Ex-vessel Loss Of Coolant Accident Research Articles

Dynamic Event Tree Analysis as a Tool for Risk Assessment in Nuclear Fusion Plants Using RAVEN and MELCOR

In the broad framework of the nuclear power plants (NPPs) industry, the dynamic probabilistic risk assessment could answer the time dependence deficiency of the event tree and fault tree analysis. The basic event tree approach relies on experts’ pre-constructed accident sequences without exploring the time-dependent nature of an accident scenario, which could strongly affect the accident sequence. Conversely, the effects of events timing can be studied by adopting a dynamic event tree (DET) approach. Developing a DET methodology requires integrating a system code capable of replicating an accident scenario and a logic-driver code able to generate the event tree sequence, trigger plant safety systems, and manage other relevant events throughout the simulation. For this purpose, Methods for Estimation of Leakages and Consequences of Release (MELCOR) and reactor analysis and virtual control environment (RAVEN) have been coupled through a Python script developed by the Sapienza University of Rome, Rome, Italy, to perform DET studies during accident transients in fusion and fission reactors. RAVEN is a software tool developed at the Idaho National Laboratory (INL), Idaho Falls, ID, USA, to act as a control logic driver and post-processing tool for different applications. MELCOR for fusion is a fully integrated design basis and severe accident code that simulates thermal-hydraulic behavior and self-consistently accounting for aerosol transport in nuclear facilities and reactor cooling systems for the evaluation of the source term in fusion reactors. The coupling between these codes will provide a wide range of NPP risk assessment analyses, establishing new best practices. In this work, a preliminary DET study has been performed, selecting as initiating event an ex-vessel loss of coolant accident (LOCA) in the water-cooled lithium lead test blanket system (WCLL TBS) to be tested in International Thermonuclear Experimental Reactor (ITER). Time-dependent parameters such as the intervention of the plasma shutdown system and the closure of the main system isolation valves have been sampled to study evolving system scenarios.

Preliminary accident analyses of ex-vessel LOCA and TES pipe break for CFETR HCCB blanket system

Safety analysis is one of the most important parts in the design of nuclear power reactor. It is a necessary step for China Fusion Engineering Test Reactor (CFETR) to make the research in safety accidents analysis of main coolant systems and tritium systems. The consequences of nuclear accidents can be used to prove the safety of plant and optimize the systems design. Now, the main design parameters of CFERR have been determined, and the conceptual design of the main systems have been done. This paper presents two design basis accidents (ex-vessel loss of coolant accident (LOCA) and TES pipe break) analysis results in primary helium cooling system (HCS), and tritium extraction system (TES). According to the accident analysis results, the radioactive release assessments are given. The analysis results can provide reference for the following parameter optimization design of CFETR systems.

Stress Analysis of HCPB BB under Ex-Vessel LOCA Condition

Helium Cooled Pebble Bed Breeding Blanket (HCPB BB) is a kind of concept for the European demonstration fusion reactor (DEMO). The blanket attachment system plays an important role in the mechanical connection of the BB and vacuum vessel. Typically, the mechanical and thermal loads should meet the requirement to avoid collapse of the system with off-normal conditions, e.g., under ex-vessel Loss of Coolant Accident (LOCA. This paper investigates the loading requirement corresponding to the maximum stress that can sustain to avoid the LOCA condition. Firstly, a model of the BB is constructed using SolidWorks. Then, stress analysis is carried out based on the cross section of the blanket. Through simulation, the critical condition for the LOCA case and the maximum stress value for the model are obtained. According to the relevant size dimension from the reference, the blanket’s cross section is drawn, and one can get the stress field under the ex-vessel LOCA through stress analysis. The stress distribution under the ex-vessel LOCA condition is simulated to find out the maximum stress field that the blanket can sustain through this paper. The significance is to predict the possible conditions leading to an accident and find possible methods to avoid them.

Preliminary sensitivity analysis for an ex-vessel LOCA without plasma shutdown for the EU DEMO WCLL blanket concept

