In-box Loss Of Coolant Accident Research Articles

Preliminary accident analyses of in-box LOCA for CFETR HCCB blanket system

Safety analysis is an important part of China Fusion Engineering Test Reactor (CFETR) design. The blanket is the key component of the fusion reactor. The helium cooling system (HCS) for blankets is the main energy exchange system, and the tritium extraction system (TES) is an important part of the fuel system in CFETR. The preliminary design of CFETR was completed in 2020. This paper presents the analysis results of the In-box loss of coolant accident (LOCA) in CFETR and gives the optimization ideas for HCS and TES design. The accident analysis and optimization method can provide a reference for the iterative design.

Preliminary thermal optimization and investigation of the overall structural behaviour of the EU-DEMO water-cooled lead lithium left outboard blanket segment

The conceptual design phase of the EU-DEMO reactor has been recently launched, with the aim of evolving the DEMO pre-conceptual layout towards a more robust and articulated geometric configuration able to cope with most of the design requirements and to show further margins for the passing of the current potential show-stoppers. Hence, the achievement of the conceptual design of the Water-Cooled Lead Lithium Breeding Blanket (WCLL BB) is one of the milestones the EUROfusion consortium aims to achieve in the close future. To this purpose, within the framework of the research activities promoted by EUROfusion, a research campaign has been launched at the University of Palermo, in close cooperation with ENEA-Brasimone, in order to investigate, from the thermal and thermomechanical standpoints, the WCLL BB Left Outboard Blanket (LOB) segment performances. The scope of the analysis carried out has been the advancement of the WCLL LOB segment design in view of the prescribed thermal and structural design requirements. To this end, the nominal WCLL BB operating conditions as well as accidental steady-state scenarios derived from the in-box loss of coolant accident and the upper vertical plasma disruption event have been considered. In the first part of the study, the internal Double-Walled Tubes (DWTs) layout of the WCLL LOB equatorial region have been preliminarily optimized so to find a geometric configuration able to cool the segment structure maintaining the maximum temperature below the suggested limit of 550 °C. Then, a purposely set-up analytical procedure has allowed obtaining a set of interpolating functions able to accurately reproduce the calculated thermal field and to extrapolate it to the whole WCLL LOB segment. Afterwards, the thermo-mechanical analysis of the entire segment has been performed under the selected loading scenarios. The structural behaviour of the segment has been assessed in compliance with the RCC-MRx code, adopting the set of criteria on the basis of the nature of the considered loading scenario. The obtained results, herewith presented and critically discussed, have allowed introducing significant design improvements with respect to the WCLL LOB segment reference geometric layout, coming from the pre-conceptual design phase, paving the way for future, and more detailed, structural investigations.

Analysis of the Thermo-Mechanical Behaviour of the EU DEMO Water-Cooled Lithium Lead Central Outboard Blanket Segment under an Optimized Thermal Field

Within the framework of the EUROfusion research activities on the DEMO Water-Cooled Lithium Lead (WCLL) Breeding Blanket (BB) design, a research study was performed to preliminarily optimize, from the thermal point of view, the WCLL Central Outboard Blanket (COB) segment in order to investigate its structural behaviour under a realistic thermal field. In particular, a study of thermal analyses was performed to optimize the Double Walled Tubes and Segment Box cooling channels’ geometric configurations along the poloidal extension of the WCLL COB segment, in order to obtain a spatial temperature distribution fulfilling the thermal design requirement. Then, the thermo-mechanical analysis of the WCLL COB segment under Normal Operation (NO, representing nominal conditions), Upper Vertical Displacements Event (UVDE, representing a plasma disruption event) and Over-Pressurization (OP, representing an in-box loss of coolant accident) scenarios were carried out, assuming the previously obtained thermal field, to realistically predict displacement and stress fields. Finally, a stress linearization procedure allowed comparing the stress values obtained in some critical regions of the structure with the criteria prescribed by the reference design standard RCC-MRx. A theoretical–numerical approach based on the Finite Element Method (FEM) was followed using the commercial code Abaqus v. 6.14.

