HT2005: Reactor Physics T13: Reactor Types1 Types of Nuclear Reactors Generations of Reactors Principles of Classification Survey of Reactor Types Light

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


Reactor Generations

1942-Dec-2 Chicago Pile-1


Principles of Classification

• Energy spectrum utilized for fission

• Prime purpose

• Fuel

• Coolant

• Coolant system

• Moderator

• Heterogeneity (?)


Heterogeneity• All reactors used today for power generation are

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

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

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

• Molten salt reactor, 235UF4+ThF4, 233UF4. In 1960’s → 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 to

develop homogenous reactors• molten salt accelerator-driven concept (?)


Homogeneous Reactor Experiment


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


Purpose

Power Ship propulsion Production Research Specialized

LWR PHWR (CANDU) HTGRAGRLMRLGR (RBMK) LMFBRHWLWR OLR

Submarines Navy Aircarrier Icebreakers PWR-cargo

Plutonium Tritium

WPR TRIGA Studsvik

Isolated areas Space propul. Heat for chem


Fuel

UO2 (3-4%)sintered pellets zirconium alloy

Nat. Upelletstubes

UO2

spherical part.carbon, silicon

U-Pu-Zr alloy pellets

steel cladding

U-Al alloyplates

Al cladding + Al-Si

LWR CANDU HTGR LMR TRIGA


Coolant

Light WaterH2O

Heavy Water D2O

Gas air, CO2, He

Liquid MetalNa, Na-K, Pb-Bi

LWR CANDU HTGRPebble bed

LMR


Moderator

Light WaterH2O

Heavy Water D2O

GraphiteC

LWR CANDU LGR (RBMK)GCR

HTGR


1-loop BWR

2-loop PWR

3-loop LMR


Commonly Accepted Classification

PWRBWR

German

USA

Na

Pb-Bi

NuclearReactors

PowerReactors

SpecializedReactors

ResearchReactors

Ship PropulsionReactors

ProductionReactors

CANDU AGR Other

PWR Fast

LMR

HTGR LWR

Unat

C+H2OUnat

C+Gas

WPR

TRIGA


• Fuel• Coolant• Moderator• Control system (rods)

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

Components of Conventional Reactors


90U (2 ppm)4.5∙109 yr

90Th (6 ppm)14.05∙109 yr

238 23992 94U Pu

99.5%

23592 U

0.7%

23.5min 2.3day238 1 239 239 23992 0 92 93 94U U Np Pun

23.3min 27.4day232 1 233 233 23390 0 90 91 92Th Th Pa Un

Fuels

Natural Elements

233U7∙108 yr

Artificial Nuclides

239Pu24∙103 yr

235

233

239

0.0064

0.0026

0.0020


Between thorium (Z=90) and bismuth (Z=83), the isotope with the longest half-life is 226Ra (T½=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 and 239Pu 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.


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

(b

arns

)

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 +/- 0.169064Lepton => 0.0130345 +/- 0.00112065Photon => 0.00124304 +/- 0.000114618* DECAY MODES:Alpha decay, possibly followed by Gamma emissionQ_value => 4.081 MeV +/- 0.004 MeV Branching => 1 +/- 0Fission decay, followed by Gamma emissionQ_value => 169.4 MeV 4.1 MeV Branching => 1.4e-11 5e-12* NUCLIDE MAT. NUMBER: 9061The following number of lines are present:Gamma: 2 Alpha: 3 Electron: 8 X-rays: 7

23.3min 27.4day232 1 233 233 23390 0 90 91 92Th Th Pa Un

232Th

91 2 14.05 10 yrT


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

(b

arns

)

233 U

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 +/- 0.0181258Lepton => 0.00759652 +/- 0.000610716Photon => 0.00122537 +/- 0.000110582* DECAY MODES:Alpha decay, possibly followed by Gamma emissionQ_value => 4.9089 MeV +/- 0.0012 MeV Branching => 1 0* NUCLIDE MAT. NUMBER: 9234The following number of lines are present:Gamma: 116 Alpha: 37 Electron: 235 X-rays: 7

233U

31 2 159.25 10 yrT

233

235

0.0026 1 1.0013

0.0064 1 1.0032

k

k


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

(b

arns

)

