ML24011A027

Fact Sheet: Reactor Concepts Manual - Boiling Water Reactor Systems

ML24011A027
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Issue date:	01/11/2024
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Reactor Concepts Manual Boiling Water Reactor Systems

Boiling Water Reactor (BWR)

Systems

This chapter will discuss the purposes of some of the major systems and components associated with a boiling water reactor (BWR) in the generation of electrical power.

USNRC Technical Training Center 3-1 0400 Reactor Concepts Manual Boiling Water Reactor Systems

Containment /D rywe ll Reactor Vessel Steam Line Turbine Building Throttle GeneratorElectrical Valve Steam Dryer SeparatorMoisture Turbine

Reactor Core Condenser

Jet Pump

To/From River RecirculationPump

Pump Containment Suppression Chamber

Boiling Water Reactor Plant

Inside the boiling water reactor (BWR) vessel, a steam water mixture is produced when very pure water (reactor coolant) moves upward through the core abso rbing heat. The major difference in the operation of a BWR from other nuclear systems is the steam voi d formation in the core. The steam-water mixture leaves the top of the core and enters the two stag es of moisture separation, where water droplets are removed before the steam is allowed to enter the st eam line. The steam line, in turn, directs the steam to the main turbine causing it to tu rn the turbine and the attached el ectrical generator. The unused steam is exhausted to the condenser where it is condensed into water. The resulting water is pumped out of the condenser with a series of pum ps and back to the reactor vessel. The recirculation pumps and jet pumps allow the operator to vary coolant flow through the core and change reactor power.

USNRC Technical Training Center 3-2 0400 Reactor Concepts Manual Boiling Water Reactor Systems

BWR Reactor Vessel Assembly

The reactor vessel assembly, shown on page 3-4, consists of the reactor vessel and its internal components, including the core support structures, core shroud, moisture removal equipment, and jet pump assemblies. The purposes of the reactor vessel assembly are to:

• House the reactor core,

• Serve as part of the reactor coolant pressure boundary,

• Support and align the fuel and control rods,

• Provide a flow path for circulation of coolant past the fuel,

• Remove moisture from the steam exiting the core, and

• Provide a refloodable volume for a loss of coolant accident.

The reactor vessel is vertically mounted within the drywell and consists of a cylindrical shell with an integral rounded bottom head. The top head is also rounded in shape but is removable via the stud and nut arrangement to facilitate refueling operations. The vessel assembly is supported by the vessel support skirt (20) which is mounted to the reactor vessel support pedestal.

The internal components of the reactor vessel are supported from the bottom head and/or vessel wall.

The reactor core is made up of fuel assemblies (15), control rods (16), and neutron monitoring instruments (24). The structure surrounding the active core consists of a core shroud (14), core plate (17), and top guide (12). The components making up the remainder of the reactor vessel internals are the jet pump assemblies (13), steam separators (6), st eam dryers (3), feedwater spargers (8), and core spray spargers (11). The jet pump assemblies are lo cated in the region between the core shroud and the vessel wall, submerged in water. The jet pump assemblies are arranged in two semicircular groups of ten, with each group being supplied by a separate recirculation pump.

The emergency core cooling systems, penetrations number 5 and 9, and the reactor vessel designs are compatible to ensure that the core can be adequately cooled following a loss of reactor coolant. The worst case loss of coolant accident, with respect to core cooling, is a recirculation line break (penetrations number 18 and 19). In this event, reactor water level decreases rapidly, uncovering the core. However, several emergency core cooling systems automatically provide makeup water to the nuclear core within the shroud, providing core cooling.

The control cell assembly (page 3-5) is representative for boiling water reactor 1 through 6. Each control cell consists of a control rod (7) and four fuel assemblies that surr ound it. Unlike the pressurized water reactor fuel assemblies, the boiling water reactor fuel bundle is enclosed in a fuel channel (6) to direct coolant up through the fuel assembly and act as a b earing surface for the control rod. In addition, the fuel channel protects the fuel during refueling ope rations. The power of the core is regulated by movement of bottom entry control rods.

