Double-ended Guillotine Break Research Articles

Analysis of Double-Ended Guillotine Break Accident of Surge Line in ACP100 Based on Coupling Method: Development and Validation of the Coupling Method

The ACP100 is a small modular reactor (SMR) designed and built in China, featuring an integrated primary loop and passive safety systems. In SMRs, the interaction between the reactor coolant system, containment vessel, and other systems is highly interdependent. Therefore, it is necessary to use high-precision analysis codes, which can be achieved by coupling multiple codes. This paper is the first part of a study which investigates the LOCA in the ACP100 using a coupled platform. In this paper, a coupling platform was developed for transient analysis of SMR accidents, based on the system code NUSOL-SYS and the Integrated Severe Accident Analysis (ISAA) code. An inter-process communication module was developed adopting shared memory and event objects to exchange data between the two codes. A coupling interface defining the data to be exchanged was proposed. Both NUSOL-SYS and ISAA were modified to perform synchronous time step control. The coupling platform were validated through a hypothetical scenario and Edwards’ pipe blowdown experiment, demonstrating exact consistency and high accuracy. This coupling platform offers a new method for SMR accident analysis, providing a foundation for future work.

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.

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.

Validation of TRACE 5.0 for steam line break with passive auxiliary feedwater system via the ATLAS facility

The TRAC/RELAP Advanced Computational Engine (TRACE) reactor systems code was validated for the Passive Auxiliary Feedwater System (PAFS) using the experimental data from the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS) facility. The accident scenario which emulates a Double-Ended Guillotine Break (DEGB) in the main steam line of Steam Generator 1, was meticulously analyzed, with a focus on transient events and steady-state conditions with an error margin of less than 2%. Comprehensive comparative analyses of key thermal hydraulic parameters, including system pressures, temperatures, flow rates, and water levels, were conducted. Discrepancies were systematically identified, investigated, and effectively resolved. The experimental results underscore the successful role of passive safety systems in mitigating potential accident impacts, emphasizing their pivotal contribution to the overall safety and reliability of nuclear reactor systems. Furthermore, the simulation results effectively mirror the experimental data, demonstrating TRACE's capability to adequately simulate the behavior of a passive feedwater system during a main steam line break.

SBLOCA analysis of a floating nuclear reactor under ocean conditions using MARS-KS moving reactor model

With increasing demands to reduce greenhouse gas emissions from the maritime industry, nuclear power has emerged as a potential solution. Among these options, offshore floating reactors have been demonstrated and developed by various organizations and countries. The offshore floating reactors offer advantages such as a long refueling period for enhanced safety and high mobility, enabling access to remote areas. However, when operating on a moving platform, offshore floating reactors are exposed to external forces that can affect the thermal-hydraulic properties within the reactor core and various hydraulic components. To address the safety concerns associated with offshore floating reactors, evaluating their thermal-hydraulic performance under different ocean motions is necessary. Therefore, this study focused on conducting accident analyses in an offshore floating reactor called BANDI-60, specifically investigating a Small Break Loss of Coolant Accident caused by a double-ended guillotine break in the Direct Vessel Injection line using the system analysis code MARS-KS. Various ocean conditions were considered, including static inclination, combined inclination with heaving, and rolling motion. Among these conditions, static inclination, especially when the break was aligned downward with the inclination direction, was identified as the most significant factor affecting the integrity of the reactor core.

Analysis of LBLOCA of APR1400 with 3D RPV model using TRACE

It is very difficult to capture the multi-dimensional phenomena such as asymmetric flow and temperature distributions with the one-dimensional (1D) model, obviously, due to its inherent limitation. In order to overcome such a limitation of the 1D representation, many state-of-the-art system codes have equipped a three-dimensional (3D) component for multi-dimensional analysis capability. In this study, a standard multi-dimensional analysis model of APR1400 (Advanced Power Reactor 1400) has been developed using TRACE (TRAC/RELAP Advanced Computational Engine). The entire reactor pressure vessel (RPV) of APR1400 has been modeled using a single 3D component. The fuels in the reactor core have been described with detailed and coarse representations, respectively, to figure out the impact of the fuel description. Using both 3D RPV models, a comparative analysis has been performed postulating a double-ended guillotine break at a cold leg. Based on the results of comparative analysis, it is revealed that both models show no significant difference in general plant behavior and the model with coarse fuel model could be used for faster transient analysis without reactor kinetics coupling. The analysis indicates that the asymmetric temperature and flow distributions are captured during the transient, and such nonuniform distributions contribute to asymmetric quenching behaviors during blowdown and reflood phases. Such asymmetries are directly connected to the figure of merits in the LBLOCA analysis. Therefore, it is recommended to employ a multi-dimensional RPV model with a detailed fuel description for a realistic safety analysis with the consideration of the spatial configuration of the reactor core.

