Advanced Boiling Water Reactor Research Articles

Efficiency analysis of nuclear power plants: A comprehensive review

Nuclear power plants play a significant role in global electricity generation, offering a reliable and low-carbon energy source. Maximizing the efficiency of nuclear power plants is crucial for optimizing energy output and reducing operational costs while ensuring safety and environmental sustainability. This paper presents a comprehensive review of the efficiency analysis conducted on nuclear power plants, covering various aspects such as thermal efficiency, conversion efficiency, fuel utilization, and overall plant performance. The review synthesizes the existing literature, discusses key factors influencing efficiency, and highlights potential areas for future research and improvements. It is revealed from the present review that the efficiency of the Boiling Water Reactor (BWR) of the Nuclear Power Plant is 32% and which can be increased to 33% by using advanced BWR. Also, the efficiency of the Pressurized Water Reactor (PWR) of the Nuclear Power Plant is 33% and which can be increased to 36.5% by using the PWR with Gas Burner and Reheating. On the other hand, the efficiency of the Small Modular Reactor (SMR) of the Nuclear Power Plant is 33.4% and which can be increased to 35.5%, 37.4% and 45% by using SMR with Gas Burner, SMR with Reheating and SMR with CCGT, respectively. Therefore, it is predicted that the Small Modular Reactor (SMR) is the most economic Nuclear Power Plant as it has higher thermal efficiency when it is incorporated with Combined Cycle Gas Turbine (CCGT).

Developing a Reactivity-Equivalent Physical Transformation to Simulate an Axially Heterogeneous Boiling Water Reactor

Hitachi is advancing their designs for a conceptual reactor called the resource-renewable boiling water reactor (RBWR), a concept reactor similar to the advanced boiling water reactor with a harder neutron spectrum. This design aims to minimise construction costs and waste production as well as to utilise separated plutonium and minor actinide fuel. However, the axial heterogeneity of the design poses calculation difficulties. The aim of this work is to use a known method, reactivity-equivalent physical transformation (RPT), for calculating fuel with double heterogeneity and apply it to a BWR-type fuel pin. This could reduce the calculation time needed for optimisation of the design of the RBWR. The objective of the study is to use SCALE 6.2 to produce an equivalent axial pin model by comparison with the burnup and neutron spectra of a radial model of the fuel. This model can then be used for 2D burnup calculations, and in future work will be used for the generation of two-group and multigroup cross-sections for further deterministic calculations as part of a two-step approach for analysis of the RBWR. The RPT method has been extensively tested on spherical fuel, and SCALE is a standard industry code. The initial radial model is a hexagonal assembly with 20% enriched UO2 fuel in a zircaloy cladding, surrounded by light water moderator. The derived axial model has a water density distribution taken from Hitachi’s RBWR designs. Criticality over 70 GWd/tU burnup is estimated using the model. The application of the RPT to the BWR pin was shown to be possible, but to have limitations with the introduction of additional radial complexity. For a single pin, excellent agreement between the radial and axial models could be found across a range of water densities, but in the case of an assembly level calculation distinct equivalence models were required for each water density. In addition, the produced RPT model is validated using SCALE’s 3D Monte Carlo module, KENO.

Proposal of quantification method of dynamic system reliability model of digital RPS using Markov state-transition model

ABSTRACT In safety systems of the latest nuclear power plants such as the Advanced Boiling Water Reactor (ABWR) and the latest Pressurized Water Reactor (PWR), digital circuits are used for start-up signals of a reactor protection system (RPS) and emergency core cooling systems (ECCS). And several studies have been conducted to develop the method to build a system reliability model of RPS to be combined with probabilistic risk assessment (PRA) so far. However, these studies have not been sufficiently progressed in terms of the quantitative evaluation and coupling with the conventional PRA method. Despite of the efforts of Task group of Digital system reliability failure mode taxonomy (DIGREL, later becoming Working Group on Digital Instrumentation and Control [WGDIC]) of Organization for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) in this field, the situation has not changed sufficiently. From these backgrounds, this paper presents a rational quantitative evaluation method of the dynamic system reliability model of the digital RPS which can represent time-dependency and is consistent with the conventional PRA method.

