Intermediate Heat Transport System Research Articles

Design Features and Simulation of the New-Build HELOKA-US Facility for the Validation of the DEMO Helium-Cooled Pebble Bed Intermediate Heat Transport and Storage System

For the EU-DEMO Helium-Cooled Pebble Bed (HCPB) concept, an indirect coupled design (ICD) with a molten salt (MS) loop as an intermediate heat transport and storage system (IHTS) is considered for the conceptual design phase. The IHTS with an energy storage decouples the primary heat transport system (PHTS) that undergoes pulse and dwell power cycles from the power conversion system (PCS), and thus can provide stable power to the turbine and grid. However, the maintenance of stable He and MS parameters during transitions from dwell to pulse and vice versa is challenging for the design of the MS loop, and the real performance of the helium–MS heat exchanger (He/MS HX) shall be verified. To investigate such components and conditions, a new R&D infrastructure HELOKA-US (Helium Loop Karlsruhe—Upgrade Storage) is under construction for the validation of prototypical components and the MS loop operation under stationary and transitional conditions. This paper provides the design features of Phase 1a of the project and the simulation results with EBSILON on the power generation phase.

Analysis of EU-DEMO WCLL Power Conversion System in Two Relevant Balance of Plant Configurations: Direct Coupling with Auxiliary Boiler and Indirect Coupling

Among the Key Design Integration Issues (KDIIs) recently selected for the DEMOnstration Fusion Power Plant (DEMO), the operation of the Balance of Plant (BoP) Power Conversion System (PCS) has been recognized as a crucial aspect, due to the typical pulsed regime characterizing the fusion power plant. In the framework of the DEMO Water-Cooled Lead-Lithium Breeding Blanket (WCLL BB) concept, three BoP solutions have been recognized to be able to overcome this issue. They rely on different coupling options between the Primary Heat Transfer Systems (PHTSs) and the PCS: an Indirect Coupling Design (ICD) with Intermediate Heat Transport System (IHTS) and Energy Storage System (ESS), a Direct Coupling Design (DCD) with AUXiliary Boiler (AUXB), and a DCD with small ESS. The present paper deals with a preliminary feasibility assessment of the first two solutions. The analysis, carried out with the GateCycleTM code, referred to a preliminary design phase, devoted to the sizing of the main components, and to a second phase focused on the cycle optimization. The study demonstrated the feasibility of the two BoP concepts. They are able to produce a satisfactory average electric power (>700 MW) with an acceptable average net electric efficiency (33.6% for both concepts). For each solution, the main strengths and weaknesses are compared and discussed.

Numerical analysis of the temperature distribution of the EM pump for the sodium thermo-hydraulic test loop of the GenIV PGSFR

The temperature distribution of an electromagnetic pump was analyzed with a flow rate of 1380 L/min and a pressure of 4 bar designed for the sodium thermo-hydraulic test in the Sodium Test Loop for Safety Simulation and Assessment-Phase 1 (STELLA-1). The electromagnetic pump was used for the circulation of the liquid sodium coolant in the Intermediate Heat Transport System (IHTS) of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) with an electric power of 150 MWe. The temperature distribution of the components of the electromagnetic pump was numerically analyzed to prevent functional degradation in the high temperature environment during pump operation. The heat transfer was numerically calculated using ANSYS Fluent for prediction of the temperature distribution in the excited coils, the electromagnet core, and the liquid sodium flow channel of the electromagnetic pump. The temperature distribution of operating electromagnetic pump was compared with cooling of natural and forced air circulation.The temperature in the coil, the core and the flow gap in the two conditions, natural circulation and forced circulation, were compared. The electromagnetic pump with cooling of forced circulation had better efficiency than natural circulation even considering consumption of the input power for the air blower. Accordingly, this study judged that forced cooling is good for both maintenance and efficiency of the electromagnetic pump.

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.

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.

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.

Magnetic field analysis of an electromagnetic pump for sodium thermohydraulic test in the Sodium Test Loop for Safety Simulation and Assessment — Phase 1

This study analyzed the magnetic field on an electromagnetic pump with a flow rate of 1380 L/min and a developed pressure of 4 bar designed for the sodium thermohydraulic test in the Sodium Test Loop for Safety Simulation and Assessment — Phase 1 (STELLA-1). The electromagnetic pump was used for the circulation of the sodium coolant in the intermediate heat transport system (IHTS) of the Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) with an electric power of 150 MWe, which is being developed in Korea. The magnetic field distribution in the narrow channel and the inner core of the electromagnetic pump was calculated through a numerical analysis method using ANSYS Maxwell to predict its performance characteristics. The end effect by the magnetic field distortion at the pump entrance and exit caused by the finite core length of the electromagnetic pump, which could result in the decrease of the developed pressure, was analyzed. The end effect analysis of the magnetic field considered the modified winding of coils in the outer core of the electromagnetic pump for the end effect reduction. The results showed that the electromagnetic pump performance would be thus improved.