The Huge Missing Factor in Leak-Before-Break Analysis: How a Circumferential Through-Wall-Crack in a Pipe System Changes the Flexibility and Reduces the Applied Moments

Abstract In typical leak-before-break (LBB) analyses in the nuclear industry, the uncracked piping normal operating forces and moments are applied in a cracked-pipe analytical procedure to determine normal leakage, and the combined forces and moments under normal operating condition and safe shutdown earthquake seismic loading are used in a fracture analysis to predict margins on “failure.” The International Piping Integrity Research Program (IPIRG) performed in 1990–1998 provided some insights to typical LBB behaviors where pipe system tests were conducted with simulated seismic loadings. The test results showed a large margin on LBB, which was also recognized in 2011 when the Argentinian Atucha II plant was analyzed using a robust full FE model. It was found that when circumferential through-wall cracks were put in the highest stressed locations, the applied moment dropped for both normal operating and N + safe shutdown earthquake (SSE) loading as the crack length increased. The through-wall crack size for causing a double-ended guillotine break (DEGB) was greater than 90% of the circumference. Similar results were also found for a petrochemical pipe system where thermal expansion stresses are much higher than the primary stresses. Even with very low toughness materials, the critical crack size leading to DEGB was greater than 80% of the circumference. The implication of this work is that pragmatically there is much higher margin for DEGB failure in nuclear plant operation, and efforts would be better focused on the potential for a small-break loss-of-coolant accident (SB-LOCA).

Preliminary Accident Analysis of Ex-Vessel LOCA for the European DEMO HCPB Blanket Concept

For the helium-cooled pebble bed breeding blanket concept improved in 2016 and the associated primary heat transfer system (PHTS) following EU DEMO Baseline 2015, an ex-vessel loss-of-coolant accident (LOCA) has been investigated with the assumption of a double-ended guillotine break of a main pipe in an outboard (OB) loop of the PHTS. The break leads to helium blowdown into the tokamak cooling room. A fast plasma shutdown followed by a plasma disruption is activated after the detection of the LOCA due to the design basis accident. Regarding three affected first-wall (FW) areas in one or two OB loops, three main cases are considered. If the FW temperature reaches the defined temperature limit of 1000°C, the FW is assumed to be failed such that an in-vessel LOCA results. In total five scenarios are simulated using MELCOR 1.8.6 for fusion with respect to the affected FW areas, mitigated or unmitigated plasma disruption conditions, the options of the dry or wet suppression tank, and the transport of source terms performed in the case of the beyond design basis accident without the plasma shutdown. The transient results are discussed for the time evolution of the accident sequences, pressurization in the systems, temperature behavior in volumes and structures, and tritium and dust transport behavior.

Pressure suppression system influence on vacuum vessel thermal-hydraulics and on source term mobilization during a multiple first Wall – Blanket pipe break

Being the Vacuum Vessel Pressure Suppression System (VVPSS) one of the most important passive safety systems to be foreseen in DEMO plant, design and integration challenges have to be faced to ensure that best performance within safety requirements are always achieved. In this framework, parametric safety analyses have been performed to support VVPSS design activities; in particular to determine the minimum flow area required by the suppression system pipework to limit the vacuum vessel pressure below the limit imposed as a requirement by design. The selected Postulated Initiating Event (PIE) is a double-ended guillotine break in the Primary Heat Transfer System (PHTS) feeding pipe of the Breeding Zone (BZ) during plasma activity. Coolant is discharged inside the VV upper port volume and an unmitigated disruption occurs, causing further breaks in the first wall (FW) cooling channels. Considering that limiters could be introduced in the future design of the EU DEMO reactor to prevent damages to plasma-facing components, the same parametric study has been performed considering limiters accident mitigation effects. Because discharge flow area toward the suppression pool could also affect the source terms mobilization, the transport of radioactive products (e.g. tritium, tungsten dust and activated corrosion products) has been simulated with the fusion version of the MELCOR code (ver. 1.8.6).

