Coolant Loop Research Articles

Tritium Recovery from Liquid Blanket Systems for Fusion Reactor

Systems to recover tritium (T) from fusion liquid blankets of LiPb (Li17Pb83) and Flibe (Li2BeF4) under safety conditions are discussed based on available data of T transport properties. We recently performed experiments on hydrogen isotopes permeating through LiPb in Kyushu University and introduced them in the present paper. Solubilities, diffusivities and permeabilities of H and D through Li17Pb83 are determined by means of a transient permeation method. Their isotope effects between H and D in the Li-Pb eutectic alloy are found to be independent of D/H atomic ratio and a function of only temperature. Tritium recovery apparatuses of a liquid-gas counter-current flowing bed or a T permeation window are designed in the liquid blanket system composed of the primary coolant of Flibe or LiPb and the secondary coolant of He or water. Simultaneous transfer of heat and T needs new configuration in order to satisfy the necessary conditions of low T leak and high heat transfer, which make it possible to operate the blanket effectively and safely. Tritium permeation rates through the primary LiPb or Flibe breeder and secondary water or He coolant loops are estimated, which is the most important factor to operate the blanket safely.

Numerical Simulation of the Mechanically Pumped Two-Phase Cooling Loop

Two-phase mechanically pumped cooling loop (MPCL) has emerged as a highly effective means for dissipating large amounts of heat from a small heat transfer area and provided a robust solution for significant design with flexibility, precise temperature control, and is othermalization. In this paper, the attempt of design optimization of the work fluid is introduced, by employing thermal simulation analysis with SINDA/FLUINT.

Application of PACTITER v3.3 Code to the ACPs Assessment of ITER Neutral Beam Injectors Primary Heat Transfer System

Activated Corrosion Products (ACPs) will be present in the various coolant loops of ITER: in-vessel and vacuum vessel, test blanket modules, auxiliary heating or diagnostic equipments. ACPs impact occupational exposure, routine effluents to the environment, and potential releases during accidents. Hence, the ACP inventory evaluation is an important task for ITER public and occupational safety.PACTITER v3.3 code is a computational tool derived from PACTOLE series of codes, modified in some modeling and computing capabilities. ITER Organization has included it as reference computer code for the ACP assessment. In the framework of its verification and validation activity, PACTITER v3.3 was used to assess the ACP inventory of the ITER Neutral Beam Injectors (NBIs) Primary Heat Transfer System (PHTS). This paper will document the preliminary results of this assessment, focusing on the impact of operation scenarios parameters (i.e. water chemistry, materials corrosion properties, etc.) and piping architecture.

Elastic-Plastic Analysis of the Transition Joint for a High Performance Divertor Target Plate

The use of tungsten as a plasma-facing material necessitates a transition joint to the structural material of the primary coolant loop at some location in order to transport the coolant to the heat exchanger. A critical issue in transition joints is the thermal expansion mismatch between materials, which can lead to unacceptably high thermal stresses. Detailed 2D and 3D analyses were performed to study the behavior of a transition from tungsten to ferritic steel (FS) with an intermediate layer of tantalum, located outside of the high heat flux region. This paper describes the results of FEM analyses including primary and secondary stresses under various time-dependent loading conditions such as warm and cold shutdown, and allowing for inelastic behaviors leading to stress relaxation and ratcheting. The results show that the transition joint satisfies the design requirement on maximum accumulated principal strain during operation.

Development and lab-test of a mobile adsorption air-conditioner

This paper is focused on the description of a mobile adsorption chiller for cooling in a truck cabin and its experimental performance. The prototype, designed and built at CNR – ITAE, consists of a double-bed adsorber connected with an evaporator and a condenser and driven by the low grade thermal energy coming from the engine coolant loop. Overall volume and weight of the machine are 170 dm3 and 60 kg, respectively. Performance assessment was preliminarily carried out at ITAE laboratory under the typical EU car air conditioning conditions. The experimental results obtained showed that the system is able to deliver an Average Cooling Power ACP = 1–2.3 kW and a Cooling COP = 0.25–0.45, at a desorption temperature of about 90 °C (Tev ∼ 8–14 °C and Tcon ∼ 28–35 °C). Afterwards, the cooling system was integrated in a real truck cabin (IVECO STRALIS) and tested on a laboratory test bench, where the engine cooling loop was simulated through an ideal heat source, so demonstrating that the chiller is able to provide cooling comfort to the passengers.

