Reactor Core Isolation Cooling Research Articles

Experimental investigation of thermal stratification development in boiling water reactor suppression pools during reactor core isolation cooling system operation

The Reactor Core Isolation Cooling (RCIC) system provides make-up water to the reactor pressure vessel during isolation events by passing steam from the vessel through a turbine which drives a pump to return water to the vessel. Exemplary performance of the RCIC system during the Fukushima Daiichi nuclear accident of 2011 has promoted interest in its potential performance.To investigate the development of thermal stratification within suppression pools due to RCIC system exhaust, a facility was constructed to produce the highest-resolution 3-D temperature measurements available in a large water pool during pool mixing by steam injection.Thermal stratification was most strongly influenced by suppression chamber pressure. Increased steam flowrate promoted mixing, which delayed the onset of thermal stratification. Containment response is discussed with respect to thermal stratification, which delays the time of warm water ingestion and cavitation of the RCIC pump, and pool mixing, which reduces the containment pressurization rate.

Reliability Analysis of Safety-Critical Systems using Optimized Petri Nets

Unified modeling language (UML) has emerged as a powerful tool for designing and modeling safety-critical systems. However, UML is not useful in illustrating the dynamic behavior of a critical system. UML does not consider the important critical aspects of the reliability of a safety-critical system, such as non-liveliness, deadlock, stability, and throughput. Therefore, we propose a framework based on UML and Petri Net (PN). In this framework, UML is used to capture all the safety-critical system requirements, whereas PN is used to deliver an in-depth analysis of the reliability aspects of a safety-critical system. Since the PN model suffers from the state space explosion problem, the converted PN may have a large number of redundant states, resulting in a high amount of time required for reliability analysis. Therefore, six reduction algorithms have been derived in this framework to overcome this limitation. The proposed framework is validated with 32 safety-critical system instances of the Nuclear Power Plant on the Reactor Core Isolation Cooling System.

Mitigation of long-term station blackout accident of advanced boiling water reactors

Taiwan Power Company proposed ultimate response guidelines (URGs) to mitigate the so-called ‘Fukushima type accident’ of nuclear power plants. As stipulated in URGs, operators will depressurize the reactor coolant system if necessary to bring in low-pressure water. Containment venting is then used to dissipate the heat generated from the decay. These collective actions are called DIVing (depressurizing, injecting and venting) in URGs. The actions are designed to avoid large-scale evacuation of the public around the damaged plant during an accident. In this study, the effectiveness of DIVing in preventing radioactive emission to the environment is evaluated using RELAP5 codes. The plant analyzed is Lungmen Nuclear Power Station, which has two 1350-MWe advanced boiling water reactors (ABWR) designed by General Electric. In the Standard Operating Procedure of DIVing in URGs, controlled depressurization is executed to bring the pressure to the lowest operating pressure of Reactor Core Isolation Cooling (RCIC) system. The core water level is kept high while RCIC is in operation.. Emergency depressurization is carried out when RCIC fails. The executing time of DIVing as specified in URGs is too conservative. The study identifies the acceptable initial states for operators to carry out DIVing actions. The success criterion is to keep the peak cladding temperature below 1088 K (1500°F) during the transient, which is the temperature of cladding breach and the initiation of gap release of radionuclide from fuel rod.In this study, depressurization transients in the station blackout accident of ABWR are simulated using RELAP5/3D. The results show that the peak cladding temperature obtained during a depressurization transient depends on the time of depressurization, initial pressure, initial water level and injection rate of low-pressure water. When the water level and injection rate are fixed, a region of successful depressurization on the depressurization time vs pressure plot is delineated based on the results of RELAP5 simulations. Regions of successfully depressurization for typical combinations of injection rate and water level are identified, and acceptable region’s boundary is linear.A computation aid, Guidance for Emergency Depressurization (GfED), was developed based on the simulation results of the RELAP5/3D code to assist operators in determining whether the emergency depressurization under given conditions can be performed successfully. GfED uses low-pressure injection flow rate, water level and initial pressure upon emergency depressurization as input parameters to calculate the latest time that is acceptable for emergency depressurization. The technique of linear interpolation is adopted. By comparing the result and value input by the operators, GfED can predict whether it is safe to depressurize.GfED also guides operators to determine the initial condition of a successful depressurization. Three of the four input parameters are fixed in the guidance, and the limiting value of the fourth parameter of a successful depressurization is calculated. These values are informative for operators to manoeuvre the system while the reactor core isolation cooling system is still running.