Preliminary thermal-hydraulic analysis of the EU-DEMO Helium-Cooled Pebble Bed fusion reactor by using the RELAP5-3D system code

In the frame of the activities promoted and encouraged by the EUROfusion Consortium aimed at developing the EU-DEMO fusion reactor, great emphasis has been placed at a very early stage of the design to incorporate the provisions needed to improve the overall plant safety and reliability performances as well as to analyse possible mitigation actions. In this framework, the research activity has been focused on the representative and safety relevant cooling loop of the Helium Cooled Pebble Bed (HCPB) Breeding Blanket (BB) Primary Heat Transfer System (PHTS), purposely selected by the safety team, in order to assess its thermal-hydraulic behaviour during normal operational conditions (ramp up/down and steady state) and to preliminarily investigate the consequences of an ex-vessel LOCA accidental scenario ensuing a Double-Ended Guillotine (DEG) break in the hot leg. The research activity has been carried out following a theoretical-computational approach based on the finite volume method adopting the RELAP5−3D system code along with the ANSYS CFX computational fluid dynamic code, which were implicitly integrated to achieve a more detailed and realistic simulation of the EU-DEMO reactor thermal-hydraulics. Models, assumptions and outcomes of this preliminary study are herein presented and critically discussed.

Numerical Study of the Air Ingress Accident of the HTR-PM and Possible Mitigation Measures

The double-ended guillotine break (DEGB) of the horizontal coaxial gas duct of a high-temperature gas-cooled reactor is an extremely hypothetical accident, which could cause the air to enter into the primary circuit and react with graphite in the reactor core. The performance of the HTR-PM plant under this extremely hypothetical accident has been studied by the system code TINTE in this work. The results show that the maximum fuel temperature will not reach the temperature design limitation, and the graphite oxidation will not cause unacceptable consequences even under some conservative assumptions. Moreover, nitrogen and helium injected from the fuel charging tube were studied as the possible mitigation measures to further alleviate the consequences of this air ingress accident. The preliminary results show that only the flow rate of nitrogen injected reaches a certain value, which can effectively alleviate the consequences, while for helium injection, both high and small flow rate can prevent or cut off the natural circulation and alleviate the consequences. The reason is that helium is much lighter than nitrogen, and the density difference between the coolant channel and the reactor core is small when helium is injected. Considering the injection velocity, the total usage amount, and the start time of gas injection, helium injected with a small flow rate is suggested.

Analysis of the upflow conversion modification and influence on selected LOCA accidents in a PWR plant

The baffle jetting phenomenon, intensified by pressure difference across baffle plates, may cause excessive fuel rod vibrations and possible damage. The downward flow direction through the baffle-barrel region opposite to the upward flow through the core in pressurized water reactors affects increase of pressure gradient. Gaps between the baffle plates widen with reactor operating time and water jets become more pronounced. During the 2015 outage the NPP Krško has undergone the upflow conversion (UFC) modification in order to minimize baffle jetting and prevent fuel rod failure. The modification consisted in altering reactor vessel internals in such a way that the coolant downflow path in the baffle-barrel region was converted to an upflow path.The impact of the UFC modification on the nuclear power plant behaviour was analyzed in the paper. The double ended guillotine break of one direct vessel injection (DVI) line was selected as a representative scenario because the DVI line break is exhibiting the highest peak cladding temperature among the small break LOCA cases. The thermal-hydraulic analysis was carried out with the RELAP5/Mod3.3 code for core average channel conditions, while additional FRAPTRAN runs were performed to evaluate peak cladding temperatures for high power and high burnup fuel assemblies. Distribution of pressure and velocity fields in the reactor pressure vessel baffle-barrel region and change in bypass flow rates due to modification were determined by the ANSYS-Fluent calculation. The accident analysis has been performed for both the downflow configuration (pre-UFC) and the upflow configuration (post-UFC) to assess the influence of the modification. Additional cases with different reactor coolant pump trip times, higher core decay heat and reduced capacity of safety systems were also analyzed to evaluate the impact on reactor safety.