Abstract In this early development phase of the DEMO design the uncertainty affecting many operational and design parameters can modify main outcomes of accident scenario aiming at studying the critical conditions for the vacuum vessel and the contiguous containment volumes. The aim of this paper is to perform a preliminary sensitivity analysis of an accident progression predicted by MELCOR code considering selected parameters as a figure of merit to predict possible code outcomes. The uncertainty band will be evaluated through sensitivity analyses programmed, collected and statistically manipulated through RAVEN software tool. MELCOR and RAVEN have been internally coupled through a new Python code interface developed by Sapienza University of Rome, to perform sensitivity and uncertainty quantification analyses during severe accident transient. The Beyond Design Basis Accident (BDBA) analysis of an ex-vessel loss of coolant accident (LOCA) for the water-cooled lithium lead (WCLL) blanket concept has been simulated with the fusion version of MELCOR 1.8.6 code. The postulated initiating event (PIE) is a double-ended break in the first wall (FW) cooling system distributor ring, with simultaneous failure of the plasma shutdown system. An in-vessel breach of the coolant system occurs because of FW failure, with consequent unmitigated plasma shutdown transient. Sensitivity analysis results have shown that the FW temperature at which plasma in-vessel breach occurs is strongly correlated with the mass of hydrogen produced. The same parameter has also an impact on the overall accident scenario, such as the trigger of VVPSS rupture disks and thus source term mobilization.

Development of plant concept related to tritium handling in the water-cooling system for JA DEMO

The conceptual design of Japan’s fusion demonstration plant (JA DEMO) is now being developed. In this paper, an overall plant system concept related to tritium handling in the water-cooling system is developed to give a concrete shape to the present JA DEMO concept as an electric power plant. The basic condition of tritium permeation from the in-vessel components to the primary cooling system is evaluated to be 5.7 g-T/day. The tritium concentration of the primary coolant is assumed to be 1 TBq/kg similar to the heavy water reactor condition. The capacity of the water detritiation system (WDS) is assessed, and the bypass feed water from the primary cooling loop is evaluated to be 94 kg/h under the tritium extraction efficiency of 0.96. Based on those specific parameters, the existing WDS in the heavy water reactor is found to be applicable to that of JA DEMO. Configuration of the primary heat transfer system (PHTS) is also discussed. Based on the heavy water reactor experience, tritium permeation through a steam generator (SG) to the secondary cooling system in PHTS is evaluated at 11.77TBq/year/loop (318 Ci/year/loop), which is found to be less than the restricted amount of tritium disposal for a pressurized water reactor in Japan. The key effect of the heavy water reactor experience is reduction of tritium permeation by oxide layer formed on SG pipes. Finally, a confinement concept of tritium release from PHTS is discussed under the condition of ex-vessel loss of coolant accident (LOCA). A pressure suppression system is installed to prevent the upper tokamak hall from pressurizing at the ex-vessel LOCA, and the tritium leakage from the upper tokamak hall is consequently restrained. The resultant early public dose at the plant site boundary can be reduced to 1.8 mSv, which is negligibly smaller than 100 mSv of the no-evacuation limit recommended by IAEA.

First thermal-hydraulic analysis of a CO2 cooled pebble bed blanket for the EU DEMO

Over the years, the Helium Cooled Pebble Bed (HCPB) Breeding Blanket (BB) has been intensively studied for the EU DEMO. However, several feasibility issues remain open for a HCPB-type DEMO reactor. Some of these issues are linked with the use of He as coolant and are related to the large size of the Primary Heat Transfer System pipework, the resulting large coolant inventory and therefore large expansion volumes required in an ex-vessel loss of coolant accident, the limited operational experience with relevant size He-turbomachinery, the large circulating power and possible considerable pipe leakage, among others. Due to the larger density of CO2, the use of this gas as primary coolant for DEMO can lead to key advantageous features, mitigating most of the issues present for He-cooling and resulting in a higher net efficiency than that of HCPB, as reported in a previous study. Therefore, a CO2-Cooled Pebble Bed (CCPB) has been proposed as an alternative coolant to He for the EU DEMO. After identifying that CO2 have a negligible influence on the neutronic performance, making the CCPB’s TBR almost equal to the HCPB’s one (TBR ≈ 1.15), a first set of thermal-hydraulic analyses with the commercial code of ANSYS CFX are reported here. The analyses are based on the newly proposed design of breeding zone (BZ) in the enhanced HCPB fuel-breeder pin concept for the EU DEMO. Despite the lower heat transfer capability of CO2 compared to He, the fuel-breeder pin design BZ improves the thermo-hydraulic performance to a point that this coolant is able to meet the materials’ temperature requirements. The results show that the CCPB can satisfy the basic thermal hydraulic requirements and that CO2 is also a realistic option as primary coolant for gas-cooled fusion reactors, with the key advantage of the existing >60 years operational experience in fission industry with this coolant.

