Steam Generator Tube Rupture Event Research Articles

The LIFUS5 separate effect test facility experimental programme

The LIFUS5 facility is a separate effect test facility aimed at investigating the heavy liquid metal-water interaction. It has been designed and constructed to withstand high pressure and temperature (i.e. up to 200 bar and up to 500 °C) and to record the fast pressure transients in heavy liquid metal melt. This type of transient is typical of a Steam Generator Tube Rupture event in a pool type Gen IV Heavy Liquid Metal cooled Fast Reactor system, and should be investigated from the safety point of view because it could potentially induce, beyond the damaging of the internal structures (HX tube bundle, above core structures, Fuel Assembly, Control Rods, etc.) several negative effects on the operation of the reactor. These include an insertion of positive reactivity into the system or reduced cooling efficiency due to steam dragging into the core. It will also have an effect on the chemistry control of the cooling. All these effects compromise the safety and the reliability of the system. The twenty years experimental programme and the different LIFUS5 configurations are presented in the paper, as well as the experimental campaigns and test matrix characterizing the heavy liquid metal-water interaction phenomena and data for codes validation.

Experimental Analysis of Steam Generator Tube Rupture in CIRCE Facility for MYRRHA Configuration

One of the main safety issues of Generation IV (Gen IV) heavy liquid-metal fast reactors is the postulated steam generator tube rupture (SGTR) accident. This event is characterized by primary and secondary coolant interaction, referred to in the literature as a coolant-coolant interaction event having a nonzero probability to occur. This accident scenario could affect the safety of a pool-type reactor, as a consequence of water secondary coolant flashing into the primary coolant liquid metal. The SGTR event needs to be experimentally characterized to evaluate the pressure waves effect, tube rupture propagation (domino effect), oxide precipitation and slug and plug formation, cover gas pressurization of the reactor, and steam flow paths through the pool and eventually the core, entailing the risk of positive reactivity insertion (due to positive local void coefficient). The design phase of the Gen IV MYRRHA plant has dealt with postulated SGTR safety issues in the framework of the MAXSIMA project, which is supported by the European Commission. A relevant contribution to this research activity was provided by the Italian Agency for New Technologies, Energy and Sustainable Economic Development Research Center Brasimone, where a new test section has been designed, assembled, instrumented, and implemented in the large-scale integral-effects pool facility CIRCE for investigating the SGTR event in a relevant configuration for the heat removal system of MYRRHA. This research reactor is not oriented to steam production for running a turbogenerator (no electric production), thus the heat removal system is referred to as the primary heat exchanger (PHX) and not as a steam generator. Four full-scale portions (four bundles of 31 tubes) of the MYRRHA PHX were adopted to carry out four independent SGTR experiments. Water flowed upward in the central tube of the bundle and two rupture positions were investigated at the lower and upper levels, named the bottom and middle scenarios, respectively. After the rupture, water was injected at 16 bar and 200°C into lead bismuth eutectic alloy at 350°C. The experimental results showed a remarkable repeatability and were presented in terms of (1) CIRCE vessel pressurization up to 2.7 and 4 bar absolute for the middle and bottom scenarios, respectively; (2) vapor flow paths through the bundle and its cooling effect up to 120°C and 140°C for the middle and bottom tests, respectively; and (3) strain measurements on tubes and bundle shells up to 2800 μm/m. The integrity of the tubes surrounding the ruptured one and the effectiveness of implemented safety device (rupture disks) pressure relief were significant engineering feedback for MYRRHA designers. The acquired high-quality data also constitute a database increase for future code verification and validation and possible new model development. The performed experimental analysis provided the awareness that a suitable design of a depressurization system (e.g., rupture disks) could allow for addressing postulated SGTR events in the MYRRHA configuration with confidence and safety.