Two-zone stratified wetwell model development and implementation for RELAP-7

The Reactor Core Isolation Cooling (RCIC) system consists of a turbine, pump and wetwell and is used in many US Boiling Water Reactors (BWRs) as an important heat removal system. The wetwell is a major heat sink within containment. Experimental investigations have shown that thermal stratification may occur in the wetwell during RCIC operation. Current systems-level analysis codes lack stratified wetwell models. This paper describes the development of a stratified wetwell model for the RELAP-7 code.This paper uses a simple two-zone model for the wetwell water that can capture thermal stratification. This model assumes that a buoyant plume transports heat and mass from the steam injection site to the upper water zone. The buoyant plume assumption requires that steam is injected at a low enough rate so it adds negligible momentum. Model validation against experimental data shows similar trends of temperature field development in the suppression pool.

An innovation idea for SBLOCA mitigation strategy: Self-coagulation system in comparison with human hemostatic mechanism

The loss of coolant accident (LOCA), as one of the design basis accidents (DBAs), is a hypothetical accident that is usually considered in the design of nuclear power plant. LOCA is caused by small/large breaks in the reactor primary coolant and pressure boundary and may result in a loss of reactor coolant at a rate in excess of the reactor makeup system capability. Currently, the typical mitigation measures for LOCA are through the vessel high/low pressure coolant injection (HPCI/LPCI), core spray system, or reactor core isolation cooling (RCIC) system, to protect reactor from fuel melting and core degradation. However, is it possible to directly repair the small breaks passively? Inspired by the hemostasis and coagulation mechanism of human blood vessels, the authors are considering to develop a technology which would allow the reactor self-coagulation during the small break loss of coolant accident (SBLOCA). With the assistance of this innovative technology, the speed of losing coolant through reactor primary loop breaks could be significantly reduced or even the loss is stopped. With more and longer existing coolant in the core, the possibility of core degradation will also be significantly reduced. This innovative technology needs to be designed feasible and simple, which can be directly used into the existing nuclear power plants to improve the inherent safety of the cooling system, benefit the economy and safety of the nuclear power plant. In this paper, the authors compare the reactor primary loop of nuclear power plant with the blood circulation system of human beings from comprehensive and multi-angle perspectives, discuss the similarity, feasibility, and economy of nuclear power plant self-coagulation system based on the existing hemostasis and coagulation mechanism of human blood vessels. If possible, this technology will play a revolutionary positive role in alleviating DBAs related to SBLOCA.

Development of the reactor core isolation cooling system model for the extended station black-out accident

To accurately predict the behavior of the Reactor Core Isolation Cooling System (RCIC) in the context of the FLEX strategies under the extended station black-out accident (SBO), we present a mechanistic model of the RCIC in this work. The control volume formulation of angular momentum conservation is employed to model the Terry turbine in the RCIC system. The two-phase nozzle flow model developed in the previous work is utilized to simulate the convergent-divergent nozzles in the Terry turbine. A nozzle flow table is generated to cover the operation range of the RCIC during beyond design basis accidents (BDBAs). The mechanistic RCIC turbine-pump model and the nozzle flow table are incorporated in a TRACE BWR model. Simulations of short-term and long-term SBO transients are performed to demonstrate dynamic RCIC responses. A system analysis code with the proposed mechanistic model of the RCIC system will have a better prediction of system safety response during SBO accidents. With the addition of such more realistic models, the system code will be capable of providing expanded understanding of safety margin, which can help help to form the technical basis for the successful implementation of the FLEX and SAMG measures under Extended Loss of AC Power (ELAP) conditions for BWR or PWR reactor designs.

