Advanced Boiling Water Reactor Research Articles

Investigation of potential operation issues of human-system interface in lungmen nuclear power project

The Lungmen Nuclear Power Project (LMNPP) with 1350 MWe twin units is the first advanced boiling water reactor (ABWR) project owned by the Taipower Company in Taiwan. Video display units are the main human-system interface for operators to manipulate, i.e., so-called "soft" control, and to know the status of the equipment and plant information. Several concerns arise when compared to hard controls. These concerns led us to conduct an investigation of the operator's performance and the potential issues regarding the operator's teamwork. The results of the investigation show that the most important task is to establish an intensive training program that addresses the knowledge and skill requirements of the operators to meet the task characteristics and the responses of the plant processes.

Accurate Tools to Model Advanced SVEA Fuel Designs

The introduction of the SVEA-96 Optima generation of advanced boiling water reactor fuel designs implies a further increment in complexity and heterogeneity that needs to be supported by accurate calculation tools. In order to take advantage of the improved economics offered by these modern fuel designs while simultaneously assuring safe and reliable reactor operation, both the reload design process and the online core monitoring procedures must be based on appropriate calculation methods. The modeling of transition cores involving the gradual introduction of these new fuel designs poses a severe challenge for the current core physics methods. Recognizing this, Westinghouse has engaged in a continuing process of improving its core physics calculation packages. This development program is supported by a comprehensive validation effort to demonstrate the accuracy and reliability of the improved methods as well as to identify areas requiring further development. The purpose of this paper is to summarize some of the results of this program.

Different variations of a passive safety containment for a BWR with active and passive safety systems

The paper presents probable variations of passive safety boiling water reactor (BWR). In order to improve safety and economy of passive safety BWR, the authors thought of use of a kind of improved Mark III type containment. The paper presents the basic configuration of the passive safety BWR that has an improved Mark III type containment. We tentatively call this passive safety BWR advanced safer BWR + (ASBWR +) and the containment Mark X containment in the paper. One of the merits of the Mark X containment is double containment function against fission products (FP) release. Another merit is very low peak pressure at severe accidents without active cooling systems. The third merit is coolability by natural circulation of outside air. Therefore, the Mark X containment is very suitable for passive safety BWRs. It does not need a reactor building (R/B) as the secondary containment, because it is a double containment by itself. The Mark X containment is a general concept and also useful for half-passive safety BWRs that have both active and passive safety systems. In those examples, active safety systems and passive safety systems function independently and constitute in-depth hybrid safety (IDHS). The concept of the IDHS is also presented in the paper.

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 2: Results of shaking table tests

A 1/8-scale model was constructed of a reinforced concrete containment vessel (RCCV) used in the latest advanced boiling water reactors (ABWR). Shaking table tests were conducted on it with input motions corresponding to or exceeding a design earthquake assumed for a real Nuclear Power Plant. The objectives of the tests were to verify the structural integrity and the leak-proof functional soundness of the RCCV subjected to design earthquakes, and to determine the ultimate strength and seismic margin by an excitation that led to the model's collapse. The model, the test sequence and the pressure and leak test results were addressed in Part 1. The shaking table test method, the input motions and the test results, including the transition of the model's stiffness, natural frequencies and damping factors and the effects of vertical input motions and internal pressure on the model's characteristics and behavior, the load–deformation, the ultimate strength, the failure mode of the reinforced concrete portion and the liner plate are described here. The seismic safety margin that was evaluated by the energy input during the failure test to a design basis earthquake will be described in Part 3. The analytical results of simulation using the multi-lumped mass model will be described in Part 4.

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 1: Test model and pressure test

In Japan, the Nuclear Power Engineering Corporation (NUPEC), sponsored by the Ministry of Economy, Trade and Industry (METI), has been conducting a series of seismic reliability proving tests using full-scale or close to full-scale models to simulate actual important equipment that is critical for seismic safety of nuclear power plants. The tests are intended to validate the seismic design and reliability with a sufficient margin even under destructive earthquakes. A series of tests was carried out on a reinforced concrete containment vessel (RCCV) for advanced boiling water reactor (ABWR) from 1992 to 1999. A large-scale high-performance shaking table at Tadotsu Engineering Laboratory, the largest in the world, was used for this test. Part 1 reports the test model and the results of pressure and leak tests. Part 2 describes test procedures, input waves and the results of verification tests such as changes of stiffness, characteristic frequency and damping ratio, the failure of the model and the load deflection. Part 3 shows the seismic safety margin that was evaluated from the energy input during the failure test to a design basis earthquake. Part 4 reports simulation analysis results by a stick model with lumped masses.