Preliminary Design of a Heat Pipe-Cooled Blanket for CFETR

Blankets are a difficult problem for fusion engineering design. Because of the complex flow channels, the design, production, processing, and accident maintenance of blankets are all huge challenges for traditional water/gas-cooled blankets. Blankets are the bridge for heat transfer and tritium production. A high-performance blanket with simplified structure is obviously beneficial for engineering, safety, and the economy. In this study, gravity heat pipes instead of coolant flow channels are adopted to remove the heat. Compared with coolant-cooled systems, heat pipes may be simpler and more reliable. The in-vessel and in-box loss of coolant accident (LOCA) will not occur because there is no coolant in the blanket. Moreover, a damaged heat pipe may be replaced easily compared to a damaged water-cooled blanket. In this study, a hypothetical heat pipe-cooled blanket for the China Fusion Engineering Test Reactor (CFETR) was proposed and one module of the blanket was analyzed by numerical simulation. The results were compared with those of a water-cooled blanket, and the temperature distribution of the heat pipe-cooled blanket is more uniform. This study verified the preliminary feasibility of heat pipe-cooled blankets and provided a fresh idea for blanket design.

Application of Inelastic Method and Its Comparison with Elastic Method for the Assessment of In-Box LOCA Event on EU DEMO HCPB Breeding Blanket Cap Region

The Helium Cooled Pebble Bed (HCPB) breeding blanket, being developed by the Karlsruhe Institute of Technology (KIT) and its partners is one of the two driver blanket candidates to be selected for the European demonstration fusion power plant (EU DEMO). The in-box Loss of Coolant Accident (LOCA) is a postulated initiating event of the breeding blanket (BB) that must be accounted within the design basis. In this paper, the BB cap region is analyzed for its ability to withstand an in-box LOCA event. Initially, an assessment is performed using conventional elastic design codes for nuclear pressure vessels. However, it is thought that the elastic rules are not ‘equipped’ to assess the material damage modes which are essentially inelastic. Therefore, a non-linear inelastic analysis is further performed to better understand the damage in the material. Two predominant inelastic failure modes are thought to be relevant and addressed: exhaustion of ductility and plastic flow localization. While the design of HCPB BB has been predominantly based on the elastic design-by-analysis studies, results from the present study show that the elastic rules may be overly conservative for the given material and loading and could lead to inefficient designs. To our knowledge, this study is the first attempt to investigate the structural integrity of the European DEMO blankets under in-box LOCA conditions using the inelastic methods.

Pressure wave behavior and its effects on structure under In-box LOCA in a helium-cooled lead lithium blanket of hydrogen fusion reactors

The helium-cooled lead lithium (PbLi) blanket is considered as one of the candidate blanket concepts selected for the hydrogen fusion DEMO reactors and beyond, which has the advantages of simple structure, strong heat removal capacity and high tritium breeding ratio. However, due to the harsh environment such as high-energy neutron irradiation, high thermal load and great pressure gradient, there is a high possibility that one or some of the thousands of coolant channels will break in the breeding zone, which is so-called In-box Loss of Coolant Accident (In-box LOCA). When the accident occurs, the high pressure helium will rapidly inject into the lead lithium flow channel, generating a complex two-phase flow and great pressure shock effect, which may cause the peak pressure to exceed the design limit and threaten the integrity of the blanket structure. Therefore, it is of great significance to perform the transient analysis of in-box LOCA to improve the safety of the blanket and avoid the leakage of radioactive materials. In this paper, a two-way coupling model for fluid-solid interaction was established based on the ANSYS Workbench, and the model were validated through the experimental data obtained by injecting the high pressure helium gas into liquid lithium lead. Then the validated model was applied to the transient pressure wave propagation analysis and structural stress analysis of the Dual-Functional Lithium Lead (DFLL) blanket in order to explore the integrity of blanket structure under In-box LOCA. In addition, the effects of break location on pressure and structural stress was also investigated through six cases. The study found that the transient pressure in the DFLL blanket gone through three stages in any case: step rise, oscillate, and flatten out. Pressure peaks occurred during oscillations and their values were strongly dependent on the break location. The closer to the inlet/outlet, the higher the peak pressure was. The maximum pressure reached more than twice of the inlet pressure (up to ~16 MPa). As a result, the structural stress in some local areas has exceeded the allowable limits, and the corresponding suggestions for improvement have also been put forward. This study can provide guidance for safety design, operation and accident mitigation measures of helium-cooled lead lithium blankets.

