13_ReactorTypes

7/30/2019 13_ReactorTypes

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HT2005: Reactor Physics T13: Reactor Types 1

Types of Nuclear Reactors

Generations of Reactors

Principles of Classification Survey of Reactor Types

Light Water Reactors

Burners, Converters, and Breeders


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Reactor Generations

1942-Dec-2 Chicago Pile-1


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Principles of Classification

Energy spectrum utilized for fission

Prime purpose

Fuel

Coolant

Coolant system

Moderator Heterogeneity (?)


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Heterogeneity

All reactors used today for power generation are

HETEROGENEOUS, i.e. fuel, coolant and/or moderator arephysically different entities with non-uniform and anisotropiccomposition.

The homogeneous reactor, is defined as a reactor whose small-scale composition is uniform and isotropic.

ORNL, Homogeneous Reactor Experiment: HRE-1, HRE-2(1950s); thermal breeder, fuel: uranyl sulfate (UO2SO4) in heavywater (D2O). HRE were dropped liquid fuel.

Molten salt reactor, 235UF4+ThF4,233UF4. In 1960s fast breeder.

A further homogenous reactor approach, a liquid- metal thermal

breeder using uranium compounds in molten bismuth. At present there is no prospect in sight of a major program todevelop homogenous reactors

molten salt accelerator-driven concept (?)


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Homogeneous Reactor Experiment


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Energy

Thermal (0.5 eV) Intermediate (103 eV) Fast (105 eV)

Power reactors:

BWR, PWR,VVER, RBMK;

thermal breeder,

Spaceship reactors

(accident in Can.)

BN350, BN600,

Super Fenix,Naval reactors

Energy Spectrum Classification


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Purpose

Power Ship propulsion Production Research Specialized

LWR

PHWR (CANDU)

HTGR

AGRLMR

LGR (RBMK)

LMFBR

HWLWR

OLR

Submarines

Navy

Aircarrier

IcebreakersPWR-cargo

Plutonium

Tritium

WPR

TRIGA

Studsvik

Isolated areas

Space propul.

Heat for chem


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Fuel

UO2 (3-4%)

sinteredpelletszirconium alloy

Nat. Upelletstubes

UO2spherical part.carbon, silicon

U-Pu-Zr alloypellets

steel cladding

U-Al alloyplates

Al cladding + Al-Si

LWRCANDU

HTGR LMRTRIGA


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Coolant

Light Water

H2O

Heavy Water

D2O

Gas

air, CO2, He

Liquid Metal

Na, Na-K, Pb-Bi

LWRCANDU

HTGR

Pebble bed

LMR


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Moderator

Light Water

H2O

Heavy Water

D2O

Graphite

C

LWRCANDU

LGR (RBMK)

GCR

HTGR


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1-loop

BWR

2-loopPWR

3-loop

LMR


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Commonly Accepted Classification

PWRBWR

German

USA

Na

Pb-Bi

Nuclear

Reactors

Power

Reactors

Specialized

Reactors

Research

Reactors

Ship PropulsionReactors

Production

Reactors

CANDU AGR Other

PWR Fast

LMR

HTGR LWR

UnatC+H2O

UnatC+Gas

WPR

TRIGA


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Fuel

Coolant Moderator

Control system (rods)

A main distinction between different types of reactors lies in thedifferences in the choices of fuel, coolant, and moderator.

Components of

Conventional Reactors


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90

U (2 ppm)

4.5109 yr90Th (6 ppm)

14.05109 yr

238 239

92 94U Pu

99.5%

235

92U

0.7%

23.5min 2.3day238 1 239 239 239

92 0 92 93 94U U Np Pun

23.3min 27.4day232 1 233 233 233

90 0 90 91 92Th Th Pa Un

Fuels

Natural Elements

233U

7108 yr

Artificial Nuclides

239Pu

24103 yr

235

233

239

0.0064

0.0026

0.0020


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Between thorium (Z=90) and bismuth (Z=83), the isotope with the

longest half-life is 226Ra (T=1600 years), and therefore there are

no fuel candidates, quite apart from the issue of fissionability.

Uranium and Thorium are the only natural elements available

for use as reactor fuels. In addition, 233U and239Pu can be

produced from capture on 232Th and 238U in reactors. Of fissile

materials, only U is both fissile and found in nature in useful

amounts.


