Effect of ideal internal MHD instabilities on NBI fast ion redistribution in ITER 15 MA scenario

Transport of fast ions is a crucial issue during the operation of ITER. Redistribution of neutral beam injection(NBI) fast ions by the ideal internal magnetohydrodynamic(MHD) instabilities in ITER is studied utilizing the guiding-center code ORBIT(White R B and Chance M S 1984Phys. Fluids 27 2455). Effects of the perturbation amplitude A of the internal kink, the perturbation frequency f of the fishbone instability, and the toroidal mode number n of the internal kink are investigated, respective...     

Quench analysis of an ITER TF coil

The first realistic application of the recently developed 4C code is presented, aimed at showing its capability to simulate in an integrated fashion relevant transients for the superconducting coil operation in the International Thermonuclear Experimental Reactor (ITER), both at the system and at the conductor levels. The quench initiation and propagation in an ITER TF coil is considered, including the coil fast discharge phase until the opening of the relief valves. The 14 coil pancakes cooled by alternating clockwise/counter-clockwise SHe flow, the radial plates and the case, the 96 case cooling channels, and the external cryogenic circuit up to the LHe bath are included in the model. The results of the analysis are discussed with particular reference to quench propagation in the winding, hot spot temperature and peak pressurization, mass flow rate evolution in the different system components. The accuracy of the analysis is guaranteed by suitable convergence studies.     

Development of a thermal-hydraulic analysis code for CARR

The China advanced research reactor (CARR) being built in Beijing, China, is a multipurpose research reactor for a variety of fields. Theoretical calculation of thermal hydraulic characteristics of CARR is presented in this paper. The theoretical analysis consists of initial steady and transient accidental analyses. Point reactor neutron kinetics model with six groups of delayed neutron is adopted for the solution of reactor power. All possible flow and heat transfer conditions are considered and the corresponding optional models are supplied in the theoretical calculations. A new simple and convenient model is proposed for the resolution of the transient behaviors of main pump instead of the complicated four-quadrant model. Gear method and Adams predictor–corrector method are adopted alternately for a better solution to such ill-conditioned differential equations corresponding to detail process. The initial multi-channel analysis shows that the effects of geometrical size on flow distribution play dominant role and the effects of core power distribution may be neglected. The temperature fields of fuel elements under asymmetrical cooling condition are also obtained, which are the bases for further study on transient-induced stress analysis, etc. Accidental analyses show that the activity of emergency cooling system apparently reduces the peak temperatures of fuel and coolant, peak quality and other operation parameters. Thus it effectively ensures the safety in operation of CARR. Because of the adoption of modular programming techniques, this code is expected to be applied to accidental analysis of other types of reactors by easily modifying the corresponding function modules. Also, this code is expected to be validated against experimental data.     

Development of a thermal-hydraulic subchannel analysis code for motion conditions

Since the conventional subchannel analysis codes are designed for the land-based reactor core, a thermal-hydraulic subchannel analysis code was developed to evaluate thermal-hydraulic characteristics of the reactor core under motion conditions. The verification of the code was performed with experimental data and commercial codes. The ISPRA 16-rod tests were used to evaluate the steady-state prediction performance of the code, and the simulation results agree well with the test data. COBRA-EN code was applied to check the transient prediction performance of the code, and there is a good agreement between the predictions with both codes. An additional forces model for motion conditions was proposed in the code, and CFX-14.0 code was applied to verify the model. The results show that the code can be used in the thermal-hydraulic analysis of the reactor core under motion conditions. To illustrate the capabilities of the code, a fuel bundle under a complex motion condition was simulated, and the results are reasonable.     

Development of a thermal-hydraulic analysis code for annular fuel assemblies

A thermal-hydraulic analysis code which is capable of modeling both internally and externally cooled annular fuel pins was developed. The coolant flow distribution in the annular fuel-based assemblies is adjusted by a pressure drop model allowing for conditions such as non-equal velocity and non-saturated phases. The heat transfer fraction is determined by the ratio of cross-sectional areas distinguished by the radius at which the first derivative of the temperature within the annular fuel equals zero. The code predictions have been compared with calculations from Korea Atomic Energy Research Institute (KAERI) and MIT. The heat transfer fraction difference between the code and RELAP was about 3.9%, and the Departure from Nucleate Boiling Ratio (DNBR) prediction of the code agreed well with the MIT’s result in the region below 3 m. For the application of the code, thermal-hydraulics of thorium-based fuel assemblies loaded with annular seed pins were compared with those of the existing thorium-based assemblies. The pressure drop in the assembly generally increased in the case of annular fuel due to the larger wetted perimeter. In the inner subchannels of the seed pins, mass fluxes were high due to the grid form losses in the outer subchannels. About 43% of the heat generated from the seed pin flowed into the inner subchannel and the rest into the outer subchannel. The minimum DNBRs (MDNBRs) of the annular fuel-based assemblies were higher than those of the existing ones. Because interchannel mixing cannot occur in the inner subchannels, temperatures and enthalpies were higher in the inner subchannels.     

