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Progress towards the validation of SIMMER-III code model for lead-lithium water chemical interaction
Abstract
Research activities are continuing between the University of Pisa and ENEA Brasimone Research Center to understand better the phenomena and processes that occur during a postulated in-box LOCA in the Water-Cooled Lead Lithium Breeding Blanket (WCLL BB) and during the system's safety response. Various activities are also going on to strengthen the reliability of numerical tools and validate computer models, codes, and procedures for their implementations.
The study presented here aims to assist in the validation process of the SIMMER III code model, to simulate the Lead-lithium and water interaction under conditions like those expected for the WCLL BB during accident conditions. In particular, the current work has been performed by numerically reproducing the experimental Test E5.2, executed in the separate effect test facility LIFUS5/Mod3 (installed at ENEA Brasimone Research Center). The code used for the simulation was SIMMER III computational code, in a version modified by the University of Pisa to take into account also the chemical interaction between the two fluids.
A comparison of significant parameters computed by the SIMMER code with those obtained in the experimental test during the transient has been carried out, and it is presented and described in this paper.
... The set-up and qualification of a numerical tool for predicting PbLi/water interaction. This requires code model improvements [7,8], the extension of the available (to date) experimental database [9][10][11][12], and validation activities [13][14][15][16], carried out during FP8 in separate effect test facility LIFUS5/Mod3; 2. ...
... The set-up and qualification of a numerical tool for predicting PbLi/water interaction. This requires code model improvements [7,8], the extension of the available (to date) experimental database [9][10][11][12], and validation activities [13][14][15][16], carried out during FP8 in separate effect test facility LIFUS5/Mod3; 2. The addressing of postulated in-box LOCA events by means of an approach that permits prevention of the occurrence, evaluation of the consequences, and investigation of countermeasures to mitigate the transient. This last point requires extensive numerical simulations and appropriate tools, i.e., the development of a coupling technique to perform transient analyses considering the chemical, thermohydraulic, and structural effects due to PbLi/water interaction [17][18][19][20][21] and numerical code to expand the simulation capabilities in fusion applications [22,23], validated against Integral Test Facility experiments [24] currently in progress as part of the FP9 Project. ...
... The experimental data provided by the separate effect test facility LIFUS5/Mod were used to continue the validation activity of SIMMER codes, both in 2D and 3 versions [10,[13][14][15][16]. A standard methodology, widely used in fission-related activities, applied [8], consisting in a three-steps analysis. ...
The Water-Cooled Lithium–Lead blanket concept is a candidate breeding blanket concept for the EU DEMO reactor and it is going to be tested as one of the Test Blanket Modules (TBM) inside the ITER reactor. A major safety issue for its design is the interaction between PbLi and water caused by a tube rupture in the breeding zone, the so-called in-box LOCA (Loss of Coolant Accident) scenario. This issue has been investigated in the framework of FP8 EUROfusion Project Horizon 2020 and is currently ongoing in FP9 EUROfusion Horizon Europe, defining a strategy for addressing and solving WCLL in-box LOCA. This paper discusses the efforts pursued in recent years to deal with this key safety issue, providing a general view of the approach, a timeline, research and development, and experimental activities. These are conducted to master dominant phenomena and processes relevant to safety aspects during the postulated accident, to enhance the predictive capability and reliability of selected numerical tools, and to validate and qualify models and codes and the procedures for their applications, including coupling and chains of codes.
... SIMMER-III is a software designed specifically to analyze core disruptive accidents (CDAs) in SFRs [66]. It is a sophisticated computer program that employs multiple dimensions to replicate the complex dynamics of liquid metal coolant in various accident scenarios, including fuel pin failure, coolant boiling, sodium expansion, and core disruption [67]. ...