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10-3 10-2 10-1 100 101 102 103 104 105 106 107

Energy (eV)

10-2

10-1

100

101

102

103

104

s(

barns)

232 Th

capture

fission

T 1/2 => 14.05 Gy +/- 60.00 My [4.434E+17s +/- 1.893E+15s]SPIN => 0 +

MEAN DECAY ENERGIES (MeV)

Baryon => 4.07742 +/- 0.169064

Lepton => 0.0130345 +/- 0.00112065

Photon => 0.00124304 +/- 0.000114618

* DECAY MODES:

Alpha decay, possibly followed by Gamma emission

Q_value => 4.081 MeV +/- 0.004 MeV Branching => 1 +/- 0

Fission decay, followed by Gamma emissionQ_value => 169.4 MeV 4.1 MeV Branching => 1.4e-11 5e-12

* NUCLIDE MAT. NUMBER: 9061

The following number of lines are present:

Gamma: 2 Alpha: 3 Electron: 8 X-rays: 7

23.3min 27.4day232 1 233 233 233

90 0 90 91 92Th Th Pa Un

232Th

9

1 2 14.05 10 yrT


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10-3 10-2 10-1 100 101 102 103 104 105 106 107

Energy (eV)

10-2

10-1

100

10

1

102

103

104

s

(barns)

233U

capture

fission

T 1/2 => 159.25 ky +/- 200.00 y [5.026E+12s +/- 6.312E+09s]

SPIN => 5/2 +

* MEAN DECAY ENERGIES (MeV)

Baryon => 4.90413 +/- 0.0181258

Lepton => 0.00759652 +/- 0.000610716

Photon => 0.00122537 +/- 0.000110582

* DECAY MODES:Alpha decay, possibly followed by Gamma emission

Q_value => 4.9089 MeV +/- 0.0012 MeV Branching => 1 0




233U

3

1 2 159.25 10 yrT 233

235

0.0026 1 1.0013

0.0064 1 1.0032

k

k


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10-3 10-2 10-1 100 101 102 103 104 105 106 107

Energy (eV)

10-2

10-1

100

10

1

102

103

104

s

(barns)

235U

capture

fission

T 1/2 => 703.81 My +/- 500.01 ky [2.221E+16s +/- 1.578E+13s]

SPIN => 7/2 -


Baryon => 4.46298 +/- 0.287321

Lepton => 0.0475365 +/- 0.00378036

Photon => 0.167808 +/- 0.00347714

* DECAY MODES:

Alpha decay, possibly followed by Gamma emissionQ_value => 4.679 MeV +/- 0.0025 MeV Branching => 1 0

Fission decay, followed by Gamma emission

Q_value => 176.4 MeV +/- 4.6 MeV Branching => 2e-10 +/- 1e-10




235U

6

1 2 703.81 10 yrT

235

233

239

0.0064

0.0026

0.0020


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10-3 10-2 10-1 100 101 102 103 104 105 106 107

Energy (eV)

10-2

10-1

100

101

102

103

104

s(

barns)

238U

capture

fission

T 1/2 => 4.47 Gy +/- 5.00 My [1.410E+17s +/- 1.578E+14s]

SPIN => 0 +


Baryon => 4.25996 +/- 0.236186

Lepton => 0.0105454 +/- 0.000870616

Photon => 0.00125402 +/- 0.000119247

* DECAY MODES:Alpha decay, possibly followed by Gamma emission

Q_value => 4.2703 MeV +/- 0.0039 MeV Branching => 0.999999


Q_value => 173.6 MeV +/- 3.4 MeV Branching => 5.4e-07 +/- 2e-08




9

1 2 4.47 10 yrT

238U

mg of U3 3ppb

tonne Seawater


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10-3 10-2 10-1 100 101 102 103 104 105 106 107

Energy (eV)

10-2

10-1

100

101

102

103

104

s(

barns)

239Pu

capture

fission

T 1/2 => 24.11 ky +/- 40.00 y [7.609E+11s +/- 1.262E+09s]

SPIN => 1/2 +


Baryon => 5.23678 +/- 0.0389047

Lepton => 0.00738587 +/- 0.000585825

Photon => 0.000707559 +/- 6.46626e-05

* DECAY MODES:


Q_value => 5.2435 MeV +/- 0.0007 MeV

Branching => 0.000115 +/- 2e-06


Q_value => 5.2434 MeV +/- 0.0007 MeVBranching => 0.999885 +/- 2e-06


Q_value => 184.3 MeV +/- 3.6 MeV

Branching => 4.4e-12 +/- 4e-13




239Pu

3

1 2 24.11 10 yrT 239

235

0.0020

0.0064


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Enrichment in 235U, a concentration of 3% - 4% is typical in

reactors used for electricity generation, but there is a trendtowards higher enrichments and greater burnup of the fuel. The

fuel is solid. For the most part it is in an oxide form, UO2, but

metallic fuel is a possibility and has been used in some reactors.

