Coolant Accident Research Articles

CFD simulation of thermal stratification and mixing in a Nordic BWR pressure suppression pool

Boiling Water Reactor (BWR) employs the Pressure Suppression Pool (PSP) as a heat sink to prevent overpressure of the reactor vessel and containment. Steam can be injected into the PSP through spargers in normal and accident conditions and through blowdown pipes in case of a loss of coolant accident (LOCA). There is a safety limit on the maximum PSP temperature at which such steam injection might cause dynamic loads on the containment structures. The performance of the pool can be affected if thermal stratification is developed when temperature of the hot layer grows rapidly while cold layer remains inactive. Simulation of pool behavior during realistic accident scenarios requires validated models that can sufficiently address the interaction between phenomena, safety systems and operational procedures. Direct modeling of steam injection into a water pool in long-term transients is computationally expensive due to the need to resolve simultaneously the smallest space and time scales of individual steam bubbles and the scales of the whole PSP. To enable PSP analysis for practical purposes, Effective Heat source and Effective Momentum source (EHS/EMS) models have been proposed that avoid the need to resolve steam-water interface. This paper aims to implement mechanistic approaches previously developed by authors for the simulation of transient thermal stratification and mixing phenomena induced by steam injection through spargers in a Nordic BWR PSP. The latest version of the EHS/EMS models using the ‘Unit cell’ approach has been validated against integral effect pool tests and applied to plant simulations. Several scenarios with boundary conditions corresponding to postulated accident sequences were simulated to investigate the possibility of stratification development and the effects of activation of different systems (e.g., blowdown pipes, high momentum nozzle) on the pool behavior.

Effect of inclination on SB LOCA transient of a floating nuclear reactor in MARS-KS analysis using moving reactor model

As the imperative to reduce Green House Gas emissions grows within the maritime sector, nuclear power emerges as a promising solution. Especially, offshore floating nuclear power plants have garnered significant attention and development efforts from various institutions and nations. These types of reactors provide unique benefits such as extended refueling periods for advanced safety and exceptional mobility, facilitating access to remote regions. However, the external forces are induced by operating on a moving condition, that can impact the thermal–hydraulic characteristics of various hydraulic components in the entire reactor system. Addressing the safety issues of offshore floating reactors requires conducting thermal–hydraulic analyses under diverse oceanic motions. To achieve this objective, this study centered on analyzing accidents in an offshore floating reactor, specifically the BANDI-60, targeting the Small Break Loss of Coolant Accident induced by a double-ended guillotine break in the Direct Vessel Injection line. This study utilizes the system thermal–hydraulic code MARS-KS moving reactor model considering various tilting conditions and a BANDI-60 nodalization with a quarter-divided reactor pressure vessel and containment vessel volumes in the circumferential direction. The inclination conditions include static inclination along the x and y directions. Significantly, static inclination, particularly when the break aligned downward with the inclination direction, emerged as the most critical condition affecting reactor core integrity. Additionally, the containment vessel volume affects the equilibrium pressure after a break occurs, with smaller volume resulting in smaller peak cladding temperature values in static inclined conditions.

Investigation of containment spraying characteristics based on the correction of droplet evaporation shielding effects

Historically, the spray models in the accident analysis program are mostly based on the isolated droplet and ignore the shielding effects among droplets, which theoretically introduces some errors. This paper investigates the effect of multi-droplet shielding on spray evaporation by taking the containment spray systems as the research object. By discussing the correction factors, the law of droplet group shielding effects can be studied, and then the correction of the spray package can be realized by the self-programming. After that, the transient response characteristics of temperature and pressure in the containment before and after the model correction are studied by modeling the loss of coolant accident and main steam line break accident, respectively. And, this research tries to analyze the influence of the cell normalization model as well as spray parameters on the calculation results. The results reveal that the program can achieve more accurate prediction after being modified by the shielding effects, and the multi-cell model can capture the temperature stratification effects more accurately under accident conditions. Additionally, the spray temperature and spray sizes have an insignificant effect on the in-shell thermal hydraulics response characteristics, but the increase of the spray flow rate has a better effect on the mitigation of the temperature stratification effects and the suppression of the in-shell pressure. The research in this paper can provide some reference or research ideas for accurate modeling and safety analysis of accidents.

