and multi walled carbon nanotubes (CNTs) are utilized. Graphical outcomes have been acquired for both single-walled carbon nanotubes (SWCNTs) and multi-walled carbon nanotubes (MWCNTs). The heat transport system is examined in the presence of thermal convective condition. Proper variables lead to a strong nonlinear standard differential framework. The associated nonlinear framework has been tackled by an optimal homotopic strategy. Diagrams have been plotted so as to examine how the temperature and velocities are influenced by different physical variables. The coefficients of skin friction and Nusselt number have been exhibited graphically. Our results indicate that the skin friction coefficient and Nusselt number are enhanced for larger values of nanoparticle volume fraction.

열공급 연계망에서 열생산비용의 최소화를 위한 최적 열생산 및 열수송 결정

Environmental problems are becoming increasingly serious and important. As a countermeasure to prevent air pollution, district heating systems with high energy efficiency and low pollution are considered as a good alternative. The heating supply system of district heating is divided into heat production, heat transportation, and heat consuming subsystems. First, heat production system is analyzed by capacity and cost. Second, heat transportation system is analyzed to determine the traffic volume. Finally, the heat consuming system is analyzed by atmospheric temperature and the usage characteristics. Based on the analysis of the three subsystems, a cost model to determine the heat tranfer amounts is proposed. A case study with data of Korea District Heating Corporation is done. Additionally, sensitivity analysis is made to investigate the effect of changing market situation to the optimal solution.

Aerosol depositional characteristics in piping assembly under varying flow conditions

In the event of a nuclear reactor accident, large amount of radioactivity in the form of fission products may get released to the piping assembly of primary heat transport system. These fission products mostly in the form of aerosol particles get deposited on the inner surface of the piping system due to various depositional processes. The removal processes in the complex piping system are controlled to a large extent by the thermal-hydraulic conditions like temperature, pressure and flow rates. These parameters generally vary with time and therefore must be carefully monitored to predict the aerosol behavior in the piping system. Experimental determination of the deposition fraction, interpretation of the role of controlling parameters and development/validation of theoretical models are key areas gaining constant attention among researchers. In the present work, experiments were conducted in a piping assembly consisting of bends and horizontal-vertical orientation at two different carrier gas flow rates. Deposition fractions in the test assembly were estimated and the role of different dynamical processes (thermophoresis, gravitation and bend impaction) was interpreted. The computational fluid dynamics modeling approach is used for theoretical simulation in this work. Aerosol behavior in terms of number concentration, particle size distribution, and particle deposition in the piping system was simulated with the computational fluid dynamics software ANSYS Fluent 16.0 and the results were compared with experimental measurements (wherever applicable).

Experimental investigation of radiation heat transfer in coolant channel under impaired cooling scenario for Indian PHWR

Postulated accidents like large break LOCA leads to expulsion of coolant in the primary heat transport system thus voiding of reactor core initially. However, the reactor is shut down at the onset of LOCA along with coolant injection from ECCS to remove the decay heat of 2–3% of nominal power. Further postulation of failure of ECCS leads to rapid increase in the temperature of fuel pins and the coolant channel as well. The moderator present around coolant channels limits the rise in temperature. The assessment of temperature distribution in the fuel pins of the bundles of a channel under high temperature is quite important from hydrogen generation and fission product release point of view. A pseudo steady state experiment has been carried out to obtain the temperature distribution of different components of a simulated coolant channel for large capacity Indian pressurized heavy water reactor. The experiment simulates a postulated LOCA with Loss of ECCS scenario for the coolant channel. The experimental results shows that for 1% decay power the simulated fuel pins attained a maximum pseudo steady state temperature of 900 °C–1000 °C, thus establishing moderator as a heat sink. An insignificant circumferential temperature variation in the channel components is observed which indicates weak effect on natural convective heat transfer within coolant channel. Estimation shows that around 85 percent of total decay heat provided through electrical power is removed by the moderator.

Preliminary CAD implementation of EU-DEMO primary heat transfer systems for HCPB breeding blanket option

This paper focuses on the 3D CAD implementation of the pipework and the main equipment of the Primary Heat Transfer System of EU-DEMO fusion power plant. In particular, the systems related to the Helium-Cooled Pebble Bed Breeding Blanket option are considered here. During the pulse operation, the breeding blanket modules will be the main thermal power source; Divertor and the Vacuum Vessel will contribute in the definition of the total reactor power. All the In-Vessel generated power is rejected to the Power Conversion System through a molten salt Intermediate Heat Transport System. The latter is equipped with an Energy Storage System to allow for continuous operation also during the dwell time. The layout of the pipework considered the major design constraints, such as the reinforced concrete structures and the dimensions of the main pieces of equipment. The size of the pipes was selected based on operating pressure, temperature and allowable velocity. Results provide a first estimation of length and weight of the piping systems as well as their space occupancy. These data are useful to evaluate the coolant inventory, as well as the expected pressure drops and plant cost.

