Binding Energy per nucleon
most stable nuclei
Radioactive decay is the release of
binding energy as a nucleus decays
to a more stable isotope.
Excess binding energy is released
in any nuclear reaction in which
heavy-mass nucleus is broken into
two intermediate-mass nuclei.
Combining two light-mass nuclei
into a heavier nucleus also
releases binding energy.
Fission Probability
[K&R, 1993]
Neutron Cross Section
[Culp, 1991]
2
Types of Neutron Reactors
Fundamentals of Nuclear Science and Engineering, 2nd ed., Shultis and Faw, CRC Press, 2008.
3
4
U.S. Nuclear Power Reactors
Babcock & Wilcox
lowing general description applies to Oconee Unit 1, al
though it is generally applicable to all U.S. B&W plants;
the concepts are applicable to all PWRs.
Containment building
Fig. 2 is a vertical section of the reactor containment
building. The structure is post-tensioned, reinforced con
crete with a shallow domed roof and a flat foundation slab.
The cylindrical portion is prestressed by a post-tensioning
system consisting ofhorizontal and vertical tendons. The
dome has a three way post-tensioning system. The foun
dation slab is conventionally reinforced with high
strength steel. The entire structure is lined with 0.25 in.
(6.3 mm) welded steel plate to provide a vapor seal.
The containment building dimensions are: inside di-
ameter 116 ft (35.4 m); inside height 208 ft (63.4 in); wall
thickness 3.75 ft (1.143 m); dome thickness 3.25 ft (0.99
m); and foundation slab thickness 8.5 ft (2.59 m). The
building encloses the reactor vessel, steam generators,
reactor coolant loops, and portions of the auxiliary and
safeguard systems. The interior arrangement meets the
requirements for all anticipated operating conditions and
maintenance, including refueling. The building is de
signed to sustain all internal and external loading con
ditions which may occur during its design life.
Reactor vessel and steam generators
Also shown in Fig. 2 are the reactor vessel and the two
vertical steam generators, each of which produces ap
proximately 5.6 x 106 lb/li (705.6 kg/s) of steam at 910
1
OnceThrough
Steam Generators (2)
fl
/
/
Reactor coolant
Pumps (4)
I
Pressurizer
Reactor Vessel
Steam, 40th ed., Babcock & Wilcox
Fig. 3 B&W nuclear steam supply system.
47-4
Steam 40 / Nuclear Installations for Electric Power
Thium, was designed and subsequently built. Italy and
r Japan also built versions of these British designs.
The Soviet Union developed a graphite-moderated,
boiling water-cooled, enriched uranium reactor from its
early Obninsk power station design. A major difference
in this reactor from similar Western designs was that it
did not include a sturdy reactor containment building
The early Soviet goals included the development of na
*val propulsion reactors based upon the PWR concept
which was then applied to civilian use. This PWR design
was exported to several Soviet bloc countries and Fin
land. The Soviet PWR was similar to the U.S. designs
except it generally provided significantly lower electri
cal output
The excellent moderator characteristics of deuterium
oxide (heavy water) became the basis for the very suc
cessful Canadian deuterium (CANDU) reactors. These
natural uranium-fueled reactors were first operated in
1962 (25 MW Nuclear Power Demonstration reactor at
1: Rolphton, Ontario), and subsequent designs increased
output to 800 MW. CANDU designs have been built in
India, Pakistan, Argentina, Romania and South Korea.
In Japan, electric utilities were experimenting with
several types ofimported nuclear generation plant designs.
They ordered gas-cooled reactors from the U.K., and BWR
and PWR designs from the U.S. The Japanese government
also began sponsorship of a breeder reactor design.
The Germans were actively involved in designs pri
marily associated with high temperature gas reactor
J concepts. In 1969, the formation of Kraftwerk Union in
troduced German PWR and BWR designs based on U.S.
• technology. The intent was to construct a series of these
plants to achieve economies of scale. However, unlike the
French program, few plants were constructed. Several
German PWRs were exported to Brazil, Spain, Switzer
land and the Netherlands
The energy crisis of the mid 1970s caused significant
reductions in worldwide electric usage growth rates, and
Steam 40/ Nuclear Installations for Electric Power
structure between reactor buildings 1 and 2. Correspond
ing equipment for Unit 3 is located separately. The fol
Fig. 2 PWR containment building, sectional view.