Performance evaluation of a Terry GS-2 steam impulse turbine with air-water mixtures

• A Terry GS-2 turbine was tested under gas mass fractions from 1.00 to 0.05. • Terry turbine performance degrades with decreasing gas mass fraction. • Results demonstrate capabilities for extended use in nuclear plant systems. Terry steam turbines are widely used in various industries because of their robust design. Within the nuclear power generation industry, they are used in the Reactor Core Isolation Cooling System to remove decay heat during reactor isolation events. During the Fukushima Daiichi nuclear power station disaster in Japan in 2011, the Reactor Core Isolation Cooling System and associated Terry turbine operated for over 70 hours in Unit 2; this runtime is well beyond the expected operating duration. Theories suggest the turbine was subjected to a two-phase inlet flow, which could degrade the turbine performance. In this work, an experimental test rig was constructed to test a full-scale Terry model GS-2 steam turbine under two-phase air/water flows. Steady-state efficiency and torque performance maps of the turbine were developed over a range of turbine inlet pressures (1.38–4.83 bar or 20–70 psia), air mass fractions (0.05–1.0) and rotational speeds up to 4000 RPM. Turbine performance followed expected trends with torque varying linearly and efficiency varying quadratically with rotational speed. In addition, high-speed images of the two-phase flow entering the turbine were also analyzed to understand how changes in inlet pressure and air mass fraction affect the flow regime and homogenization. The present tests with air–water two-phase mixtures are an important step towards providing an understanding of the full-scale Terry turbine’s behavior and performance curves under two-phase conditions. The results of this work will be combined with air/water and steam/water data gathered using a small-scale Terry ZS-1 steam turbine in order to understand the scaling relationship between large and small size Terry turbines and fluid pairs. The combined data set will enable further development of analytical models over a wide range of conditions and may be used to provide technical justification for expanded use of the Terry turbines in nuclear power plant safety systems and other systems.

Accident tolerant fuels (FeCrAl Cladding & Coating) performance analysis in Boiling Water Reactor (BWR) by the MELCOR 1.8.6 UDGC

Since the accident at Fukushima, one major goal of reactor safety research has been the development of more accident tolerant technologies that can mitigate or delay fuel degradation during a Beyond Design Basis Accident (BDBA). One major effort has been focused on increasing the capability of the fuel to be more tolerant of damage during an accident, i.e., Accident Tolerant Fuel (ATF) materials. In this work, we present the development of a generic BWR plant model, the modification of MELCOR to model ATF materials and the use of ATF materials (specifically FeCrAl alloy) as a coating on Zircaloy cladding or as a substitute material for cladding and fuel assembly canister material and its effect on severe accident progression, specifically, a Station Blackout accident. The analysis indicates that significant fuel degradation via fuel heat-up, clad oxidation, and hydrogen generation was delayed up to an hour if FeCrAl alloy was used as a clad and canister material. And, combined with the passive safety systems (i.e., the Reactor Core Isolation Cooling system, RCIC), the extended operation of these systems delayed fuel degradation further. However, an adverse effect should be emphasized for the monolithic FeCrAl design—it generated more hydrogen than the designs based on the Zircaloy due to the high reaction rate at a high temperature of FeCrAl. The design of the FeCrAl-coated-Zircaloy avoids this defect. Therefore, it is a promising choice to combine some of the beneficial traits of both materials.

Reliability and performance analysis of safety-critical system using transformation of UML into state space models

Non-functional requirement analysis is the most critical factor for safety-critical system construction as it reduces the risk of catastrophic loss of assets by taking measurable actions in the design phase. Reliability and performance analysis are the two main components of non-functional requirement analysis. In this paper, a reliability analysis framework is devised, which maps the Unified Modeling Language (UML) state chart model into the state-space model to analyze dynamic behavior and state transition probabilities of a safety-critical system. A system that does not meet the target reliability and performance requirements is considered to be a failure. The proposed framework is validated with 32 safety-critical system instances of the Nuclear Power Plant on the Reactor Core Isolation Cooling System module.