The fuel usually is in cylindrical pellets with typical dimensions

on a centimeter scale, but some designs for future reactors are

based on fuel in submillimeter microspheres embedded in

graphite. The goal here is enhanced raggedness at high

temperatures.


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a

and vs U enrichment


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Moderators

1H

Widely used

H2O or D2O

2He

Not used

Gas Press.

3

He absorbs

3Li

Not used

6Li absorbs

4Be

Was used

9Be toxic,

expensive

5B

Impossible

10B absorbs

6C

Widely used

Must be pure

No use to consider Z > 6

Only D2O and C can be used with Unat


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Coolants

Functiontransfer heat

Objective: power density, temperatureLimitations: in PWR: below saturation T,

Tin=293, Tout = 315; in LMFBR, T=140; inHTGR, T=500.

After shutdown

Coolant is either gas or liquid: H2O, D2O,He, CO2,Na, Na-K, Pb, Pb-Bi.

Coolant is moderator Classification: LWR, HWR, GR, LMR


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Control materials are materials with large thermal neutron-absorption cross sections, used as controllable poisons to

adjust the level of reactivity. They serve a variety of

purposes:

To achieve intentional changes in reactor operatingconditions, including turning the reactor on and off

To compensate for changes in reactor operating conditions,

including changes in the fissile and poison content of the fuel

To provide a means for turning the reactor off rapidly, in case

of emergency

Control Materials


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Total cross section of10B, Cd and In.


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Energy variation of

absorption cross-

sections for elements

of extremely highcross-sections.


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Control Materials

Commonly used113

Cd (12.2%, 20000 b)and 10B (19.9%, 3800 b)

Control rodsCd, B (boron steel)

PWR: c. rods B4C or Cd (5%)+AgInBWR: c. rods B4C

Water solutionB

Burnable poison (B or Gd)


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World Inventory of

Reactor Types (Dec. 1994)

Type

Number Capacity (GWe) Usual

Fuel

Moderat

or

Coolant First

develOper. Cons Oper. Cons

PWR 245 39 215.7 36.8 UO2 enr H2O H2O USA

BWR 92 6 75.9 6.0 UO2

enr H2O H

2O USA

PHWR 34 16 18.6 7.9 UO2 nat. D2O D2O Canada

LGR

(RBMK)

15 1 14.8 0.9 UO2 enr C H2O USA/

USRR

GCR 35 0 11.7 0 U, UO2 C CO2 UK

LMFBR 3 4 0.9 2.4 UO2+

PuO2

None Liq. Na Various

Total 424 66 338 54


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PWR BWR HTGR LMFBR GCFR PHW

MWt3400

(2700)

3600

(2900)3000 2400 2500 1600

MWe 1150 1200 1170 1000 1000 500

Eff 33.7 33.5 39.0 39.0 39.5 31.0Fuel UO2 UO2 UC,ThO2 PuO2, UO2 PuO2, UO2 UO2

Cool H2O H2O He Na He D2O

Moder. H2O H

2O Graphite D

2O

Height 366 376 634 91 148 410

Diam. 377 366 844 222 270 680

Burnup 33,000 27,500 98,000 100,000 100,000 10,000

Typical Data


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PWR. The pressurized water reactor accounts for almost two-

thirds of all capacity and is the only LWR used in some

countries, for example, France, the former Soviet Union, and

South Korea.

BWR. The boiling water reactor is a major alternative to the

PWR, and both are used in, for example, Sweden, the United

States and Japan.PHWR. The pressurized heavy water reactor uses heavy water

for both the coolant and moderator and operates with natural

uranium fuel. It has been developed in Canada and is commonly

referred to as the CANDU. CANDU units are also in operationin India and are being built in Romania and South Korea.