Loss of coolant accident analysis under restriction of reverse flow

Abstract This paper analyzes a new method for reducing boiling water reactor fuel temperature during a Loss of Coolant Accident (LOCA). The method uses a device called Reverse Flow Restriction Device (RFRD) at the inlet of fuel bundles in the core to prevent coolant loss from the bundle inlet due to the reverse flow after a large break in the recirculation loop. The device allows for flow in the forward direction which occurs during normal operation, while after the break, the RFRD device changes its status to prevent reverse flow. In this paper, a detailed simulation of LOCA has been carried out using the U.S. NRC's TRACE code to investigate the effect of RFRD on the flow rate as well as peak clad temperature of BWR fuel bundles during three different LOCA scenarios: small break LOCA (25% LOCA), large break LOCA (100% LOCA), and double-ended guillotine break (200% LOCA). The results demonstrated that the device could substantially block flow reversal in fuel bundles during LOCA, allowing for coolant to remain in the core during the coolant blowdown phase. The device can retain additional cooling water after activating the emergency systems, which maintains the peak clad temperature at lower levels. Moreover, the RFRD achieved the reflood phase (when the saturation temperature of the clad is restored) earlier than without the RFRD.

ROSA/LSTF test and RELAP5 code analyses on PWR steam generator tube rupture accident with recovery actions

Abstract An experiment was performed for the OECD/NEA ROSA-2 Project with the large-scale test facility (LSTF), which simulated a steam generator tube rupture (SGTR) accident due to a double-ended guillotine break of one of steam generator (SG) U-tubes with operator recovery actions in a pressurized water reactor. The relief valve of broken SG opened three times after the start of intact SG secondary-side depressurization as the recovery action. Multi-dimensional phenomena specific to the SGTR accident appeared such as significant thermal stratification in a cold leg in broken loop especially during the operation of high-pressure injection (HPI) system. The RELAP5/MOD3.3 code overpredicted the broken SG secondary-side pressure after the start of the intact SG secondary-side depressurization, and failed to calculate the cold leg fluid temperature in broken loop. The combination of the number of the ruptured SG tubes and the HPI system operation difference was found to significantly affect the primary and SG secondary-side pressures through sensitivity analyses with the RELAP5 code.

Analysis of double-ended guillotine break at a direct vessel Injection line of ATLAS

Abstract A double-ended guillotine break at a direct vessel injection line of the ATALS facility has been evaluated with the MARS-KS thermal hydraulic system analysis code. The results are validated against the ATLAS experimental data provided under the First Domestic Standard Problem for Code Assessment (DSP-01) project, indicating that the calculation can generally explain the important phenomena observed in the experiment. The major difference is that the simulation predicts a 2nd peak for the peak cladding temperature, which was not measured during the experiment. Further investigation for the core and downcomer liquid levels reveals that the 2nd peak happens because of incorrect prediction of the downcomer wall temperature caused by the lack of information for heat loss through the downcomer wall in the experiment. This is confirmed by a calculation with the downcomer wall temperature measured at the experiment modeled as a boundary condition. The results with modified downcomer wall boundary conditions show a good prediction of the PCT behavior, indicating that additional heat loss measurements at the downcomer are required for a better understanding of the complex phenomena occurring in the ATLAS downcomer.

Instrumenting full-scale Boron Injection Test Facility to support Atucha-2 NPP licensing

Abstract The Atucha-2 Pressurized Heavy Water Reactor is equipped with a back-up shutdown system based on the fast injection of boron into the moderator tank. Such system had initially been designed to cope with a 10%-area (0.1A) break Loss Of Coolant Accident (LOCA) scenario, but based on upgraded licensing requirements the design had to be revised and possibly improved against a double-ended guillotine (2A) break LOCA. In particular, the boron injection had to be proven fast enough to allow a timely shutdown of the reactor, even in the case of a failure of the primary shutdown system (control rods). A full-scale test facility was built for such “design validation” purpose, in the framework of a cooperation program between the University of Pisa – San Piero a Grado Nuclear Research Group (GRNSPG) and the utility Nucleoelectrica Argentina S.A. (NA-SA). A special instrumentation system, based on conductivity probes designed on purpose by the Helmholtz Zentrum Dresden-Rossendorf (HZDR), was adopted for the measurement of the injection delay, as well as for the monitoring of pressure at several key locations. Care was taken to reproduce the relevant NPP conditions as closely as possible to those expected on the basis of extensive safety analyses performed adopting a Best Estimate Plus Uncertainty (BEPU) approach. In this respect, not only the test facility is full-scale, but also the key components (such as the fast opening air valves, the boric acid tanks, the rupture device, the injection lance) were directly borrowed from the Atucha-2 NPP. This paper provides an overview of the test facility, with particular emphasis on the Authors’ contributions to its design, implementation and operation. Then, it highlights the final outcomes of the experimental campaign carried out by NA-SA, namely: allowing to improve the design of the boron injection system (especially as to some fluid–structure interaction issues) and – what was the main goal – demonstrating that the system’s performance is fast enough to assure a timely and safe shutdown of the reactor, thus contributing to the successful completion of the NPP licensing process.