Validation of intermediate heat and decay heat exchanger model in MARS-LMR with STELLA-1 and JOYO tests

The MARS-LMR code has been developed by the Korea Atomic Energy Research Institute (KAERI) to analyze transients in a pool-type sodium-cooled fast reactor (SFR). Currently, KAERI is developing a prototype Gen-IV SFR (PGSFR) with metallic fuel. The decay heat exchangers (DHXs) and the intermediate heat exchangers (IHXs) were designed as a sodium-sodium counter-flow tube bundle type for decay heat removal system (DHRS) and intermediate heat transport system (IHTS), respectively. The IHX and DHX are important components for a heat removal function under normal and accident conditions, respectively. Therefore, sodium heat transfer models for the DHX and IHX heat exchangers were added in MARS-LMR. In order to validate the newly added heat transfer model, experimental data were obtained from the JOYO and STELLA-1 facilities were analyzed. JOYO has two different types of IHXs: type-A (co-axial circular arrangement) and type-B (triangular arrangement). For the code validation, 38 and 39 data points for type A and type B were selected, respectively. A DHX performance test was conducted in STELLA-1, which is the test facility for heat exchangers and primary pump in the PGSFR. The DHX test in STELLA-1 provided eight data points for a code validation. Ten nodes are used in the heat transfer region is used, based on the verification test for the heat transfer models. RMS errors for JOYO IHX type A and type B of 19.1% and 4.3% are obtained, respectively. It was reported that a bypass flow of 20% was allowed in type A test. When the bypass flow model was added to MARS-LMR, the RMS error for type A reduced to 12.6%. The RMS error for the STELLA-1 DHX was estimated to be 3.5%. Through the comparative studies with various alternative heat transfer models in the tube and shell-side, the base heat transfer model in MARS-LMR was found to give good prediction for the STELLA-1 and JOYO experiment results.

소듐냉각고속로 원형로 중간열전달계통 고온배관의 파단전누설 예비평가

소듐냉각고속로 원형로 중간열전달계통 고온배관의 파단전누설 예비평가

전산유체역학을 이용한 소듐-소듐 열교환기 설계코드의 검증

A Prototype Gen-IV Sodium-cooled Fast Reactor which is one of the <TEX>$4^{th}$</TEX> generation nuclear reactors is in development by Korea Atomic Energy Research Institute. The reactor is composed of four main fluid systems which are categorized by its functions, i.e., Primary Heat Transport System, Intermediate Heat Transport System, Decay Heat Removal System and Sodium-Water Reaction Pressure Relief System. The coolant of the reactor is liquid sodium and sodium-to-sodium heat exchangers are installed at the interfaces between two fluid systems, Intermediate Heat Exchangers between the Primary Heat Transport System and the Intermediate Heat Transport System and Decay Heat Exchangers between the Primary Heat Transport System and the Decay Heat Removal System. For the design and performance analysis of the Intermediate Heat Exchanger and the Decay Heat Exchanger, a computer code was written during previous step of research. In this work, the computer code named "SHXSA" has been validated preliminarily by computational fluid dynamics simulations.

Thermal hydraulic performance analysis of a double tube bundle steam generator for a liquid metal reactor

In this study, thermal-hydraulic performance of a double tube bundle steam generator (DTBSG) using helically coiled tubes was analyzed numerically. For this purpose a one-dimensional thermal-hydraulic analysis computer program was developed. The DTBSG contains hot fluid and cold fluid tube bundles, a medium fluid and a pump for a medium fluid circulation in an SG. Heat is transferred from the hot side to the cold side through a medium fluid which fills the SG shell side. The fluids of the hot, cold and medium side are sodium, water/steam, lead–bismuth respectively. The assessment was made for four different types of the DTBSG at 200 MWt. It showed there is an optimum medium flow rate for each type and the separated bundle type have the best heat transfer efficiency (heat transfer rate per unit area) among the various DTBSG. However, it requires quite a large pumping power and the velocity of the medium fluid. The effect of a medium flow rate change is relatively small in the integrated bundle types while it is very large in the separated type. The developed program can be used for the optimum thermal design of a double tube bundle steam generator, which is one of the candidates for an Intermediate Heat Transport System (IHTS) simplification concept.