USNRC Technical Training Center 3-3 0400 Reactor Concepts Manual Boiling Water Reactor Systems

BWR 6 Reactor Vessel

USNRC Technical Training Center 3-4 0400 Reactor Concepts Manual Boiling Water Reactor Systems

BWR 6 Fuel Assembly

USNRC Technical Training Center 3-5 0400 Reactor Concepts Manual Boiling Water Reactor Systems

Cont ainment /Drywel l

Safety/Relief Valve

Main Steam Lin e To Main Tur bine Steam Dryer As s e m b l y Main Feedwater Line Steam Separator As s e m b l y

Recirculation Loop Reactor (Typical of 2)

Core

Filt er/

Demineralizer

Recirculation Pump Reactor Water Cleanup Pump Non-Regenerative Heat Exchanger

Regenerative Containment Suppressi on Chamber Heat Exchanger

Reactor Bu ilding C oolin g Water

Reactor Water Cleanup System

The purpose of the reactor water cleanup system (RWC U) is to maintain a high reactor water quality by removing fission products, corrosion products, and othe r soluble and insoluble impurities. The reactor water cleanup pump takes water from the recircula tion system and the vessel bottom head and pumps the water through heat exchangers to cool the flow. The water is then sent through filter/demineralizers for cleanup. After cleanup, the water is returned to the reactor vessel via the feedwater piping.

USNRC Technical Training Center 3-6 Rev 0200 Reactor Concepts Manual Boiling Water Reactor Systems

Containment/D rywell Reactor Ve ss e l Stea m Line Electrical Generator

Steam Dryer Ass emb ly

TurbineBypass Line

ReactorCore

Jet Pump

To/FromRiver

RecirculationPump

Residual Heat Removal Residual Heat Heat Exchanger Containment Suppression Chamber Remo val Pu mp

Service Water

Decay Heat Removal

Heat is removed during normal power operation by generating steam in the reactor vessel and then using that steam to generate electrical energy. When th e reactor is shutdown, the core will still continue to generate decay heat. The heat is removed by bypassing the turbine and dumping the steam directly to the condenser. The shutdown cooling mode of the residual heat removal (RHR) system is used to complete the cooldown process when pressure decr eases to approximately 50 psig. Water is pumped from the reactor recirculation loop through a heat exch anger and back to the reactor via the recirculation loop. The recirculation loop is used to limit the number of penetrations into the reactor vessel.

USNRC Technical Training Center 3-7 Rev 0200 Reactor Concepts Manual Boiling Water Reactor Systems

Cont ainment /Drywel l

Safety/Relief Valve

Main Steam Lin e To Main Tur bine Steam Dryer As s e m b l y Main Feedwater Line Steam Condensate Separa t or Storage Tan k As s e m b l y

Recir cula ti on Loo p Reactor (Typical of 2)

Core

RCIC Recirculation RCIC Turbine Pump Pump

Containment Suppressi on Chamber

Reactor Core Isolation Cooling

The reactor core isolation cooling (RCIC) system provides makeup water to the reactor vessel for core cooling when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. The RCIC system consists of a turbine-driven pump, piping, and valves necessary to deliver water to the reactor vessel at operating conditions.

The turbine is driven by steam supplied by the main st eam lines. The turbine exhaust is routed to the suppression pool. The turbine-driven pump supplie s makeup water from the condensate storage tank, with an alternate supply from the suppression pool, to the reactor vessel via the feedwater piping. The system flow rate is approximately equal to the steaming rate 15 minutes after shutdown with design maximum decay heat. Initiation of the system is accomplished automatically on low water level in the reactor vessel or manually by the operator.

USNRC Technical Training Center 3-8 Rev 0200 Reactor Concepts Manual Boiling Water Reactor Systems

Cont ainment /Drywell

Main Steam Lin e To Main Tur bine St eam Dryer Assembly Main Feedwater Line St eam Se pa ra t or Assembly

Recir cul a ti on Loo p Reactor (Typical of 2)

Core

Poison Tank Recirculation (Boron)

Pump Explosive Valve

Containment Suppr ession Chamber

Standby Liquid Control System

The standby liquid control system injects a neutron poison (boron) into the reactor vessel to shutdown the chain reaction, independent of the control rods, and maintains the reactor shutdown as the plant is cooled to maintenance temperatures.

The standby liquid control system consists of a heated storage tank, two positive displacement pumps, two explosive valves, and the piping necessary to in ject the neutron absorbing solution into the reactor vessel. The standby liquid control system is ma nually initiated and provide s the operator with a relatively slow method of achieving reactor shutdown conditions.

USNRC Technical Training Center 3-9 Rev 0200 Reactor Concepts Manual Boiling Water Reactor Systems

Emergency Core Cooling Systems

The emergency core cooling systems (ECCS) provide core cooling under loss of coolant accident conditions to limit fuel cladding damage. The emer gency core cooling systems consist of two high pressure and two low pressure systems. The high pressure systems are the high pressure coolant injection (HPCI) system and the automatic depressuri zation system (ADS). The low pressure systems are the low pressure coolant injection (LPCI) mode of the residual heat removal system and the core spray (CS) system.