Revisiting the thermodynamic properties of the ZrCr2 Laves phases by combined approach using experimental and simulation methods

A comprehensive study of ZrCr2 Laves phases is important for both fundamental research and technological applications. The ZrCr2 compound is a typical precipitate observed in Zr-based alloys including the newly developed Cr-coated Zr cladding known as Accident Tolerant Fuel cladding. The brittle behavior of the Laves phases and the low melting point at the interface zone between the precipitates and the Zr matrix are considered as governing the thermomechanical behaviors of these newly developed nuclear fuel claddings for normal operating conditions but more dramatically in case of Loss Of Coolant Accident (LOCA) conditions. The aim of the present study is to investigate the various ZrCr2 Laves polymorphs and their thermodynamic features particularly where conflicts exist in the literature. Based on diffraction experiments of annealed samples, it has been established that the transformation from the C14 high-temperature form to the C15 low-temperature form is of displacive nature without the C36 polymorph forming as an intermediate phase. The stable and metastable domains of C14 and C15 and their thermodynamic properties have been determined. The specific heat of both C14 and C15 types was measured over a wide range of temperatures from 2 to 1063 K by coupling relaxation calorimetry and DSC. The experimental data were fitted using a modified Einstein model. The room temperature entropies of the C14 and C15 phases were evaluated. The enthalpy of formation was measured for the first time using drop solution calorimetry in liquid Al at 1173 K in a Tian–Calvet calorimeter. The experimental value is in good agreement with Density Functional Theory (DFT) calculations. Finally, all these new results are discussed regarding the abundant literature on the ZrCr2 Laves phases.

Thermal safety characteristics analysis of helium xenon gas cooled small reactor system under LOCA accidents

The thermal safety characteristics of the system after a small break loss of coolant accident (SB-LOCA) in a helium xenon gas-cooled small reactor will determine whether the reactor will experience a meltdown and the duration of the meltdown. At present, research on SB-LOCA both domestically and internationally is mostly focused on water-cooled reactors, lacking relevant research on helium xenon gas-cooled reactors (He-Xe GCR). Therefore, it is of great significance to research the thermal safety characteristics of He-Xe GCR after a breach loss of coolant accident. This article improves the gas turbine model based on the original SB-LOCA analysis program for pressurized water reactors. And added a calculation model for the characteristics of helium xenon mixed gas, a heat transfer model for helium xenon mixed airflow, and a compressor model. Finally, a Brayton Cycle Reactor System Analysis program (BRESA) was developed for SB-LOCA analysis of He-Xe GCR. Explore the impact of the size and location of fractures on the thermal safety characteristics of the system using BRESA. Research has found that as the size of the rupture gradually increases, the decrease in reactor core power is greater, and the reactor power is shut down earlier. Comparative analysis of the impact of different breach positions on the residual heat removal characteristics reveals that a breach accident on the high-pressure side of the system poses a greater threat to system safety. Therefore, when setting the safety threshold for the Brayton cycle system, the focus should be on the hazards of high-voltage side breaches. This research achievement assists with the safe operation strategy of He-Xe GCR in the future.