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 accident tolerant FeCrAl-ODS fuel cladding for BWRs in Japan

The FeCrAl-oxide dispersion strengthened (ODS) alloy is a promising candidate alloy for the accident tolerant fuel (ATF) cladding of light water reactors (LWRs) and has been recently developed in Japan. This paper reports on the progress of the development of accident tolerant FeCrAl-ODS fuel claddings for boiling water reactors (BWRs) in Japan. Both experimental and analytical studies were conducted to evaluate the influence of implementation of the FeCrAl-ODS fuel cladding to the current BWRs. In the experimental study, key material properties of FeCrAl-ODS fuel cladding have been obtained and accumulated using bar, sheet and tube-shaped materials to support the evaluations in the analytical study. In the evaluation at normal operating conditions, strength and ductility , corrosion property , tritium permeability, wear property, resistance to iodine stress corrosion cracking (SCC) and weldability were examined. A preliminary assessment of compatibility of the FeCrAl-ODS fuel cladding with the current recycling system in Japan was also conducted. In the evaluation of the design basis accident and the beyond design basis accident, strength, steam oxidation property, resistance to water quenching during a loss-of-coolant accident (LOCA) and the LOCA burst property were examined. In order to evaluate the influence of implementation of the FeCrAl-ODS cladding to the current BWRs, the core characteristics and the fuel behavior were evaluated in the analysis study at the normal operating condition. The analysis for the 9 × 9-type and the 10 × 10-type fuel assemblies and the reactor type of the Advanced BWR (ABWR) revealed a good applicability of FeCrAl-ODS fuel cladding. Finally, the challenges and perspectives found in the program are highlighted to enhance international collaborations to promote the development of the FeCrAl-ODS fuel cladding.

Fuel performance analysis of reduced moderated boiling water reactor for transuranic waste incineration

Reduced moderated boiling water reactors (BWR) could serve as near-term alternative to sodium fast reactors for the application of transuranic waste incineration. One such concept currently proposed by Hitachi is the Resource-Renewable Boiling Water Reactor for Transuranic Burning (RBWR-TB2) that utilizes the proven advanced BWR (ABWR) technology. FRAPCON-3.5 EP with its capability to model the physical phenomena at high temperature and high burnup conditions is used to compare the fuel performance of RBWR-TB2 and ABWR fuel rods. The comparison suggests that because of high axial peaking factors and relatively flat power history, fuel temperature is significantly higher in fissile zones of the RBWR-TB2 leading to higher fission gas release and cladding deformation. The higher local fuel burnup in fissile zones of RBWR-TB2 also leads to large but acceptable fuel swelling, accelerated cladding oxidation, and PCMI. In order to assist future design optimizations, sensitivity analyses on important fuel design parameters are also performed.

Uncertainty studies on hydrogen source term with MAAP5 code

Abstract In the present study, a methodology is developed to quantify the uncertainties of special model parameters of the integral severe accident analysis code MAAP5. Here, the in-vessel hydrogen production during a core melt accident for Lungmen Nuclear Power Station of Taiwan Power Company, an advanced boiling water reactor, is analyzed. Sensitivity studies are performed to identify those parameters with an impact on the output parameter. For this, multiple calculations of MAAP5 are performed with input combinations generated from Latin Hypercube Sampling (LHS). The results are analyzed to determine the 95th percentile with 95% confidence level value of the amount of in-vessel hydrogen production. The calculations show that the default model options for IOXIDE and FGBYPA are recommended. The Pearson Correlation Coefficient (PCC) was used to determine the impact of model parameters on the target output parameters and showed that the three parameters TCLMAX, FCO, FOXBJ are highly influencing the in-vessel hydrogen generation. Suggestions of values of these three parameters are given.

Writing ‘usable’ Nuclear Power Plant (NPP) safety cases using bowtie methodology

Historically, the majority of Nuclear Power Plants (NPP) Safety Cases have been produced and implemented under highly prescriptive regulatory regimes, with emphasis placed on demonstrations of the robustness of the facility’s design basis against a set of deterministic criteria and technical standards and rules set by the Regulatory Body. This has resulted in Safety Cases that are technically sound, but at the same time too complex and not easily accessible to and usable by persons responsible for ensuring safe operations; i.e. operations and maintenance staff who are in direct control of the plant as well as managers who are accountable for safety – the key end-users.Shortcomings regarding the ‘usability’ of the Safety Cases are not new and have been the subject of discussion in recent years. They are deeply rooted in the way these documents are produced and implemented. It means that in order to overcome these difficulties attention should be focused primarily on the Safety Case process, affording it the same importance that is given to the final product – the documented Safety Case. This paper explores the advantages that incorporation of the Bowtie risk management methodology into the Safety Case (production) strategy can bring to the delivery of a fit-for-purpose, accessible and usable Safety Case, supporting current efforts undertaken by the nuclear industry to ensure ‘Right First Time Safety Cases’.The paper also suggests that the conduct of Bowtie workshops - in the author’s opinion, the most important part of the Bowtie building process – provides vital input from the people who have most knowledge and experience of the plant and its current operational status. It enables the Bowtie diagrams to reflect the ‘operational reality’ in the plant, by providing a real picture of the actual barrier condition and, more importantly, to identify those barriers that need to be strengthened. The latter will have a direct effect on the usability of the resultant Safety Case product during plant operation.A generic Advanced Boiling Water Reactor (ABWR) design has been taken as an example to highlight the claimed advantages. The paper reflects how Bowtie diagrams can be built within a workshop setting that depicts how a particular accident scenario can be managed through appropriate barriers and controls. It also shows how the use of Bowtie diagrams can make the Safety Case process a true ‘aid’ to thinking and deliver a final product that is accessible and easy-to-understand by key end-users.