Thermal analysis of the cryostat feed through for the ITER Tokamak TF feeder

In Tokamaks,the toroidal field (TF) coil feeder is an important component that is used to supply the cryogens and electrical power for the TF coils.As a part of the TF feeder,the cryostat-feed through (CFT) is subject to low temperatures of 9 and 80 K inside and room temperature of 300 K outside.Based on the features of the International Thermonuclear Experimental Reactor TF feeder,the thermal performance of the CFT under the nominal conditions is studied.Taking into account the conductive,convective and radiation heat transfer,the finite element model of the CFT is built.Transient thermal analysis is performed to determine the temperatures of the CFT on the 9th day of cooldown.The model is assessed by comparing the cooling curves of the CFT after 9 days.If the simulation and experimental results are the same,the finite element model can be considered as calibrated.The model predicts that the cooling time will be approximately 26 days and the temperature distribution and heat load of the main components are obtained when the CFT reaches thermal equilibrium.This study provides a valid quantitative characterization of the CFT design.     

Status of the ITER magnets

The first 2 years of the ITER IO has seen substantial progress towards the construction of the magnets, in three main areas. Firstly, the design has been developed under the conflicting constraints to minimise construction costs and to maximise plasma physics performance. Building construction momentum while updating the design to take account of new physics assessments of the coil requirements has been challenging. Secondly, with a stabilising design, it has been possible for the Domestic Agencies to launch the first industrial procurement contracts. And thirdly, critical R&D to confirm the performance of the Nb3Sn cable in conduit design is proceeding successfully.The design consolidation has been accompanied by design reviews involving the international community. The reviews conducted by magnet experts have enabled a consensus to be built on choosing between some of the design options in the original ITER basic design in 2001. The major design decisions were to maintain the circular Nb₃Sn conductor embedded in radial plates for the toroidal field (TF) coils and to maintain NbTi-based conductors for the PF coils. Cold testing, at low current, is also being introduced for quality control purposes for all coils.     

Experimental validation of the thermal-hydraulic code SACATRI

A sub-channel analysis steady state thermal-hydraulic code (SACATRI) was developed for the Moroccan TRIGA MARK II research reactor. The main objective of the thermal-hydraulic study of the whole reactor core is to evaluate the main safety parameters of the reactor core, and to ensure that they are within the safety limits for any operating conditions. The thermal-hydraulic model used in SACATRI is based on four partial differential equations that describe the conservation of mass, energy and momentum. In order to assess the thermal-hydraulic mathematical model of SACATRI, the present paper focuses on the quantification of the physical model accuracy to judge if the code is capable to represent the thermal-hydraulic behaviour of the reactor core with sufficient accuracy. The methodology adopted is based on the comparison between responses from SACATRI computational model and experimentally measured responses performed on the IPR-R1 TRIGA research reactor. The results showed good agreement between SACATRI predictions and the experimental measurements where the discrepancies observed (simulation-experiment) are less than 6%.     

ITER纵场磁体过渡馈线结构力学分析

运用数值方法计算了不同等离子体运行时刻纵场磁体过渡馈线(CFT)超导母线上的电磁载荷,并确定了磁感应强度最大的时刻,采用增量有限元法对过渡馈线进行非线性力学分析,得到不同工况下结构上的应力分布及变形情况。分析结果表明,带有万向节的过渡馈线结构具有足够的强度来承受运行过程中的各种载荷,从而证明了结构设计的合理性。     

Critical issues for the manufacture of the ITER TF coil winding pack

Research and trials by the Japan Atomic Energy Agency (JAEA) focus on the remaining technical issues in the ITER TF coil winding pack (WP) manufacturing process. Specific issues include the feasibility of automatically measuring conductor length during automatic winding with a high degree of accuracy (±0.02%) and a fabrication process to comply with the demanding tolerances (up to 1 mm distortion in flatness and 1.5 mm in-plane shrinkage) of the radial plate (RP) due to cover plate (CP) welding. The authors developed a new technique to measure conductor length very accurately by combining an ordinary encoder and a newly developed optical system. A simulation based on test results of CP welding using a RP mock-up indicates that a flatness of 1 mm is achievable, but the in-plane shrinkage of the RP is approximately 5 mm. One possible solution is to fabricate the RP larger than required to allow for in-plane shrinkage. Another solution is to reduce the thickness or length of the welding. The feasibility of these solutions to most of the major technical issues suggests that it is time for full qualification testing of the fabrication process in a dummy double-pancake trial.     