Assessment of SIMMER-III code in predicting Water Cooled Lithium Lead Breeding Blanket “in-box-Loss of Coolant Accident”

The in-box Loss of Coolant Accident is a major safety concern for the Water Cooled Lithium Lead Breeding Blanket design. SIMMER-III code has been modified by University of Pisa and ENEA C.R. Brasimone to perform deterministic safety investigation of such accidental scenario. Thanks these modifications, this version of the code has unique features for dealing with the PbLi water reaction related phenomena. The reliability of the code requires a systematic validation activity, carried out applying a standard methodology based on a three-steps procedure and through qualitative and quantitative evaluations. The methodology was applied to the LIFUS5 campaign. Post-test analyses highlighted open issues of test execution and of experimental data, as well as code limitations and capabilities. Nevertheless, the modified version of SIMMER-III code for fusion application is able to predict the relevant thermal-hydraulic phenomena during PbLi/water interaction.

Analysis of Test D1.1 of the LIFUS5/Mod3 facility for In-box LOCA in WCLL-BB

The in-box Loss of Coolant Accident (LOCA) scenario is considered as one of the most affecting safety concerns for the Water-Cooled Lead Lithium Breeding Blanket (WCLL-BB) modules of the DEMOnstration (DEMO) reactor, which is sequentially followed by a multi-phase multi-component physical and chemical interaction. Therefore, the transient behavior of such accidents has to be carefully investigated during the design phase of the plant, to evaluate the consequences and to adopt the necessary mitigating countermeasures. This also requires a numerical predictive tool, which is capable to model such transients and predict the relevant phenomena under an operational condition and the connected safety parameters i.e. system pressure, temperature, chemical products mass, and volume fractions of all the existing components. Consequently, the SIMMER-III code was firstly improved at the University of Pisa by implementing the chemical reaction between PbLi eutectic alloy and water. In addition to this, an experimental campaign and a test-matrix have been recently designed according to the LIFUS5/Mod3 facility to perform a series of experiments and code post-test analyses.In the present work, the experimental data of the first LIFUS5/Mod3 test is used for the validation of the chemical model implemented in SIMMER-III through a comprehensive sensitivity study. The applied methodology for the code validation is based on a three-step procedure including qualitative analysis, quantitative analysis and the results from sensitivity analyses. The qualitative accuracy evaluation is performed through a systematic comparison between experimental and calculated time trends based on the engineering analysis, the resulting sequence of main events and the identification of phenomenological windows and of relevant thermo-hydraulic aspects. Afterwards, the accuracy of the code prediction is evaluated from a quantitative point of view by means of selected, widely used, figures of merit. Finally, the results from the sensitivity cases are analysed and quantified, to determine the effects of the most influencing code input options and transient parameters.Furthermore, the analysis is followed by applying the Fast Fourier Transform Method (FFTM) to the experimental signals and all the sensitivity calculations. The comparison shows a very good agreement for pressure transient between the experimental and numerical data, while for the temperature and the hydrogen production the results fall into acceptable criteria, which means that the code is reliable in capturing and predicting the transient values but not perfectly match with the experimental signals.

Steady State and Transient Analysis of Novel Design Helium Cooled Ceramic Blanket (HCCB) System of China Fusion Engineering Test Reactor (CFETR)

An upgraded form of China fusion engineering test reactor (CFETR) was investigated for the safety performance. In the current study, modification of the designs were presented with relative tolerance. The steady state were calculated for the new design using Relap5 code. Two accidents were simulated i.e., in-vessel and In-box loss of coolant accident. These accidents were simulated in helium cooled ceramic blanket (HCCB) system for the purpose to investigate the safety measures of the CFETR. It is utmost important to ensure the safety performance of the reactor. In this research, sudden break at blanket system was assumed and calculated different parameters including temperature, pressure and coolant fluxes to observe the differences in pattern during the accident under limited time domain. The research is very important because the design of HCCB is new and there is a need to conduct steady state and transient state of the reactor in order to make sure and authenticate the design and to safer the reactor.