A small modular reactor (SMR) is a nuclear reactor that is characterized by its smaller size and capacity when compared to traditional large-scale nuclear reactors. An SMR is often categorized as having an electrical output of less than 300MW and is built to be more mobile, safe, and extensible to deploy. It has been established that SMRs can provide economic and flexibility advantages in a variety of industries thanks to the development, study, and use of multiple types of SMRs in recent years. The goal of this paper is to present a comprehensive overview of several SMR types, including light water reactors (LWRs), liquid metal-cooled reactors (LMRs), molten salt reactors (MSRs), and gas-cooled reactors (GCRs). Each type of reactor will be reviewed in terms of its structural design, modeling control implementation, applications, and impacts concerning the power system.
... In this setup, protons with a highenergy spectrum strike the spallation nuclear object (often lead or its eutectic alloy) to generate neutrons through a break-up of a bombarded nucleus process [3]. Research and development activities are also going on in the field of demonstration power plant fusion reactors and other advance generation-IV nuclear reactors [4][5][6][7][8][9][10][11][12][13][14][15][16], which also use lead alloy-based fluids (lead-lithium eutectic) and other advance fluids for multiple purposes (coolant, breeder, and neutron multiplier) [17]. Fast neutron spectrum nuclear reactors that use molten lead or lead-bismuth eutectic (LBE) coolant are known as lead-cooled fast reactors. ...
The differential scanning calorimetry (DSC) method has recently emerged as a sophisticated and precise technique for promising contributions to the thermal analysis of various materials, including heavy liquid metal (HLM) coolants. However, there is a lack of experimental studies on the thermal properties of lead-based fluids, such as lead-bismuth eutectic (LBE) and lead-lithium eutectic, which are potential candidates for use as coolants, breeders, and neutron multipliers in advanced nuclear systems like the fourth-generation lead-cooled fast reactor. The available experimental data on the thermal properties of LBE and other lead-based fluids is limited, and the measurements have significant uncertainty. In addition, the composition of components used in the previous studies is inconsistent, and the environmental conditions were often unknown. Therefore, to fill these gaps and advance the thermal properties measurement technique for heavy liquid metal coolants, ENEA Brasimone, in collaboration with DICI-UNIPI, has installed a DSC instrument setup. The experiments performed at the installed DSC setup are focused on measuring some essential thermal properties of LBE using DSC. The experience gained from this work will facilitate the measurement of other fluids based on lead alloy, especially lead-lithium eutectic, a potential candidate for breeder, coolant, and neutron multiplier in DEMO fusion reactors. This study represents the first effort to advance the DSC approach for accurately measuring the thermal characteristics of heavy liquid metals that are highly reactive, such as lead-lithium, which has significant potential in advanced nuclear systems.
Liquid metals are employed as a coolant in liquid metal fast reactors (LMFR) and are considered breeders and coolants for future power-producing fusion reactors. The interaction of liquid metals with other coolants, such as air and water, is one of the possible occurrences in liquid metal-cooled reactors. Although the SIMMER-III computer code is currently in the validation and verification phase, it is a prospective candidate code for correctly investigating possible accidents. This study aims to determine the stability and well-posedness of the SIMMER-III code Eulerian-Eulerian two-fluid model (TFM) under any such accident scenario. This work also considers the influence of the virtual mass force and diffusion forces in momentum on TFM stability in accelerated liquid metal, steam, and non-condensable gaseous flows throughout all flow regimes. The characteristics method is used to assess the ill-posed nature of TFM for all types of accidents and accident scenarios in a hypothetical simplified system model with several components (Lead–Lithium, non-condensable gases, and water vapor). It has been discovered that the analysis findings vary from the air and water two-component two-phase flows (characteristics roots spectrum and error growth rate patterns) and are particularly sensitive to the diffusion and virtual mass coefficients. This is because liquid metals have a higher density than liquid water and steam, resulting in strong virtual mass forces and weak diffusion forces in liquid-metal and gas two-phase flows. Because of the extensive range of possible interactions between fluxes of different components, producing an accurate representation of diffusion in multi-component mixtures is difficult. Because of this, it is strongly suggested that the virtual mass coefficient and diffusion coefficients be handled more accurately for these kinds of flows. The values of these coefficients significantly affect how accurate proposed TFM predictions are, but there hasn't been much research on how to estimate them. The study also sheds light on the model's accuracy and highlights the areas where the model's predictions will be mathematically trustworthy.