An analysis on the steam generator tube rupture events with core damage

This paper investigates risk significance of Steam Generator Tube Rupture (SGTR) events accompanying a core damage and a direct release of radionuclides into the atmosphere bypassing the containment. Analyses on a spontaneous steam generator tube rupture (S-SGTR) event leading to a core damage and a consequential steam generator tube rupture (C-SGTR) event caused by a station black out (SBO) event are performed for APR1400 using MELCOR 2.2. Both events led to a direct radiological release to the atmosphere. The amount of radiological release was more than that of the Chernobyl and the Fukushima accident.Operator actions to mitigate the radiological releases are investigated. For the S-SGTR scenario, recovery of either auxiliary feed water (AFW) or safety injection alone was not able to prevent significant radiological releases. Combined action of opening of atmospheric dump valve (ADV) and supply of AFW to the steam generators (SGs) resulted in an aggressive depressurization of the primary system and an actuation of safety injection tank. Recovery of liquid inventory and minimizing the U-tube leakage prevented core damage. For C-SGTR scenario, delayed operator action until the time of U-tube creep rupture was not able to prevent significant radiological releases. C-SGTR had to be prevented by a supply of AFW to SGs or rapidly depressurizing the reactor coolant system before an occurrence of U-tube creep rupture.

Evaluation of cancer risk after a release from a hypothetical nuclear reactor steam generator tube rupture accident (SGTR)

This work aims to model a potential consequence after the release of radioactive material from the inventory of a PWR nuclear reactor in a case of an event known as steam generator tube rupture (SGTR). Such a release contains the radioactive materials forming the contamination plume simulated by the software HotSpot. The event simulation was performed varying the atmospheric stability state. The radiation dose was then evaluated in its contribution to the radiological risk through the BEIR V equations. The results were taken in a timely manner offering flexibility for the early stage decision-making. Furthermore, the findings suggest that the atmospheric stability impacts the risk evaluation due to its influence to the radiation dose, which in turn is a key parameter for the risk evaluation. At last, it is expected that the findings from this work encourage further studies aimed to model consequences for a SGTR event.

An analysis on the consequences of a severe accident initiated steam generator tube rupture

An analysis on the thermal hydraulic behavior, core damage progression and radiological releases of a steam generator tube rupture event initiated by a station black out event for APR1400 is provided in this paper. The analysis was performed by using MELCOR, where counter-current natural circulation, mixing in the steam generator inlet plenum, creep ruptures of reactor coolant system components and U-tube, and mitigation actions by the operator are modeled. A station black out accident with an opening of atmospheric steam dump valves (ADVs) at the time of severe accident management strategy entry condition with a failure of a supply of auxiliary feed water to the steam generators selected as a base case resulted in a creep rupture of U-tubes with a direct release of radioactive materials into the atmosphere. It was shown that a mitigation action by the operator for an aggressive cooldown of the reactor coolant system using ADVs and supplying the water to the steam generator even after the creep rupture of U-tubes was effective. It arrested core melt progression and significantly reduced the amount of radiological releases to the atmosphere.

GEN-IV LFR development: Status & perspectives

Since Lead-cooled Fast Reactors (LFR) have been conceptualized in the frame of Generation IV International Forum (GIF), great interest has focused on the development and testing of new technologies related to Heavy Liquid Metal (HLM) nuclear reactors. In this frame, ENEA developed one of the larger European experimental fleet of experimental facilities aiming at investigating HLM thermal-hydraulics, coolant chemistry control, corrosion behavior for structural materials, and at developing components, instrumentations and innovative systems, supported by experiments and numerical tools.The present work aims at highlighting the capabilities and competencies developed by ENEA so far in the frame of the liquid metal technologies for GEN-IV LFR.In particular, an overview on the ongoing R&D experimental program will be depicted considering the actual fleet of facilities: CIRCE, NACIE-UP, LIFUS5, LECOR and HELENA.CIRCE (CIRColazione Eutettico) is the largest HLM pool facility presently in operation worldwide. Full scale component tests, thermal stratification studies, operational and accidental transients and integral tests for the nuclear safety and SGTR (Steam Generator Tube Rupture) events in a large pool system can be studied.NACIE-UP (NAtural CIrculation Experiment-UPgraded) is a loop with a HLM primary and pressurized water secondary side and a 250 kW power Fuel Pin Simulator working in natural and mixed convection.LIFUS5 (lithium for fusion) is a separated effect facility devoted to the HLM/Water interaction. HELENA (HEavy Liquid metal Experimental loop for advanced Nuclear Applications) is a pure lead loop with a mechanical pump for high flow rates experiments. LECOR (LEad CORrosion) is a corrosion loop facility with oxygen control system installed.All the experiment actually ongoing on these facilities are described in the paper, depicting their role in the context of GEN-IV LFR development.