Impact from the Adoption of Advanced Materials on a Sodium Fast Reactor Design

To improve the economic characteristics of fast reactors, researchers are developing advanced structural materials for application to reactor components. These advanced materials provide higher strength at elevated temperatures. Coupled thermal-hydraulic and structural analyses have been carried out to investigate the benefits of the advanced structural materials for a specific fast reactor design: the Advanced Burner Reactor (ABR) developed at Argonne National Laboratory. The benefits of the advanced materials, in terms of increased design margins, possible longer lifetime, thinner structures, and higher operating temperatures, were calculated for the major ABR structural components, including the reactor vessel, the core support structure, the intermediate heat exchanger, the intermediate heat transport system piping, and the steam generator. For each structure, the possible reduction in the component thickness was calculated and was converted into estimates of the commodities savings provided by the use of the advanced materials. Overall, a significant material mass saving of ~40% was calculated for the considered fast reactor structures.

ICONE19-43935 Development of a Helical-Coil Double Wall Tube Steam Generator for 4S Reactor

The 4S, Super-Safe Small and Simple, is a small-sized sodium-cooled fast reactor. A fast reactor usually uses sodium as a coolant to transfer heat from core to turbine/generator system. The heat of the intermediate heat transport system and that of the water steam systems are exchanged by the steam generator (SG) tubes. If the tube failure occurs, a sodium/water reaction could be occurred. To prevent the reaction and enhance safety, a helical-coil-type double wall tube with wire mesh interlayer and continuous monitoring systems of tube failure are applied to the SG of the 4S. The development and general features of this type double wall tube were described in Ref.1) and Ref.2). Those paper summarized following results; The tubes studied in these references were straight type. To establish this SG, development of manufacturing method of helical-coil-type double wall tube and validation of the tube failure monitoring system are needed. In this study, three demonstration tests have been performed; welding test of the double wall tube to manufacture the tubes with 70-80m length, assembling test of the helical-coil tube, and confirmation test of the tube processing system using the fabricated helical-coil tubes. As a result, following technologies have been successfully established. (1) Development of the welding techniques for manufacturing of the helical-coil-type double wall tube with wire mesh interlayer. (2) The confirmation test for manufacturing the helical coil tube of the SG.

DESIGN STUDY OF AN IHX SUPPORT STRUCTURE FOR A POOL-TYPE SODIUM-COOLED FAST REACTOR

The IHX (Intermediate Heat eXchanger) for a pool-type SFR (Sodium-cooled Fast Reactor) system transfers heat from the primary high temperature sodium to the intermediate cold temperature sodium. The upper structure of the IHX is a coaxial structure designed to form a flow path for both the secondary high temperature and low temperature sodium. The coaxial structure of the IHX consists of a central downcomer and riser for the incoming and outgoing intermediate sodium, respectively. The IHX of a pool-type SFR is supported at the upper surface of the reactor head with an IHX support structure that connects the IHX riser cylinder to the reactor head. The reactor head is generally maintained at the low temperature regime, but the riser cylinder is exposed in the elevated temperature region. The resultant complicated temperature distribution of the co-axial structure including the IHX support structure may induce a severe thermal stress distribution. In this study, the structural feasibility of the current upper support structure concept is investigated through a preliminary stress analysis and an alternative design concept to accommodate the IHTS (Intermediate Heat Transport System) piping expansion loads and severe thermal stress is proposed. Through the structural analysis it is found that the alternative design concept is effective in reducing the thermal stress and acquiring structural integrity.

DEVELOPMENT OF A TWO-DIMENSIONAL THERMOHYDRAULIC HOT POOL MODEL AND ITS EFFECTS ON REACTIVITY FEEDBACK DURING A UTOP IN LIQUID METAL REACTORS

The existence of a large sodium pool in the KALIMER, a pool-type LMR developed by the Korea Atomic Energy Research Institute, plays an important role in reactor safety and operability because it determines the grace time for operators to cope with an abnormal event and to terminate a transient before reactor enters into an accident condition. A two-dimensional hot pool model has been developed and implemented in the SSC-K code, and has been successfully applied for the assessment of safety issues in the conceptual design of KALIMER and for the analysis of anticipated system transients. The other important models of the SSC-K code include a three-dimensional core thermal-hydraulic model, a reactivity model, a passive decay heat removal system model, and an intermediate heat transport system and steam generation system model. The capability of the developed two-dimensional hot pool model was evaluated with a comparison of the temperature distribution calculated with the CFX code. The predicted hot pool coolant temperature distributions obtained with the two-dimensional hot pool model agreed well with those predicted with the CFX code. Variations in the temperature distribution of the hot pool affect the reactivity feedback due to an expansion of the control rod drive line (CRDL) immersed in the pool. The existing CRDL reactivity model of the SSC-K code has been modified based on the detailed hot pool temperature distribution obtained with the two-dimensional pool model. An analysis of an unprotected transient over power with the modified reactivity model showed an improved negative reactivity feedback effect.