Modeling and Simulation of the ABWR with Transient Analyses for Validation of RETRAN-3D/MOD003

Two advanced boiling water reactors (ABWRs) whose electric output power is 1356 MW have been commercially operated since 1996 and 1997 by the Tokyo Electric Power Company (TEPCO) in Japan. Features of an ABWR are reactor internal pumps (RIPs) placed in the lower plenum and downcomer, peripherally bottom-mounted on the reactor pressure vessel—which should require different modeling from the jet pumps and two recirculation pumps in the primary outer-loop recirculations of BWR-5.Efforts focused on modeling and simulating the ABWR with transient analyses by the point-kinetics model with the local reactivity modified by local importance weighting of the squared nodal power during start-up tests using the RETRAN-3D code, version MOD003 without three-dimensional kinetics. The core and reactor pressure vessel including ten ABWR RIPs and the steam lines were modeled, and simulations were carried out for the cases of the one-pump trip test, the changing-setpoint tests, the main-steam-isolation-valve-closure test, and the generator load rejection test with bypass.The analytical simulation with RETRAN-3D/MOD003 well reproduced the measured data of the ABWR in operation for the RIP trip and the transient tests and could demonstrate its validation for applying to the ABWR with modeling of RIPs.

Technology Development Issues for High-Conversion BWRs Using Island-Type MOX Fuel

The high-conversion boiling water reactor (HCBWR) has been studied as one of the next-generation BWRs. The HCBWR can be improved by the use of island-type fuel, which has mixed-oxide rods in the bundle interior and uranium rods only in the small region of the periphery, to have inherently negative void coefficient (i.e., negative void coefficient in infinite lattice configuration). The proposed reactor concept also has the sustainability to extend the light water reactor’s period by ~180 yr and the compatibility with a conventional BWR system such that only substitution of fuel bundles and control rods is required. As an example case, the high-conversion advanced boiling water reactor II (ABWR-II) is evaluated.

ABWR-II Core Design with Spectral Shift Rods for Operation with All Control Rods Withdrawn

An innovative reactor core concept applying spectral shift rods (SSRs) is proposed to improve the plant economy and the operability of the 1700-MW(electric) Advanced Boiling Water Reactor II (ABWR-II). The SSR is a new type of water rod in which a water level is naturally developed during operation and changed according to the coolant flow rate through the channel. By taking advantage of the large size of the ABWR-II bundle, the enhanced spectral shift operation by eight SSRs allows operation of the ABWR-II with all control rods withdrawn. In addition, the uranium-saving factor of 6 to 7% relative to the reference ABWR-II core with conventional water rods can be expected due to the greater effect of spectral shift. The combination of these advantages means the ABWR-II with SSRs should be an attractive alternative for the next-generation nuclear reactor.

Basic concept of a near future BWR

This paper shows a basic concept of a near future boiling water reactor (BWR) aiming at evolutional safety and cost savings with minimum change from the current advanced BWR (ABWR). The plant output is uprated to 1500 MWe from 1356 MWe. This power uprate can bring about potential of 11% cost saving per MWe base. Safety improvement as a next generation large reactor is also achieved. The advanced reinforced concrete containment vessel (ARCCV) is used for the containment vessel to improve safety for severe accidents. The peak pressure of the containment at severe accidents can be kept close to the design pressure. The advanced passive containment cooling system (APCS) is also provided and can accomplish no primary containment vessel (PCV) venting. The advanced emergency core cooling system (AECCS) consists of four divisions in the front line. The advanced passive cooling system (APCS) is also provided. The combination of the four divisional emergency core cooling system (ECCS) and the passive safety system improves the plant performance in probabilistic safety assessment (PSA). This plant concept is designed based on the heritage of the current ABWR. No more major research and development (R&D) are necessary. Therefore, construction and operation is possible in the early 2010s.