The in-box LOCA (Loss of Coolant Accident) represents a major safety concern to be addressed in the design of the WCLL-BB (water-cooled lead-lithium breeding blanket). Research activities are ongoing to master the phenomena and processes that occur during the postulated accident, to enhance the predictive capability and reliability of numerical tools, and to validate computer models, codes, and procedures for their applications. Following these objectives, ENEA designed and built the new separate effects test facility LIFUS5/Mod3. Two experimental campaigns (Series D and Series E) were executed by injecting water at high pressure into a pool of PbLi in WCLL-BB-relevant parameter ranges. The obtained experimental data were used to check the capabilities of the RELAP5 system code to reproduce the pressure transient of a water system, to validate the chemical model of PbLi/water reactions implemented in the modified version of SIMMER codes for fusion application, to investigate the dynamic effects of energy release on the structures, and to provide relevant feedback for the follow-up experimental campaigns. This work presents the experimental data and the numerical simulations of Test E4.1. The results of the test are presented and critically discussed. The code simulations highlight that SIMMER code is able to reproduce the phenomena connected to PbLi/water interaction, and the relevant test parameters are in agreement with the acquired experimental signals. Moreover, the results obtained by the first approach to SIMMER-RELAP5 code-coupling demonstrate its capability of and strength for predicting the transient scenario in complex geometries, considering multiple physical phenomena and minimizing the computational cost.
The scope of this work comprises the study of the mathematical nature of the RELAP5 two-fluid model (TFM) for two-component (liquid water, steam and non-condensable gass (helium and air)) and two phases (liquid and vapor) mixture flows, and to predict the accuracy of the model under these flow conditions. The method of characteristics is applied to check the non-hyperbolic nature of conservation equations of RELAP5 for some sample fluid conditions. The sample problem selected here is a small break in helium-water heat exchanger primary tubes (containing high pressure and high-temperature helium gas) into the secondary shell (water) side high-temperature gas-cooled reactor, that causes mixing of helium-steam and liquid water and results in two-component two-phase flow. This sample problem is selected to address the RELAP5 model behavior and its sensitive nature under multi-component two-phase analysis. Linear stability analysis is also performed to check the error growth rates for the sample fluid conditions.
Keywords Multi-component two-phase flow; RELAP5; method of Characteristics; Two fluid models; high-temperature gas-cooled reactor (HTGCR); eigenvalue analysis; thermal-hydraulics
Highlights
Helium-liquid water and water vapor mixture flow TFM stability
Non-hyperbolic region of the two components and two-phase mixture model
Effect of various thermal-hydraulic parameters on the hyperbolicity TFM
The sensitivity of nuclear reactor safety analysis
Insight into the mathematical properties of the model.
The scope of this work comprises of the study of mathematical nature of this code model and to predict the accuracy of the model in nuclear reactor safety analysis. Method of characteristics is applied to check the non-hyperbolic nature of conservation equations for all normal and accident conditions of LWR. The analysis also gives information about the soundness of the model and to identify the regions where the solutions obtained from it will be numerically convergent. The characteristics of equations of non-hyperbolic nature are complex. It implies that results thus obtained (by finite difference method) have to be interpreted very carefully in view of the sensitive nature of reactor safety analysis. The present analysis shows that governing equations of the code exhibit complex characteristics for some operating conditions thus implying non-hyperbolicity under those conditions. Results are less accurate under such conditions, so sensitivity analysis plays an important role. The sensitivity of closure relationship on conservation equation's stability is also checked. The analysis is performed in MATLAB environment for four different systems, (a) a simplified vertical pipe (b) Pressurized Water Reactor (PWR, 15.5 MPa), (c) Boiling Water Reactor (BWR, Pressure 7.0 MPa ) and (d) Natural circulation reactor (AHWR, pressure 0.7 MPa). These results can also be extended for other thermal hydraulic systems.The importance of method of characteristics (in reactor thermal safety analysis) is clearly evident here.