NRC’s methodology to estimate fuel dispersal during a large break loss of coolant accident

Recent experimental findings indicate that under certain transient conditions, high-burnup fuel operating at sufficiently high power can fragment and escape from burst fuel cladding, as discussed in the Nuclear Regulatory Commission’s Research Information Letter (RIL) 2021–13. The escaped fuel fragments could subsequently be distributed throughout a reactor coolant system (RCS), potentially challenging core coolability during a loss of coolant accident (LOCA). Consequently, the NRC has developed a methodology to estimate the mass of fuel that could be dispersed during a large-break LOCA. The methodology involves several NRC-sponsored computer codes, including SCALE/Polaris for lattice physics, PARCS/PATHS for full-core neutronics, TRACE for core and RCS thermal hydraulics, and FAST for steady-state and transient fuel performance calculations. Output from FAST is combined with the models in RIL 2021–13 to estimate fuel dispersal to the coolant.NRC staff have exercised this methodology using a prospective Westinghouse four loop core design with high burnup and extended enrichment that was obtained from the Department of Energy’s NEAMS program. Results demonstrate that the NRC methodology can be exercised to provide fuel dispersal estimates that can subsequently be used to study the potential consequences of FFRD. However, this exercise has also identified several areas for improvements to both the codes and to the modeling approaches used here, including the pin power reconstruction methods in PARCS, the modeling approach used to connect the Cartesian and cylindrical vessel components in TRACE, transient fission gas release and fuel rod upper plenum models in FAST, and tighter coupling between FAST and TRACE.

Corium interface flow dynamics investigation during severe accident in pressurized water reactors using compressive advection interface capturing method

Past nuclear accident occurrences raised strong concerns which led to research on nuclear safety. One of the major causes of nuclear accidents is the impeded circulation of core coolant, leading to decay heat removal cessation and rapid temperature rise. If uncontrolled, this results in critical heat flux, loss of coolant accidents, and core dryout. Detailed melted core relocation (i.e., nuclear fuel, graphite, and zircaloy) needs to be investigated through interface capture and multimaterial flow model coupling, which have not been done in previous studies. This work aims to investigate the impacts of temperature and core material composition on the flow dynamics during core relocation. In this study, mass fraction is discretized using a streamlined upwind Petrov–Galerkin method spatially and a modified Crank–Nicolson method temporally to accurately capture fluid interfaces using a high-order accurate flux-limiter. Two core material composition cases (individual material properties case and bulk material properties case) were considered to assess the impact of temperature and core materials composition on both flow dynamics and computational time. Temperature has a significant impact on core material transport and corium flow dynamics during core relocation. Bulk materials properties case has greater impact of temperature on its corium resulting in faster materials transport, but with higher computation time.

Tailored high-temperature corrosion behavior of Cr coatings using high power impulse magnetron sputtering on ZIRLO alloys for accident-tolerant fuel application

To further enhance the oxidation resistance of Zirconium alloys during Loss of Coolant Accident (LOCA), protective Cr coatings were deposited in synchronized pulse bias mode through high power impulse magnetron sputtering (HiPIMS). The results showed the weight gain of the Cr coatings with synchronized pulse bias was reduced by approximately 61.7 % compared to that of the Cr coatings without synchronized pulse bias after 90 min of steam oxidation at 1200 °C. This was mainly attributed to the formed intense (200) preferred orientation of the Cr coatings with synchronized pulse bias during deposition, accelerating the recrystallization growth of the residual Cr layer in a high-temperature steam environment, thereby reducing the rapid diffusion pathways for oxygen. Moreover, the synchronized pulse bias induced Cr grains refinement, favoring the formation of a more compact Cr2O3 scale, and therefore efficiently inhibiting the diffusion of external oxygen to the underlying alloy.

Scaling analysis of nuclear power plant (APR1400) utilizing ATLAS integral effect test result of A5.1 (Counterpart test for LSTF 1% cold-leg break LOCA)