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CANDU reactors


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LGR (RBMK). The light-water-cooled, graphite-moderated

reactor uses water as a coolant and graphite for moderation. The

worlds only currently operating LGRs are the RBMK reactors in

the former Soviet Union. There were four such units at theChernobyl plant at the time of the accident there, two of which

are still in operation.

Ignalina


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GCR. The gas-cooled, graphite-moderated reactor uses a CO2

coolant and a graphite moderator. Its use is largely limited to the

United Kingdom; it is sometimes known as the Magnox reactor. A

larger second-generation version is the advanced gas-cooled,

graphite- moderated reactor (AGCR).


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HTGR. The high-temperature gas-cooled reactor uses helium

coolant and a graphite moderator. The only operating HTGR in the

United States (Fort St. Vrain) has been shut down, and there are no

operating HTGRs elsewhere.

HWLWR. The heavy-water-moderated, light-water-cooled reactor

is an unconventional variant of the heavy water reactor, and only

one has been in recent operation, a 148-MWe plant in Japan.


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LMFBR. The liquid-metal fast breeder reactor uses fast neutrons

and needs no moderator. A liquid metal is used as coolant, now

invariably liquid sodium. There are 2 liquid- metal reactors in

operation + 2 newly shut-down (FrancePhenix and Superphenix,

and one each in Kazakhstan- BN-600 and Russia BN-60) and

several others under construction or reconstruction, including

Monju reactor in Japan, which suffer sodium leakage accident in

1994.


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HT2005: Reactor PhysicsT13: Reactor Types 39

M j R


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Monju Reactor


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LIGHT WATER REACTORSPWRs and BWRs

The two types of LWR in use in the world are the pressurized

water reactor (PWR) and the boiling water reactor (BWR). In the

PWR, the water in the reactor vessel is maintained in liquid form

by high pressure. Steam to drive the turbine is developed in a

separate steam generator. In the BWR, steam is provided directly

from the reactor.


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PWR


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BWR


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Typical conditions in a PWR, temperatures of the cooling water

into and out of the reactor vessel are about 292o

C and 325o

C,respectively, and the pressure is about 155 bar. For the BWR,

typical inlet and outlet temperatures are 278 oC and 288oC,

respectively, and the pressure is only about 72 bar. The high

pressure in the PWR keeps the water in a condensed phase; the

lower pressure in the BWR allows boiling and generation of

steam within the reactor vessel. Neither the PWR or BWR has

an overwhelming technical advantage over the other, as

indicated by the continued widespread use of both.

The future is not clear-cut, however, and in Japan 3 BWRswere under construction at the end of 1994, compared to only 1

PWR.

C f i


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Components of a Light Water Reactor

The reactor pressure vessel is a massive cylindrical steel tank.

Typically for a PWR, it is about 12 m in height and 4.5 m in

diameter. It has thick walls, about 20 cm, and is designed towithstand pressures of up to 170 atmospheres.

A major component, or set of components, is the system for

converting the reactors heat into useful work. In the BWR, steam

is used directly from the pressure vessel to drive a turbine. This is

the step at which electricity is produced. In the PWR, primary

water from the core is pumped at high pressure through pipes

passing through a heat exchanger in the steam generators. Water

fed into the secondary side of the steam generator is converted

into steam, and this steam is used to drive a turbine. Thesecondary loop is also closed. The exhaust steam and water from

the turbine enter a condenser and are cooled in a second heat

exchanger before returning to the steam generator.


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The pressure vessel and the steam generators are contained within

a massive structure, the containment building, commonly made of

strongly reinforced concrete. In some designs, the concrete

containment is lined with steel; in others, there is a separate inner

steel containment vessel. The containment is intended to retain

activity released during accidents and is also believed capable of

protecting a reactor against external events such as an airplane

crash. In the Three Mile Island accident, the containment verysuccessfully retained the released radioactivity, although it may be

noted that the physical structure was not put fully to the test, since

there was no explosion or buildup of high pressures. At Chernobyl

there was no containment, with disastrous results. In principle, if a

reactor is sufficiently protected against accidents, a containment is

unnecessary. Nonetheless, it is widely thought to be an essential

safety feature, providing an important redundancy.


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Reactor Cores

PWR

BWR


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BWR-fuel bundle

SVEA-95


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Schematic diagram

of containment

building with

enclosed reactor

vessel and steam

generator for an

illustrative BWR.The output of the

steam generator

drives the turbines,

external to the

containment

building.