A major safety issue in the Water-Cooled Lead-Lithium Breeding Blanket (WCLL-BB) system foreseen for fusion reactor is the interaction concerning the primary coolant (water) and the neutron multiplier (PbLi), due to a hypothetical tube rupture in the coolant circuit. This scenario involves an exothermic chemical reaction between PbLi and water with the production of hydrogen, in addition to critical interactions in a complex multiphase system in non-thermal equilibrium. In recent years the PbLi/water reaction was successfully implemented in the SIMMER-III code and validated against data from the LIFUS5/Mod3 experimental campaign. However, due to limitations of SIMMER-III, this work was restricted to the prediction of the phenomena inside the vessel, neglecting the simulation of the injection line. Nevertheless, since the injection line may actually have an important effect on the development of the transient, the simulation of the whole facility would be highly desirable. Indeed, the University of Pisa recently developed a coupling methodology between the SIMMER-III and RELAP5/ Mod3.3 codes and applied it to simple single-phase cases. In this paper the complete simulation of the LIFUS5/ Mod3 facility is presented, with the injection line modelled through RELAP5. Furthermore, all the complex aspects of the phenomena inside the reaction tank were included: the multiphase system and the interaction between water and PbLi with the chemical reaction and the production of hydrogen were modelled by SIMMER. Preliminary results are presented, showing that the coupling methodology can be effectively employed for the prediction of the chemical and thermal-hydraulic behaviour of complex loop experimental facilities.
One of the four breeding blanket concepts for European DEMO nuclear fusion reactor is the Water-Cooled Lithium Lead Breeding Blanket (WCLL BB). The WCLL in-box LOCA (Loss Of Coolant Accident) is a major safety concern of this component, therefore transient behavior shall be investigated to support the design, to evaluate the consequences and to adopt mitigating countermeasures. To fulfill this objective, at first, SIMMER-III code was improved by implementing the chemical reaction model between PbLi and water. Then, SIMMER-III Verification and Validation (V&V) procedures have been established and conducted to obtain a qualified code for deterministic safety analysis. The verification activity was successfully completed, while the validation activity requires further effort according to the R&D plan set up in the framework of the EUROfusion Project. In view of this, an experimental campaign and a test matrix has been designed in LIFUS5/Mod3 facility performing pre-test analyses of Test D1.1.
The preliminarily-defined test matrix will be used for the validation SIMMER-III according to a standard procedure. At the present stage, a pre-test numerical analysis was performed to support future experimental tests. The presented work aims to support the upcoming experimental activity in terms of setting up Boundary & Initial condition, specifying the most important parameters to be measured during tests and calculated by SIMMER-III code during transient and obtaining the best nodalization for the post-testing simulation. In particular, a qualitative analysis of obtained results was performed according to the available data time trends and based on engineering considerations. It aims to interpret the resulting sequence of main events and the identification of phenomenological windows and aspects, relevant to pressure transient and hydrogen production due to the chemical reaction between heavy liquid metal and water.
The new experimental facility LIFUS5/Mod3 has been designed, manufactured and installed to investigate the phenomena connected with the thermodynamic and chemical interaction between lithium-lead and water in case of in-box LOCA (Loss of Coolant Accident) of the WCLL breeding blanket concept and to validate the chemical model implemented in SIMMER code for fusion application. In order to fulfill these objectives, the necessary step is to obtain data, suitable to code validation, by means of an experimental campaign in LIFUS5/Mod3 facility, executed with controlled initial and boundary conditions. Thus, specific instrumentation and dedicated data acquisition system are installed on the facility to provide meaningful and reliable data.
The final aim of the LIFUS5/Mod3 campaign is the SIMMER code validation, applying the standard methodology to post-test analyses. Besides, the expected outcomes of the tests are the improvement of the knowledge of physical behavior and of understanding of the phenomena, the investigation of the dynamic effects of energy release towards the structures and of the chemical reaction with the consequent hydrogen production, and the enlargement of the database for code validation.