Recent progress in developing a feasible and integrated conceptual design of the WCLL BB in EUROfusion project

The water-cooled lithium-lead breeding blanket is in the pre-conceptual design phase. It is a candidate option for European DEMO nuclear fusion reactor. This breeding blanket concept relies on the liquid lithium-lead as breeder-multiplier, pressurized water as coolant and EUROFER as structural material. Current design is based on DEMO 2017 specifications. Two separate water systems are in charge of cooling the first wall and the breeding zone: thermo-dynamic cycle is 295–328 °C at 15.5 MPa. The breeder enters and exits from the breeding zone at 330 °C. Cornerstones of the design are the single module segment approach and the water manifold between the breeding blanket box and the back supporting structure. This plate with a thickness of 100 mm supports the breeding blanket and is attached to the vacuum vessel. It is in charge to withstand the loads due to normal operation and selected postulated initiating events. Rationale and progresses of the design are presented and substantiated by engineering evaluations and analyses. Water and lithium lead manifolds are designed and integrated with the two consistent primary heat transport systems, based on a reliable pressurized water reactor operating experience, and six lithium lead systems. Open issues, areas of research and development needs are finally pointed out.

Fission product release in thoria and urania defective fuel experiments FDO-680 and FDO-681

FDO-680 (thoria as ThO2-4%UO2) and FDO-681 (urania) were circa 1975 intentionally defected fuel experiments conducted in the National Research Experimental (NRX) reactor X-2 defect loop in Chalk River, Canada. The intent of these experiments was to determine the release rates of fission gases and radioiodines from a defected ThO2-4%UO2 element and to determine the extent to which a moderately powered ThO2-4%UO2 element would deteriorate when defected. Detailed analysis of the FDO-680 and FDO-681 fission product release data has only recently been undertaken. The objective of analyzing this data is to address a gap in defected thoria fuel behaviour knowledge enroute to commercialisation of thoria fuels. The analysis suggests: 1) thoria fuel has superior stability when defected relative to urania, 2) thoria results in a lower fission product burden in the Primary Heat Transport System (PHTS) than urania under steady-state conditions, and 3) the fission product transport models developed for urania cannot be directly applied to thoria.

Assessment of Mechanical Properties for Dissimilar Metal Welds: A Nondestructive Approach

Several engineering components are fabricated by joining similar or dissimilar materials. In the present investigation, welded joints between low alloy steel and austenitic stainless steel were considered. This assembly is one of the critical components in the heat transport system of water-cooled reactors of nuclear power plants. 309L austenitic stainless steel and IN 182 alloys were used for buttering low alloy steel during fabrication of joints. Weld metal was 308L austenitic stainless steel and IN 182, respectively, for the two assemblies. Buttering by 309L SS was performed by gas tungsten arc welding, and the same was done by shield metal arc welding at the time of applying IN 182. Joining of 304L SS and buttered assembly in both the cases was carried out by shield metal arc welding. Evaluation of Vickers micro-hardness was done across the weld centerline for both joints. Microstructure of different regions was examined for the joints. Apart from conventional destructive tests, ultrasonic evaluation of joints was also carried out to ensure weld integrity. In this respect, ultrasonic velocity in longitudinal and shear mode was measured along the transverse direction of the weld. Young’s modulus was determined using ultrasonic tests as well as from the average hardness of different regions. The obtained Young’s modulus by two different techniques revealed satisfactory co-relation. The Young’s modulus obtained from ultrasonic measurements was used to determine the yield strength of different regions across the welds. This technique proved to be useful in identifying the failure prone area across the welded joint.

Performance of heat transport systems: least square method generated correlations of non-dimensional variables

Waste heat and renewable energy are increasingly being recognized as important sources of heat for district heating systems and for industrial clients, and as a means to reduce fossil fuel consumption. One of the most important challenges of using this heat is to transport it from the source to the load in an efficient manner. The objective of this paper is to establish explicit relations between the performance and the essential design variables of a system that includes a heat source, a heat transport sub-system, and a heat load. For this purpose we have used dimensional analysis, a factorial design of experiment methodology with the least square method and obtained linear correlations between two performance indicators (the system effectiveness and its exergy efficiency) and non-dimensional groups which combine the physical and operational characteristics of the system. It has also been shown that the economic desirability of the system as measured by the Internal Rate of Return increases linearly with the system effectiveness. A sensibility analysis has determined the non-dimensional groups which have the most important effect on each performance indicator. The obtained correlations have been applied to solve the following two design problems: (1) find the optimum values of design parameters such as the pipe insulation thickness and the mass flow rate of the heat transport fluid in order to maximize the exergy efficiency of the system and (2) find the optimum distribution of a fixed total thermal conductance in order to maximize the system effectiveness and/or its exergy efficiency.