Criticality Analysis for BWR Spent Fuel Based on the Burnup Credit Evaluation from Full Core Simulations

This study performed criticality analysis for the GBC-68 storage cask loaded with boiling water reactor (BWR) spent fuel at the discharged burnups obtained from the full-core simulations. The analysis was conducted for: (1) different reloading scenarios; (2) target burnups; and (3) two fuel assembly types—GE14 and SVEA100—to estimate the impact each of the three factors has on the cask reactivity. The BWR spent fuel composition was estimated using the results of the nodal analysis for the advanced boiling water reactor (ABWR) core model developed in this study. The nodal calculations provided realistic operating data and axial burnup and coolant density profiles, for each fuel assembly in the reactor core. The estimated cask’s keff were compared with the fresh fuel and peak reactivity standards to identify the benefit of the burnup credit method applied to the BWR spent fuel at their potential discharge burnups. The analysis identified the significant cask criticality reduction from employing the burnup credit approach compared to the conventional fresh fuel approach. However, the criticality reduction was small compared to the peak reactivity approach, and could even disappear for low burnt fuel assemblies from non-optimal reloading patterns. In terms of cask manufacturing, the potential financial benefit from using the burnup credit approach was estimated to be USD 3.3 million per reactor cycle.

A Comparison of Advanced Boiling Water Reactor Simulations between Serpent/CTF and Polaris/DYN3D: Steady State Operational Characteristics and Burnup Evolution

High fidelity modelling for nuclear power plant analysis is becoming more common due to advances in modelling software and the availability of high-performance computers. However, to design, develop and regulate new light water nuclear reactors there are, up until now, limited requirements for high fidelity methods due to the already well-established computational methods already being widely accepted. This article explores the additional detail which can be obtained when using high fidelity methods through Monte Carlo/Sub-channel analysis compared to industrial methods of cross-section/nodal analysis using the Advanced Boiling Water Reactor as a case study. This case study was chosen due to the challenges in modelling two phase flow and the high levels of heterogeneity within the fuel assembly design. The article investigates how to implement such an approach, from a bottom-up procedure, by analysing each stage of the modelling process.

Technical and non-technical challenges for the Lungmen nuclear power plant project in Taiwan

ABSTRACT The Lungmen Nuclear Power Plant (LNPP) Project is the fourth nuclear power plant project in Taiwan, with two sets of 1,350 MW Advanced Boiling Water Reactors designed by General Electric Company (GE). This project, with a planned construction duration of six years, was extended to seventeen years due to political issues, protests, and other interference. The construction work of the project was completed, with the contract terminated, on 31 May 2015. Although the LNPP project, besieged with controversy, was successfully and safely delivered, it was forced to ‘stop and store’ by the government after extreme anti-nuclear protestations. The technical (cost, schedule, QA/QC, etc.) and non-technical (stakeholder communication) challenges faced in this project are herein addressed. Project management methods are adopted to handle and review the results of resolution of the challenges. The contribution of this paper includes: (1) disclosing the construction management system of the LNPP project; (2) sharing solutions to the technical and non-technical challenges; and (3) learning lessons from the lack of public communication. The conclusions and suggestions proposed might be of help to activating the LNPP, as well as for other nuclear power plant projects under similar situations in Taiwan and beyond.