Risk effect of the hybrid-SIT system under SGTR accidents

A hybrid safety injection tank (H-SIT) is designed to passively inject coolant into a reactor coolant system under any pressure condition without depressurization, enabling it to be applied in various accident types. In certain accident conditions, if long-term cooling is not achievable because the required active mitigation components are unavailable, the H-SIT can replace the function of the failed components for at least a certain period. Since the H-SIT has limited inventory, related mitigation strategies need to be carefully developed to effectively enhance accident mitigation strategies and contribute to overall plant safety. This study focuses on the risk effect of the H-SIT system under a steam generator tube rupture (SGTR) accident with safety injection system (SIS) failure, a case typically considered to cause core damage. In such cases, the use of the H-SIT provides for a greater diversity of mitigation options and allows more time to repair the failed components. In order to address plant dynamics in a realistic manner, this study considers a variety of secondary-side cooling performances with corresponding repair probabilities. Multiple-tube rupture cases are also considered. The analysis results demonstrate that the H-SIT provides preferred strategies for managing the risks of both SGTR and MSGTR events.

Pressure drop-flow rate curves for single-phase steam in Combustion Engineering type steam generator U-tubes during severe accidents

Characteristic pressure drop-flow rate curves are generated for all row numbers of the OPR1000 steam generators (SGs), representative of Combustion Engineering (CE) type SGs featuring square bend U-tubes. The pressure drop-flow rate curves are applicable to severe accident natural circulations of single-phase superheated steam during high pressure station blackout sequences with failed auxiliary feedwater and dry secondary side which are closely related to the thermally induced steam generator tube rupture event. The pressure drop-flow rate curves which determine the recirculation rate through the SG tubes are dependent on the tube bundle geometry and hydraulic diameter of the tubes. The larger CE type SGs have greater variation of tube length and height as a function of row number with forward flow of steam favored in the longer and taller high row number tubes and reverse flow favored in the short low row number tubes. Friction loss, natural convection heat transfer coefficients, and temperature differentials from the primary to secondary side are dominant parameters affecting the recirculation rate. The need for correlation development for natural convection heat transfer coefficients for external flow over tube bundles currently not modeled in system codes is discussed.

PWRの蒸気発生器伝熱管複数本破損事象における熱水力的挙動に関する検討

PWRの蒸気発生器伝熱管複数本破損事象における熱水力的挙動に関する検討

De-agglomeration mechanisms of TiO2 aerosol agglomerates in PWR steam generator tube rupture conditions