A Comparison Study of Creep-Fatigue Defect Growth Evaluations for a SFR IHTS Piping

A defect behavior is one of the principal concerns to be dealt with for a structural integrity since defects may lead to the failure of a SFR(Sodium-cooled Fast Reactor) structure under high temperature loading conditions. A thin-walled and large diametric piping is adopted for the IHTS (Intermediate Heat Transport System) piping system of a SFR to enhance the system's economy and to reduce the thermal stress level. The structural material for the piping is Type 316 austenitic stainless steel. In this study, a defect growth evaluation of IHTS piping was performed for a semi-elliptical surface defect subjected to a combined mechanical loading and thermal loading of a high temperature above 500°C including a cyclic loading condition corresponding to a reactor refueling cycle. The creep-fatigue defect behavior evaluations were carried out by following the French Assessment Procedure A16, the UK Assessment Procedure R5 and the JNC procedure. The results were compared with each other and their strong and weak points were discussed. Also the degree of conservatism of each procedure was reviewed.

CONCEPTUAL DESIGN OF THE SODIUM-COOLED FAST REACTOR KALIMER-600

The Korea Atomic Energy Research Institute has developed an advanced fast reactor concept, KALIMER-600, which satisfies the Generation IV reactor design goals of sustainability, economics, safety, and proliferation resistance. The concept enables an efficient utilization of uranium resources and a reduction of the radioactive waste. The core design has been developed with a strong emphasis on proliferation resistance by adopting a single enrichment fuel without blanket assemblies. In addition, a passive residual heat removal system, shortened intermediate heat-transport system piping and seismic isolation have been realized in the reactor system design as enhancements to its safety and economics. The inherent safety characteristics of the KALIMER-600 design have been confirmed by a safety analysis of its bounding events. Research on important thermal-hydraulic phenomena and sensing technologies were performed to support the design study. The integrity of the reactor head against creep fatigue was confirmed using a CFD method, and a model for density-wave instability in a helical-coiled steam generator was developed. Gas entrainment on an agitating pool surface was investigated and an experimental correlation on a critical entrainment condition was obtained. An experimental study on sodium-water reactions was also performed to validate the developed SELPSTA code, which predicts the data accurately. An acoustic leak detection method utilizing a neural network and signal processing units were developed and applied successfully for the detection of a signal up to a noise level of -20 dB. Waveguide sensor visualization technology is being developed to inspect the reactor internals and fuel subassemblies. These research and developmental efforts contribute significantly to enhance the safety, economics, and efficiency of the KALIMER-600 design concept.

ICONE15-10701 The Comparison Study of the Creep-Fatigue Defect Growth Evaluations for a SFR IHTS Piping

The defect behavior is one of the principal concerns to be dealt with for the structural integrity since defects may lead to the failure of SFR(Sodium-cooled Fast Reactor) structure under high temperature loading conditions. A thin-walled and large diametric piping is adopted for the IHTS (Intermediate Heat Transport System) piping system of SFR to enhance the system economy and to reduce the thermal stress. The structural material for the piping is Type 316 austenitic stainless steel. In this study, the defect growth evaluation of IHTS piping was performed for the semi-elliptical surface defect subjected to the combined mechanical loading and the thermal loading of a high temperature above 500・・ including a cyclic loading condition corresponding to the reactor refueling cycle. The creep-fatigue defect behavior evaluations are carried out following the French Assessment Procedure A16, the UK Assessment Procedure R5/R6 and the Japanese procedure. The results are compared each other and the strong and weak points are discussed. Also the degree of conservatism of each procedure is reviewed.

An analytical model for dynamics of a sodium/water reaction bubble in an LMFBR steam generator and the coupled response of the intermediate heat transport system

An analytical model and computational methods are presented to predict the dynamics of a sodium/water reaction bubble in the bulk of liquid sodium in the steam generator of a liquid metal reactor. The rate of reaction is based on a simple model including a parameter that must be deduced from test data. The computations include the coupled fluid response of the intermediate heat transport system. The analytical model was programmed in a computer code, and the computer code was used to analyze the results of large-scale SWR mockup tests. Comparison of the analytical results and the test data indicated generally good agreement despite the simplicity adopted in the reaction rate equation. Representative comparison results are presented and discussed. Implication of the empirical parameter included in the analytical model is discussed also.