Steam, 40th ed., Babcock & Wilcox
47-3
r
Babcock & Wilcox
—
Steam, 40th ed., Babcock & Wilcox
Fig. 4 Reactor containment building, ground floor plan view.
psi (62.7 bar) and 595F (313C). This provides GOF (33C)
of superheat. The B&W nuclear steam supply system
(NSSS) (Fig. 3) is the only large scale commercial design
that provides the added benefit of superheated steam
output. A plan view of the containment building is shown
in Fig. 4. The concrete shielding around the NSSS com
ponents is also shown. Table 1 provides a listing of the
important design parameters for the Oconee-type NSSS.
Adso provided are the design parameters for a larger B&W
Primary
Flow Loop
NSSS. The first of these la
into operation in 1987 at th
station in West Germany. T
der construction in the U.S
The NSSS for a typical se
water reactor consists of th
steam generators, coolant r
piping to connect these com
simplified schematic provi’
sists of two water flow circ
uses pressurized water tot,
core to the steam generati
reactor coolant pump. Th
primary loop pressure hi
generation in the reactor e
The steam generators prov
mary coolant loop and the
flow loop. In the B&W desi
water enters the steam ged
heated steam, which is son
duce power. The primary R
supplying the energy to go
information is included in
Fig. 6 shows a reactor vet
steam generators inside a ci
selis constructed oflowaflc
less steel cladding. Externt
is 20.75 ft (6.32 m) with an I
m) including the closure I
head is held in place by Mi
high tensile alloy steel stui
vessel and closure head is
The vessel has six maji
flow (two outlet hot leg anc
ant water enters the vest
nozzles located above the
flows downward in the am
tor vessel and the interna
the core and upper plenul
the upper portion of the v
hot leg nozzles. Fkr 7 fini
a nuclea
•r
i’c”?
r—U IImnd1 o.’-fl_
primary loop press1j rnw’n
Fig. 4 Reactor containment building, ground floor plan view.
psi (62.7 bar) and 59SF (313C). This provides 60F (33C)
of superheat. The B&W nuclear steam supply system
(NSSS) (Fig. 3) is the only large scale commercial design
that provides the added benefit of superheated steam
output. A plan view of the containment building is shown
in Fig. 4. The concrete shielding around the NSSS com
ponents is also shown. Table 1 provides a listing of the
important design parameters for the Oconee-type NSSS.
Also provided are the design parameters for a larger B&W
Steam, 40th ed., Babcock & Wilcox
generation in the reactor
The steam generators provide the Ii
mary coolant loop and the power p
flow loop. In the B&W design, subec
water enters the steam generator an
heated steam, which is sent to the si
duce power. The primary side water
supplying the energy to generate U,
information is included in Table 1 a
Fig. 6 shows a reactor vessel being
steam generators inside a containme
sel is constructed oflow alloy steel wi
less steel cladding. External diamet
is 20.75 ft (6.32 m) with an internal h
m) including the closure head. Th
head is held in place by sixty R 5 ir,
high tensile alloy stee
vessel and closure he
The vessel has sb
flow (two outlet hot leg and tour mu
ant water enters the vessel throu;
nozzles located above the midpoint
flows downward in the annular spa
tor vessel and the internal therma
the core and upper plenum, down
the upper portion of the vessel and
hot leg nozzles. Fig. 7 shows a react
a nuclear power plant site.
Steam Outlet
Inlet
To Turbine
—
—---
———
—.
—
——- -.-
From Last
Feedwater Heater
Baffle
t Loop
system Containment
515—Safety Injection System
4
4
4
———
Secondary coolant Loop
Fig. 5 Simplified schematic of primary and secondary loops.
Steam 40 / Nuclear Installations for Electric Power
Fig. 6 Reactor vessel pk
Reactor Concepts Manual
Pressurized Water Reactor Systems
CONTAINMENT
MOISTURE SEPARATOR
REHEATER (MSR)
SAFETY
VALVES
THROTTLE
VALVE
MAIN STEAM
ISOLATION VALVE
(MSIV)
HP
LP
LP
S/G
HP
HEATER
PZR
MAIN
TURBINE
ELECTRIC
GENERATOR
MAIN
CONDENSER
CIRC.
WATER
CIRC.
WATER
LP
HEATER
CORE
LP
HEATER
REACTOR COOLANT
SYSTEM
RCP
CLEAN UP
SYSTEM
MAIN
FEEDWATER
PUMP
CONDENSATE
PUMP
CONTAINMENT SUMP
The major secondary systems of a pressurized water reactor are the main steam system and the
condensate/feedwater system. Since the primary and secondary systems are physically separated from each other
(by the steam generator tubes), the secondary system will contain little or no radioactive material.