Verification study of the FORE-2M nuclear/thermal-hydraulic analysis computer code

The verification of the LMFBR core transient performance code, FORE-2M, was performed in two steps. Different components of the computation (individual models) were verified by comparing with analytical solutions and with results obtained from other conventionally accepted computer codes (e.g., TRUMP, LIFE, etc.). For verification of the integral computation method of the code, experimental data in TREAT, SEFOR and natural circulation experiments in EBR-II were compared with the code calculations. Good agreement was obtained for both of these steps. Confirmation of the code verification for undercooling transients is provided by comparisons with the recent FFTF natural circulation experiments.     

A two-fluid code for the thermohydraulic transient analysis of CICC superconducting magnets

We present here a finite element computer model (Mithrandir) for the transient thermohydraulics of compressible helium in a Cable-In-Conduit Conductor (CICC) with central cooling hole, as presently envisaged for superconducting magnets of the International Thermonuclear Experimental Reactor (ITER). In the model the He in the hole and that in the cable bundle are treated as separate fluids, each characterized by its own flow and thermodynamic properties, coupled by exchanges of mass, momentum and energy. Results for the simulation of a quench both with and without a wall delimiting the central cooling hole are discussed. Time and space convergence of the code are demonstrated numerically.     

Development of a thermal-hydraulic analysis code TAC-DS for spent fuel dry-storage system for high-temperature pebble bed reactor

We report the development of a thermal-hydraulic analysis code (called TAC-DS: Thermal-hydraulic Analysis Code for Dry-storage System). The spent fuel dry-storage system of High-Temperature Reactor Pebble-bed Modules in China is simulated using the TAC-DS to confirm the design basis and to analyze the transient behavior following an accident involving blower failure. The TAC-DS includes mathematical models for the air-coolant system, heat conduction within spent fuel canisters, and thermal radiation between heat structures. The time-dependent hydrodynamic model of the TAC-DS is formulated using one-dimensional mass, momentum and energy equations, and solved using semi-implicit finite-difference scheme. The complicated heat transfer models of heat structure are incorporated into the hydrodynamic system implicitly with enclosure correlations. Code is written in Fortran 90. A validation calculation is performed by solving a simplified model. Thermal performance of the buffer storage region in the system under forced ventilation scenario is studied with TAC-DS to validate the design requirement, as well as to provide the initial condition for the transient analysis. Blower failure accident is studied to assess the performance of the safety features during the transient accident. Since the code is modular, TAC-DS can be easily modified and applied to other spent fuel dry-storage system in the future.     

Assessment of TAPINS code for application to thermal-hydraulic analysis of an integral reactor,REX-10

The thermal-hydraulic analysis program for integral reactor system (TAPINS) is a thermal-hydraulic system code developed by Seoul National University for transient analysis of an integral reactor, REX-10. Specialized for a fully passive integral pressurized water reactor, TAPINS adopts a one-dimensional four-equation drift-flux model for two-phase flows. It also consists of component models for the core, the helical-coil steam generator, and the steam-gas pressurizer. This paper presents the developmental assessment of TAPINS to validate its applicability to the thermal-hydraulic analysis of REX-10. Assessment problems are determined by taking into account thermal-hydraulic phenomena expected during design basis accidents of REX-10, including the loss-of-feedwater accident and the small-break loss-of-coolant accident. To confirm the predictive capability of TAPINS for these phenomena, the TAPINS model is validated against four sets of separate effects problems, including the pressurizer insurge test, the subcooled boiling experiment, the critical flow test, and the Edwards pipe problem. In addition, the calculation results of TAPINS are compared with the experimental data obtained from a series of integral effects tests using a scaled apparatus of REX-10. From the validation results, it is demonstrated that TAPINS can provide the reasonable prediction on the thermal-hydraulic responses of REX-10 during the transient and accident conditions.     

Validation of coupled neutronic/thermal-hydraulic code RELAP5-3D for RBMK-1500 reactor analysis application

This paper summarizes RELAP5-3D code validation activities carried out at the Lithuanian Energy Institute, which was performed through the modeling of RBMK-1500 specific transients taking place at Ignalina NPP. A best estimate RELAP5-3D model of the INPP RBMK-1500 reactor has been developed and validated against real plant data, as well as with the calculation results obtained using the Russian STEPAN/KOBRA code. The obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters, as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data, which demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core. The performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500 reactor allowed to improve the model, which in the future would be used for the safety substantiation calculations of RBMK-1500 reactors. Future activities are discussed.     