Water radiolysis in fusion neutron environments

Cooling loop water chemistry is important for it dictates corrosion, and key safety and environmental issues. Martensitic and austenitic steels with plasma facing components (first wall, divertor), blankets, and vacuum vessel, have been used in the design of PPCS (power plant conceptual studies), DEMO and ITER. Under radiolysis, a rise in oxidation potential enhances material corrosion, with a corresponding increase in build-up of gamma field from transport of activated corrosion products to the heat-exchanger/boiler units. The cooling system requires scheduled maintenance that involves component checking and changing. Workers could potentially be exposed to an unusual ORE (occupational radiological exposure) due to deposits of active corroded material around the cooling loop. The activation transport and deposition code transport of activation (TRACT) uses approximate concentrations (based on empirical data) of injected hydrogen for suppression of radiolysis in fusion power plant cooling channels. While it is possible to measure at some point outside the radiation-zone the concentrations of the stable end-products, their concentrations inside the radiation-zone, along with the concentrations of the short-lived intermediates, can only be estimated through computer simulation. This paper presents the time–dependent transport code water radiolysis H2ORAD with example calculations, based on ITER and generic Power Plant studies. H2ORAD will be used to replace the empirical data of injected hydrogen for suppression of radiolysis in TRACT.

Transient behaviour of a parallel pump in nuclear research reactors

A theoretical analysis of the fast transients of a parallel pump, based on inertia of the rotating parts and inertia of the fluid, is proposed. It leads to total torque, total head, and total system resistance during transient periods. The equations indicate that an increase in coolant inertia increases the acceleration head. While an increase in the moment of inertia of rotating parts decreases the acceleration head. The model is used to analyze the behaviour of the Tehran Research Reactor (TRR) primary coolant loop parallel pump during a fast startup. The results of present model are compared with similar studies and good agreement is observed.

NaK release model for MASTER-2009

Sodium–potassium droplets from the primary coolant loop of Russian orbital reactors have been released into space. These droplets are called NaK droplets. Sixteen nuclear powered satellites of the type RORSAT launched between 1980 and 1988 activated a reactor core ejection system, mostly between 900 and 950km altitude. The core ejection causes an opening of the primary coolant loop. The liquid coolant consists of eutectic sodium–potassium alloy and has been released into space during these core ejections. The NaK coolant has been forming droplets up to a diameter of 5.5cm. NaK droplets have been modeled before in ESA's MASTER Debris and Meteoroid Environment Model. The approach is currently revised for the MASTER-2009 upgrade. A mathematical improvement is introduced by substituting the current size distribution function by the modified Rosin–Rammler equation. A bimodal size distribution is derived which is based on the modified mass based Rosin–Rammler equation. The equation is modified by truncating the size range and normalizing over the finite range between the size limits of the smallest and the biggest droplet. The parameters of the model are introduced and discussed. For the validation of the NaK release model, sixteen release events are simulated. The resulting size distribution is compared with radar measurement data. The size distribution model fits well with revised published measurement data of radar observations. Results of orbit propagation simulation runs are presented in terms of spatial density.

ANALYSIS OF RESIDUAL STRESS FOR NARROW GAP WELDING USING FINITE ELEMENT METHOD

Reactor coolant loop (RCL) pipes circulating the heat generated in a nuclear power plant consist of so large diameter pipes that the installation of these pipes is one of the major construction processes. Conventionally, a shield metal arc welding (SMAW) process has been mainly used in RCL piping installations, which sometimes caused severe deformations, dislocation of main equipments and various other complications due to excessive heat input in welding processes. Hence, automation of the work of welding is required and narrow-gap welding (NGW) process is being reviewed for new nuclear power plants as an alternative method of welding. In this study, transient heat transfer and thermo-elastic-plastic analyses have been performed for the residual stress distribution on the narrow gap weldment of RCL by finite element method under various conditions including surface heat flux and temperature dependent thermo-physical properties.