The manner in which the emergency core cooling system s operate to protect the core is a function of the rate at which reactor coolant invent ory is lost from the break in the nuclear system process barrier. The high pressure coolant injection system is designed to operate while the nuclear system is at high pressure.

The core spray system and low pressure coolant inj ection mode of the residual heat removal system are designed for operation at low pressures. If the break in the nuclear system pro cess barrier is of such a size that the loss of coolant exceed s the capability of the high pressure coolant injection system, reactor pressure decreases at a rate fast enough for the low pressure emergency core cooling systems to commence coolant injection into the reactor vessel in time to cool the core.

Automatic depressurization is provided to automatica lly reduce reactor pressure if a break has occurred and the high pressure coolant injection system is i noperable. Rapid depressurization of the reactor is desirable to permit flow from the low pressure emer gency core cooling systems so that the temperature rise in the core is limited to less than regulatory requirements.

If, for a given break size, the high pressure coolant injection system has the capacity to make up for all of the coolant loss, flow from the low pressure emergency core cooling systems is not required for core cooling protection until reactor pressure has decreased below approximately 100 psig.

The performance of the emergency core cooling syst ems as an integrated package can be evaluated by determining what is left after the postulated break and a single failure of one of the emergency core cooing systems. The remaining emergency core coo ling systems and components must meet the 10 CFR requirements over the entire spectrum of break locati ons and sizes. The integrated performance for small, intermediate, and large sized breaks is shown on pages 3-11 and 3-12.

USNRC Technical Training Center 3-10 Rev 0200 Reactor Concepts Manual Boiling Water Reactor Systems

e CI S Single HP stem AD Failur Sy ate all or edi Sm erm Int Break LOCA

S CI stem CS AD HP Sy

maining EC ly Re mp stem 1 CS System 2 CS mp stem 1 CS System 2 CS nce Any RHR Pu Sy Any RHR Pu Sy all Break On Sm rforma

re oling hieved Co Co Ac tegrated Pe In stem 2 CS stem 2 RHR CS Sy Sy EC

CS

stem 1 CS stem 1 RHR Sy Sy maining EC Re

stem 2(1) RHR stem 2(1) CS Sy Sy

rge CA La Break LO

e Single stem 1(2) RHR stem 1(2) CS Failur Sy Sy

USNRC Technical Training Center 3-11 Rev 0200 Reactor Concepts Manual Boiling Water Reactor Systems

Diesel Diesel Generator Generator A A

Shutdown Board Shutdown Board

A

T C HPCI System

RHR Syste m 1 Feedwater Line Recirculation Pump B

C Reactor B Vessel

A D

Core Spray Core Spray Syste m 1 Syste m 2

Steam Line Recirculation Pump A ADS

B

To Suppression D P ool RHR Syste m 2

Emergency Core Cooling System Network

USNRC Technical Training Center 3-12 Rev 0200 Reactor Concepts Manual Boiling Water Reactor Systems

Cont ainment /Dr y wel l

Sa fety/Relief Valve

Main Steam Lin e To Main Tur bine St ea m Dryer Assembly Main Feedwater Line St ea m Condensate Se pa ra t or Storage Tan k Assembly

Recirculation Loop Reactor (Typical of 2)

Core

HPCI Recirculation HPCI Turbine Pump Pump

Containment Suppression Chamber

High Pressure Emergency Core Cooling Systems

The high pressure coolant injection (HPCI) system is an independent emergency core cooling system requiring no auxiliary ac power, plant air systems, or external cooling water systems to perform its purpose of providing make up water to the reactor ve ssel for core cooling under small and intermediate size loss of coolant accidents. The high pressure coolant injection system can supply make up water to the reactor vessel from above rated reactor pressure to a reactor pressure below that at which the low pressure emergency core cooling systems can inject.

The automatic depressurization system (ADS) consis ts of redundant logics cap able of opening selected safety relief valves, when required, to provide reactor depressurization for events involving small or intermediate size loss of coolant acci dents if the high pressure coolant injection system is not available or cannot recover reactor vessel water level.