Optimal Main Heat Transport system configuration for a nuclear power plant

Selection of an optimal Main Heat Transport (MHT) system configuration for a nuclear power plant (NPP) is an activity of high significance, as it influences the safety and operational aspects. This paper brings out the methodology for selection of optimal MHT configuration of NPPs which provides enhanced safety along with ease of operation. The criterion for selection of NPP-MHT configuration have been developed for enhancing safety and ease of operation. As a case study, an innovative MHT configuration has been arrived for a channel type Boiling Water Reactor (BWR) which further improves safety. Some of the multiple failure Design Extension Conditions (DEC) scenarios without SCRAM which lead to severe core damage, for commonly deployed MHT configurations, do not encounter any core heat up with this optimal MHT configuration, for a prolonged duration with Emergency Core Cooling System (ECCS). To demonstrate the effect of MHT system configuration on safety, the Natural Circulation Advanced Boiling Water Reactor (NCABWR) has been considered in this study. Though NCABWR is considered for demonstration purpose, similar methodology can be used for devising or conceptualizing an optimal MHT system configuration for any other NPP. The NCABWR is a four inter-connected loop channel type BWR with natural circulation based passive MHT cooling at all power levels. Following the recent DEC events, more emphasis is laid on very low or no radiological impact in the public domain and no requirement of evacuation in the spirit of “more good than harm”. New NPPs like NCABWR come closer to these requirements. With the mostly passive safety systems, NCABWR can avoid and mitigate DEC. The MHT system configuration for such NPPs need to be selected optimally based on the safety enhancement. In this study, optimal NCABWR MHT configuration has been further evolved based on all the Postulated Initiating Events (PIEs) including multiple failures, such as Anticipated Transients without Scram (ATWS), 200% inlet header break Loss of Coolant Accident (LOCA) without reactor trip etc. Several MHT configurations were conceptualized based on comprehensive overall design, layout requirements and reactor specific PIEs like a LOCA with a postulated break in the four partitioned ECCS header compartment. LOCA in one of the four ECCS header compartment required a mandatory interconnection among all the four MHT loops through the Common Reactor Inlet Header (CRIH) without any partition. This analysis and related issues signaled towards requirement of development of an optimal MHT configuration. The objective of the present study is to configure the MHT components and the ECCS in such an optimal formation so as to provide an enhanced performance in terms of heat removal during adverse conditions. This paper brings out clearly the important aspects needed to be considered while selecting the MHT configuration for the NPPs with enhanced safety even for ATWS-DEC scenarios, without compromising the ease of operation. The development of advanced accident tolerant fuels with improved performance, an improved/optimal MHT NPP configuration is inevitable to ensure safety.

Uncertainty Quantification Using the MAAP5 Code of In-Vessel Hydrogen Generation in a Severe Accident at an Advanced Boiling Water Reactor

MAAP5 is an integral severe accident analysis program that simulates the responses of a light water reactor power plant during a severe accident. This program has been used extensively for probabilistic safety assessments, verification and validation of mitigation actions specified in severe accident management guidelines, and source term quantification. In this study, the uncertainty of in-vessel hydrogen generation predicted by the MAAP5 code was quantified. The surrogate plant that was analyzed is the Lungmen Nuclear Power Station of the Taiwan Power Company. The plant employs an advanced boiling water reactor. We performed sensitivity studies to identify the important model parameters that affect the target output parameters. A range and distribution were assigned to these parameters on the basis of experimental results and expert judgment. The number of input parameters in the analysis was 27. Multiple MAAP5 calculations were performed with an input combination generated from Latin hypercube sampling. The calculation results were analyzed parametrically and nonparametrically to determine the 95th percentile with the 95% confidence level value of the amount of in-vessel hydrogen generation. The Pearson correlation coefficient was used to determine the effect of the model parameters on the target output parameters. The analysis results provide guidance for code applications. The only parameters that pass the threshold of 0.362 for hydrogen generation in the core are FCO and TCLMAX. For hydrogen generation in the lower plenum, FOXBJ is the only input parameter that passes the threshold.

Development of an equilibrium loading pattern and whole-core fuel performance assessment in the Advanced Boiling Water Reactor (ABWR) with UO2 and U3Si2 fuels

