Reactor Core Isolation Cooling System 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.

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.

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.

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.

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.

Analysis of Fukushima unit 2 accident considering the operating conditions of RCIC system

A severe accident in Fukushima occurred on March 11, 2011 and units 1, 2 and 3 were damaged severely. A tsunami following an earthquake made the supply of electricity power stop, and the safety systems, which use AC or DC power in plants could not operate properly. It is supposed that the degree of core degradation of unit 2 is less serious than in the other plants, and it was estimated that the operation of reactor core isolation cooling (RCIC) system at the initial stage of the accident minimized the core damage through decay heat removal. Although the operating conditions of the RCIC system are not known clearly, it can be important to analyze the accident scenario of unit 2. In this study, best case of the Fukushima unit 2 accident was presented considering the operating conditions of the RCIC system. The effects of operating condition on core degradation and fission product release rate to environment were also examined. In addition, importance of torus room flooding level in the accident analysis was discussed. MELCOR 1.8.6 was used in this research, and the geometries of plant and operating conditions of safety system were obtained from TEPCO through OECD/NEA BSAF Project.

Two-phase flow degradation on Fukushima-Daiichi Unit 2 RCIC turbine performance

After the Fukushima accident, several investigation reports, including experiments and simulations have been done for each of the affected units to completely understand the accident progression and use their results to improve the knowledge of severe accident management and the severe codes performance. In Unit 2, the major uncertainties are related with the reactor core isolation cooling (RCIC) system performance during the accident progression especially focused in the RCIC turbine, which is assumed to work in two-phase flow. The main objective of this study is to analyze the RCIC turbine performance under two-phase flow scenarios under the assumption that the power produced by the turbine is lower than expected due to the liquid phase in the flow. A degradation coefficient quantifying the turbine power reduction is developed as a function of the flow quality by using the sonic speed reduction at critical flow conditions principle obtained by applying the non-homogeneous equilibrium model (NHEM). The degradation coefficient was applied to RELAP/ScdapSIM severe accident code showing a drastic reduction of the turbine-generated power during two-phase flow and obtaining a RCIC system behavior closer to the Tokyo electric power company (TEPCO) investigation report conclusions.

Simulation of thermal stratification in BWR suppression pools with one dimensional modeling method

The suppression pool in a boiling water reactor (BWR) plant not only is the major heat sink within the containment system, but also provides the major emergency cooling water for the reactor core. In several accident scenarios, such as a loss-of-coolant accident and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; the pool temperature distribution also affects the NPSHa (available net positive suction head) and therefore the performance of the Emergency Core Cooling System and Reactor Core Isolation Cooling System pumps that draw cooling water back to the core. Current safety analysis codes use zero dimensional (0-D) lumped parameter models to calculate the energy and mass balance in the pool; therefore, they have large uncertainties in the prediction of scenarios in which stratification and mixing are important. While three-dimensional (3-D) computational fluid dynamics (CFD) methods can be used to analyze realistic 3-D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, resulting in a long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code (Berkeley mechanistic MIXing code in C++) has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by one-dimensional (1-D) transient partial differential equations and substructures (such as free or wall jets) are modeled with 1-D integral models. This allows very large reductions in computational effort compared to multi-dimensional CFD modeling. One heat-up experiment performed at the Finland POOLEX facility, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, is used for validation. Comparisons between the BMIX++, GOTHIC, and CFD calculations against the POOLEX experimental data are discussed in detail.

A study of the effectiveness of defense echelon design of ABWR/LMNPP plant for mitigating LOCA accident

This study adopted a physical digital High Pressure Core Flooder (HPCF) automatic control system for Advanced Boiling Water Reactor (ABWR) design. The HPCF control system was connected to the PCTRAN/ABWR nuclear power plant (NPP) simulator by a communication interface. The HPCF logic control behavior was clearly identifiable and the defense echelon of HPCF digital control system was also analyzed in this work. In order to understand the defense echelon function of Emergency Core Cooling System (ECCS), the dynamic behavior of ECCS was also analyzed by the test facility.The defense echelons of ECCS include the following – Reactor Core Isolation Cooling System (RCIC), HPCF, Low Pressure Flooder mode of Residual Heat Removal System (RHR/LPFL), and Automatic Depressurization System (ADS). In the study, HPCF was implemented and simulated by a commercial grade digital I&C platform (National Instruments), RCIC, HPCF and the plant model were simulated by PCTRAN/ABWR. An interface transferred the computer code parameters to digital input, digital output, analog input, and analog output signals.Six case studies were performed, which were (1) ECCS is initiated by high drywell pressure signal, (2) ECCS is initiated by low reactor water level signal, (3) ECCS initiation follows the Final Safety Analysis Report (FSAR) condition of Lungmen Nuclear Power Plant (LMNPP), (4) ECCS fails without operators manual action, (5) ECCS fails with operators manual action on HPCF, and (6) ATWS mitigated with ARI, ECCS Initiated by High Dry Well Pressure Signal. The results obtained from the system development process, has contributed additionally towards the research which is conducted at Institute of Nuclear Energy Research (INER) and also benefits all other Instrumentation and Control (I&C) system improvement and development of Taiwan Nuclear Instrumentation and Control Systems.

ICONE19-43754 RETRAN Analysis of Feedwater Controller Failure Transient for Lungmen ABWR

This study used RETRAN-3D program to analyze the FWCFWB transient at 100% rated power/111% rated core flow and under EOC operating condition where the ΔCPR response was worst. Important system operational actions included (1) the turbine and the feedwater pumps tripped on high water level (L8), (2) fast bypass valves opening which inhibited reactor scram and RIPs trip on turbine trip, (3) reactor scram and the four RIPs which are not connected to the Motor Generator (M-G) sets tripped on low water level (L3), (4) low-low water level (L2) initiation of the Reactor Core Isolation Cooling System (RCIC). The ΔCPR of this event was 0.05 and the results were compared with those from FSAR. It showed that there was a significant difference in the change rate of vessel water level, but other transient trends were in good agreement with FSAR. A test of all bypass valves failed to open was also covered in this study. In this situation, the reactor scram occurred on high water level turbine trip, and the ΔCPR would rise to 0.17.

Analysis of Failures on Demand of Reactor Core Isolation Cooling System in BWRs by Hierarchical Bayes Method

Failures on demand of a reactor core isolation cooling (RCIC) system in BWRs are the most frequent events of limiting conditions for operation during 1982-2006 in Japan, according to data gathered in Nuclear Information Archives (NUCIA). In this work, probabilities of failures of the RCIC system are analyzed by using the hierarchical Bayes method. The failures on demand of the RCIC system are classified into two groups; one is related to the demand at a periodical inspection test, which is performed almost every 13 months at the end of the periodical inspection of the nuclear power plant, and the other is related to the monthly surveillance test during plant operation. The hierarchical Bayes analysis shows the characteristics of probabilities of failures of each Japanese plant and also that probabilities of failures at the periodical inspection test are quite different from those at the surveillance test, comparing Japanese nuclear power plants with American ones. This paper provides a new approach to analyzing sparse failure data taken from nuclear power plants in Japan.

Modeling the Steam Condensing Mode on a BWR

In this paper the improved mathematical model of the Reactor Core Isolation Cooling System (RCIC) is presented. The controls of the model were tonned and tested in the Simulation Laboratory of the lIE. The formal analysis of the control loops was realized to find the ultimate frequency and gain to tune them with the Ziegler-Nichols settings.