Preliminary steady and transient analysis for the CFETR helium cooled solid blanket system with RELAP

Abstract The helium cooled solid breeder (HCSB) blanket, which is one of the blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR), has been developed in University of Science and Technology of China (USTC). And the blanket cooling system (HCS) is also designed for HCSB. The steady-state results show that the blanket system has a good relevance to the design parameters. The transient analysis of possible accidents should proceed to evaluate its safety performance. The accident analysis in this paper adopts the ITER-like conditions and the ex-vessel loss of coolant accident (LOCA) case is investigated with RELAP5 code. The results show that the ex-vessel LOCA with detection couldn't cause serious consequence, however the ex-vessel LOCA without detection has potential to melt the blanket that we should take measures to avoid it.

Thermal–Hydraulic Analysis for One Water-Cooled Blanket Module of CFETR Based on RELAP5

The water-cooled ceramic breeder blanket (WCCB) is one of the blanket candidates for the Chinese fusion engineering test reactor (CFETR). The conceptual of WCCB for CFETR under 200-MW fusion power has been designed based on pressurized water-cooled reactor technology. RELAP5 code, which is mature and often used in transient thermal–hydraulic analysis in pressurized water reactor, was selected as the simulation tool. In this paper, the nodal model for RELAP5 was developed corresponding to typical WCCB module, i.e., the coolant passages inside the module were nodalized as hydrodynamic components, and the associated module components were simulated with 1-D heat structures. The steady-state characteristic under full power was analyzed. The stable fluid, wall temperature distributions, and pressure drops were studied. The results were compared with those of the 2-D and 3-D computational fluid dynamics analyses by FLUENT. Furthermore, the transient characteristics under three accidental scenarios, in-vessel loss of coolant accident (LOCA), in-box LOCA, and ex-vessel LOCA, were analyzed, respectively. Simulation results show that the cooling of stiffening plates and side walls for breeder zone is very important for the blanket module. All the pressurization of the related volumes is within the design limits. The temperature of the structures does not exceed the limits for a relatively long time that allows mitigation measures to be taken. As for in-box LOCA, the exclusion of Be–water reaction makes the temperature nonconservative and still needs further study.

Transient thermal analysis and structural assessment of an ex-vessel LOCA event on the EU DEMO HCPB breeding blanket and the attachment system

In the framework of the EUROfusion Consortium the Helium Cooled Pebble Bed (HCPB) breeding blanket (BB) is being developed. A blanket attachment system is also under development, which allows the integration of the BB into the EU DEMO tokamak. The blanket attachment system is thereby responsible for the mechanical connection of the BB to the DEMO vacuum vessel (VV). It transfers the mechanical loads between BB and VV. The high mechanical and thermal loads on the blanket segments during normal and off-normal conditions especially pose strong requirements to the blanket attachment system. This article focuses on the off-normal event of an ex-vessel Loss Of Coolant Accident (LOCA) in the HCPB BB. The performance of its blanket attachment system is assessed as this event will induce a large temperature increase and thus an additional thermal expansion in the blanket, posing big challenges for the blanket attachment system and the blanket supporting structure itself. A 3D transient thermal analysis under an ex-vessel LOCA condition of the HCPB BB is presented. Furthermore structural assessment of HCPB blanket segments and the current blanket attachment system with respect to RCC-MRx rules is reported, giving also an estimation of the resulting reaction loads to the VV during such an accidental event. The Finite Element Method (FEM) software ANSYS 18.0 is used for the thermal and structural analysis.

Preliminary safety assessment on two LOCAs under ITER-like condition for the optimized CFETR helium cooled solid breeder blanket

A conceptual design of helium cooled solid breeder (HCSB) blanket, one of three blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR), has been performed recently. The transient analysis for different possible accidents should be carried out to assess its safety performance. In this paper, the ITER-like conditions are adopted since the associated system for CFETR is missing, such as helium cooling system, plasma shutdown condition, pump behavior, etc. The complete model is recreated including the optimized typical outboard HCSB blanket (NO.12) and its ancillary helium cooling loop, and accident analyses of two loss of coolant accident (LOCA) cases are investigated using RELAP5. The influences of different break areas under in-vessel LOCA are compared, and the accident consequence after small area break is chosen to be investigated. Regarding the ex-vessel LOCA, the influences of different break locations are thoroughly analyzed. Since the plasma cannot terminate passively after ex-vessel pipe break, the plasma termination behaviors are investigated with different shutdown time. The computational results show that with the safety criteria for ITER the HCSB blanket can be cooled down effectively by the helium cooling system (HCS) and the integrity of pressure barriers can be guaranteed for both accidents.