A semi-empirical approach for predicting two-phase flow discharge through branches of various orientations connected to a horizontal main pipe

The subdivision of two-phase flow in branching conduits consisting of a large horizontal main pipe with upward, downward, or lateral branches of reduced diameter is of great interest in various technological fields. For example, these conduits are important in light-water nuclear reactors (LWRs) in the case of a small break loss-of-coolant accident (SBLOCA) in a leg of the reactor's primary coolant loops, as well as for breaks or valve malfunctions in a large pipeline. In these kinds of circumstances, the relevant phenomenology often involves phase stratification coupled with possible liquid entrainment or gas pool-through phenomena. Therefore, these phenomena were studied in depth to evaluate the pressure drop across conduit elements resulting from the redistribution of flow phases and the discharged mass among them. In the past, several experiments have been performed along with studies in flow modelling. As a result, several formulae and models for branch exit quality and consequent discharge mass flow rate predictions have been proposed. In recent years, we have been engaged in extensive research on this subject, resulting in a new semi-empirical formulation to express branch exit quality in terms of the physical processes taking place in the conduits, the operating parameters and the branch geometry. In this paper, we applied these correlations to representative sets of experiments previously reported in the literature, comprising a wide range of branch-to-main-pipe diameter ratios and operating pressures, which proved our formulation to be very accurate.

Modelling of an adsorption system driven by engine waste heat for truck cabin A/C. Performance estimation for a standard driving cycle

Abstract This paper presents the main characteristics of an innovative cooling system for the air conditioning of a truck cabin, as well as a first estimation of its performance during a standard driving cycle, obtained with a specifically developed vehicle-engine-cooling system overall model. The innovative cooling system consists of a water–zeolite adsorption–desorption system, which employs the waste heat from the engine to produce the cooling of the vehicle cabin. The developed global model is completely dynamic and is able to: reproduce the operation of the engine through a standard driving cycle, evaluate the waste heat available at the engine hydraulic loop; calculate the sequential operation of an adsorption–desorption system, calculate the condensed water per cycle, the cooling effect produced at the evaporator, and finally, the temperature and humidity evolution of the air inside the cabin. The model was validated by experimental data. The experimental tests were performed in a lab-scale adsorption chiller prototype specifically designed and realized to be driven by the low grade waste heat (80–90 °C) from the engine coolant loop of a truck. The experimental activity carried out showed that the chiller is able to generate up to 5 kW of peak cooling power at 10 °C (35 °C of condensation temperature) with a COP of 0.6. The obtained results show that the system could be able to provide a significant amount of the required cooling.

Titanium-Water Loop Heat Pipe Operating Characteristics Under Standard and Elevated Acceleration Fields

An experiment has been developed to examine the behavior of a titanium-water loop heat pipe under standard and elevated acceleration fields. The loop heat pipe was mounted on a 2.44 meter-diameter centrifuge table on edge with heat applied to the evaporator via a mica heater and heat rejected using a high-temperature polyalphaolefin oil coolant loop. The loop heat pipe was tested under the following parametric ranges: heat load at the evaporator 100 ≤ Q in ≤ 600 W, heat load at the compensation chamber 0 ≤ Q cc ≤ 50 W, radial acceleration 0 ≤ a r ≤ 10 g. For stationary operation, the evaporative heat transfer coefficient decreased monotonically with heat load whereas the thermal resistance decreased to a minimum then increased. Heat input to the compensation chamber was found to increase the evaporative heat transfer coefficient and decrease the thermal resistance for Q in = 500 W. Transient periodic flow reversal in the loop heat pipe was found for some cases, which was likely due to vapor bubble formation in the primary wick. Operation in an elevated acceleration environment revealed that dry out was dependent on both Q in and a r , and the ability for the loop heat pipe to reprime after an acceleration event that induced dry out was influenced by the evaporator temperature. The evaporative heat transfer coefficient and thermal resistance were found not to be significantly dependent on radial acceleration. However, the evaporator wall superheat was found to increase slightly with radial acceleration at high heat loads.