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Steam Separator in BWR

BURNERS, CONVERTERS, AND BREEDERS


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Achievement of High Conversion Ratios in Thermal Reactors

Difficulty of Reaching a Conversion Ratio of Unity with 235U

The limiting condition for a breeder reactor is that the conversionratio C be at least 1. This means that , the number of neutrons

produced for each neutron absorbed in 235U, must be 2 or more.

For thermal neutrons absorbed in 235U, =2.075. Were there no

losses, this would suffice for breeding: 1 neutron for continuing

the chain reaction, 1 neutron for production of 239Pu, and 0.08

neutrons free to be wasted. Such efficient utilization cannot be

achieved, however, because there is absorption in the moderator

and other non-fuel materials, as well as escape of neutrons from

the core. There fore, a 235U -fueled thermal breeder reactor is notpractical. Nonetheless, the production of 239Pu is significant,

increasing the overall energy output from the reactor fuel beyond

that gained from the U alone.

Th l B d


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Thermal Breeders

233 = 2.296; 235 = 1.65

Eugene Wigners group in 1945235U+232Th233U; then one uses a selfsustaining cycle 233U+232Th;

c(

232

Th) > c(

238

U) Conclusion: 233U breeders are possible

Thermal Breeders were abandoned in favorof Fast Breeders

It is conceivable: interest may revive; onlyfew breeder programs nowadays (all fast)

Hi h C i R ti


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High Conversion Ration

Motivation: Pu production; extension of fuel

resources High capture in 238U must be compensated by

smaller losses in moderator

Moderator: C instead of H2

O:

Possibility to use Unat C is less efficient as moderator more captures in

238U when slowing down

Less absorption in 12C than in H2O

Weapon-grade Pu (Graphite): Windscale, RBMK

D2O moderator has the same advantages


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Resonance Absorption

10-3 10-2 10-1 100 101 102 103 104 105 106 107

Energy (eV)

10-2

10-1

100

101

102

103

104

s

arns

238U

capture

fission

Fast Breeder Reactors


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Fission in Pu by fast (1 Mev) neutrons

No loss while slowing down

High and

Ratio c/f< 0.03 almost all absorptions result

in fission = = 3 (235

U = 2.3) Detailed analysis: = 3 = 1(F)+1(B)+1(L)

Thermalization undesireable coolant

with large A (23

Na, Pb+Bi) Fuel: UO2(80%)+PuO2(20%); DU is used


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Status of Fast Breeder Programs

The main incentive for the development of fast breeder reactors is

the extension of uranium supplies. A fast breeder economy would

extract much more energy per tonne of uranium than is obtainedfrom other reactors, e.g., the LWRs. Further, with more energy per

unit mass, it becomes economically practical to use more expensive

uranium ores, increasing the ultimate uranium resource.

the argument, it has been pointed out that a liquid-metal fast reactor

(LMR) can be used to destroy unwanted plutonium and other heavy

elements in weapons stockpiles or nuclear wastes. In this reversal of

motivation, the LMR would be used to consume unwanted

plutonium, rather than to produce plutonium as a fuel. There is

flexibility in this, because as LMR technology and facilities aredeveloped, they could be turned to either purpose.


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France has led in the development and deployment of breeder

reactors, with two completed reactors, the 233-MWe Phenix, put

into operation in 1973, and the 1200-MWe Superphenix at Creys-

Malville, which first generated electricity in 1986. The Superphenix

was shut down for almost two years beginning in May 1987 because

of leaks in the sodium- filled spent-fuel storage tank. Although it

resumed some operation in 1989, troubles recurred, and from 1989

through the summer of 1995 operation of Superphenix has beenintermittent and trouble-plagued, with long periods of shutdown.

The emphasis has been shifted from operation of Superphenix as a

breeder reactor to its use for study of the safety of sodium-cooled

fast reactors and the destruction of heavy elements in fast reactors.In early summer 1997 French government took a decision to

shutdown SuperPhenix and this decision was realized in 1998.