A methodology for thermo-mechanical assessment of in-box LOCA events on fusion blankets and its application to EU DEMO HCPB breeding blanket

Abstract Within the framework of EUROfusion research activities, the Karlsruhe Institute of Technology (KIT) is developing the Helium Cooled Pebble Bed (HCPB) breeding blanket (BB) for European demonstration fusion power plant (DEMO), with the support of BME and Wigner-RCP (Hungary) and CIEMAT (Spain). During the development of the BB concept in the past years, the in-box Loss Of Coolant Accident (LOCA) has been identified as a main design driver for BB. During this event, a leakage of pressurized coolant within the blanket box happens (in-box LOCA), pressurizing the blanket module and endangering its structural integrity. This accident has not been investigated in a detailed approach yet. In this paper, a more precise methodology is introduced taking into account the temperature evolution of the structure for assessing the structural integrity of BB under in-box LOCA and has been applied to the HCPB BB. The methodology and results are presented and critically discussed.

Effect of irradiation hardening on brittle fracture performance of box-shaped blanket fabricated by F-82H

Fusion reactor blankets designed for DEMO are based on a water cooled ceramic breeder. The unstable fracture of the box-shaped blanket would occur by the superimposition of three factors as follows: (1) it is the reduction of fracture toughness for structural materials, such as F-82H, caused by the significant irradiation hardening, (2) it is the formation of cracks which act as the initiation point of the fracture, and (3) it is the excessive tensile load caused by the inner pressure to In-Box loss of coolant accident (LOCA). In the present study, application of Weibull stress criterion to fracture assessment is to investigate the effect of irradiation hardening on the brittle fracture performance of box-shaped blanket fabricated by F-82H. Both structural analysis of the box-shaped blanket simulating In-Box LOCA and deformation behavior analysis of irradiated toughness specimens were conducted by elastic-plastic finite element analysis (FEA). Weibull stress values calculated from FEA of irradiated toughness specimens were higher than those from FEA of irradiated structural components. It, therefore, is considered that the irradiation hardening of structural materials would not promote the brittle fracture at the corner of rectangular coolant tubes of the blankets caused by the load of In-Box LOCA.

Preliminary accident analysis of Loss of Off-Site Power and In-Box LOCA for the CFETR helium cooled solid breeder blanket

As one of three candidate tritium breeding blanket concepts for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of helium cooled solid breeder (HCSB) blanket was recently proposed. In this paper, the preliminary thermal-hydraulic and safety analyses of the typical outboard equatorial blanket module (No.12) have been carried out using RELAP5/Mod3.4 code. Two design basis accidents are investigated based on the steady-state initialization, including Loss of Off-Site Power and In-Box Loss of Coolant Accident (LOCA). The differences between circulator coast down and circulator rotor locked under Loss of Off-Site Power are compared. Regarding the In-Box LOCA, the influences of different break sizes and locations are thoroughly analyzed based on a relatively accurate modeling method of the heat structures in sub-modules. The analysis results show that the blanket and the combined helium cooling system (HCS) are designed with sufficient decay heat removal capability for both accidents, which can preliminarily verify the feasibility of the conceptual design. The research work can also provide an important reference for parameter optimization of the blanket and its HCS in the next stage.

Design improvement of blanket box structure with fillet against water ingress

The Water Cooled Ceramic Breeder Test Blanket module (TBM) has been developed for a demonstration of the major functions under the fusion reactor environment in ITER. Testing of the TBM shall neither hinder the ITER operation, nor threaten safety of operation even at water ingress into a box structure of the TBM (In-Box Loss of Coolant Accident, In-Box LOCA). In present study, the structural responses on the TBM under In-Box LOCA were evaluated by using Finite Element Method. Based on the analysis results, the TBM design was modified. The filled region and Tritium Breeding Ratio of the modified TBM design adding the fillet with 35mm of radius increased by up to 20%, 12%, respectively, compared to the original TBM model without fillet.