Study on heat transfer and pressure drop characteristics of internal heat exchangers in CO2 system under cooling condition

In order to study the heat transfer and pressure drop on four types of internal heat exchangers (IHXs) of a CO2 system, the experiment and numerical analysis were performed under a cooling condition. The configuration of the IHXs was a coaxial type and a micro-channel type. Two loops on the gas cooler part and the evaporator part were made, for experiment. And the section-by-section method and Hardy-Cross method were used for the numerical analysis. The capacity and pressure drop of the IHX are larger at the micro-channel type than at the coaxial type. When increasing the mass flow rate and the IHX length the capacity and pressure drop increase. The pressure drop of the evaporator loop is much larger than that of the gas cooler loop. The performance of the IHX was affected with operating condition of the gas-cooler and evaporator. The deviations between the experimental result and the numerical result are about ±20% for the micro-channel type and ±10% for the coaxial type. Thus, the new CO2 heat transfer correlation should be developed to precisely predict a CO2 heat transfer.

Estimation of the Doppler Coefficient from a Lower Power Transient Observed in a Zero-Power Reactor Physics Test of a PWR (I)—Methodology—

A method for estimating the Doppler coefficient from low power transient data that are measured in a routine experiment conducted in a zero-power reactor physics test of a PWR power plant has been developed. A dynamics identification model was introduced by considering two dynamics components for reactor kinetics and heat removal in the primary coolant loops. The dynamics identification method in the time domain was applied under consideration that the effect of reactivity feedback observed in the experiment was very small and measured transient data contained some intermittent intervals unavailable for dynamics identification. Unknown parameters, including the Doppler coefficient, contained in the dynamics model were determined stepwise through five estimation steps. In these steps, gamma ray noises were removed from the NIS signal, the reactivity feedback component was extracted from reactivity transient, and fuel temperature transient was determined in such way as to reproduce the observed low power transient in the digital simulation. In the final step, the Doppler coefficient was estimated from the extracted reactivity feedback component using the quantities evaluated in the preceding steps. The estimated value agreed well with the value evaluated with a core design code.

Evaluating accuracy of standard homogenization and need for generalized equivalence theory for ACR ®-lattice checkerboard configurations

The Advanced CANDU ® 1 CANDU ® (CANada Deuterium Uranium) is a registered trademark of Atomic Energy of Canada Limited (AECL). 1 Reactor ACR ® 2 ACR ® (Advanced CANDU Reactor) is a registered trademark of Atomic Energy of Canada Limited (AECL). 2 is a new type of CANDU reactor that is slightly-enriched-uranium fuelled, light-water cooled, and heavy-water moderated. The characteristics of the ACR lattice therefore differ somewhat from those of the standard (natural-uranium fuelled and heavy-water cooled) CANDU lattice. Consequently, the validity of the older neutronic-analysis methodology employed for the natural-uranium CANDU lattice and, in particular, that of the standard homogenization method used for that type of lattice, need to be investigated for the ACR lattice. Of particular interest are configurations in which neutronic properties of lattice cells alternate in a checkerboard pattern. Such patterns occur in CANDU reactors because of alternating fuelling directions in neighbouring channels and because neighbouring channels belong to different coolant loops or different passes within the same loop. The work presented in this paper assesses the accuracy of the standard homogenization method when applied to checkerboard ACR-lattice configurations and investigates whether the use of Generalized Equivalence Theory (GET) yields more accurate results. The conclusion is that the most accurate results, aside from using exact (or, at least, leakage-corrected) macroscopic cross sections and discontinuity factors, can be obtained by using exact (or, at least, leakage-corrected) macroscopic cross sections and no discontinuity factors. The expected errors in the latter case, assuming a 6-cm or finer Cartesian mesh, are less than 1% in bundle power and less than 1 mk in k eff .

Estimation of the Doppler Coefficient from a Lower Power Transient Observed in a Zero-Power Reactor Physics Test of a PWR (I) —Methodology—

A method for estimating the Doppler coefficient from low power transient data that are measured in a routine experiment conducted in a zero-power reactor physics test of a PWR power plant has been developed. A dynamics identification model was introduced by considering two dynamics components for reactor kinetics and heat removal in the primary coolant loops. The dynamics identification method in the time domain was applied under consideration that the effect of reactivity feedback observed in the experiment was very small and measured transient data contained some intermittent intervals unavailable for dynamics identification. Unknown parameters, including the Doppler coefficient, contained in the dynamics model were determined stepwise through five estimation steps. In these steps, gamma ray noises were removed from the NIS signal, the reactivity feedback component was extracted from reactivity transient, and fuel temperature transient was determined in such way as to reproduce the observed low power transient in the digital simulation. In the final step, the Doppler coefficient was estimated from the extracted reactivity feedback component using the quantities evaluated in the preceding steps. The estimated value agreed well with the value evaluated with a core design code.