The Advanced Boiling Water Reactor (ABWR) is an established evolutionary water reactor that has successfully achieved design certification in a number of countries. However, core design information is scarce in the open literature and this makes studying the suitability of new fuel types difficult and complicates comparisons with other reactor systems. This study therefore aimed to address important data gaps in the open literature relating to the latest generation of BWR core designs. Furthermore, this paper describes an assessment which has been carried out comparing the performance of U3Si2 and standard UO2 fuel deployed in the ABWR.An ABWR core design model has been created using the reactor physics software package CASMO-SIMULATE, with the method used to develop the model and the key set of parameters employed to determine the viability of the core design detailed. A major part of the optimisation process involves the application of simulated annealing to generate loading pattern candidates. Whilst simulated annealing has been employed routinely in loading pattern optimisation, one of the novel aspects in this study relates to each candidate adhering to a predetermined batch map to develop a viable equilibrium loading pattern.NNL's whole-core fuel performance framework NEXUS, which utilises ENIGMA as the fuel performance engine, was used in conjunction with the developed core model to investigate how standard UO2 fuel behaves in a modern BWR design and therefore highlight the most limiting fuel performance characteristics. The NEXUS results highlighted that, for the core designed here, the most limiting fuel performance characteristic for UO2 fuel was rod internal pressure.In addition, a first preliminary whole-core fuel performance assessment was carried out for the proposed enhanced Accident Tolerant Fuel candidate U3Si2 using the same power histories from the UO2 core design to enable a comparison of fuel behaviour independent of power history discrepancies. The assessments for U3Si2 have focused largely on temperature as, due to lack of U3Si2 in-pile experimental measurements, fuel performance parameters that are heavily dependent on swelling and fission gas release models currently have large uncertainties regarding their validity. Therefore, we begin to quantify the potential temperature characteristics associated with U3Si2 relative to UO2, in particular the peak homologous fuel temperature (that is the maximum fuel temperature throughout irradiation as a fraction of its melting point) for U3Si2, which was found to be 0.51 compared with 0.61 for UO2 fuel operating under the same conditions. As a further comparison between UO2 and U3Si2, simple single pin power-to-melt calculations were performed using ENIGMA. It was found that power-to-melt for UO2 is around 88 kW/m and the power-to-melt for U3Si2 is around 230 kW/m.

Numerical investigation on the influence of the type of PI controller combined with MAAP code on the steady-state performance of a nuclear power plant

The Advanced Boiling Water Reactor (ABWR) was adopted as a physical model in this work to investigate the influence of the type of PI controller combined with MAAP 5.03 code on the Core Shroud Water Level (CSWL). For all cases, the proportional gain (kp) is equal to 1. The numerical results reveal that when the integral gain (ki) and kp are the same, the feedwater flow modified by PI controller is sensitive. Accordingly, CSWL is always larger than the Level-2, which does not incur a scrammed reactor. Besides, CSWL modified by the PI controller also consists with the designed water level. However, ki≪kp causes the feedwater flow modified by PI controller to be insensitive. To avoid a scrammed reactor, a P controller (k≫1) is added in front of the PI controller to increase the response speed of CSWL, which improves the tardy speed of feedwater injected into the reactor. Although ki≫kp does not cause a scrammed reactor, a P controller (k≪1) added in front of the PI controller is able to reduce the vibration of CSWL; hence, CSWL is close to the designed water level. To exactly simulate the steady-state performance of a nuclear power plant under the normal operation, our results can offer MAAP code users an important reference to design the PI controller.

Safety Features of Advanced and Economic Simplified Boiling Water Reactors

The Advanced Boiling Water Reactor (ABWR) and the Economic Simplified Boiling Water Reactor (ESBWR) are two kinds of contemporary, advanced, commercially available nuclear power reactors. Reactor internal pumps in an ABWR improve performance while eliminating the large recirculation pumps in earlier BWRs. The utilization of natural circulation and passive safety systems in the ESBWR design simplifies nuclear reactor system designs, reduces cost, and provides a reliable stability solution for inherently safe operation. The conceptually reliable stability solution for inherently safe ESBWR operation is developed by establishing a sufficiently high natural circulation flow line, which has a core flow margin at least 5% higher than the stability boundary flow at 100% rated power of a conventional BWR, and then by designing a high flow natural circulation system to achieve this high natural circulation flow line. The performance analyses for the ESBWR Emergency Core Cooling System (ECCS) show that: (1) the core remains covered with a large margin and there is no core heat up in the ESBWR for any break size, (2) the long-term containment pressure increases gradually with time, in the order of hours, and the peak pressure is below the design value with a large margin, and (3) the margins depend on the containment volumes and water inventories. These safety design features ensure inherently safe ESBWR operation. Enhanced safety features based on lessons learned from the Fukushima nuclear accident are added in ABWR’s and ESBWR’s safety designs. The major enhancements are the further prevention of station blackout and loss of ultimate heat sink.