Russia has one LMFBR in operation (BN-60) and Kazakhstan has

l h d i h l d


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recently shutdown its BN-600 LMFBR. Japan has completed a

280-MWe prototype LMFBR - Monju, but its operation was set

back in December 1995 by a leak of the sodium coolant and a fire-

like chemical reaction of sodium with the air.The United States breeder reactor program has been marked by

indecision and opposition, with successive projects started and

abandoned. The latest apparent casualty was the main U.S. breeder-

related project of the past decade the investigation at ArgonneNational Laboratory of fast LMRs as part of the integral fast reactor

program. Advocates of this program stressed its potential to offer a

high degree of safety against reactor accidents and to destroy

nuclear wastes in an on-line process. The breeding potential was

often secondary in these arguments, and the planned LMR need nothave operated as a breeder, namely, with a conversion ratio greater

than unity. Nonetheless, the basic configuration of the system was

similar to that of a breeder reactor.

Generation IV - Initiative


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Neutron

Spectrum

Fuel

Cycle Size Applications R&D

Sodium Fast

Reactor (SFR)

Fast Closed Med to

Large

Electricity,

Actinide Mgmt.

Advanced Recycle

Lead-alloy Fast

Reactor (LFR)

Fast Closed Small

to

Large

Electricity,

Hydrogen

Production

Fuels, Materials

compatibility

Gas-Cooled Fast

Reactor (GFR)

Fast Closed Med Electricity,

Hydrogen, AM

Fuels, Materials,

SafetyVery High Temp.

Gas Reactor

(VHTR)

Thermal Open Med Electricity,

Hydrogen,

Process Heat

Fuels, Materials,

H2 production

Supercritical Water

Reactor (SCWR)

Thermal,

Fast

Open,

Closed

Large Electricity Materials, Safety

Molten Salt Reactor

(MSR)

Thermal Closed Large Electricity,

Hydrogen, AM

Fuel, Fuel

treatment,

Materials, Safety

and Reliability


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The END

Classifications of Reactors


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Classifications of ReactorsThermal Reactors and Fast Reactors

Reactors designed to operate with slow, thermalized neutrons (to

take advantage of increase of cross-sections with neutron energydecrease) are termed thermal reactors. However, it is also

possible to operate a reactor with fast neutrons, at energies in

the neighborhood of 1 MeV or higher. These reactors are called

fast-neutron reactors or just fast reactors. The only prominent

example of a fast reactor is the liquid-metal breeder reactor.

Homogeneous and Heterogeneous Reactors

All reactors used today for power generation are

HETEROGENEOUS, i.e. fuel, coolant and/or moderator arephysically different entities with non-uniform and anisotropic

composition.


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The homogeneous reactor, was defined as a reactor whose small-

scale composition is uniform and isotropic.

One variant of this was known as the aqueous homogenousreactor because the fuel was mixed with water (H2O or D2O). In

the so-called homogeneous reactor experiment, two small

reactors, called HRE-1 and HRE-2, were built at Oak Ridge

National Laboratory (ORNL) in the 1950s. For HRE-2, the fluid

was uranyl sulfate (UO2SO4) in heavy water (D2O), with the

uranium highly enriched in 235U. This program had the potential

of developing a thermal breeder reactor, but although HRE-2

operated uninterruptedly for over 100 days at 5 MWel, some

difficulties developed, and the program was dropped in favor ofalternative liquid-fuel projects.


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One alternative was the molten-salt reactor. The fluid was a

mixture of fluoride compounds, including the fissile component235UF4 and the fertile component ThF4. After initial operation,233UF4 was successfully tried as an alternative to

235UF4. Like

HRE-2, this reactor was designed to be a thermal breeder

reactor. There was initial success in the reactor operation, but a

decision was made in the 1960s to abandon development ofthermal breeders in favor of fast breeder reactors. A further

homogenous reactor approach, a liquid- metal thermal breeder

using uranium compounds in molten bismuth, was also

investigated but was abandoned without construction of even atest reactor. At present there is no prospect in sight of a major

program to develop homogenous reactors, with an exception of

molten salt accelerator-driven concept.


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Superphenix


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The END

F l


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Fuels

Few natural nuclides that can be used as reactor fuels

Uranium (Z=92). This is the main fuel in actual use, especially 235Uwhich is fissile. In addition, 238U is important in reactors, primarily

as a fertile fuel for239Pu production, and 233U could be used as a

fissile fuel, formed by neutron capture in 232Th.

Protactinium (Z=91). The longest-lived isotope of Pa (

231

Pa) has ahalf-life of 3.3 x104 yr, and therefore there is essentially no Pa in

nature. Further, there is no stable A =230 nuclide that could be used

to produce 231Pa in a reactor.