Ex-Vessel Loss of Coolant Accident Analysis of ITER Divertor Cooling System Using Modified RELAP/SCADAPSIM/Mod 4.0

The initial design of ITER incorporated the use of carbon fiber composites in high heat flux regions and tungsten was used for low heat flux regions. The current design includes tungsten for both these regions. The present work includes thermal hydraulic modeling and analysis of ex-vessel loss of coolant accident (LOCA) for the divertor (DIV) cooling system. The purpose of this study is to show that the new concept of full tungsten divertor is able to withstand in the accident scenarios. The code used in this study is RELAP/SCADAPSIM/MOD 4.0. A parametric study is also carried out with different in-vessel break sizes and ex-vessel break locations. The analysis discusses a number of safety concerns that may result from the accident scenarios. These concerns include vacuum vessel (VV) pressurization, divertor temperature profile, passive decay heat removal capability of structure, and pressurization of tokamak cooling water system. The results show that the pressures and temperatures are kept below design limits prescribed by ITER organization.

Thermal hydraulic responses of the Primary Heat Transfer System of the WCCB blanket to accident cases for CFETR

The Water Cooled Ceramic Breeder (WCCB) blanket is one of the blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR). In this work, the Primary Heat Transfer System (PHTS) of the WCCB blanket was designed based on the configuration of the blanket sectors, employing two identical loops at this stage. Each loop consists of a steam generator, a pressurizer and two pumps, feeding water coolant into each blanket modules individually of 8 blanket sectors. One of the loop was modeled using RELAP5/MOD3.3 under normal condition and accident cases. The operational mode of PHTS was carefully chosen so as to obtain a more stable hydraulic behavior under steady state, due to the anisotropy of geometry structures and heat sources. Enveloping accidental cases, including Loss of Flow Accident (LOFA), in-vessel Loss of Coolant Accident (LOCA), and ex-vessel LOCA, were selected to preliminarily evaluate the safety performance of the system. The results show that the integrity of the confinement barriers can only be ensured in limited period of time during LOCAs. Additional safety facilities are needed and mitigation methods: are discussed.

Accident analyses of selected postulated events for safety assessment of Indian LLCB TBS in ITER

The Lead Lithium cooled Ceramic Breeder (LLCB) TBM) is being developed by India for testing in ITER machine. Thermal hydraulic safety analyses are the essential part of engineering design of LLCB Test Blanket System (TBS) for safe operation in ITER environment. The current work involves thermal hydraulic analyses and preliminary results of selected accident scenarios involving ancillary systems, the First Wall Helium Cooling System (FWHCS) and the Lead Lithium Cooling System (LLCS). A modified version of RELAP/MOD4.0 safety analysis code is used to simulate the transient scenarios. A detailed nodalization is carried out with all the major components in FWHCS and LLCS along with associated heat structures and connected confinement volumes. The accident analysis is performed for the scenarios including the ex-vessel Loss Of Coolant Accident (LOCA) in FWHCS into Port Interspace, Loss Of Fluid Accident (LOFA) and Loss Of Heat Sink (LOHS) accident in FWHCS and LLCS. Transient analyses results of each accident case are reported here and it is found that current design meets most of the thermal hydraulic safety requirements while changes in the design are proposed for certain accident scenarios.

LOCA Accident for the DEMO Helium Cooled Blanket

The exploitation of Fusion as energy source requires also the demonstration of a limited impact in terms of risk to the staff, to the public, and to the environment, well below the limits established by international committees and national safety authorities. Therefore, a systematic safety analysis has to follow the design development to demonstrate that the safety objectives are met for each proposed solution. This analysis points out the dominant accident sequences and outlines the possible prevention, protection and mitigation actions and their associated systems. This analysis points out the dominant accident sequences and outlines the possible prevention, protection and mitigation actions and their associated systems. One of the most challenging accidents is a large break Loss of Coolant Accident (LOCA) of the Primary Heat Transfer System (PHTS) outside the Vacuum Vessel (VV), due to the possible consequences in terms of radiological releases to the environment. However, because of the relative small radiological inventory and to the lower decay heat density, the risk associated with a break of the primary cooling loop in a fusion reactor is lower than the risk of the same event in a fission reactor. Nevertheless the consequent peak of pressure in the Expansion Volume located within the Tokamak Building could severely impact the confinement function, hence the overall safety of the plant. For this purpose a numerical assessment of a blanket PHTS ex-vessel LOCA has been carried out considering two possible layout solutions. This analysis has been performed employing MELCOR 1.8.2 and aims to support the design of the Blanket and its PHTS with some safety-related considerations.