The main steam system starts at the outlet of the steam generator. The steam is routed to the high pressure main
turbine. After passing through the high pressure turbine, the steam is piped to the moisture separator/reheaters
(MSRs). In the MSRs, the steam is dried with moisture separators and reheated using other steam as a heat
source. From the MSRs, the steam goes to the low pressure turbines. After passing through the low pressure
turbines, the steam goes to the main condenser, which is operated at a vacuum to allow for the greatest removal
of energy by the low pressure turbines. The steam is condensed into water by the flow of circulating water
through the condenser tubes.
At this point, the condensate/feedwater system starts. The condensed steam collects in the hotwell area of the
main condenser. The condensate pumps take a suction on the hotwell to increase the pressure of the water. The
condensate then passes through a cleanup system to remove any impurities in the water. This is necessary
because the steam generator acts as a concentrator. If the impurities are not removed, they will be left in the
steam generator after the steam forming process, and this could reduce the heat transfer capability of the steam
generator and/or damage the steam generator tubes. The condensate then passes through some low pressure
feedwater heaters. The temperature of the condensate is increased in the heaters by using steam from the low
pressure turbine (extraction steam). The condensate flow then enters the suction of the main feedwater pumps,
which increases the pressure of the water high enough to enter the steam generator. The feedwater now passes
through a set of high pressure feedwater heaters, which are heated by extraction steam from the high pressure
turbine (heating the feedwater helps to increase the efficiency of the plant). The flow rate of the feedwater is
controlled as it enters the steam generators.
USNRC Technical Training Center
4-20
0603
Reactor Concepts Manual
Pressurized Water Reactor Systems
SPRAY NOZZLE
SAFETY NOZZLE
RELIEF
NOZZLE
MANWAY
UPPER HEAD
INSTRUMENTATION
NOZZLE
LIFTING
TRUNNION
(LOAN BASIS)
SHELL
LOWER HEAD
HEATER SUPPORT
PLATE
INSTRUMENTATION
NOZZLE
ELECTRICAL HEATER
SUPPORT SKIRT
SURGE NOZZLE
Cutaway View of a Pressurizer
USNRC Technical Training Center
4-18
0603
Power Plant Engineering, Black & Veatch, 1996
ower Plant Engineering
8.
Typical Design Parameters and Data for Components
of a Large Four-Loop PWR
RELIEF
NOZZLE
olant pump
ature
height
ngrpm
diameter
erator (U-tube)
height
shell diameter (outside)
shell diameter (outside)
pressure (tube side)
ng pressure (tube side)
e (shell side)
pressure (shell side)
How rate
coolant flow rate
coolant inlet temperature
coolant outlet temperature
ight
SPRAY
NOZZLE
Is)
3
97,600 gpm (6.15 m
m)
ft
(85.3
280
2,500 psia (17.2 MPa)
650° F (343° C)
28 ft (8.5 m)
1,189
6ft-5 in. (196 cm)
7,000 hp (5,220 kW)
NOZZLE
MAN WAY
UPPER HEAD
INSTRUMENTATION
NOZZLE
TEMPORARY
LIFTING
TRUNNION
67 ft-8 in. (20.6 m)
14 ft-7¼ in. (4.5 m)
11 ft-3 in. (3.4 m)
2,500 psia (171.2 MPa)
2,250 psia (15.5 MPa)
1,000 psia (6.9 MPa)
1,200 psia (8.3 MPa)
3.8 x 106 lbih (480 kgls)
35 x 106 lb/h (4,419 kg/s)
621°F (327° C)
558°F (292°C)
346 tons (314 tonnes)
SHELL
stinghouse 1984.
ssure parts are carbon or low-alloy steel. Areas
reactor coolant contact are clad with stainless steel
tghouse and Combustion Engineering use U-tube
ierators. Thç steam generators used by Babcock &
t vertical, straight tube and shell exchanges. Reac
it enters at the top of the steam generator and slows
d through the Inconel tubes and exits at the bottom
tin generator. Feedwater enters the steam generator
middle and flows down an annulus between the
Derator shell and tube baffle and enters near the
rhe water, as it passes up the steam generator, is
Ito steam. The steam is superheated before leaving
generator. Since the steam is superheated, moisture
s and dryers are not used.
ses or decreases in the generating plant electrical
;e fluctuations in the reactor coolant pressure. A
r on one of the RCS loops is used to limit pressure
luring load transients and to maintain constant cool
m pressure during steady-state operation. Figure
,ws a typical pressurizer. The pressurizer contains
ELECTRICAL
HEATER
HEATER
SUPPORT
PLATE
LOWER HEAD
INSTRUMENTATION
NOZZLE
SURGE
Fig. 23-12. Pressurizer for a large PWR design. (From Westinghouse.