Code assessment of ATLAS integral effect test simulating main steam-line break accident of an advanced pressurized water reactor

ABSTRACT KAERI has been operating an integral effect test facility, Advanced Thermal–Hydraulic Test Loop for Accident Simulation (ATLAS), for accident simulations of advanced pressurized water reactors. As an integral effect test database for major design basis accidents has been accumulated, a domestic standard problem (DSP) exercise using ATLAS was proposed in order to transfer the database to domestic nuclear industries and to contribute to improving the safety analysis technology for pressurized water reactors (PWRs). As the third DSP exercise, a double-ended guillotine break of the main steam-line at an 8% power without loss of off-site power was decided as a target scenario. Seventeen domestic organizations joined this DSP exercise. They include universities, government, and nuclear industries. The participants of DSP-03 were classified into three groups and each group has focused on the specific subject related to the enhancement of the code assessment; (1) scaling capability of the ATLAS test data by comparing with the code analysis for a prototype, (2) multi-dimensional thermal–hydraulic phenomena anticipated during the steam-line break transient, (3) effect of various models in the one-dimensional safety analysis codes.

Analysis of the equipment and instrumentation qualification criteria using 3D containment models

Containment safety analyses of Design Basis Accidents (DBAs) rely on the use of Lumped Parameters (LP) codes, and therefore, on pressure and temperature averaged values. In those analyses the full containment building is usually modeled with a single or few computational cells. During the latest years, many efforts have been done to develop 3D containment models which bring the possibility of more detailed simulations. Equipment and instrumentation Environmental Qualification in containment are based on pressure and temperature enveloping profiles calculated with LP models. These enveloping profiles calculated with LP models are the solution in a single cell that represents a large volume; however, 3D models volumes are subdivided in thousands of cells and could provide a more accurate solution. In this paper, generic equipment and instrumentation Environmental Qualification (EQ) criteria are assessed during a Double Ended Guillotine Break Loss Of Coolant Accident in a PWR-W 3D containment model developed with GOTHIC 8.1 (QA).

The influence of nuclear graphite oxidation on air ingress accident of HTR-PM

The high temperature gas-cooled reactor pebble bed module (HTR-PM) has wide potential applications in the development of advanced nuclear energy. The double-ended guillotine break (DEGB) of the horizontal coaxial gas duct accident is a kind of beyond design basis accident of HTR-PM, which will result in air ingress into the reactor core. Even though the rare probability of this accident, it still has got a lot of concerns in the safety analyses of HTR-PM, due to the possible severe consequences of the air ingress accident, which would cause oxidation of the graphite material in the reactor. The oxidation of matrix graphite of spherical fuel elements may lead to the exposure of the TRISO fuel particles and the increase of the radioactivity release. The oxidation of the bottom reflector may cause the decrease of mechanical strength of the nuclear grade graphite. Different kinds of nuclear graphite of the bottom reflector will influence the consequences of the DEGB of the horizontal coaxial gas duct accident. This paper mainly analyzed the influence of the nuclear grade graphite used as the bottom reflector on the consequences of this accident by using the system analysis TINTE code.

Comparative CFD simulations of a hydrogen fire scenario

Hydrogen leakage and fire ignition and propagation are safety concerns in several industrial plants. In a nuclear fusion power plants the separation of hydrogen and tritium takes place in different steps, among which one or more electrolyzers are foreseen. A fire scenario could take place in case of leakage of hydrogen. In such cases, it is important to prevent the spreading of the fire to adjacent rooms and, at the same time, to withstand the pressure load on walls, to avoid radioactivity release in the surrounding environment. A preliminary study has been carried out with the aim of comparing CFD tools for fire scenario simulations involving hydrogen release. Results have been obtained comparing two codes: ANSYS Fluent© and FDS. The two codes have been compared both for hydrogen dispersion and hydrogen fire in a confined environment. The first scenario is aimed to obtaining of volume fraction 3D maps for the evaluation of the different diffusion/transport models. In the second scenario, characterized by a double-ended guillotine break, the fire is supposed to be ignited at the same time of the impact. Simulations have been carried out for the first 60 seconds. Hydrogen concentration, temperature and pressure fields are compared and discussed.