Boiling Water Reactors

Boiling Water Reactors
Copyright
Contributors
About the Authors
Preface of JSME series in Thermal and Nuclear Power Generation
Preface
Editing working group for volume 4: Boiling water reactors
Abbreviations
History of BWR development
	Nuclear energy development in Japan
		Primary energy supply
		Electric power generation
		Nuclear power generation
		Nuclear power generation legislation
			References
	Establishment and realization of BWR technologies
		Established stage
			Introduction
			Development of BWR by the Argonne National Laboratory in the United States
		Realizing stage [3,6-22]
			The early stage of GEs BWR development
			The further stage of GEs BWR development
			ABWR development with international cooperation
			Next BWR development
			References
	Improvement and standardization program in Japan
		Technology importation
		First improvement and standardization program
			Countermeasure for SCC [9,11,12,17,18]
			Improved primary containment vessel (improved PCV)
		Second improvement and standardization program [12,16]
			Improvement of fuel-core design and CRD
		Third improvement and standardization program
			Development program of ABWR [9,13,20-27]
			References
	Improvement of system and construction
		Reduction of the construction period of BWRs [1-3]
			Large block and module technologies
			Management by computer and information technologies
		Improvement of ABWR system and construction [4-8]
			Large block-module construction method
			All-weather construction method
			References
	Construction experience and operation performance
		Introduction period [1-4]
		Improvement and standardization programs of Japan [2-7]
		Recent status
			References
Features of BWR plant
	Introduction
		Reference
	Reactor
		Overview
		Reactor system
			Reactor pressure vessel (RPV)
			Reactor internals
			Core and fuel
			Control rods and control rod drive
		Reactivity control system
			Control rod drive system
			Standby liquid control system
		Core monitoring system
			References
	Reactor coolant system and connectedsystems
		Overview
		Nuclear boiler system
			Main steam system
			Feedwater system
		Reactor recirculation system
		Reactor water cleanup system
		Residual heat removal system
		Leak detection system
			References
	Engineered safety features
		Overview
		Containment system
			Overview
			Primary containment vessel (PCV)
			Primary containment isolation system (PCIS)
			Primary containment vessel gas control systems
				Overview
				Atmospheric control system (AC)
				Flammability control system (FCS)
			Containment heat removal system
			Secondary containment
			Standby gas treatment system (SGTS)
		Emergency core cooling system
			Overview of ECCS
			Reactor core isolation cooling (RCIC) system
			High-pressure core flooder system
			Low-Pressure Flooder (LPFL)
			References
	Instrumentation and controls
		Introduction
		Overall architecture (example of ABWR)
		Major control systems and auxiliary control systems
			Major control systems
			Auxiliary control systems
		Safety systems
		Process computer system
		Human-machine interface
	Electric power
		Overview
		Function
		Configuration/main equipment (example of ABWR)
			Grid connection
			Transformers
			Auxiliary medium-voltage distribution buses
			Emergency diesel generators
			DC power supply system
			AC instrumentation power supply system
	Auxiliary system
		Overview
		Auxiliary system
		Fuel pool cooling and cleanup system [1]
		Reactor building cooling water system [2]
		Reactor building service water system [2]
		Turbine building cooling water system [2]
		Turbine building service water system [2]
		Makeup water condensate system [2]
		Instrument air system [2]
		High-pressure nitrogen gas supply system [2]
		Sampling system [2]
		Heating ventilating and air conditioning system [2]
			References
	Steam and power conversion systems
		Overview
		Steam and power conversion systems
			Turbine generator [1]
			Main steam system, auxiliary steam system, and turbine bypass system [1]
			Extraction steam system [1]
			Turbine gland steam system [1]
			Feedwater heater drain and vent system [1]
			Condenser [1]
			Circulating water system [1]
			Condensate and feedwater system [2]
			Off-gas system [2]
			References
Nuclear reactor dynamics and thermal hydraulics of reactor core and fuel assembly
	Reactor internals and coolant flow paths in a reactor pressure vessel