The interaction between lithium-lead and water is a major concern of Water Cooled Lithium Lead (WCLL) breeding blanket design concept, therefore deterministic safety analysis of the in-box LOCA postulated accident is of primary importance. The paper presents the preliminary code assessment process of the modified version of SIMMER-III code for fusion application by means of LIFUS5 Test#3 post-test analyses. A series of sensitivity calculations are performed to overcome uncertainties in the test data and experiment execution, and to investigate the capability of the code in predicting the phenomena occurring during PbLi/water interaction. Results show agreement between numerical results and experimental data. Besides, differences are observed in the first second of the transient due to imperfect knowledge of initial and boundary conditions, and test execution procedure.
The key objective of Test Blanket Module (TBM) program is to develop the design technology for DEMO and future power producing fusion reactor. The proposed First Wall of Test Blanket System (TBS) is a generalized concept for testing in ITER, an experimental fusion reactor being constructed presently in France. The First Wall of TBM (TBM FW) directly faces the plasma and is cooled by First Wall Helium Cooling System (FWHCS), it is considered as a critical component from ITER safety point of view. The scope of this work comprises of thermal hydraulic analysis of the FWHCS of a generalized TBS and the assessment of Postulated Initiating Events (PIEs) on the ITER safety with the help of thermal-hydraulic code RELAP/SCDAPSIM/MOD4.0. The three reference accidents: In-Vacuum Vessel (VV) Loss of Coolant Accident (In-Vessel LOCA), Ex-Vessel LOCA and Loss of Flow Accident (LOFA) in FWHCS are selected for the safety assessment. This safety assessment addresses safety concerns resulting from FWHCS component failure, such as VV pressurization, TBM FW temperature profile, pressurization of Port Cell (PC) and Tokomak Cooling Water System Vault Annex (TCWS-VA) and passive decay heat removal capability. The analysis show that the critical parameters are under the design limit and having large safety margins, in the investigated accident scenarios. A comparative analysis is also carried out with the previous results to validate the results. Keywords Accident analyses, Test Blanket System (TBS), Safety analyses, RELAP5
The interaction between heavy liquid metal (HLM) and water is a safety concern for the preliminary designs of lead fast reactor (i.e. LFR) and of subcritical transmutation system prototypes (i.e. XT-ADS). Current pool-type configurations have steam generators (SG) inside the reactor vessel. This implies that the primary to secondary leak (e.g. steam generator tube rupture) shall be considered as a postulated initiating event. The issue is addressed for CIRCE facility in ICE (Integral Circulation Experiment) configuration. CIRCE facility is a large pool system aimed at studying key operating principles of Lead Bismuth Eutectic (and Lead) systems. The configuration ICE was carried out to perform integral experiments, simulating the coupling between a high-performance heat source (electrically heated fuel bundle) and the heat exchanger, which was representative of the preliminary design of the XT-ADS heat exchanger. A Failure Mode and Effect Analysis (FMEA) is applied in order to get a complete picture of all the failure modes pertaining to this system, to determine their effects and to classify them according to their severity. The outcome of the analysis has identified as major hazard, relative to the CIRCE facility in the ICE configuration, the risk related to the LBE/water reaction, although with a very low probability, with the potential for a suddenly and dangerous pressurization (beyond the failure threshold) within the main vessel. A SIMMER-III code model of the system has been setup to provide deterministic results of the scenario. The results are supported by means of a LBE/water interaction experiment executed in LIFUS5 facility. LIFUS5 is a separate effect test facility dedicated to the investigation of LBE/water interaction. SIMMER-III code pre-test and post-test analyses are performed to define the boundary conditions of the experiment and to demonstrate the reliability of the code in simulating the phenomena of interest. The activity contributes to solving the safety issue raised for the operation of CIRCE facility and it provides a sample approach for addressing the safety studies needed in the development of the lead fast reactor and of the subcritical transmutation system.