Revisiting the analysis of passive plasma shutdown during an ex-vessel loss of coolant accident in ITER blanket

In this contribution, the analysis of passive safety during an ex-vessel loss of coolant accident (LOCA) in the first wall/shield blanket of ITER has been studied with AINA safety code.In the past, this case has been studied using robust safety arguments, based on simple 0D models for plasma balance equations and 1D models for wall heat transfer. The conclusion was that, after first wall heating up due to the loss of all coolant, the beryllium evaporation in the wall surface would induce a growing impurity flux into core plasma that finally would end in a passive shut down of the discharge.The analysis of plasma-wall transients in this work is based in results from AINA code simulations. AINA (Analyses of IN vessel Accidents) code is a safety code developed at Fusion Energy Engineering Laboratory (FEEL) in Barcelona. It uses a 0D/1D architecture, similar to that used for past analyses.Initial results show good agreement with previous studies, plasma shutdown happens before significant melting occurs in first wall. However, after a parametric survey a new scenario was identified with the potential not only to melt beryllium but also the copper heat sink. If the coolant tubes were compromised in such scenario, the in-vessel radioactive inventory could be mobilized. http://nerg.upc.edu/

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 and safety analysis for Chinese helium-cooled solid breeder TBM cooling system

The thermal–hydraulic behavior and safety performance of the Chinese helium-cooled solid breeder (CH HCSB) test blanket module (TBM) with helium cooling system (HCS) has been studied using RELAP5/Mod3.4 code. According to accident analysis specification for TBM, two design basis accidents including loss of off-site power and TBM first wall (FW) ex-vessel coolant pipe break are investigated. The influences of different break locations and plasma termination behaviors are analyzed comprehensively. The results show that natural circulation is established in helium cooling circuit and the TBM can be cooled effectively after loss of off-site power. It is much more critical when the pipe break occurs at the downstream side of the circulator compared with that of upstream side of the circulator. In case of a more serious accident that the ex-vessel break extends to the TBM FW, the results reveal that TBM could be cooled down by natural circulation and radiation. In addition, at the beginning of ex-vessel loss of coolant accident (LOCA), large temperature difference between break and intact TBM FW pipes is found. The accidental results finally show that the integrity of the FW can be guaranteed if the plasma is terminated with a 3s delay time by fusion power shutdown system (FPSS) in the case of ex-vessel LOCA.

Preliminary safety analysis of ex-vessel LOCA for the European HCPB TBM system

Ex-vessel loss of coolant accident caused by a double-ended pipe break of the helium coolant system inside port cell is considered as one of the most critical accident for the European Helium Cooled Pebble Beds Test Blanket Module (HCPB TBM) system. The resulting rapid helium blow-down causes an immediate block of the TBM cooling, which requires a prompt plasma shutdown. Even after the plasma shutdown the temperature can increase over the design limit and the accident sequence can lead up to a break of the TBM box protection after the failure of different protection systems. Thus air ingresses in the vacuum vessel from the damaged TBM system and steam from the surrounding ITER blanket and divertor structures. The evaluation of this sequence is very important for the definition of the correct protection strategy of the system. To consider all these different events a methodology has been developed in KIT combining different codes for a complete analysis of the accident. In particular, this paper shows an application of MELCOR code to model beryllium–steam reaction in a particular accidental sequence for the long term cooling.

ITER divertor ex-vessel pipe break

This is a study of a divertor (DV) ex-vessel Loss Of Coolant Accident (LOCA), performed for the ITER reactor. One base case and two parameter studies are performed. The calculations are performed with the fusion MELCOR code. The initiating event, for all three analyses, is a double-ended pipe break in the DV cooling loop, which is postulated to occur in a large diameter pipe at the inlet of the main pump. The break flow is discharged into the Tokamak Cooling Water System (TCWS) vault. This is followed by plasma disruption and a break of the DV heat structure cooling pipes, discharging coolant into the Vacuum Vessel (VV). The results show, that the pressures and temperatures are kept below design limits. The results also show that activated corrosion products (ACP) and tritium are released into the TCWS vault from the coolant water, and that VV dust does not enter into the DV coolant loop, and thus does not reach the TCWS vault.

Safety approach for plasma anomaly events in fusion reactor: Code development and assessments

A safety assessment method for plasma transients, including thermal response of in-vessel components, was provided. We developed a plasma physics model for safety analysis which has been implemented in a safety analysis code (SAFALY). The SAFALY code consists of a zero-dimensional plasma dynamics model and a 1.5-dimensional thermal behavior model of in-vessel components in the thickness direction. The code can treat hydraulic accidents using the results from a hydraulic code and analyze a passive plasma shutdown due to the impurity release from the wall. The abnormal events in the International Thermonuclear Experimental Reactor (ITER) calculated here are: (i) loss of plasma control, and (ii) large ex-vessel loss of coolant accident (LOCA) in the first wall and large ex-vessel LOCA in the divertor. The plasma transient and thermal response of in-vessel components in such events have been assessed using the code.