		Unique basic characteristics of the BWR core
			Application of negative void reactivity
			BWR core configuration and basic design concept
		Reactor core support structure and other reactor internals
		Coolant flow paths and the BWR operating map
			Coolant flow paths
			Operating map
			References
	Advances of reactor core and fuel assembly
		High burnup fuel design
			Introduction
			Reliability improvement (1970s)
			Operational improvement
			Economical improvement-Step I fuel and core
			Economical improvement-Step II fuel and core
			Economical improvement-Step III fuel and core
			Summary
				References
	MOX fuel design
		Thermal-hydraulic design
		Thermal-hydraulic design basis of the reactor core
		Nuclear thermal-hydraulic stability
		Flow-induced vibration
			References
		Introduction
		Basic information about Pu
		Characteristics of Pu should be considered for utilization
		MOX fuel assembly design
		MOX core design
		Summary
			References
	Countermeasures and cause of fuel rodfailure
		Overview of fuel failures in BWRs
		Countermeasures and cause of fuel rod failure
			References
	Proving test on the thermal-hydraulicperformance of BWR fuel assembly
		Introduction
		Proving test on thermal-hydraulic performance of a BWR fuel assembly
		Void fraction measurement test for BWR fuel assembly [11-13]
		Development of thermal-hydraulic correlations based on the full-scale BWR fuel assemblies data
			References
	Advances in reactor core and fuel assemblyanalysis
		Nuclear analysis in BWRs
			2D lattice calculation
			3D core calculation analysis
			Validation with measurements
				References
				General reference for nuclear analysis
			Thermal-hydraulic system analysis code
			Thermal-hydraulic subchannel analysis code
				References
	Advances in containment vessel design
		Thermal hydraulics of severe accidents
			Introduction
			Initiation of fuel melt
			Progression of core melt
			Water-Zircaloy reaction accelerating fuel melt
			Melting relocation inside the RPV
			Melting jet structure and behaviors (from the RPV bottom to the PCV floor)
			FP aerosol behaviors [3,4]
		Accident management for BWR
			Summary of AM
			Defense in depth
			International event scale (INES)
			Selection of BWR AM measures
			Typical BWR core damage sequence
			In-vessel phenomena (from core melt to RPV bottom leak)
			Ex-vessel phenomena after RPV failure
			AMs for existing BWR
			AMs for the recently operated and planned plants (also with PWR)
			References
	Advances in safety analysis codeand safety systems
		Various BWR analysis codes
			Importance of nuclear analysis codes
			Best estimate code and evaluation model code
			Verification and validation (VandV) of simulation [3,4]
			BWR analysis code (EM code) [5]
			LOCA analysis code (BE code)
			SA progression analysis code
			Computational fluid dynamic (CFD) analysis code
			Large-scale test facility for code verification and obtaining correlations
		BWR safety systems for severe accident
			Passive safety concept
			Reinforcement for passive safety
			Lineup of passive safety systems
			References
Fukushima Daiichi nuclear power plant accident and analysis evaluation
	Outline of accident
	Event progress and analysis evaluation at Unit 1
	Event progress and analysis evaluation at Unit 2
	Event progress and analysis evaluation at Unit 3
	Hydrogen explosion at Unit 4
	Avoiding severe accidents at Fukushima Daini NPS
		Overview of emergency response at Fukushima Daini NPS
		Fukushima Daini Unit 1 response and station behavior
			Response status at the time of tsunami arrival
			Reactor cooling water injection and PCV cooling
			RHR restoration and reactor cold shutdown
			Continuous ERC planning activities
	Lessons learned from Fukushima Daiichi accident
		Causes of severe accidents and countermeasures
		Measures for severe accidents installed in the United States and European NPPs
		Filtered containment venting system
		Special emergency heat removal system
		Tsunami protection
	New nuclear regulatory requirements in Japan
		New nuclear regulatory requirements
		Tsunami protection examples
		Tornado protection examples
	Example of compliance with new regulatory standards for PWRs that can be used as a reference for BWRs
	BWR NPS to be reviewed for new requirements or restarting
	Activities toward decommissioning Fukushima Daiichi
		Current status of reactors at Units 1 through 4