Modeling and assessment of two-phase transonic steam flow with condensation through the convergent-divergent nozzle

In the Fukushima accident, it was found that the Reactor Core Isolation Cooling (RCIC) system played a crucial role in delaying core meltdown by almost three days in Fukushima Daiichi Unit 2, because of self-regulated operation of the steam driven RCIC turbine-pump injection system. Steam flow in the convergent-divergent nozzles of the RCIC Terry turbine is two-phase non-equilibrium transonic flow with homogenous nucleation condensation. To more accurately predict the dynamic process and behavior of the transonic compressible steam flow, a one-dimensional transient two-phase analytical model is presented. A simplified four-fluid model was employed in the present work with the consideration of four separate fluid fields: vapor, liquid film, entrained droplets and condensed droplets. The mass, momentum and energy interactions between the fluids were considered and modeled. An extended seven-equation non-equilibrium critical flow model was developed to obtain the critical pressure and velocities of each phase at the nozzle throat. To predict the wetness in the divergent section, a mechanistic nucleation condensation model was integrated in the nozzle analysis model, considering the generation and consequent growth of droplets. The governing differential equations on a staggered grid were discretized using the second-order Lax-Wendroff scheme with a flux limiter, and the Semi-Implicit Method for Pressure-Linked Equation (SIMPLE) algorithm was employed to solve the discrete linear system. To demonstrate the predictability and reliability of the physical models and the numerical method proposed in the present work, three representative nozzles were modeled and simulated. The results show good agreement with the available experimental data, even for condensation shock.

System code evaluation of near-term accident tolerant claddings during boiling water reactor short-term and long-term station blackout accidents

Near-term Accident Tolerant Fuel (ATF) claddings, such as iron-chromium-aluminum (FeCrAl) and chromium (Cr)-coating, are being extensively studied as candidates to replace the current zircaloy claddings. They present excellent oxidation resistance in high-temperature steam environment that has the potential to greatly reduce the hydrogen production and increase coping time in Beyond Design Basis Accident (BDBA) scenarios. In this work, since we are investigating unmitigated BDBA conditions, coping time is treated as the delay in the time to cladding melting. Three Station Blackout (SBO) scenarios for a representative U.S. operating Boiling Water Reactor (BWR) based on Peach Bottom Atomic Power Station are analysed using system thermal-hydraulics code TRACE. They are (1) short-term SBO without Reactor Core Isolation Cooling (RCIC) blackstart, (2) short-term SBO with RCIC blackstart and (3) long-term SBO. These scenarios are defined to be very similar to the accidents studied in the State-of-the-Art Reactor Consequence Analysis (SOARCA) project. TRACE code is also modified to reflect the oxidation kinetics of FeCrAl and Cr-coating. TRACE simulations only estimated marginal increase of the coping times (1–22 min) with both near-term ATF claddings. However, the hydrogen productions from FeCrAl and Cr-coating are one to two orders of magnitude less that those from Zircaloy cladding at the time of clad melting. Furthermore, the time to generate significant amount of hydrogen gas (0.5 kg) are delayed by 5–32 min. Finally, at the time of complete cladding oxidation, the hydrogen generated by Cr-coated cladding is similar to Zircaloy, while those generated by FeCrAl claddings are reduced by 37%–48% due to its thinner cladding thickness.

Sensitivity analysis of in-vessel accident progression behavior in Fukushima Daiichi Nuclear Power Plant Unit 3

The Great East Japan earthquake and the subsequent tsunami which occurred on March 11th 2011 put the operating Units 1–3 at Fukushima Daiichi Nuclear Power Plant (NPP) in severe accident conditions resulting from loss of offsite power and AC power. It is believed that the Station Blackout (SBO) and loss of heat sink led to core meltdown in all Units 1–3. Despite past research efforts on the severe accident progression in Fukushima NPP Units 1–3, there are still knowledge gaps and uncertainties existing in understanding of the severe accident scenarios and consequences.Hence, this study aims at identifying the modeling uncertainties and addressing the sensitivity parameters in Fukushima NPP Unit 3. A more detailed Control Volume (CV) division model of the reactor core region has been developed to better simulate the thermal-hydraulic behavior of liquid water and steam, which is considered to be crucial in simulating the core uncovery and degradation process. The boundary conditions such as the water injection rates by the Reactor Core Isolation Cooling (RCIC) system, the High Pressure Core Injection (HPCI) system and Alternative Water Injection (AWI) to the reactor core were determined based on the available reactor water level and pressure measurement data. The current study suggested that the local vapor heatup behavior could influence the core melting and relocation behavior, which can lead to different core degradation scenarios. With the current modeling assumptions in MELCOR, the best estimate conditions for RPV pressure history of Unit 3 suggested that 6 SRVs could have remained open when the major core slumping took place at ca. 45:20 h (ca. 12:00, March 13) with 50 to 80 tons of water inventory in the lower plenum. The current analysis also suggested that the efficiency of the AWI to the reactor core could have been only 15% as of reported by TEPCO with the current modeling conditions if debris dryout was assumed to have occurred at around ca. 54.0 h (20:40 h, March 13th). As for lower head failure, there is still large uncertainty in predicting lower head failure time with Larson-Miller creep rupture model in the current MELCOR modeling.