ATLAS is the thermal hydraulic integral effect test loop that can simulate the same pressure and temperature condition with nuclear power plant. It was designed with a height ratio of 1/2 and a volume ratio of 1/288 with reference to APR1400. The counterpart test against LSTF which was named as A5.1 test was performed within the OECD/NEA ATLAS international joint research project to compare major thermal hydraulic phenomena between two test facilities. The main scenario for the A5.1 test is the LSTF SB-CL-32 test. The SB-CL-32 test is a 1% horizontal small break loss of coolant accident (SBLOCA) at the cold leg with cold leg injection of the emergency core cooling system (ECCS) and accident management (AM) action.From the counterpart test results, the significant differences were found in the viewpoint of the thermal–hydraulic phenomena. It was concluded that the different result were caused by the different design characteristic of each test facility because they referred the different nuclear power plant. Thus, the applicability of the ATLAS test results to its reference nuclear power plant was evaluated by utilizing ATLAS A5.1 test results which was the counterpart test with the LSTF.The transient scenario and initial and boundary conditions of the A5.1 test were adopted to APR1400 after applying scaling ratios of ATLAS. The MARS-KS code was utilized to analyzed thermal hydraulic behaviors of APR1400 and the analysis results were compared with the scaled A5.1 test result. From the result of comparison, it was concluded that the ATLAS test results predicted the thermal hydraulic behavior of the nuclear power plant well. Thus, it was confirmed that ATLAS is a well-designed integral effect test facility which is designed with an appropriate scaling method.

A thermogravimetric comparative study of the high temperature steam oxidation of bare and pre-oxidized Zircaloy-4, M5Framatome and Optimized ZIRLO™

For Loss of Coolant Accident (LOCA) analysis, high-temperature oxidation kinetics in steam of Zircaloy-4 and the two Nb-bearing alloys used in French Pressurized Water Reactors (PWRs), M5Framatome11M5 and M5Framatome are trademarks or registered trademarks of Framatome or its affiliates, in the USA or other countries. and Optimized ZIRLO™22Optimized ZIRLO is a trademark or registered trademark of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners., have been studied using Thermo-Gravimetric Analysis (TGA) technique in the temperature range of 900–1200 °C. A specific test procedure allowing the use of a high steam flow rate and good steam supply to the whole specimen surface, as well as fast heating of the specimen, has been developed. Clean kinetic data, not affected by steam starvation, were obtained on the bare, as-received alloys. For as-received claddings, oxidation kinetics are comparable and close to parabolic for the three alloys at 1100 °C and above. Below 1100 °C, a significant deviation from the parabolic kinetic regime was observed, and M5 Framatome oxidizes slower than Zircaloy-4 and Optimized ZIRLO. Thanks to the fast-heating protocol, the influence of pre-oxidation, simulating corrosion in normal reactor operation, could be investigated, avoiding evolution of the pre-oxidation scale during a slow, long heating phase. Pre-oxidized Zircaloy-4 and M5 Framatome were investigated. A strong protective effect is observed below 1200 °C, while at 1200 °C, the pre-oxidation layer appears to be much less protective, possibly due to the monoclinic to tetragonal zirconia phase transformation.

Large pipe break opening time in pressurized heavy water reactors

In the frame of the licensing process of the AtuchaII PHWR (pressurized heavy water reactor) the BEPU (Best-Estimate Plus Uncertainty) approach has been selected for issuing Chapter 15 of FSAR (Final Safety Analysis Report) dealing with accident analysis. In case of a LOCA (Loss of Coolant Accident) the assumption of instantaneous opening time is too conservative, especially for this kind of reactor with a positive coolant void reactivity coefficient.Analysis was performed to assess the ability of Atucha II nuclear power plant’s fast boron-injection system to prevent overheating of the fuel elements in the event of a Beyond Design Basis Accident (BDBA). The goal of the Analysis was to develop a Dynamic Break Opening which was used as the input for the LOCA analyses. The results show that the fast boron injection can effectively shut down the reactor, thereby preventing the overheating of the fuel elements.