A steam generator tube rupture (SGTR) in a pressurized water reactor (PWR) might be a major source of accidental release of radioactive aerosols into the environment during severe accident due to its potential to by-pass the reactor containment. In the ARTIST program, tests were carried out at flow conditions typical to SGTR events to determine the retention of dry aerosol particles inside a steam generator tube. The experiments with TiO2 agglomerates showed that for high velocities in the range of 100–350 m/s, the average particle size at the outlet of the tube was significantly smaller than at the inlet due to particle de-agglomeration. Earlier, particle de-agglomeration has not been considered significant in nuclear reactor severe accidents. However, the tests in ARTIST program have shown that there is a possibility that TiO2 aerosol particles de-agglomerate inside a tube and in the expansion zone after the tube exit under SGTR conditions. In this investigation, we measured TiO2 aerosol de-agglomeration in the tube with very high flow velocities with two different TiO2 aerosols. The de-agglomeration was determined by measuring the size of the agglomerates at the inlet and outlet of the test section. The test section was composed of tubes with three different lengths, 0.20, 2.0 and 4.0 m, followed by an expansion zone. The main results were: (i) the de-agglomerate process was relatively insensitive to the initial particle size distribution, (ii) the agglomerates were observed to de-agglomerate in all the tubes, and the resulting particle size distributions were similar for both TiO2 aerosols, (iii) at high flow rates, increasing the gas mass flow rate did not produce further de-agglomeration, and (iv) the agglomerates did not de-agglomerate to primary particles. Instead, after de-agglomeration the particles had a median outer diameter Dc = 1.1 μm. Based on analysis using computational fluid dynamics (CFDs), the de-agglomeration was caused by the turbulent shear stresses due to the fluid velocity difference across the agglomerates in the viscous subrange of turbulence. It has to be noted that the particles used in this investigation were TiO2 agglomerates, and not prototypical nuclear aerosols with significantly different characteristics. Therefore, the results of this investigation cannot be directly used to determine whether the nuclear aerosol particles may de-agglomerate in SGTR sequences. However, this investigation highlights the possibility of particle de-agglomeration under SGTR conditions, and identifies the mechanism of the de-agglomeration inside the broken tube and when the aerosol is discharged to an open space.

Identification and classification of dynamic event tree scenarios via possibilistic clustering: Application to a steam generator tube rupture event

This paper illustrates a method to identify and classify scenarios generated in a dynamic event tree (DET) analysis. Identification and classification are carried out by means of an evolutionary possibilistic fuzzy C-means clustering algorithm which takes into account not only the final system states but also the timing of the events and the process evolution. An application is considered with regards to the scenarios generated following a steam generator tube rupture in a nuclear power plant. The scenarios are generated by the accident dynamic simulator (ADS), coupled to a RELAP code that simulates the thermo-hydraulic behavior of the plant and to an operators’ crew model, which simulates their cognitive and procedures-guided responses. A set of 60 scenarios has been generated by the ADS DET tool. The classification approach has grouped the 60 scenarios into 4 classes of dominant scenarios, one of which was not anticipated a priori but was “discovered” by the classifier. The proposed approach may be considered as a first effort towards the application of identification and classification approaches to scenarios post-processing for real-scale dynamic safety assessments.

Inherent Safety of Fuels for Accelerator-Driven Systems

Transient safety characteristics of accelerator-driven systems using advanced minor actinide fuels have been investigated. Results for a molybdenum-based Ceramic-Metal (CerMet) fuel, a magnesia-based Ceramic-Ceramic fuel, and a zirconium-nitride–based fuel are reported. The focus is on the inherent safety aspects of core design. Accident analyses are carried out for the response to unprotected loss-of-flow and accelerator beam-overpower transients and coolant voiding scenarios. An attempt is made to establish basic design limits for the fuel and cladding. Maximum temperatures during transients are determined and compared with design limits. Reactivity effects associated with coolant void, fuel and structural expansion, and cladding relocation are investigated. Design studies encompass variations in lattice pitch and pin diameter. Critical mass studies are performed. The studies indicate favorable inherent safety features of the CerMet fuel. Major consideration is given to the potential threat of coolant voiding in accelerator-driven design proposals. Results for a transient test case study of a postulated steam generator tube rupture event leading to extensive coolant voiding are presented. The study underlines the importance of having a low coolant void reactivity value in a lead-bismuth system despite the high boiling temperature of the coolant. It was found that the power rise following a voiding transient increases dramatically near the critical state. The studies suggest that a reactivity margin of a few dollars in the voided state is sufficient to permit significant reactivity insertions.