USNRC Technical Training Center 3-13 Rev 0200 Reactor Concepts Manual Boiling Water Reactor Systems

Cont ainment /Drywel l Con tainm ent Spray

St eam Main Steam Lin e Dryer Assembly

Main Feedwater Line

Recirculation Loop Reactor (Typical of 2) RHR Core LPCI He a t Jet Pump Exchanger

Service Water Recirculation Pump

Con tainm ent Spray RHR Pumps (LP CI Mode)

Containment Suppression Chamber

Core Spray Pump

Low Pressure Emergency Core Cooling Systems

The low pressure emergency core cooling systems cons ist of two separate and independent systems, the core spray system and the low pressure coolant injection (LPCI) mode of the residual heat removal system. The core spray system consists of tw o separate and independent pumping loops, each capable of pumping water from the suppression pool into the reactor vessel. Core cooling is accomplished by spraying water on top of the fuel assemblies.

The low pressure coolant injection mode of the re sidual heat removal system provides makeup water to the reactor vessel for core cooling under loss of cool ant accident conditions. The residual heat removal system is a multipurpose system with several ope rational modes, each utilizing the same major pieces of equipment. The low pressure coolant injecti on mode is the dominant mode and normal valve lineup configuration of the residual heat removal system. The low pressure coolant injection mode operates automatically to restore and, if necessary, maintain the reactor vessel coolant inventory to preclude fuel cladding temperatures in excess of 2200EF. During low pressure coolant injection operation, the residual heat removal pumps take water from the suppression pool and discharge to the reactor vessel.

USNRC Technical Training Center 3-14 Rev 0200 Reactor Concepts Manual Boiling Water Reactor Systems

Boiling Water Reactor Containments

The primary containment package provided for a pa rticular product line is dependent upon the vintage of the plant and the cost-benefit analysis performed prior to the plant being built. During the evolution of the boiling water reactors, thr ee major types of containments were built. The major containment designs are the Mark I (page 3-16), Mark II (page 3-17), and the Mark III (page 3-18). Unlike the Mark III, that consists of a primary containment and a dr ywell, the Mark I and Mark II designs consist of a drywell and a wetwell (suppression pool). All three containment designs use the principle of pressure suppression for loss of coolant accidents. The prim ary containment is designed to condense steam and to contain fission products released from a loss of coolant accident so that offsite radiation doses specified in 10 CFR 100 are not exceeded and to provide a heat sink and water source for certain safety-related equipment.

The Mark I containment design consists of severa l major components, many of which can be seen on page 3-16. These major components include:

• The drywell, which surrounds the reactor vessel and recirculation loops,

• A suppression chamber, which stores a large body of water (suppression pool),

• An interconnecting vent network between the drywell and the suppression chamber, and

• The secondary containment, which surrounds th e primary containment (drywell and suppression pool) and houses the spent fuel pool and emergency core cooling systems.

The Mark II primary containment consists of a steel dome head and either a post-tensioned concrete wall or reinforced concrete wall standing on a base mat of reinforced concrete. The inner surface of the containment is lined with a steel plate that acts as a leak-tight membrane. The containment wall also serves as a support for the floor sl abs of the reactor building (secondary containment) and the refueling pools. The Mark II design is an over-under configuration. The drywell, in the form of a frustum of a cone or a truncated cone, is located directly above the suppression pool. The suppression chamber is cylindrical and separated from the drywell by a rein forced concrete slab. The drywell is topped by an elliptical steel dome called a drywell head. The drywell inerted atmosphere is vented into the suppression chamber through as series of downcomer pipes penetrating and supported by the drywell floor.

The Mark III primary containment consists of several major components, many of which can be seen on page 3-18. The drywell (13) is a cylindrical, reinfor ced concrete structure with a removable head. The drywell is designed to withstand and confine steam generated during a pipe rupture inside the containment and to channel the released steam into the suppression pool (10) via the weir wall (11) and the horizontal vents (12). The suppression pool contai ns a large volume of water for rapidly condensing steam directed to it. A leak ti ght, cylindrical, steel containment vessel (2) surround the drywell and the suppression pool to prevent gaseous and particulate fission products from escaping to the environment following a pipe break inside containment.

USNRC Technical Training Center 3-15 Rev 0200 Reactor Concepts Manual Boiling Water Reactor Systems

Mark I Containment

USNRC Technical Training Center 3-16 Rev 0200 Reactor Concepts Manual Boiling Water Reactor Systems

Mark II Containment

USNRC Technical Training Center 3-17 Rev 0200 Reactor Concepts Manual Boiling Water Reactor Systems

Mark III Containment

USNRC Technical Training Center 3-18 Rev 0200