Uranium-zirconium hydride nuclear fuel performance in the NaK-cooled MARVEL microreactor

The Microreactor Applications Research Validation and EvaLuation (MARVEL) Project is producing a high temperature NaK-cooled nuclear reactor test bed at the Idaho National Laboratory to ultimately improve integration of microreactors to end-user applications. This ambitious effort seeks to design, authorize, construct, test, and operate the reactor within five years. The computational software package chosen for high fidelity fuel performance modeling, BISON, has been upgraded to model the performance of the MARVEL reactor's selected fuel system: 304 stainless steel-clad U-ZrHx. In this manuscript, we begin by describing the MARVEL fuel element. The known properties and relationships of the fuel element constituents are recapitulated from data available in the literature. These historical relationships assume a particular U-ZrH microstructure which previously was not well-understood; therefore, the fuel's microstructure is experimentally confirmed using electron microscopy and X-ray diffraction. This information is then used to analyze the thermomechanical performance of the MARVEL fuel element during a hypothetical extreme accident scenario (a loss of flow accident during a loss of coolant pressure accident) using the nuclear fuel performance code BISON. The analysis shows that the primary phenomenon which could compromise the structural integrity of the MARVEL fuel element is the same as for TRIGA reactors: high temperature internal gas overpressurization-induced rupture of the cladding arising from excessive hoop stresses. Despite modeling the most severe MARVEL accident scenario, the highest anticipated temperature in the fuel, about 718 °C, results in hoop stresses of less than 10 MPa. Based on this analysis, a conservative steady state fuel meat temperature limit of 900 °C precludes cladding damage. The newly incorporated fuel performance simulation capabilities developed in this work may also be extended to model this fuel system under other conditions or reactors.

Effect of Si on the structure and oxidation behavior of Cr-Si coatings on Zry-4 under high temperature steam environment

Under loss of coolant accident (LOCA), zirconium alloy (Zry-4) is easily oxidized by high temperature steam, which makes its performance deteriorate. To improve the oxidation resistance of Zry-4 under LOCA, Cr-Si coatings with different Si content were deposited on Zry-4 by magnetron sputtering, and the effects of Si content on the steam oxidation resistance of the Cr-Si coatings were investigated. With the fluctuation of Si content inside the coating, there appears different lattice structure. Some amorphous phases and lattice distortions in the coatings with moderate Si content were discovered by TEM, which were beneficial to limit O diffusion in the lattice. However, in the coatings with the higher Si content, the lattice distortions transform into dislocations, which contributes to the O diffusion. The oxidation kinetics of the coated samples proved that the Cr-Si coating with lattice distortion was more consistent with the parabolic law and showed excellent oxidation resistance. In addition, the results of first-principles calculations supported the above view that the O diffusion energy in the lattice increases after the introduction of Si.

Multi-physics analysis of SBLOCA by coupled Neutronics/Thermal-Hydraulics full plant model

With the development of computational tools, the accident analysis would be performed by best estimate approaches. The employment of multi-physics codes would be considered as new calculation procedure to obtain high valuable results by improving calculations with specialized codes. The small break loss of coolant accident (SBLOCA) is an accident that has typically been analyzed using system codes with a limited number of Thermal-Hydraulics (TH) channels in the core region, neglecting the use of multi-physics calculations. This approach results in insufficient accuracy and resolution in both neutronics and TH calculations within the core zone. To address this issue, this study proposes the implementation of a Neutronics/Thermal-Hydraulics coupling with one-to-one mapping in the core zone, achieved through the development of a comprehensive plant TH model. This approach aims to enhance the accuracy and resolution of calculations, enabling the identification of local phenomena, parameters, and feedback during accident simulations, and the incorporation of these effects on the macroscopic cross-sections at each time step. The results of SBLOCA simulation reveals the effects of conservative/best estimate approach on the sequence of events, the initiation time of the safety systems, the critical controlled parameters of the power plant, and the precise understanding of the accident. Also, it was found that the radial power distribution becomes disrupted and loses its symmetry immediately after the SCRAM leading to different characteristics of steam generators (SGs) and natural circulation flow rate in each loop of primary side.