Turbopump scaling analysis and similarity level estimation for Texas A&M university RCIC system experimental test facility

The Reactor Core Isolation Cooling (RCIC) System is a safety system that provides water to the reactor pressure vessel during off-normal BWR reactor conditions, such as reactor isolation from the turbines or loss of AC power. Under loss of AC power conditions, the RCIC System is expected to fail due to battery depletion within 4–8 h of operation for many units. However, the system did not fail until about 70 h into the accident at Fukushima Dai-ichi Unit 2, which was well past the time of battery depletion. To investigate the full potential of the RCIC System, the Laboratory for Nuclear Heat Transfer Systems (NHTS) at Texas A&M University is designing and constructing an experimental RCIC System facility. A careful scaling analysis is essential to ensure proper representation of the RCIC System's key components and phenomena in the experimental testing.This paper describes a method to estimate the similarity level (SL) of the RCIC turbine-pump assembly (turbopump), along with application of the methodology using the Peach Bottom Unit 2 RCIC System as the reference full-scale system. The methodology is demonstrated with the Texas A&M University facility but can be applied to other RCIC system facilities.Zuber's H2TS (Hierarchal Two-Tiered Scaling) method has been adopted to specify the required level of detail for the scaling analysis. The analysis includes development of unitless characteristic time ratios to quantify the degree of similarity of the RCIC turbomachinery between the test facility and the prototype. The RCIC turbine and pump are focused on herein because, first, the Fukushima Dai-ichi Unit 2 RCIC System may have ceased operation at 70 h due to turbine failure, and, two, the greatest RCIC System modeling challenges are with the long-term operation of these turbomachinery components. Since the Texas A&M experiments will provide validation data for analytical models of the turbomachinery, applicability of the data must be evaluated.Based on the similarity level estimates, there is a high degree of similarity between the NHTS and Peach Bottom RCIC System turbines, with a minimum similarity level of 75%. This allows the use of the NHTS RCIC turbine for testing over the full range of Design Basis Accident conditions. Another important finding from this study is that the pump configuration initially designed into the NHTS facility had a low similarity level between the experimental and full-scale facilities. This paper identifies pump parameters in the NHTS facility that can be changed to achieve greater similarity. The similarity level analysis indicate that the NHTS RCIC turbine control volume is appropriate for representing the full-size turbomachinery and can be used to study the prototype turbine and pump behaviors under design basis operating conditions.Scaling analyses of the other RCIC System components are to follow the turbomachinery similarity evaluations. The NHTS facility is currently being prepared to test over the entire Design Basis Accident containment pressure and temperature ranges. Future efforts will also focus on modeling, experimentally and with Computational Fluid Dynamics tools, the RCIC System to provide data for beyond design basis operational conditions.

Fukushima Unit 2 Accident Simulation With MELCOR 2.1

A VTT Fukushima Daiichi Unit 3 (1F3) method for estimation of liquids and consequences of releases (MELCOR) model was modified to simulate the Fukushima Daiichi Unit 2 (1F2) accident. Five simulations were performed using different modeling approaches. The model 1F2 v1 includes only the basic modifications to reproduce the 1F2 accident. The model 1F2 v2 includes the same modifications used in 1F2 v1 plus the wet well (WW) improvement. In the 1F2 v3 model, the reactor core isolation cooling (RCIC) system logic was modified to avoid the use of tabular functions for the mass flow inlet and outlet. Because of this analysis, it is concluded that there is a strong dependency on parameters that still have many uncertainties, such as the RCIC two-phase flow operation, the alternative water injection, the suppression pool (SP) behavior, the rupture disk behavior and the containment failure modes, which affect the final state of the reactor core.