Practical modeling of fluid in ABWR spent fuel pool for seismic analysis using 3D FEM model

Seismic analysis is important for ensuring the integrity of structures in nuclear power plants. This study investigates the practical modeling method of the fluid in a spent fuel pool (SFP) of an advanced boiling water reactor (ABWR) nuclear power plant for seismic analysis using a three-dimensional finite element method (3D FEM) model of a reactor building (RB) in the plant. Three kinds of 3D FEM models of the RB were developed in this study: combined effect, impulsive, and no-fluid models. The combined effect model included both convective and impulsive effects of the fluid, whereas the impulsive model included only the impulsive effects. The no-fluid model did not include fluid. To investigate the fluid effects, the relative error ratios for maximum acceleration, von Mises stresses, and pressure fluctuations obtained from seismic analyses using these models were calculated. The relative error ratios for the impulsive model with respect to the combined effect model were less than 5%, excluding maximum pressure fluctuations; those for the no-fluid model with respect to the impulsive model were mostly less than 5%. These results indicate that the convective effects were significantly small, excluding pressure fluctuations, and the impulsive effects were also small. On the basis of these results, a practical modeling method of the fluid in an ABWR SFP is proposed and summarized in a modeling flowchart. The flowchart enables practical modeling of the fluid for seismic analysis using an ABWR RB 3D FEM model.

Safety Demonstration of a Class 1 Smart Device

Horizon Nuclear Power intends to build advanced boiling water reactors at Wylfa and Oldbury in the United Kingdom based on the Hitachi-GE (Hitachi) design. In accordance with U.K. policy for new nuclear build, Hitachi, as the reactor designer, is the requesting party to the Generic Design Assessment (GDA), during which the reactor design will be reviewed by the Office for Nuclear Regulation and the Environment Agency. An important step in the GDA process is to demonstrate the viability of the approach developed by Hitachi-GE for the assessment and justification of smart devices. This was done by means of pilot studies of safety class 1 (SC1) and SC2 devices. This paper describes the scope, criteria, process, and approach for the SC1 pilot study and summarizes the results of the study.

A passive decay-heat removal system for an ABWR based on air cooling

This paper describes the capability of an air cooling system (ACS) operated under natural convection conditions to remove decay heat from the core of an Advanced Boiling Water Reactor (ABWR). The motivation of the present research is the Fukushima Severe Accident (SA). The plant suffered damages due to the tsunami and entered a state of Station Blackout (SBO) during which seawater cooling was not available. To prevent this kind of situation, we proposed a passive decay heat removal system (DHRS) in the previous study. The plant behavior during the SBO was calculated using the system code NETFLOW++ assuming an ABWR with the ACS. However, decay heat removal under an air natural convection was difficult. In the present study, a countermeasure to increase heat removal rate is proposed and plant transients with the ACS are calculated under natural convection conditions. The key issue is decreasing pressure drop over the tube banks in order to increase air flow rate. The results of the calculations indicate that the decay heat can be removed by the air natural convection after safety relief valves are actuated many times during a day. Duct height and heat transfer tube arrangement of the AC are discussed in order to design a compact and efficient AC for the natural convection mode. As a result, a 4-pass heat transfer tubes with 2-row staggered arrangement is the candidate of the AC for the DHRS under the air natural convection conditions. The heat removal rate is re-evaluated as a function of air mass velocity at the inlet of heat transfer tube banks.

The influence of void-reactivity feedback on the bifurcation phenomena and nonlinear characteristics of a single nuclear-coupled boiling channel

This study employs the nonlinear dynamic model developed previously by the authors to investigate the effect of void-reactivity feedback on the bifurcation phenomena and nonlinear characteristics of a single nuclear-coupled boiling channel in an advanced boiling water reactor. The parametric effects on the system stability are evaluated by nonlinear analyses as well. The effects of void-reactivity feedback and outlet flow resistance will both destabilize the system, while the inlet flow resistance will generate a stable effect on the system. The strength of void-reactivity feedback has a great influence on the bifurcation phenomena and nonlinear dynamics of the single nuclear-coupled boiling channel. The system can evolve from stable state, limit cycle, periodic to chaotic oscillation through supercritical Hopf and period-doubled bifurcations as the increase in the absolute value of void-reactivity coefficient. In addition, the system can experience a series of period-doubled bifurcation to present various types of oscillations, from periodic ones to chaos, in the unstable region as the controlling parameter of steady phase change number or subcooling number under some specific conditions with a relatively strong void-reactivity coefficient of Cα=-0.38$/%. The appearance of a periodic cycle of three suggests that there may be immeasurable periodic types of nonlinear oscillations in the limited unstable space of this system.