Used with permission.)
creases above the pressurizer’s capability. Both relief and
safety valves vent to a pressure relief tank that contains
cooler water for condensing steam. A pressurizer for a large
four-loop design has a typical height of approximately 53 ft
(16.2 in), with an 8 ft (2.3.m) diameter, and contains approx
imately 75 immerser heaters with 1,800 kW of power. The
pressurizer is low-alloy steel with stainless steel interior
Reactor Concepts Manual
Pressurized Water Reactor Systems
SAFETIES
RELIEFS
RCS
SPRAY
VALVES
RCS
COOLING
SPRAY
PRESSURIZER
HEATERS
VENT
SURGE LINE
PRESSURIZER
RELIEF TANK
REACTOR COOLANT
SYSTEM (RCS)
DRAIN
Pressurizer and Pressurizer Relief Tank
USNRC Technical Training Center
4-19
0603
t[IWdI .sm-a+tari
steam then passes through moisture separators in the
upper
steam generator shell, where the entrained liquid is remo
ved.
The steam then passes through steam dryers to raise
the
steam quality to 99.75%. Table 23-8 shows typical
parame
ters and other data for reactor coolant pumps and
steam
generators used in a large four-loop PWR design. The
steam
generator U-tubes are made of Inconel, and the shell
and
MIXING
VANES
Power Plant Engineering, Black & Veatch, 1996
Steam Nozzle
with IRestrictor
DASHPOT
REGION
Secondary
Separator
DIMPLE
Manway
THIM8LE
SCREW
17 Fuel assembly with control rod cluster. (From
sed with permission.)
Upper Shell
Feedwater
Nozzle
Primary
Separator
Feedwater Ring
Antivibration
Bars
pical Reactor Fuel Parameters for a Large PWR
mr-Loop Design
Parameter
Tube Bundle
Wrapper
Tube Support Plate
Value
Tube Bundle
diameter
g thickness
il
Bter
)d arrangement
Lssembly
i a core
ol rod cluster assemblies
‘Se 1984.
12 ft (3.66 m)
0.36 in. (0.914 cm)
0.0225 in. (0.0572 cm)
Zircaloy 4
0.3088 in. (0.7844 cm)
0.496 in. (1.26 cm)
17)< 17
264
50,952
53
Lower Shell
Handhole
Channel Head
Slowdown Pipe
Tubesheet
Fig. 23-li. Steam generator for a large PWR. (From Westinghouse.
Used with permission.)
nail tflat in a I’WR, with the fuel temperatures
roughly comparable. Because Water
and vapor coexist in the core, a BWR produces
saturated steam at about 545°F (285°C).
The coolant thus serves The triple function of
coolant, moderator, and working fluid.
In its simplest form (Fig. 10-15), a boiling-waterreactor powerplant consists of
a reactor, a turbine generator, a condenser and
associated equipment (air ejector,
cooling system, etc.), and a feed pump. Slightly
subcooled liquid enters the reactor
core at the bottom, where it receives sensible heat
to saturation plus some latent heat
of vaporization. When it reaches the top of the core
, it has been converted into a very
wet mixture of liquid and vapor. The vapor sepa
rates from the liquid, flows to the
turbine, does work, is condensed by the condense
r, and is then pumped back to the
reactor by the feedwater pump.
The saturated liquid that separates from the vapor
at the top of the reactor or in
a steam separator flows downward via downcomers
within or outside the reactor and
mixes with the return condensate. This recirculat
ing coolant flows either naturally, by
the density differential between the liquid in the dow
ncomer and the two-phase mixture
in the core, or by recirculating pumps in the dow
ncomer (not shown in the figure).
This is similar to what happens in modern large fossi
l-fueledsteam generators. Modern
large boiling-water reactors are of the internal, force
d recirculation type.