Time in DNB experimental study on Cr coated zircaloy cladding

Since the Fukushima-Daiichi accident effort has been put in by the nuclear community to develop Accident Tolerant Fuel (ATF) cladding materials to reduce the risk of clad oxidation and, consequently, hydrogen generation and potential major accidents. Among designed materials, the use of a chromium coating on top of zirconium-alloy cladding showed reasonable results in preliminary tests, being a favorable cladding solution for Pressurized Water Reactors. This paper provides highly detailed temperature data obtained during time in departure from nucleate boiling (DNB) experiments under prototypical reactor conditions of temperature, pressure, mass flux, and heat flux profile on a Cr-coated zirconium-alloy cladded heater rod simulating the fuel pin. Data were obtained at mass fluxes from 1300 to 1700 kg/m2s, pressure of 16 MPa, and inlet temperature of 333 °C (subcooling degree of 14.4 °C). The change in dry time was obtained by increasing the spike in wall temperature necessary to shut down the power. These tests showed that increasing the spike in temperature to characterize the critical heat flux from 50 °C to 400 °C resulted in variations of the departure from nucleate boiling measured value within the uncertainty range of their measurements, hence, validating the use of a spike of 50 °C, a commonly used value in this kind of study, as an indicator of the critical heat flux. Within the dry region, the temperature was found to increase along the direction of the flow for a very short length, reaching a peak, from which it reduces linearly until the surface reaches a temperature that is no longer able to sustain a dry patch. It was noticed that the dry region can grow against the flow direction due to the cosine profile of the heat flux, and that small dryout events can occur without leading to departure from nucleate boiling, i.e., being fully suppressed. However, two consecutive events may sum up and cause the critical heat flux. Rewet time was found to increase with peak cladding temperature and decrease with mass flux. A maximum peak cladding temperature of 844 °C was found, with the surface remaining dry for approximately 14.1 s. The minimum film boiling temperature was measured during the rewet process, showing absolute values of around 580 °C, and difference between wall and inlet temperatures of about 240 °C. A post-inspection using microscope and X-ray images showed no damage to the cladding and heater structure, differently than what was noticed for the bare zircaloy in a previous study. This indicated that the presence of chromium coated cladding can improve the oxidation resistance of the cladding not only in short critical heat flux transients but also in longer events associated with major accidents, like a Loss of Coolant Accident.

Study on spent fuel heatup during spent fuel pool complete loss of coolant accident

Abstract After years of operation, nuclear power plants accumulate a substantial amount of spent fuel stored in spent fuel pools. Prior to the completion of dry storage facilities, this spent fuel remains in the spent fuel pools for several decades. The progression of accidents in spent fuel pools leading to fuel damage typically spans several days or even weeks, which may lead people to overlook the importance of spent fuel pool safety. However, accidents involving spent fuel pools have a characteristic of low probability but high consequences. Their significance has been reemphasized, especially after the Fukushima accident. This study conducts an analysis of the scenario where a spent fuel pool experiences a complete loss of water. A plant equipped with a BWR-4 reactor and a MARK-I containment structure is used as the reference plant. The analysis utilizes the MAAP5 code to assess the impact of different fuel cooling times and ventilation flow rate of secondary containment building on the fuel temperature variation, considering only air circulation for cooling. Based on the analysis results, the plant utilizes the normal ventilation system of the reactor building, which can maintain the fuel temperature below 565 °C with a cooling time of 1 year. Additionally, it was found in the study that adequate ventilation flow rates and sufficient cooling time can ensure that the fuel temperature remains below 565 °C, thereby preserving fuel integrity.