Thorium (Z=90). Thorium is found entirely as 232Th, which is not

fissile (for thermal neutrons). It can be used as a fertile fuel forproduction of fissile 233U.

Moderators


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The options for moderating materials are limited:

Hydrogen (Z=l). The isotopes 1H and 2H are widely used as

moderators, in the form of light (ordinary) water and heavy water,

respectively.

Helium (Z=2). The isotopes 3He and 4He are not used, because

helium is a gas, and excessive pressures would be required to obtain

adequate helium densities for a practical moderator. Moreover 3He

absorbs neutrons very strongly (see lecture about detectors)Lithium (Z=3). The isotope 6Li (7.5% abundant) has a large

neutron-absorption cross section, making lithium impractical as a

moderator.

Beryllium (Z=4). 9Be has been used to a limited extent as a

moderator, especially in some early reactors. It can be used in the

form of beryllium oxide, BeO. Beryllium is expensive and toxic.

( ) h l b i i i


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Boron (Z=5). The very large neutron-absorption cross section in10B (20% abundant) makes boron impossible as a moderator.

Carbon (Z=6). Carbon in the form of graphite is widely used as a

moderator. It is important that the graphite be pure, free of elementsthat have high absorption cross sections for neutrons.

There are no advantages in considering elements heavier than

carbon.

With natural uranium, it is not possible to use light water and it iscommon practice to use heavy water or graphite, which have

particularly high moderating ratios.

Coolants


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Coolants

The main function of the coolant in any generating plant is to

transfer energy from the hot fuel to the electrical turbine, either

directly or through intermediate steps. During normal reactoroperation, cooling is an intrinsic aspect of energy transfer. In a

nuclear reactor, cooling has a special importance, because

radioactive decay causes continued heat production even after the

reactor is shut down and electricity generation has stopped. It is still

essential to maintain cooling to avoid melting the reactor core, and

in some types of reactor accidents (e.g., the accident at Three Mile

Island) cooling is the critical issue. The coolant can be either a

liquid or a gas. For thermal reactors, the most common coolants are

light water, heavy water, helium, and carbon dioxide. The type ofcoolant is commonly used to designate the type of reactor. Hence,

the characterization of reactors as light water reactors (LWRs),

heavy water reactors (HWRs), and gas-cooled reactors (GCRs).


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The coolant may also serve as a moderator, as is the case for

LWRs and HWRs. In gas-cooled reactors, the density of the

coolant is too low to permit it to serve as a moderator, and

graphite is used. Fast reactors, in which fission is to occur

without moderation, require a coolant that has a relatively high

atomic mass number (A). Generically, these reactors are termed

liquid-metal reactors, because the coolant is a liquid metal,

most commonly sodium (A =23).

T l d t l t i l d i d b Th


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Two commonly used control materials are cadmium and boron. The

isotope 113Cd, which is 12.2% abundant in natural cadmium, has a

thermalneutron capture cross section of 20,000 b. The isotope

10B, 19.9% abundant in natural boron, has a thermal absorptioncross section of 3800 b. Cadmium and boron are effective materials

for reducing the reactivity of the reactor.

Cadmium is commonly used in the form of control rods, which can

be inserted or withdrawn as needed. Boron can be used in a controlrod, made, for example, of boron steel, or it can be introduced in

soluble form in the water of water-cooled reactors to regulate

reactor operation. It is also possible to use other materials. Control

rods for pressurized water reactors (PWRs) commonly use boron in

the form of boron carbide (B4C) or cadmium in a silverindium

alloy containing 5% cadmium. Control rods for boiling water

reactors (BWRs) commonly use boron carbide.

A sophisticated control variant is the so called burnable poison


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A sophisticated control variant is the so-called burnable poison,

incorporated into the fuel itself. During the three years or so that

the fuel is in the reactor, the 235U or other fissile material in the

fuel is partially consumed, and fission product poisons areproduced. The burnable poison helps to even the performance of

the fuel over this time period. It is a material with a large

absorption cross section, which is consumed by neutron capture as

the fuel is used. The decrease in burnable poison content with

time can be tailored to compensate for the decrease in 235U and the

buildup of fission fragment poisons. Boron and gadolinium can be

used as burnable poisons.

Control materials are needed to regulate reactor operation and

provide a means for rapid shutdown. Boron and cadmium areparticularly good control materials because of their high cross

sections for the absorption of thermal neutrons. These control

materials are usually used in the form of rods.