Experimental investigation of parameters influencing the freeze start ability of a fuel cell system

To improve the freeze start ability of a fuel cell system some significant influencing parameters are defined and investigated. Experiments with a fuel cell test system are carried out in a climate chamber at various conditions. The time interval until fuel cell stack power equals 50% of its maximum power is defined as an indicator for a successful freeze start as well as a value for comparison and evaluation of the results. The target of this work is the minimization of this freeze start time by avoiding the freezing of process water on the catalyst layer of the Membrane Electrode Assembly (MEA), since this leads to temporary performance losses. The shut down strategy of the fuel cell system is identified to be one of the main parameters influencing the freeze start. It is found that a higher degree of dryness in the stack leads to a significant improvement in the freeze start performance, since the water absorbing capacity of the membrane increases and therefore also the time until its saturation. If this saturation takes place after the temperature of the MEA reached 0 °C, no significant ice-formation occurs. It is shown that by improving the shut down strategy of the fuel cell system at TStart = −6 °C a start without performance loss can be realized. At temperatures lower than that temporary performance losses occur. Even if a lower voltage leads to a higher current and therefore to a higher water production rate, its effect on the freeze start due to the increased heat of reaction is positive. Further investigated parameters, for example the volume of the coolant loop, also affect the freeze start ability, but it can be concluded that the shut down strategy is of main importance.

Commissioning of the ATLAS thermal-hydraulic integral test facility

KAERI recently constructed a new thermal-hydraulic integral test facility for advanced pressurized water reactors (PWRs) – ATLAS. The ATLAS facility has the following characteristics: (a) 1/2-height&length, 1/288-volume, and full pressure simulation of APR1400, (b) maintaining a geometrical similarity with APR1400 including 2(hot legs) × 4(cold legs) reactor coolant loops, direct vessel injection (DVI) of emergency core cooling water, integrated annular downcomer, etc., (c) incorporation of specific design characteristics of OPR1000 such as cold leg injection and low-pressure safety injection pumps, (d) maximum 10% of the scaled nominal core power. The ATLAS will mainly be used to simulate various accident and transient scenarios for evolutionary PWRs, OPR1000 and APR1400: the simulation capability of broad scenarios including the reflood phase of a large-break loss-of-coolant accident (LOCA), small-break LOCA scenarios including DVI line breaks, a steam generator tube rupture, a main steam line break, a feed line break, a mid-loop operation, etc. The ATLAS is now in operation after an extensive series of commissioning tests in 2006.

Modular front-end cooling module architecture for future powertrains

Cooling system development is driven by upcoming CO2 reduction legislation. Valeo highlights an innovative cooling system architecture, which uses two modular main coolant radiators in the vehicle frontend. This newly developed “Ultimate Cooling” system has both low temperature and high temperature coolant loops, which are connected with each other. It allows together with an adapted control strategy to optimize the available heat dissipation capacity installed by the radiators in the vehicle front end. Results of measurements and simulation show the advantages of this new system in a demo-car with a Euro-4 diesel engine (50 kW/l) and a highly charged (80 kW/l) Euro-6 diesel engine.

Performance of the Extravehicular Mobility Unit (EMU) Airlock Coolant Loop Remediation (A/L CLR) Hardware

An EMU water processing kit (Airlock Coolant Loop Recovery A/L CLR) was developed as a corrective action to Extravehicular Mobility Unit (EMU) coolant flow disruptions experienced on the International Space Station (ISS) in May of 2004 and thereafter. Conservative schedules for A/L CLR use and component life were initially developed and implemented based on prior analysis results and analytical modeling. The examination of postflight samples and EMU hardware in November of 2006 indicated that the A/L CLR kits were functioning well and had excess capacity that would allow a relaxation of the initially conservative schedules of use and component life. A relaxed use schedule and list of component lives was implemented thereafter. Since the adoption of the relaxed A/L CLR schedules of use and component lives, several A/L CLR kit components, transport loop water samples and sensitive EMU transport loop components have been examined to gage the impact of the relaxed requirements. The intent of this paper is to summarize the findings of that evaluation, and to outline updated schedules for A/L CLR use and component life.