The In-box LOCA for the WCLL-BB is recognized as a Design Basis Accident (DBA) and is of substantial interest to the DEMOnestration reactor design, therefore, the transient response of such an accident must be carefully investigated and addressed to ensure the safe operation and integrity of the whole system. In this way, the LIFUS5/Mod3 test facility (constructed at ENEA Brasimone Research Center) has been upgraded in the period 2018–2020 to perform a series of tests. The first set of the tests were named Series D, which are characterized by injecting specified amounts of pressurized water into the Lithium lead liquid bulk. The experimental campaign aimed at validating and qualifying the SIMMER-III code as a reliable numerical tool for the safety studies of the WCLL BB.
In parallel with performing the tests, SIMMER-III Verification and Validation (V&V) was conducted according to a standard code validation procedure. SIMMER-III is a two-dimensional, multiphase, multicomponent, Eulerian, fluid-dynamics code which was firstly developed at the Japan Nuclear Cycle Development Institute (JNC). An adopted version of the original SIMMER-III code so-called Ver.3 F Mod.0.1 (which was developed at University of Pisa), was employed for the analyses. The V&V activity was successfully completed and documented as the technical reports within the past numerical analytical and experimental activities for the first three tests (D1.1, D1.2 and D1.5).
In this article, the experimental data of the Tests D1.1, D1.2 and D1.5 are used for the SIMMER-III code results comparison. A qualitative analysis of the results obtained is reported according to the time trends for the most relevant parameters. The results show that the SIMMER-III code acceptably predicts the transient and the accuracies of the relevant test parameters are in agreement with the acquired experimental signals. Although that the primarily validation results are highly promising but further code assessment, development and validation are essential to approach such a qualified system code, which is suitable for fusion safety applications. The present validation work has been successfully followed by its ongoing experimental and numerical activities as a multilateral EUROfusion project.
System thermal-hydraulic code RELAP5 is based on a two-fluid, non-equilibrium, and non-homogeneous hydrodynamic model for simulation of transient two-phase behavior. The code model includes six governing equations to incorporate the mass, energy, and momentum of the two fluids. In this paper, linear stability analysis is performed to check the ill-posedness of the RELAP5 specific two-fluid model (TFM) for all normal and accident conditions of a standard pressurized water reactor (PWR). The analysis gives information about the soundness of the model and identifies the range of parameters where the solutions obtained from the model will be numerically convergent. The linear two-phase fluid dynamic stability (by dispersion analysis) of the RELAP5 one-dimensional two-fluid model and numerical stability of the difference equation formulation (by Von Neumann method) is presented. The present analysis shows that the two-fluid model becomes ill-posed for some fluid conditions, where the results are less accurate, so sensitivity analysis plays an important role. The ill-posed nature implies that results thus obtained (by finite difference method) have to be interpreted carefully because of the sensitive nature of reactor safety analysis. It is also identified that the variation in various parameters (like slip ratio, system pressure, void fraction, and phasic velocities) can affect the error growth rates. It has been demonstrated that the basic system of one-dimensional two-phase flow equations, that possesses complex characteristics, exhibits unbounded instabilities in the short-wavelength limit and constitutes an improperly posed initial value problem. The semi-implicit numerical method, which is unconditionally stable for hyperbolic systems, becomes unstable for non-hyperbolic systems. For some of the fluid conditions, even after the introduction of artificial viscosity terms (in the difference equation formulation) that damp the high-frequency spatial components of the solution, are not sufficient for regularization of the two-fluid model. Thus, there is a need for the addition of newer terms, e.g. bubble collision, so that the existing incomplete model provides better results. It is also demonstrated that the basic TFM of RELAP5 with additional collision term makes the system unconditionally well-posed which are originally conditionally well-posed. Excellent agreement is obtained between the predicted and computed growth rates of harmonic disturbances. It is found that the instability associated with the two-fluid model discretized system of equations is related to the instability associated with the ill-posedness of the two-fluid model but is quantitatively different. Highlights 1. The linear stability of the RELAP5 one-dimensional TFM in the nuclear reactor (PWR) accident conditions is determined. 2. The numerical stability of the RELAP5 one-dimensional two-fluid model in nuclear reactor (PWR) accident conditions is determined. 3. The effect of the addition of diffusion terms on the linear stability of the basic two-fluid model is determined. 4. The effect of the addition of a bubble collision force on the linear stability of the basic two-fluid model is determined. 5. Performed sensitivity analysis of RELAP5 one-dimensional two-fluid model