		Finding contaminated water leak path for leak shutdown from PCV
		Isolation of groundwater flow from contaminated water
		Contaminated water management
		Preparation for fuel-debris removal
	Important lessons learned from Fukushima Daiichi NPS accident
	References
BWR innovations
	Trans-uranic (TRU) burner reactor and reduced-moderation water reactor
		TRU burner reactor
			Introduction
			RBWR concept
			RBWR specifications
			RBWR core characteristics
			Progressive introduction of RBWR [10]
				References
		Reduced-moderation light water reactor
			Introduction [1-4]
			Research and development of the cost-reduced low-moderation spectrum BWR
				References
	Design innovation of BWR and high-pressureBWR
		Introduction
		Objective of LSBWR design
			Natural circulation core concept
			Conceptual design of long cycle core of LSBWR
			Examination of plant operating pressure and plant thermal efficiency
		Safety system and PCV concept
		Module fabrication and construction
			Ship hull structure for reactor building
			General arrangement of LSBWR and LLBWRs building design
			Construction methodology and evaluation
		Summary of design innovation of LSBWR, LLBWR, and high-pressure BWR
			References
	Power uprate in BWR
		Current status and trend of reactor power uprates
			Benefits and safety of constant rated reactor thermal power operation
			Possibilities and issues on constant rated reactor thermal power operation
			Current status of reactor power uprate with equipment modification
		Reactor thermal power and electric power
		Reactor power uprate with constant rated reactor thermal power operation
		Relationship between reactor thermal power and electric power outputs
		Issues and safety in constant rated reactor thermal power operation
		Experiences in BWR operation with constant rated reactor thermal power operation
		Power uprate with equipment modification
			Uprate by measurement uncertainty recapture
			High accuracy leading edge flowmeter (LEFM) for nuclear reactor feedwater measurement in MU
			Inevitable issues on the accuracy of the PF in high accuracy ultrasonic flowmeters and new-concept flowmeter pos ...
			Recent implementation and issues of uprates in the United States
			References
	Post-BT standard for BWR power plant
		Introduction
		Standard for the assessment of fuel integrity under anticipated operational occurrences
			The method for predicting the change of rod temperature during post-BT operation
			The criteria of fuel integrity after BT [9,10,13,20]
			References
	Core catcher
		Overview of core melt stabilization and cooling
		Core catcher of EU-ABWR
			Concept of core catcher
			Performance evaluation test
		Core catcher for the existing BWR
			Concept of core catcher
			Performance evaluation test
			References
	Steam injector
		Introduction
		Principle and application of SI
		SI analysis model
		Visualized fundamental tests
			Test apparatus and measurements
			Test results
		Application of steam jet-type SI to PCIS
			High-pressure tests and analysis
		Application of water jet type SI to RLP
			Confirmation of analysis method
			Scale-up examination of SI for application to RLP
			High-pressure tests using scale models
		Simplified feed water system by SI
			Scaled model tests of simplified feedwater system
			Analysis for improving SI-FWH
			Transient test result of the first stage
			Advantages of SI introduction to ABWR in volume and mass reduction
		Steam injector (SI) pump-up water system to refill pool for passive containment cooling isolation condenser (PCC/I ...
			Concept of SIPOWER
			Evaluation of PCC/IC pool water level transient by SIPOWER
			Full-scale mock-up test to confirm feasibility of SIPOWER
			Air-purge analysis in PCC/IC pool for SIPOWER
			Summary of SIPOWER
			References
	Built in upper internal control rod drives(CRDs) for ABWR-III
		Introduction of merits and technical tasks for internal CRD
		Plant concepts of ABWR-III
		Power devices for the internal CRD
			Magnet coupling power connector
			Magnet coupling signal connector
		Internal CRDs mechanism
			Latch mechanism for scram operation and lift a control rod
			Development of heatproof motor
			Ceramics coil radiation durability test
			Neutron flux at the internal CRD
		Evaluation of ABWR-III conditions
			Durability test of ball bearing
		Two-phase flow and structural integrity
		LOCA and pressure transient analysis
		Aseismic analysis results
		Summary
			References
Index