Fuzzy-Adapted Recursive Sliding-Mode Controller Design for a Nuclear Power Plant Control

In this paper, a multi-input multi-output fuzzy-adapted recursive sliding-mode controller (FARSMC) is designed for an advanced boiling water reactor (ABWR) nuclear power plant, to control reactor pressure, reactor water level and turbine power. The FARSMC is intended to replace the existing conventional controllers for the power range of 70% to 100% rated power. The controller has a recursive form that treats model uncertainty and external disturbances in an implicit way. Thus there is no need to specify uncertainties and disturbances for this controller design in advance. Moreover, the chattering problem common among conventional sliding-mode controllers is completely removed by this recursive sliding-mode control (RSMC) algorithm. The performance of the resulting RSMC is further improved by parameter adaptation using fuzzy logic algorithm, resulting in fuzzy-adapted RSMC (FARSMC). To apply RSMC technique, the original nonlinear plant model is first transformed to a canonical form. Simulations of the simplified ABWR model with the designed FARSMC indicate that FARSMC may result in better performance than the existing PI controllers in that the plant transient responses to the desired output step change have shorter settling time and smaller magnitude overshoot/undershoot. The control actions of the FARSMC are milder than those from the PI controllers. Robustness of the FARSMC with respect to power level variation and capability to reject external disturbances is also achieved. Successful fuzzy adapted sliding-mode controller design and implementation on the model shows that FARSMC may be a practical choice for nuclear power plant control.

改良型沸騰水型軽水炉下部プレナム内の流動評価技術(NP2 軽水炉の設計・建設技術)

Reactor internal pumps (RIPs) are used in the advanced boiling water reactor (ABWR) nuclear power plants to circulate the reactor coolant in the reactor pressure vessel. Control rod drive housings (CRDHs) and control rod guide tubes (CRGTs) in the lower plenum were subjected to RIP discharge flow. Fluid force acting on those structures in the lower plenum must be evaluated in the next ABWR in which lower plenum structures have been improved from the current design. Computational fluid dynamics (CFD) analysis is expected to evaluate the fluid force acting on those structures. The 5 RIPs sector models of the reactor lower plenum were simulated to investigate the fluid force in the present study. Fluctuating fluid forces were calculated based on the velocity distribution obtained from the CFD analysis. Fluctuating fluid forces were applied to the structural models and its FIV stress was calculated by the vibration analysis. As a result, flow in the lower plenum went along the RPV bottom and this flow was desirable for structures. Remarkable fluid force acting on structures was not observed in the ABWR lower plenum.

Status of Research and Development of the Lead-Alloy-Cooled Fast Reactor

The U.S. Department of Energy is leading an international initiative to develop the Generation IV Nuclear Reactor. The vast majority of the nation's 104 reactors are type Generation II; Generation III designs (i.e., Advanced Boiling Water Reactor) and have only been built overseas (Korea, Japan). The Generation IV reactor concepts must meet the following criteria: (1) economic competitiveness, (2) inherent safety, (3) minimization of waste, (4) nonproliferation enhancement, and (5) provide social benefit (electricity, “hydricity”). One of the concepts under evaluation is the lead-cooled (Pb) or lead-alloy-cooled (primarily Pb-Bi) fast reactor. Lead/Lead-bismuth reactor cooling offers enhanced safety and reliability in contrast to other liquid metals through its properties, these being chemical inertness with air and water, high atomic number, low vapor pressure at operating temperatures, and high boiling temperature. The only significant drawback is “corrosiveness,” or incompatibility with fuel cladding a...

Passive System Reliability Analysis: A Study on the Isolation Condenser

This paper deals with the reliability assessment of passive systems that have been developed in recent years by suppliers, industries, utilities, and research organizations, aimed at plant safety improvement and substantial simplification in its implementation. The present study concerns the passive decay heat removal systems that use, for the most part, a condenser immersed in a cooling pool. The focus of the paper is a reliability study of the isolation condenser system foreseen for advanced boiling water reactors (BWRs) for the removal of the excess sensible and core decay heat from the BWR by natural circulation. Furthermore, an approach aimed at the thermal-hydraulic performance assessment (i.e., the natural circulation failure evaluation) from the probability point of view is given. The study is not plant-specific-related but pertains to the conceptual design of the foregoing system.