소듐냉각고속로 중간열교환기의 열수력 성능개선에 대한 수치해석적 연구

KAERI has been developed a conceptual design of the PGSFR, the pool-type sodium cooled fast reactor. The intermediate heat exchanger is located in the primary heat transport system and transfers heat generated from reactor core from primary heat transport system to the intermediate heat transport system. KAERI developed a one-dimensional code for design and performance analysis of an intermediate heat exchanger. Computational fluid dynamics simulations were performed to validate this code and analyze three-dimensional heat flow. In the case of temperature, the code and CFD results were very similar and the pressure was slightly different, the code was modified compared with the experimental results. The problem caused by temperature imbalance founded in tube-side outlet chamber of the IHX was solved by incorporating the flow mixer.

Performance evaluation of passive pulse generator for auto depressurization system of Advanced Heavy Water Reactor

The Advanced Heavy Water Reactor (AHWR) is a 300 MWe pressure tube type boiling light water cooled, heavy water moderated reactor. It has natural circulation based main heat transport system with integral steam drums. Recently, work has been initiated on a novel passive safety system named Passive Auto Depressurization System (PADS) for AHWR. This system is a part of shutdown cooling system and in case of occurrence of small break loss of coolant accident; PADS system mitigates chance of reactor core temperature rise. It actuates automatically at low steam drum level and rapidly depressurizes main heat transport system so that water injection from gravity driven water pool can be initiated to maintain coolant inventory for a sufficiently long time.The main component of PADS is Passive Pulse Generator (PPG), a passive device which is actuated at low steam drum level and generates high pressure signal within a short time span. The PPG generates a passive pressure pulse of around 11 bar amplitude with 8 min response time with threshold steam drum level fall of 100 mm. This pressure pulse is used for opening a passive valve which initiates decay heat removal through isolation condenser leading to depressurization of main heat transport system. An experimental facility has been designed towards the development and performance evaluation of PPG for PADS system of AHWR.This paper deals with the design of the experimental facility describing its piping layout, steam drum & heat exchanger vessel and associated instrumentation & control system. Test results along with RELAP5 analysis are discussed to validate PPG performance for PADS in AHWR.

Power Flow Method-Based Integrated Modeling and Optimization for Building Heat Transport and Gas Refrigeration System

AbstractThe combination of heat transport and heat-work conversion processes in thermal systems highlights the importance of the integrated modeling and optimization method. Following the analogy b...

Design optimization analysis of a large electromagnetic pump for sodium coolant transportation in PGSFR

An annular linear induction electromagnetic pump (ALIP) with a flow rate of 0.86 m3/s and a developed pressure of 3.6 bar is designed by conducting an optimization analysis of its geometrical, electromagnetic, and hydrodynamic variables. The purpose of the ALIP involves circulating liquid sodium in the intermediate heat transport system of the Prototype Generation-IV Sodium-cooled Fast Reactor (electric output: 150 MWe) that is currently under development in Korea. The efficiency and developed pressure of the ALIP are derived as a function of its design variables and include the core length, flow gap, number of poles, flow velocity, pole pitch, input frequency, and input current. In terms of electromagnetic variables that can be controlled during its operation, hydraulic efficiency is maximized at an input frequency of a few tens of Hertz. The P–Q characteristics of the optimized ALIP are analyzed and verified via a theoretical comparison with an extremely large ALIP.

Status of EU DEMO heat transport and power conversion systems

DEMO in Europe is considered to be the nearest-term reactor design to follow ITER and capable of demonstrating production of electricity, operating with a closed fuel-cycle and to be a facilitating machine between ITER and a commercial reactor. The aim of this paper is to show the design progress of the complex system’s “chain” devoted to the extraction of the plasma generated pulsed thermal power and its conversion into electricity delivered to the grid, including the Primary Heat Transport System (PHTS), the Power Conversion System (PCS) and the Intermediate Heat Transport System (IHTS) – provided by an Energy Storage System (ESS) – in between PHTS and PCS, which is introduced for the scope of smoothing the generated pulsed plasma power removed by PHTS transmitted to PCS for a more continuous conversion/production of electricity and to guaranty plant lifetime. This will be done with reference to two breeding blanket concepts: the Helium Cooled Pebble Bed (HCPB) Breeding Blanket (BB) and the Water Cooled Lithium Lead (WCLL) BB.The PHTS design criteria and the system design finalization through a descriptive 3D CAD model which shows the system inside the tokamak building are presented. The design and operational challenges as well as the on-going development plans to investigate possible design simplifications (e.g. the direct coupling of Breading Blanket PHTS to PCS) and improved technology readiness actions potentially leading to better plant reliability and cost minimisation are briefly described.