The study of the mitigation strategy of Lungmen ABWR nuclear power plant using RELAP5 and FRAPTRAN with the interface SNAP

In this study, the model of Lungmen ABWR NPP (nuclear power plant) using RELAP5 with the interface SNAP was established. This RELAP5/SNAP model performed the mitigation strategy simulation and study. The main systems of the mitigation strategy are RCIC (Reactor Core Isolation Cooling System), ADS (Automatic Depressurization System), and AC-Independent Water Addition System (ACIWA). This research focuses to assess the usefulness of the mitigation strategy under Fukushima-like conditions. There are three steps in this study. The first step is the establishment of Lungmen NPP RELAP5/SNAP model using the systems data, startup tests reports, and FSAR. The second step is the mitigation strategy simulation and analysis under Fukushima-like conditions using the RELAP5/SNAP model. In this step, the case without mitigation strategy was also performed to evaluate the mitigation strategy effectiveness. Finally, the FRAPTRAN analysis with the interface SNAP was performed in this study. This can evaluate the mechanical property and integrity of fuel rods. According to the RELAP5 results under Fukushima-like conditions, the mitigation strategy can maintain the PCT (peak cladding temperature) below the criteria (1088.7 K). And the FRAPTRAN results also present that the mitigation strategy can keep the fuel rods integrity. It indicates that Lungmen NPP would be in a safe situation when Lungmen NPP performs the mitigation strategy.

MELCOR modeling and sensitivity analysis of Fukushima Daiichi unit 2 accident considering the latest TEPCO investigations

Fukushima Daiichi accident occurred on 11 March 2011 due to the Great East Japan Earthquake and the following tsunami. Recently, Tokyo Electric Power Company (TEPCO) published investigation reports pertaining to the status of unit 2 and other issues that were confirmed. The present study modeled unit 2 accident with MELCOR 2.1 code and performed sensitivity analysis, in order to provide information towards understanding severe accident. The two-phase flow rate and its void fraction in the steam line connecting to the turbine of reactor core isolation cooling (RCIC) system were calculated using the developed RCIC operation model, and the pump injection rate was obtained as well. The suppression chamber (S/C) was divided into three layers with flow path connection to model thermal stratification and the mixing flow at SRV discharge, thereby enabling to capture the measured dry well (D/W) pressure. Through sensitivity analysis of seawater injection, it indicates that the likely seawater injection rate was about 1–2% of the total flow rate through fire engines, and the corresponding reactor pressure vessel (RPV) lower head failure time was located in 92.84 h– 96.17 h. As a result, about 74% of total fuel debris discharged into the reactor pedestal, and the release fractions of noble gas, iodine and cesium to the environment were about 0.741, 0.0167 and 0.00331, respectively. Even though the plausible accident progression was tentatively given, there are still many uncertainties concerning models, boundary conditions and the accident progression.

Transformation of deterministic models into state space models for safety analysis of safety critical systems: A case study of NPP

The state space models has been successfully applied in engineering, statistics, computer science and economics to solve a broad range of dynamical systems problems; like safety analysis, reliability analysis, performability analysis, etc. However, embedding the complete and accurate system requirements in such models is quite challenging. Analyzing model with incomplete or inaccurate requirements gives inaccurate results. UML is a proven and easy approach to capture all the system requirements. This paper proposes a methodology to transform the UML model into the state space model. The resultant model will embed all the system requirements and hence can be used to analyze the critical attributes of the systems. The methodology is validated on 13 sets of operational profile of different safety critical systems of Nuclear Power Plant and shown on Reactor Core Isolation Cooling System (RCICS).