Enhanced Mechanical Properties and Wear Resistance of FeCrAl Alloys at ~ 300°C and Higher Temperatures

Worldwide, light water reactors (LWRs) have been using zirconium (Zr)-based alloys for the cladding of the uranium dioxide fuel for more than 6 decades. Zr alloys oxidize rapidly in the presence of water and steam at temperatures > 450°C; therefore, they do not respond well to scenarios of loss of coolant accidents. There is a global effort by nuclear materials technologists to find more robust or stronger cladding materials for LWRs. One option is to use an iron-chromium-aluminum (FeCrAl) alloy since they have excellent resistance to high temperature oxidation and superior mechanical properties at LWR operation temperatures. Results show that (1) FeCrAl alloys have better mechanical properties than Zr alloy and are orders of magnitude more resistant to creep at temperatures higher than LWR normal operation conditions. (2) FeCrAl alloys have better resistance to fretting wear than Zr alloys at the normal operation conditions of LWRs.

Preliminary accident analyses of in-vessel LOCA for the chinese fusion test reactor HCCB blanket concept

For the helium-cooled ceramic breeder (HCCB) blanket concept improved in 2018 and the primary heat transfer system following China Fusion Engineering Test Reactor (CFETR) in 2019, an in-vessel loss of coolant accident (LOCA) has been investigated with the assumption of a break of the first wall (FW) coolant channels and a double-ended guillotine break of a helium cooling system (HCS) pipe in vacuum vessel (VV). The break leads to helium inject into the VV. After the accident occurs, the helium release leads a plasma disruption and pressure increase in the VV. This paper presents the analysis results of the in-vessel LOCA accidents in CFETR and gives the accident sequence compare with the break in the FW and header pipe. The accident analysis and influence compares can provide a reference for the iterative design.

Fractional scaling analysis of cladding temperature in large break loss of coolant accident

In integral test facilities, some phenomena relevant in the separate effect test are usually neglected due to scaling limitations and design compromises. The typical heat transfer phenomenon in reactor core in large break loss of coolant accident (LBLOCA) scenario is analyzed with fractional scaling analysis (FSA) method in this work. A conservation form of cladding temperature is derived on the basis of energy conservation. Then the scaling criteria and their importance are evaluated. It is found that the similarities of the criteria for convection heat transfer and heat accumulation should be guaranteed since their priorities are high. The contribution of initial pressure and distribution of coolant inventory to cladding temperature in LBLOCA scenario is too slight. This work is encouraging and may shed a light on the coupling application of hierarchical two-tiered scaling (H2TS) and FSA method in engineering application.

Preliminary accident analysis of the loss of vacuum in vacuum vessel for the European DEMO HCPB blanket concept

Design basis accidents are investigated continuously for the European DEMO reactor accompanying its development. One selected postulated initial event (PIE) is a loss of vacuum (LOVA) in vacuum vessel (VV) with large ingress of air induced by rupture in a VV penetration. It has been investigated for the helium cooled pebble bed (HCPB) blanket concept according to the DEMO baseline 2017. The associated primary heat transfer system (PHTS) and the related systems in the tokamak building, from the VV to the PHTS vault and galleries, are considered for the investigation. The LOVA is postulated to occur at a port seal of the electron cyclotron equatorial port plug on the side of the closure plate with (i) a small leak of 1.0 × 10−3 m2, or (ii) a large break size of 1.0 × 10−2 m2. Air ingress from one port cell into the VV leads to the VV pressurization and the fusion power termination followed by an unmitigated plasma disruption. A loss of off-site power for 32 h is assumed to coincide with the disruption. An in-vessel loss of coolant accident (LOCA) is considered as a consequence if the affected first wall (FW) reaches the defined temperature of 1000 °C. The radioactive inventories in the VV (tritium, W-dust) can mobilize towards the VVPSS, the affected systems in the building and the environment due to pressurization, venting and leak conditions. MELCOR 1.8.6 for fusion is applied for this deterministic safety analysis. The resulting releases of radioactivity to the environment are then provided for dose calculation using the computer systems UFOTRI and COSYMA. Outcomes of this LOVA analysis are critically discussed: the transient evolutions of different cases are compared; hydrogen production is detected in case of aggravating FW failure; the source terms (tritium, W-dust) are transported to the connected systems; and the dose results from the environmental releases are provided.