The water-cooled lithium-lead breeding blanket is in the pre-conceptual design phase. It is a candidate option for European DEMO nuclear fusion reactor. This breeding blanket concept relies on the liquid lithium-lead as breeder-multiplier, pressurized water as coolant and EUROFER as structural material. Current design is based on DEMO 2017 specifications. Two separate water systems are in charge of cooling the first wall and the breeding zone: thermo-dynamic cycle is 295–328 °C at 15.5 MPa. The breeder enters and exits from the breeding zone at 330 °C. Cornerstones of the design are the single module segment approach and the water manifold between the breeding blanket box and the back supporting structure. This plate with a thickness of 100 mm supports the breeding blanket and is attached to the vacuum vessel. It is in charge to withstand the loads due to normal operation and selected postulated initiating events. Rationale and progresses of the design are presented and substantiated by engineering evaluations and analyses. Water and lithium lead manifolds are designed and integrated with the two consistent primary heat transport systems, based on a reliable pressurized water reactor operating experience, and six lithium lead systems. Open issues, areas of research and development needs are finally pointed out.
This work attempts to investigate the thermal hydraulic safety of lithium lead ceramic breeder (LLCB) test blanket system (TBS) in International Thermonuclear Experimental Reactor (ITER) with the help of modified thermal hydraulic code RELAP/SCDAPSIM/MOD4.0. The design basis accidents, in-vessel and ex-vessel loss of coolant of first wall (FW) of test blanket module (TBM) are analyzed for this safety assessment. The sequence of accidents analyzed was started with postulated initiating events (PIEs). A detailed modeling of first wall helium cooling system (FWHCS) loop and lithium lead cooling system (LLCS) is presented. The analysis of steady-state normal operation along with 10 s power excursion before the accident is also discussed in order to better understanding of initial condition of accidents. The analysis discusses a number of safety concerns and issues that may result from the TBM component failure, such as vacuum vessel (VV) pressurization, TBM FW temperature profile, passive decay heat removal capability of TBM structure, pressurization of port cell and Tokomak cooling water system vault annex (TCWS-VA) and to check the capability of passive safety system (vacuum vessel pressure suppression system (VVPSS)). The analysis shows that in these accident scenarios, the critical parameters have reasonable safety margins.
The water-cooled lithium–lead breeding blanket is a candidate option for the European Demonstration Power Plant (DEMO) nuclear fusion reactor. This breeding blanket concept relies on the liquid lithium–lead as breeder–multiplier, pressurized water as coolant, and EUROFER as structural material. The current design is based on DEMO 2015 specifications and represents the follow-up of the design developed in 2015. The single-module-segment approach is employed. This is constituted by a basic geometry repeated along the poloidal direction. The power is removed by means of radial–toroidal (i.e., horizontal) water cooling tubes in the breeding zone. The lithium–lead flows in a radial–poloidal direction. On the back of the segment, a 100-mm-thick plate is in charge of withstanding the loads due to normal operation and selected postulated initiating events. Water and lithium–lead manifolds are designed and integrated with a consistent primary heat transport system, based on a reliable pressurized water reactor operating experience, and the lithium–lead system. Rationale and features of the single-module-segment water-cooled lithium–lead breeding blanket design are discussed and supported by thermo-mechanic, thermo-hydraulic, and neutronic analyses. Preliminary integration with the primary heat transfer system, the energy storage system, and the balance of plant is briefly discussed. Open issues, areas of research, and development needs are finally pointed out. @EUROfusion Consortium*, 2017. *For more details see http://www.euro-fusionscipub.org/disclaimer-copyright
Water-cooled lithium-lead breeding blanket is considered a candidate option for European DEMO nuclear fusion reactor. ENEA and the linked third parties have proposed and are developing a multi-module blanket segment concept based on DEMO 2015 specifications. The layout of the module is based on horizontal (i.e. radial-toroidal) water-cooling tubes in the breeding zone, and on lithium lead flowing in radial-poloidal direction. This design choice is driven by the rationale to have a modular design, where a basic geometry is repeated along the poloidal direction. The modules are connected with a back supporting structure, designed to withstand thermal and mechanical loads due to normal operation and selected postulated accidents. Water and lithium lead manifolds are designed and integrated with a consistent primary heat transport system, based on a reliable pressurized water reactor operating experience, and the lithium lead system. Rationale and features of current status of water-cooled lithium-lead breeding blanket design are discussed and supported by thermo-mechanics, thermo-hydraulics and neutronics analyses. Open issues and areas of research and development needs are finally pointed out.