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Potential of233U for a Thermal Breeder Reactor


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The number of neutrons produced is significantly higher for 233U

(=2.296) than for235U. In early 1945 Eugene Wigners group in

Chicago envisaged a cycle in which 233U is produced initially in a

reactor with 235U as the fissile fuel and 232Th as the fertile fuel.

Subsequently, a 233U 232Th cycle could in principle be self-

sustaining. Not only is n higher for 233U than for 235U, but the

capture cross section is significantly higher for232Th than for238U

at thermal energies, making the conversion ratio higher than in the235U- 238U cycle.

However, while thermal breeders based on 233U are in principle

possible, thermal breeders were abandoned in favor of the fast

breeder reactor. It is conceivable that interest in thermal breederscould revive, but at present the relatively few breeder reactor

development programs in the world are all based on fast breeders.

High Conversion Ratios without Breeding


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g g

If breeding is not achieved with thermal reactors, a high conversion

ratio can still be desirable. One motivation could be plutonium

production. Another motivation is the extension of fuel resources.As the conversion ratio increases, the energy output increases for a

given original 235U content. A high conversion ratio means a high

ratio of capture in 238U to absorption in 235U. This must be

accomplished without losing criticality. Greater losses of neutrons

to 238U can be compensated for by smaller losses in the moderator.The use of carbon instead of light water as a moderator is favorable

on two counts, if a high conversion ratio is desired (in addition to

the advantage that with a carbon moderator it is possible to use

natural uranium). Because carbon is a less effective moderator thanwater, more collisions are required to reach thermal energies, and

therefore there is more possibility of neutron capture in 238U at

intermediate energies.

Further, because of the low neutron-capture cross section in 12C, the


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loss of thermal neutrons to absorption will be less for carbon than for

light water. Together, this means a higher conversion ratio. It may be

noted that reactors designed with production of weapons-grade

plutonium in mind, as either the main or an auxiliary function, have

been mostly graphite- moderated. Examples include the Windscale

plant in England, the plutonium production reactors at Hanford, and

the RBMK reactors built in the USSR.

The conversion ratio is higher in a heavy water reactor than in a lightwater reactor for much the same reasons as for a graphite moderator.

Again, heavy-water is a less effective moderator than light water, and

it has a lower capture cross section for thermal neutrons. For both

graphite-moderated and heavy water reactors, there have beensuggestions that the 232Th 233U cycle be used, to further increase

conversion and extend the life of the uranium fuel, even without

breeding.


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To avoid thermalization of the neutrons, fast breeder reactors use a

coolant with high mass number A to reduce moderation. Liquid

metals have the best combination of high A and favorable heat-

transfer properties, and the fast breeder reactors in actual use or

under construction are liquid-metal fast breeder reactors (LMFBR).

The standard choice for the coolant is liquid sodium ( 23Na).

The fuel is made of pellets of mixed plutonium and uranium oxides,

PuO2 (about 20%) and UO2 (about 80%). Uranium depleted in 235Uis commonly used, it being available as a residue from earlier

enrichment. The fission cross section for Pu is between 1.5 and 2.0 b

over virtually the entire fast-neutron region (from 0.01 to 6 MeV),

and therefore fission in239

Pu is much more important than fission in238U. The most probable fast-neutron reactions in U are inelastic

scattering, which produces lower energy neutrons, and capture,

which produces 239Pu.



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Plutonium as Fuel for Fast Breeders

The fast breeder reactor relies on a chain reaction in which the

neutrons are not thermalized but instead produce fission at relativelyhigh energies. If239Pu is the fissile fuel, this has the advantage of a

relatively large fission cross section sf and a relatively large number

of neutrons per fission event (high n and ). For 1.0-MeV neutrons

on 239Pu, sf =1.7 b, and the ratio (a) of the capture cross section to

the fission cross section is less than 0.03. The low value of this ratio

means that almost all absorption in 239Pu leads to fission (=n=3.0).

In contrast, the fission cross section in 235U is 1.2 b at 1 MeV, and

is only 2.3. With about three neutrons per fission, breeding with

239Pu is obtained if 1 neutron is used for continuing the chainreaction, 1+ for breeding, and 1- for losses. It has been shown that it

is quite possible to achieve this.

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13_ReactorTypes