Numerical Analysis of Droplet Breakup Behavior using Particle Method

The numerical method used in this study is Moving Particle Semi-implicit (MPS) method which is based on moving particles and their interactions. Grids are not necessary. Large deformation of fluids can be calculated without grid tangling. A surface tension calculation model is developed to analyze droplet breakup. This model is verified by the simulation of vibration of an ethanol droplet. Two-dimensional numerical analyses of droplet breakup in liquid-liquid and gas-liquid systems are carried out. The correlation between the Weber number and the breakup mode observed in the calculations agrees with that in the experiments. Breakup behavior of a droplet surrounded by a vapor film is analyzed. Flow in the vapor film is considered, though boiling of water and solidification of the melt droplets are ignored. It is found that the breakup of a droplet is suppressed by the vapor film. The critical Weber number in the vapor film is obtained as 50. Molten core coolability is considered by using this result. The median diameter of stable droplets of the molten core is expected as 5 mm in a typical condition, which is consistent with FARO experiment. This result shows that in Advanced Boiling Water Reactor (ABWR) the debris bed up to 40% of the core can be cooled down in the lower head of the reactor pressure vessel.

Numerical Analysis of Droplet Breakup Behavior using Particle Method.

The numerical method used in this study is Moving Particle Semi-implicit (MPS) method which is based on moving particles and their interactions. Grids are not necessary. Large deformation of fluids can be calculated without grid tangling. A surface tension calculation model is developed to analyze droplet breakup. This model is verified by the simulation of vibration of an ethanol droplet. Two-dimensional numerical analyses of droplet breakup in liquid-liquid and gas-liquid systems are carried out. The correlation between the Weber number and the breakup mode observed in the calculations agrees with that in the experiments. Breakup behavior of a droplet surrounded by a vapor film is analyzed. Flow in the vapor film is considered, though boiling of water and solidification of the melt droplets are ignored. It is found that the breakup of a droplet is suppressed by the vapor film. The critical Weber number in the vapor film is obtained as 50. Molten core coolability is considered by using this result. The median diameter of stable droplets of the molten core is expected as 5 mm in a typical condition, which is consistent with FARO experiment. This result shows that in Advanced Boiling Water Reactor (ABWR) the debris bed up to 40% of the core can be cooled down in the lower head of the reactor pressure vessel.

Radiation Shielding for Fission Reactors

Radiation shielding aspects relating fission reactors have been reviewed. Domestic activities in the past five years have been mainly described concerning nuclear data, calculation methods, shielding and skyshine experiments, Advanced Boiling Water Reactor (ABWR), Advanced Pressurized Water Reactor (APWR), High Temperature Engineering Test Reactor (HTTR), Experimental and Prototype Fast Reactors (JOYO, MONJU), Demonstration FBR, core shroud replacement of BWR, and spent fuel transportation cask and vessel. These studies have valuable information in safety and cost reduction issues of fission reactor design for not only existing reactors but also new reactor concepts in the next century. It has been concluded that we should maintain existing shielding technologies and improve these data and methods for coming generations in the next millennium.

Hydrogen management and overpressure protection of the containment for future boiling water reactors

New design and evaluation method for hydrogen management of containment atmosphere have been developed for application in the future boiling water reactor (BWR). These are intended as a part of consideration of severe accidents in the course of design so as to assure a high level of confidence that a large release of radioactivity to the environment that may result in unacceptable social consequences can reasonably be avoided. Emphasis on hydrogen management and protection against overpressure failure is based on the insights from probabilistic safety assessments (PSAs) that late phase overpressure (and associated leakage) and molten corium concrete reaction (MCCI) need attention to ensure that containment remains intact, in case energetic challenges to the containment such as DCH (direct containment heating) or FCI (fuel coolant interactions) are practically eliminated by design or resolved from risk standpoint of view. The authors studied the use of palladium-coated tantalum for hydrogen removal from containment atmosphere in order to avoid pressurization of the containment with small free volume by non-condensable gas and steam. Its effectiveness for ABWR (advanced boiling water reactor) containment was evaluated using laboratory test data. Although further experimental studies are necessary to confirm its effectiveness in real accident conditions, the design is a promising option and one that could be backfitted upon necessity to existing plants for which pressure retaining capability cannot be altered. Also new evaluation method for flammability control under severe accident conditions was developed. This method employes a realistic assessment of the amount of oxygen and hydrogen gases generated by radiolytic decomposition of water under severe accident conditions and their subsequent transport from water to containment atmosphere.