Effects of tube rupture modeling and the parameters on the analysis of multiple steam generator tube rupture event progression in APR1400

A multiple steam generator tube rupture (MSGTR) event in APR1400 has been investigated using the best estimate thermal hydraulic system code, MARS1.4. The effects of the parameters such as the number of ruptured tubes, rupture location, affected steam generator on the analysis of the MSGTR event in APR1400 are examined. In particular, tube rupture modeling methods, single tube modeling (STM) and double tube modeling (DTM), are compared. The APR1400 is found to have the capability of allowing more than 30 min to operators for the MSGTR event of five tubes. The effects of rupture location on the MSSV lift time is not significant in the case of STM, but the MSSV lift time for tube-top rupture is found to be 25.3% larger than that for rupture at the hot-leg side tube sheet in the case of DTM. The MSSV lift time for the cases that both steam generators are affected (4C5x, 4C23x) are found to be larger than that of the single steam generator cases (4A5x, 4B5x) due to a bifurcation of the primary leak flow. The discharge coefficient of Cd is found to affect the MSSV lift time only for a smaller value of 0.5. It is found that the most dominant parameter governing the MSSV lift time is the leak flow rate. Whether any modeling method is used, it gives the similar MSSV lift time if the leak flow rate is close, except in the case where both steam generators are affected. Therefore, the system performance and the MSSV lift time of the APR1400 are strongly dependent on the break flow model used in the best estimate system code.

The effect of tube rupture location on the consequences of multiple steam generator tube rupture event

A multiple steam generator tube rupture (MSGTR) event has never occurred in the commercial operation of nuclear reactors while single steam generator tube rupture (SGTR) events are reported to occur every 2 years. As there has been no occurrence of a MSGTR event, the understanding of transients and consequences of this event is very limited. In this study, a postulated MSGTR event in an advanced power reactor 1400 (APR1400) is analyzed using the thermal-hydraulic system code, MARS1.4. The APR1400 is a two-loop, 3893 MWt, PWR proposed to be built in 2010. The present study aims to understand the effects of rupture location in heat transfer tubes following a MSGTR event. The effects of five tube rupture locations are compared with each other. The comparison shows that the response of APR1400 allows the shortest time for operator action following a tube rupture in the vicinity of the hot-leg side tube sheet and allows the longest time following a tube rupture at the tube top. The MSSV lift time for rupture at the tube-top is evaluated as 24.5% larger than that for rupture at the hot-leg side tube sheet.

Development of a Dynamic Event Tree for a Pressurized Water Reactor Steam Generator Tube Rupture Event

The DPRA-SGTR computer program was written to develop a dynamic event tree for the analysis of a steam generator (SG) tube rupture (SGTR) event. Using the dynamic event tree, a full-scope understanding of the possible responses of a plant following an SGTR event and the related actions with the emergency operating procedures (EOPs) can be analyzed. RELAP5/MOD3.2 was linked to DPRA-SGTR to calculate the thermal-hydraulic response of a Westinghouse three-loop pressurized water reactor at the Maanshan nuclear power plant. One SG tube with a double-ended break was postulated at the beginning of the accident. The plant thermal-hydraulic behaviors, status of the mitigation systems, and operator actions following the EOPs were explicitly modeled in the postulated SGTR. A total of 131 sequences were generated after an SGTR event. Among the 131 sequences, 91 sequences with a frequency sum of 8.5 × 10–6 were stopped either because of low-occurrence frequency (<1 × 10–12) or because the preset mission time was reached (30 000 s after initiating the event). Seven out of the 91 sequences with a frequency sum of 6 × 10–9 were intentionally stopped as a fatal error occurred when RELAP5 was calculating the thermal-hydraulic response.

An Evaluation of Emergency Operator Actions by an Experimental SGTR Event at the IIST Facility and a Comparison of Mihama-2 SGTR Event Record

The current emergency operating procedures (EOPs) for a three-loop pressurized water reactor on steam generator tube rupture (SGTR) incident are experimentally evaluated at the Institute of Nuclear Energy Research Integral System Test Facility. The focus is on the adequacy of EOPs to limit primary-to-secondary leakage with particular emphasis on the number of ruptured U-tubes on the severity of an incident and the response time available for operator actions. By comparing test data with plant records of a Mihama-2 SGTR event, the key thermal-hydraulic phenomena during SGTR transients can be successfully simulated. The test results indicated that the current EOPs can function well in terminating the break flow and maintaining adequate core cooling. However, the effectiveness in minimizing the radioactive release demands an early and substantial operator involvement. To mitigate the consequences of such an event, the timing to isolate the faulted steam generator (SG) and to terminate safety injection flow will be critical. Furthermore, to avoid overfilling, the water level in the faulted SG needs to be drained prior to the implementation of the cooldown process.