The ratio of the recirculation liquid to the satur
ated vapor produced is called the
recirculation ratio. It is a function of the core aver
age exit quality [Eq. (10-9), below].
Boiling-water core exit qualities are low, between
10 and 14 percent, so that recir
culation ratios in the range of 6 to 10 are com
mon. This is necessary to avoid large
void fractions in the core, which would mate
rially lower the moderating powers of
Saturated
steam to turbine
mg
(a)
(b)
Figure 10-15 Schematic of a BWR syste
m: (a) internal and (b) external recirculation.
V
THERMAL-HSSION REACTORS AND PORPLANTS
I
421
:i
This state of affairs has been corrected by one of several methods: by-pass control,
a dual-pressure cycle, an overmoderated reactor in which voiding results in a reduced
loss of neutrons due to water absorption and hence increased power, and recirculation
control [3] The last method is the one adopted in current designs. It is described in
the next section.
.
10-8 BWR LOAD FOLLOWING CONTROL
The recirculation control method is based on a direct cycle but with variable recir
culation flow in the downcomer, It is shown schematically in Fig. 10-17. Equation
(10-17) is rewritten with the help of Eqs. (lO-8a) and (10-9) as
t = ?etZi(hg
Q
—
(10-18)
hd)
hg, the saturated steam enthalpy at the system pressure, and hd, the feedwater enthalpy,
are both weak functions of load. Thus the plant load Q
t is therefore proportional to
the product of e and ñz, the flow in the downcomer.
MAIN STEAM FLOW
TO TURBINE
STEAM DRYERS
STEAM SEPARATORS
MAIN FEED FLOW
FROM TURBiNE
DRVINGFLOW
—.-
RECI RCULATION
PUMP
Figure 10-17 BWR reactor vessel internal flow paths.
Reactor Concepts Manual
Containment/Drywell
Boiling Water Reactor Systems
Reactor
Vessel
Steam
Line
Turbine Building
Throttle
Valve
Steam Dryer
&
Moisture
Separator
Electrical
Generator
Turbine
Reactor Core
Condenser
Jet Pump
To/From
River
Recirculation
Pump
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 absorbing heat. The major difference in the operation
of a BWR from other nuclear systems is the steam void formation in the core. The steam-water mixture
leaves the top of the core and enters the two stages of moisture separation, where water droplets are
removed before the steam is allowed to enter the steam line. The steam line, in turn, directs the steam
to the main turbine causing it to turn the turbine and the attached electrical 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 pumps 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 6 Reactor Vessel
USNRC Technical Training Center
3-4
0400
750
Power Plant Engineering
Power Plant Engineering, Black & Veatch, 1996
STEAM OUTLET
CORE SPRAY INLET
LOW PRESSURE
COOLANT
INJECTION INLET
CORE SPRAY SPARGER
JET PUMP ASSEMBLY
FUEL ASSEMBLIES
CORE PLATE
JET PUMP/RECIRCULATION
WATER INLET
RE CIRCU tAT ID N
WATER OUTLET
SHIELD WALL
VESSEL SUPPORT SKIRT
CONTROL ROD DRIVES
-
CONTROL ROD DRIVE
HYDRAULIC LINES
which fits into
for transfen-jnj
signs manufaci
x 9 and Jo <
designs are cc
overseas.
Three types
standard fuel
outer edge of
lower and upp
assembly dun
fuel assembly
clad tubes will
The water rod
remaining 5k
slightly enrici
into anchor ho
a BWR than f
along Ihe fuel
ing. Each fue
channel. The
the fuel rods
Control ro
typical contra
shape and co
sheath. Each
with each tub
ing. Each coi
Control rods
-
IN-C ORE
FLUX MONITOR
Fig. 23-15. Pressure vessel for a BWR. (From General EleCtric
Company. Used with permission.)
and flows back into the annulus region where it mixes with
entering feedwater.
The steam leaves the separators at the top flowing up
wards and outwards through a series of drying vanes in the
steam dryer where additional separated water is recovered.
The steam then leaves the pressure vessel at the steam outlet
nozzle and passes to the turbine—generator through the main
steam lines. The steam separator assembly sits on the top
flange of the core shroud and forms the top of the core
discharge plenum. The steam dryer assembly is supported by
pads extending from the pressure vessel wall above the
ml
1
t,A k,1A,,
il-in otoni
BAIL
HANDLE
UPPER
TIE PLATE
FUEL
BUNDLE
FUEL ROD
INTERIM