Mechanistic modelling of station blackout accidents for a generic 900 MW CANDU plant using the modified RELAP/SCDAPSIM/MOD3.6 code

CANDU (CANada Deuterium Uranium) reactors have many unique design features that play important roles during a severe accident, however analysis of such features using Light Water Reactor (LWR) specific computer codes is challenging. Severe accidents in CANDU involve complex thermo-mechanical deformation phenomena which differ from the phenomena present during LWR accidents. For example, during complete station blackout scenarios with a failure of all emergency measures, the pressure tubes may balloon or sag into contact with the surrounding calandria tubes (CTs) establishing a thermal conduction pathway for heat rejection to the large moderator water volume. As the moderator liquid evaporates or boils its level decreases until fuel channels become uncovered in the calandria vessel. The uncovered channels heat up quickly and the entire fuel channel assembly (fuel, pressure tube and calandria tube) will sag and possibly disassemble. During the disassembly process some channel components may fall to the bottom of the calandria while others may form a suspended debris bed supported by channels which are still submerged in moderator liquid. These phenomena impact event-timing, accident progression, hydrogen production and fission product release.In this work several mechanistic channel deformation models have been developed and integrated into RELAP/SCDAPSIM/MOD3.6 to provide a coupled treatment of the deformation phase for such postulated accidents. MOD3.6 is a new version of the RELAP/SCDAPSIM code being developed to support the analysis of Pressurized Heavy Water Reactors (PHWRs) under severe accident conditions. In this paper, the code system is used to simulate a postulated station blackout accident for a generic 900 MW CANDU plant. To reduce the uncertainty in the modeling of core disassembly and to overcome the memory constraints of the code, the simulation is broken into two phases with the first phase (i.e., from initiating event to the channel failure and depressurization) simulated using a full-plant RELAP5 model providing relatively high spatial fidelity of the entire heat transport system, and the second phase (i.e. continued from the end of the first phase until calandria vessel dryout) using alternative nodalization focusing on the calandria vessel and fuel channel components. The paper assesses the entire accident progression up to the point of calandria vessel dryout and performs sensitivity analysis on model parameters to assess their relative importance.

Decision support system for design of long distance heat transportation system

District Heating (DH) systems are commonly supplied using local heat sources. Nowadays, modern insulation materials allow for effective and economically viable heat transportation over long distances (over 20 km). The paper proposes a Decision Support System (DSS) for optimized selection of design and operating parameters of a long distance Heat Transportation System (HTS). The method allows for evaluation of feasibility and effectiveness of heat transportation from the considered heat sources to the given DH area. The optimized selection is formulated as the multicriteria decision-making problem. The constraints for this problem include a static HTS model, allowing considerations of the system life cycle, as well as time variability and spatial topology. Thereby, the variation of heat demand and ground temperature within the DH area, the insulation and pipe aging, as well as the terrain elevation profile are taken into account in the decision-making process. The HTS construction costs, the operating costs (pumping power), and the heat loss are considered as objective functions, while such parameters as: inner pipeline diameter, insulation thickness, temperature and pressure profiles, as well as pumping station locations are optimized during the decision-making process. Moreover, variants of pipe laying e.g. one pipeline with a larger diameter, or two parallel pipelines with smaller diameters might be considered during the optimization. The analyzed optimization problem is multicriteria, hybrid and nonlinear. The genetic solver has been proposed to solve it.

Study on Effects of Heat Loss and Channel Deformation on Thermal-Hydraulic Performance of Semicircular Straight Channel Printed Circuit Heat Exchangers

Effective and robust high-temperature heat transport systems are essential for the successful deployment of advanced high temperature reactors. The printed circuit heat exchanger (PCHE) is a strong potential candidate for the intermediate or secondary loop of high temperature gas-cooled reactors (HTGRs) due to their high power density and compactness. For high-temperature PCHE applications, the heat loss, which is difficult to be insulated completely, could lead to the degradation of heat exchanger performance. This paper describes an analytical methodology to evaluate the thermal-hydraulic performance of PCHEs from experimental data, accounting for extraneous heat losses. Experimental heat exchanger effectiveness results, evaluated without accounting for heat loss, exhibited significant data scatter while the data were in good agreement with the ε-NTU method once the heat loss was accounted for. The deformation of PCHEs would occur during the diffusion-bonding fabrication process or high temperature operations due to the thermal deformation. Computational assessment of the PCHE performance test data conducted at the Ohio State University showed that the deformation of flow channels caused increase of pressure loss of the heat exchanger. The computational fluid dynamics (CFD) simulation results based on the nominal design parameters underestimated the pressure loss of the heat exchanger compared to the experimental data. Image analysis for the flow channel inlet and outlet was conducted to examine the effect of channel deformation on the heat exchanger performance. The CFD analysis based on the equivalent channel diameter obtained from the image analysis resulted in a better prediction of PCHE pressure loss.