The availability of a qualified system code for the deterministic safety analysis of the in-box LOCA postulated accident is of primary importance. Considering the renewed interest for the WCLL breeding blanket, such code shall be multi-phase, shall manage the thermodynamic interaction among the fluids, and shall include the exothermic chemical reaction between lithium-lead and water, generating oxides and hydrogen. The paper presents the implementation of the chemical correlations in SIMMER-III code, the verification of the code model in simple geometries and the first validation activity based on BLAST Test N°5 experimental data.
The water-lithium lead interaction implies a direct energy release, which leads to temperature and pressure increase, due to a combined thermal and chemical reaction, and an indirect form of energy release, the hydrogen production, due to secondary chemical reaction involving the initial reaction products. Review and understanding of the knowledge acquired in past studies, experimental works and numerical activities are needed in view of the renewed interest in the Water Cooled Lithium Lead blanket concept and safety issues connected with the fusion reactor design. This paper presents a review of the studies carried out in the past to characterize the potential safety concerns associated with the use of water and lithium-lead eutectic alloy, the main experimental campaigns, and numerical simulations of BLAST Test No. 5 performed by SIMMER-III code. As results, no code was found able to perform a satisfactory post-test analysis of separate effect experiments, without engineering assumptions. Therefore, a code model for the exothermic reaction and hydrogen production, and experimental data are needed for solving the WCLL blanket safety issues associated with the water-PbLi interaction.
In the frame of the THINS Project, an experimental campaign was performed on LIFUS5/Mod2 facility at ENEA RC Brasimone, aiming to investigate the water–LBE interaction. Such a phenomenon occurs as consequence of a postulated Steam Generator Tube Rupture event in a HLMFR system. Four tests were performed injecting sub-cooled water at 40 bar into a reaction vessel partially filled by low pressure LBE at 400 °C.
The post-test activity was performed by the SIMMER-III code in order to improve the understanding of the involved phenomena and to confirm the code capabilities in simulating the water–LBE interaction. The calculated data showed a qualitatively agreement with the measured values and a faster reaction kinetics due to the modelling assumptions.
The second, revised edition of this comprehensive handbook by one of the leading experts in the field of hydraulic engineering has been completely updated and will be a valued addition to the literature. Partial Contents: General Information and Elements of Aerodynamics and Hydraulics of Pressure Systems; Flow in Straight Tubes and Conduits; Flow at the Entrance into Tubes and Conduits; Flow through Orifices with Sudden Changes in Velocity and Flow Area; Flow with a Smooth Change in Velocity; Flow with Changes in Stream Direction; Merging of Flow Streams and Division into Flow Streams; Flow through Barriers Uniformly Distributed over the Channel Cross Section; Flow through Pipe Fittings and Labyrinth Seals; Flow Past Obstructions in a Tube; Flow at the Exit from Tubes and Channels; Flow through Various Types of Apparatus.
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