Model validation of GAMMA code with heat transfer experiment for KO TBM in ITER

By considering the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He cooled molten lithium (HCML) test blanket module (TBM) for testing in ITER. A performance analysis for the thermal–hydraulics and a safety analysis for the KO TBM have been carried out using a commercial CFD code, ANSYS-CFX, and a system code, GAMMA (GAs multicomponent mixture analysis), which was developed by the gas cooled reactor in Korea. To verify the codes, a preliminary study was performed by Lee using a single TBM first wall (FW) mock-up made from the same material as the KO TBM, ferritic martensitic steel, using a 6 MPa nitrogen gas loop. The test was performed at pressures of 1.1, 1.9 and 2.9 MPa, and under various ranges of flow rate from 0.0105 to 0.0407 kg/s with a constant wall temperature condition. In the present study, a thermal–hydraulic test was performed with the newly constructed helium supplying system, in which the design pressure and temperature were 9 MPa and 500 °C, respectively. In the experiment, the same mock-up was used, and the test was performed under the conditions of 3 MPa pressure, 30 °C inlet temperature and 70 m/s helium velocity, which are almost same conditions of the KO TBM FW. One side of the mock-up was heated with a constant heat flux of 0.3–0.5 MW/m² using a graphite heating system, KoHLT-2 (Korea heat load test facility-2). Because the comparison result between CFX 11 and GAMMA showed a difference tendency, the modification of heat transfer correlation included in GAMMA was performed. And the modified GAMMA showed the strong parity with CFX 11 calculation results.     

Fatigue assessment of the ITER TF coil case based on JJ1 fatigue tests

The material of the TF coil case in the ITER requires to withstand cyclic electromagnetic forces applied up to 3 × 10⁴ cycles at 4.2 K. A cryogenic stainless steel, JJ1, is used in high stress region of TF coil case. The fatigue characteristics (S–N curve) of JJ1 base metal and welded joint at 4.2 K has been measured. The fatigue strength of base metal and welded joint at 3 × 10⁴ cycles are measured as 1032 and 848 MPa, respectively. The design S–N curve is derived from the measured data taking account of the safety factor of 20 for cycle-to-failure and 2 for fatigue strength, and it indicates that an equivalent alternating stress of the case should be kept less than 516 MPa for the base metal and 424 MPa for the welded joint at 3 × 10⁴ cycles. It is demonstrated that the TF coil case has enough margins for the cyclic operation. It is also shown the welded joint should be located in low cyclic stress region because a residual stress affects the fatigue life.     

1-D thermal-hydraulic analysis of the high temperature superconducting current leads for the ITER magnet system from 5 K to 300 K

The magnet system of ITER includes high temperature superconducting (HTS) current leads with a maximum current of 68 kA for the toroidal field (TF) coils, 55 kA for the poloidal field (PF)/central solenoid (CS) coils and 10 kA for the control coils (CC), respectively. Although different in terms of size and operative conditions, the ITER HTS current leads have been all designed on the basis of an established concept, which was successfully developed for the LHC at CERN and proven by the so-called 70 kA “demonstrator” lead made by KIT and by the ITER pre-prototypes made by ASIPP in China. A broad R&D campaign has been undertaken by ASIPP and CERN in order to find optimized designs for each component of the leads. Nevertheless, a comprehensive picture of the performance of the entire HTS current leads is not yet available. In this paper, a steady state, full length, thermal-hydraulic 1-D modeling is applied to the study of the three types (TF, PF/CS, CC) of ITER HTS current leads. The results of this predictive analysis are then compared with relevant ITER requirements. It was found that the present design of the HTS current leads will fulfill these specifications.     

Transient electrical behaviour of the ITER TF coils during fast discharge and two fault cases

Improvements of high voltage design criteria and quality assurance for ITER coils are indispensable taking into account the problems occurred during high voltage tests of the ITER TF model coil. One important aspect to consider is the transient electrical behaviour because fast changes of voltages may cause local overloading and destruction of the insulation system. This paper will present the calculation of the terminal voltages within the ITER TF coil system and the voltage stress of the insulation within an individual ITER TF coil for the fast discharge and two fault cases. Proposals for the high voltage tests are discussed based on the calculated voltage stress of the two fault cases and the experiences gained during the ITER TF Model Coil test to ensure appropriate dielectric quality of the ITER TF coils.