Sensitivity study on in-box LOCA for a Korean HCCR TBM in ITER

Korea has designed a Helium-Cooled Ceramic Reflector (HCCR)-based Test Blanket System (TBS) for International Thermonuclear Experimental Reactor (ITER). Among seven selected reference accidents in Korean TBS, in-box loss of coolant accident (LOCA) is one of them. This is initiated by a double-ended break of the coolant pipe in the Breeding Zone (BZ), pressurizing the BZ box structure, causing pressurization of the Tritium Extraction System (TES) and purging of pipelines. When the accident is detected, the Plant Safety System (PSS) isolates the Helium Cooling System (HCS) and TES, and requests plasma shutdown to Fusion Power Shutdown System (FPSS). To prevent aggravating failure of the system, the safety function is automatically activated when the accident is detected, the device being the isolation valve of HCS and TES. One important observation of this accident is that instant isolation is not a good measure to take. In terms of the possibility of aggravating failure, system isolation is an important safety procedure but isolated TES volume is exposed to high pressure and temperature conditions in the early move of the accident transient. The result of system safety analysis shows that delayed isolation keeps the system safe for a while. In this article, given the preliminary accident analysis results for the current HCCR TBS, case studies were performed regarding the delayed isolation timing effect. For this transient simulation, Korean nuclear fusion reactor safety analysis code (GAMMA-FR) was used.

Current status of accident analysis for Korean HCCR TBS

Korea has decided to test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in ITER and design of the TBM with its ancillary systems, i.e. Test Blanket System (TBS), is under progress. Since the TBM is operated at elevated temperature with high heat load, safety consideration is essential in design procedure. In this paper, preliminary accident analysis results for the current HCCR TBS design on selected scenarios are presented as an important part of safety assessments. To simulate transient thermo-hydraulic behavior, GAMMA-FR code which has been developed in Korea for fusion applications was used. The main cooling and tritium extraction circuit systems, as well as the TBM, were simulated and the main components in the TBS were modeled as the associated heat structures. The important accident scenarios were produced and summarized in the paper considering the HCCR TBS design and ITER conditions, which cover in-vessel Loss Of Coolant Accident (LOCA), in-box LOCA, ex-vessel LOCA, Loss Of Flow Accident (LOFA), Loss Of Heat Sink Accident (LOHSA) and purge pipe rupture case. The accident analysis based on the selected scenarios was performed and it was found that the current design of the HCCR TBS meets the thermo-hydraulic safety requirements.

Thermal-Hydraulic LOCA Analysis of Chinese Helium-Cooled Solid Breeder Test Blanket Module for ITER

The Chinese helium-cooled solid breeder (CH-HCSB) test blanket module (TBM) is designed to be tested in ITER, and its aim is to validate the feasibility of a DEMO fusion reactor. The thermal-hydraulic transient analysis has to testify that the TBM and its helium cooling system (HCS) will not impact the safe operation of ITER under both normal and accidental conditions. In order to simulate the transient accidents, the TBM and HCS are modeled using the RELAP5/MOD3 system code. The steady-state results indicate that the designed TBM inlet/outlet temperatures are obtained and the temperature of first-wall (FW) structural material is below the limit. An ex-vessel loss-of-coolant accident (LOCA) will induce the melting of FW beryllium armor after ~80 s of LOCA initiation, and some controlling measures have to be taken before melting. The pressurization of the vacuum vessel induced by an in-vessel LOCA is within the allowable value of the ITER design. Because of pressurization of the purge gas system, the tritium extraction system has to be isolated from the TBM quickly when an in-box LOCA happens. Based on the results, the design of the CH-HCSB TBM could be further modified in order to assure the safety of the TBM and ITER, from an engineering point of view.