Experimental investigation of BWR Suppression Pool stratification during RCIC system operation

In Boiling Water Reactor (BWR) nuclear power plants with the Mark I containment, the condition of the Suppression Pool can be a large influence on overall plant safety. When the Reactor Core Isolation Cooling (RCIC) System is operating, steam from the reactor drives the RCIC turbine and is then exhausted to the Suppression Pool. When subcooled, the pool can readily condense the steam, warming it up in the process. However, if hot spots or thermal stratification appear, this can limit the Suppression Pool’s ability to perform its safety functions, and can be a limiting factor for RCIC System operation.In order to better understand the RCIC system and its true limits of long-term operation, an experimental model of the system was constructed at the Laboratory for Nuclear Heat Transfer Systems at Texas A&M University (TAMU). These tests provide confirmation of thermal stratification in the Suppression Pool from RCIC System operations, and show a significant degree of dependence on pressure in the airspace above the pool. In the TAMU facility, vertical thermal stratification was limited to 21°C when fully vented to atmospheric pressure, while pre-pressurization led to stratification well in excess of 60°C.

Reactor core isolation cooling system analysis of the Fukushima Daiichi Unit 2 accident with RELAP/ScdapSIM

ABSTRACTThe reactor core isolation cooling (RCIC) system is an auxiliary system of a boiling water reactor (BWR) that provides makeup water in the case of a severe accident. During the Fukushima accident, the extended operation of the RCIC had a large influence on the accident progression and delayed the core meltdown by almost 70 h. During the Fukushima accident, the water level in the reactor pressure vessel (RPV) was assumed to rise enough to flood the main steam line (MSL), which caused the water to move through the RCIC steam turbine and reduce the overall system water injection capability. A RELAP/ScdapSIM analysis was carried out by using RCIC nodalization to reproduce the Fukushima accident and evaluate the impact of the RCIC system on the accident progression. A coefficient based on the critical flow model was included in the RELAP/ScdapSIM source code to reproduce the degradation suffered by the turbine due to the presence of water. Although highly simplified, the analysis demonstrated the RCIC system's feedback capability, which allows the RCIC to control the plant conditions for a long period of time without any human interaction.

Suppression pool testing at the SIET laboratory: experimental investigation of critical phenomena expected in the Fukushima Daiichi suppression chamber

In the unlikely event of a nuclear power plant long duration station black-out, as in the Fukushima Daiichi (1F) severe accident (SA), it was recognized that the suppression chamber (S/C) functions of heat sink and fission product (FP) scrubbing will degrade, resulting in the S/C pressure increase, reduction of the scrubbing efficiency and subsequent necessity of venting operations. Consequently, a relatively large amount of FPs, in particular highly volatile elements (e.g. CH3I), is likely to be dispersed into the environment. As a method to evaluate the degradation of the pool characteristics under discharge of steam and non-condensable gases through vent pipes and steam through different quencher geometries of make-up systems, an experimental campaign was recently started at the SIET research laboratory in Italy. Two different quencher geometries, representing vent pipes and the reactor core isolation cooling (RCIC) exhaust pipes in 1F2 and 1F3, were adopted. Several combinations of steam and air mass flow rates were tested to scale down the main conditions occurred during the 1F SA. Measurements of pool water temperature in different locations and visualization with high-speed camera represent the main outcome of the experimental activity. The preliminary results have demonstrated that a relatively small concentration of air in the steam flow is able to suppress the occurrence of chugging of the steam, with reduced mixing in the pool. Both RCIC quenchers adopted induced large chugging at the bottom of the pool which are effective to avoid temperature stratification, thanks to the large water recirculation and vertical mixing within the pool. At decreased subcooling, mixing in the pool ceases and the quenchers with holes disposed in the vertical direction, as in the RCIC exhaust pipe of the 1F unit 3, introduce intense stratification that drastically reduces the condensation efficiency of the S/C pool. Quencher of 1F2 RCIC does not present stratification possibly dependent on the distance of the pipe outlet to the pool floor. Given the reduced size of the pool compared to the plant scale, the observed phenomena should not be extrapolated for the whole S/C. The objective of the ongoing experimental activity is to construct a database based on the high-speed filming, measurements of major quantities such as water temperature, steam pressure and FPs concentration to foster the development of physical models for both lumped parameter SA codes and detailed computational fluid dynamics software, in an effort to enhance the understanding of the complex phenomena following the 1F accident.