Full-Load-Rejection Design Analysis for the Lungmen Advanced Boiling Water Reactor Using RETRAN

The very first work on RETRAN model development and its application to plant design study for the Lungmen nuclear power station (LMNPS) is presented. Lungmen is the fourth nuclear power plant of the Taiwan Power Company (TPC). LMNPS has two advanced boiling water reactor (ABWR) units, each with a thermal power of 3926 MW. The preliminary safety analysis report (PSAR) of LMNPS is currently under review. The Lungmen RETRAN-02/MOD5 model was developed to provide support to TPC in the PSAR review and system design study. An analysis of generator load rejection with failure of all bypass valves was performed against the analysis in the PSAR to benchmark the Lungmen RETRAN model. One of the specific designs of LMNPS is that the reactor has the capability to withstand a full-load rejection or a turbine trip event without a reactor scram. An analysis of generator load rejection with all bypass valves was done to evaluate this design. The results show that this design is appropriate.

Modeling and Simulation of the First ABWR for Validation of RETRAN-3D

The advanced boiling water reactor (ABWR) has ten reactor-internal pumps peripherally mounted on the bottom of a reactor vessel. Analytical simulation of reactor-internal pumps unique to the ABWR requires new modeling because of the difference in core flow characteristics between the reactor-internal pumps and the two external-recirculation pumps of the primary outer loops with the jet pumps in a current boiling water reactor. Efforts in this work focused on modeling and simulation of reactor-internal pumps and core flow of the ABWR, using the RETRAN-3D code, the computer program for transient thermal-hydraulic analysis of a complex fluid flow system, without multidimensional kinetics. Included are modeling of the core and reactor pressure vessel with ten reactor-internal pumps, and simulation of the events of reactor-internal-pumps trip during the startup-phase tests, which are unable to be done in the simulation of a current BWR. Sensitivity analyses on the recirculation flow control and the slip model were also performed. The predictions by the RETRAN-3D code successfully tracked the measured data of reactor-internal-pump trip during the startup-phase test. The present analytical simulations could demonstrate the validation of the RETRAN-3D code applicable to the ABWR with the pump model of reactor-internal pumps in the program.

Safety system consideration of a supercritical-water cooled fast reactor with simplified PSA

Abstract The probabilistic safety of the supercritical-water cooled fast reactor (SCFR) is evaluated with the simplified probabilistic safety assessment (PSA) methodology. SCFR has a once-through direct cycle where all feedwater flows through the core to the turbine at supercritical pressure. There are no recirculation loops in the once-through direct cycle system, which is the most important difference from the current light water reactor (LWR). The main objective of the present study is to assess the effect of this difference on the safety in the stage of conceptual design study. A safety system configuration similar to the advanced boiling water reactor (ABWR) is employed. At loss of flow events, no natural recirculation occurs. Thus, emergency core flow should be quickly supplied before the completion of the feedwater pump coastdown at a loss of flow accident. The motor-driven high pressure coolant injection (MD-HPCI) system cannot be used for the quick core cooling due to the delay of the emergency diesel generator (D/G) start-up. Accordingly, an MD-HPCI system in an ABWR is substituted by a turbine-driven (TD-) HPCI system for the SCFR. The calculated core damage frequency (CDF) is a little higher than that of the Japanese ABWR and a little lower than that of the Japanese BWR when Japanese data are employed for initiating event frequencies. Four alternatives to the safety system configurations are also examined as a sensitivity analysis. This shows that the balance of the safety systems designed here is adequate. Consequently, though the SCFR has a once-through coolant system, the CDF is not high due to the diversity of feedwater systems as the direct cycle characteristics.