Automating strategies of emergency operation for optimal shutdown in pressurized water reactors

In this paper, the automating strategies of emergency operation are proposed for achieving optimal shutdown in pressurized water reactors. These strategies can make emergency operation optimal, and as well they considerably lengthen the operator response time. Decision making and control are investigated in order to develop the automating strategies. In decision-making, diagnostic trees are established to automate the diagnostic tasks for selecting appropriate emergency operations, and the decision-making procedure is developed to automate some decisions which must be made on a plant- and event-specific basis. In control, cooldown is planned by establishing operational goals, forming strategies, and specifying operational means and constraints. The reference set points of the controlled parameters are adaptively adjusted through fuzzy reasoning for optimal cooldown and depressurization. The validation of the proposed strategies was carried out using the micro-simulator for the Kori Unit 2 with steam generator tube rupture events. The results indicated that the automated emergency operation successfully drove the plant at full power to a cold shutdown state with all the operational constraints satisfied.

Accident sequence precursor analyses for steam generator tube rupture events that actually occurred

Steam generator tube ruptures (SGTRs) at pressurized water reactors are identified as one of the risk significant events in the probabilistic risk assessment studies. In addition, operating experience indicates that SGTRs might result in complex transients, some of which involved additional anomalies such as the delayed recognition of event, resulting in the retarded isolation of the ruptured SG and/or failure to timely equalize the primary and secondary pressures. In order to identify the risk significant anomalies and to obtain generic insights useful for examining alternative mitigation measures for STGR, the present study systematically analyzed ten actual and one potential SGTR events using an accident sequence precursor model consistently. The analysis results show that the SGTR event involving the delayed identification of tube rupture or failure to timely depressurize the reactor could have a relatively high possibility of leading to core damage and would be a significant precursor, though the models and failure probabilities applied might be further examined. This implies the importance of improving the capability to detect SGTR and the operating procedures. It is also shown that some of the other anomalies observed would largely contribute to the possibility of core damage, which points out the need of examining alternative measures for recovering from such conditions.

The development and application of the accident dynamic simulator for dynamic probabilistic risk assessment of nuclear power plants

This paper describes the principal modelling concepts, practical aspects, and an application of the Accident Dynamic Simulator (ADS) developed for full scale dynamic probabilistic risk assessment (DPRA) of nuclear power plants. Full scale refers not only to the size of the models, but also to the number of potential sequences which should be studied. Plant thermal-hydraulics behaviour, safety systems response, and operator interactions are explicitly accounted for as integrated active parts in the development of accident scenarios. ADS uses discrete dynamic event trees (D-DET) as the main accident scenario modelling approach, and introduces computational techniques to minimize the computer memory requirement and expedite the simulation. An operator model (including procedure-based behaviour and several types of omission and commission errors) and a thermal-hydraulic model with a PC run time more than 300 times faster than real accident time are among the main modules of ADS. To demonstrate the capabilities of ADS, a dynamic PRA of the Steam Generator Tube Rupture event of a US nuclear power plant is analyzed.

Modelling operator rediagnosis during an SGTR event

A parametric analysis has been performed analysing the hypothetical steam generator tube rupture (SGTR) accident combined with the total loss of secondary heat sink. The analyses have been performed by Relap5/Mod2/36.05 and Mod3 5M5 computer code versions using universal Krško NPP input deck, developed at IJS. Two break sizes have been chosen for the SGTR event: two and five double-ended broken tubes have been assumed. Total loss of secondary heat sink has been assumed from the beginning of the calculation with no possibility of initiating secondary-side feed and bleed. The ways of cooling down the plant after the postulated accidents have been investigated. A scenario has been analysed, when the operator applies rediagnosis procedure, during the period of loss of secondary heat sink threat, and manages SGTR consequences first. The Krško nuclear power plant emergency operating procedures have been verified for this case