235 U

capture

fission

T 1/2 => 703.81 My +/- 500.01 ky [2.221E+16s +/- 1.578E+13s]SPIN => 7/2 -* MEAN DECAY ENERGIES (MeV)Baryon => 4.46298 +/- 0.287321Lepton => 0.0475365 +/- 0.00378036Photon => 0.167808 +/- 0.00347714* DECAY MODES:Alpha decay, possibly followed by Gamma emissionQ_value => 4.679 MeV +/- 0.0025 MeV Branching => 1 0Fission decay, followed by Gamma emissionQ_value => 176.4 MeV +/- 4.6 MeV Branching => 2e-10 +/- 1e-10* NUCLIDE MAT. NUMBER: 9240The following number of lines are present:Gamma: 49 Alpha: 17 Electron: 127 X-rays: 7

235U

61 2 703.81 10 yrT

235

233

239

0.0064

0.0026

0.0020


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

(b

arns

)

238 U

capture

fission

T 1/2 => 4.47 Gy +/- 5.00 My [1.410E+17s +/- 1.578E+14s]SPIN => 0 +* MEAN DECAY ENERGIES (MeV)Baryon => 4.25996 +/- 0.236186Lepton => 0.0105454 +/- 0.000870616Photon => 0.00125402 +/- 0.000119247* DECAY MODES:Alpha decay, possibly followed by Gamma emissionQ_value => 4.2703 MeV +/- 0.0039 MeV Branching => 0.999999Fission decay, followed by Gamma emissionQ_value => 173.6 MeV +/- 3.4 MeV Branching => 5.4e-07 +/- 2e-08* NUCLIDE MAT. NUMBER: 9249The following number of lines are present:Gamma: 2 Alpha: 3 Electron: 6 X-rays: 2

91 2 4.47 10 yrT

238U

mg of U3 3ppb

tonne Seawater


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

(b

arns

)

239 Pu

capture

fissionT 1/2 => 24.11 ky +/- 40.00 y [7.609E+11s +/- 1.262E+09s]SPIN => 1/2 +* MEAN DECAY ENERGIES (MeV)Baryon => 5.23678 +/- 0.0389047Lepton => 0.00738587 +/- 0.000585825Photon => 0.000707559 +/- 6.46626e-05* DECAY MODES:Alpha decay, possibly followed by Gamma emissionQ_value => 5.2435 MeV +/- 0.0007 MeV

Branching => 0.000115 +/- 2e-06Alpha decay, possibly followed by Gamma emissionQ_value => 5.2434 MeV +/- 0.0007 MeV

Branching => 0.999885 +/- 2e-06Fission decay, followed by Gamma emissionQ_value => 184.3 MeV +/- 3.6 MeV

Branching => 4.4e-12 +/- 4e-13* NUCLIDE MAT. NUMBER: 9440The following number of lines are present:Gamma: 188 Alpha: 69 Electron: 468 X-rays: 7

239Pu

31 2 24.11 10 yrT

239

235

0.0020

0.0064


Enrichment in 235U, a concentration of 3% - 4% is typical in reactors used for electricity generation, but there is a trend towards 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.


σa and vs U enrichment


Moderators

1H

Widely used

H2O or D2O

2He

Not used

Gas → Press.

3He 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


Coolants• Function – transfer heat

– Objective: power density, temperature– Limitations: in PWR: below saturation T,

Tin=293, Tout = 315; in LMFBR, ΔT=140; in HTGR, ΔT=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


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 operating conditions, 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


Total cross section of 10B, Cd and In.


Energy variation of absorption cross-sections for elements of extremely high cross-sections.


Control Materials

• Commonly used 113Cd (12.2%, 20000 b) and 10B (19.9%, 3800 b)

• Control rods – Cd, B (boron steel)– PWR: c. rods B4C or Cd (5%)+AgIn

– BWR: c. rods B4C

• Water solution – B

• Burnable poison (B or Gd)


World Inventory of Reactor Types (Dec. 1994)

Type Number Capacity (GWe) Usual Fuel

Moderator

Coolant First devel

Oper. 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 H2O 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


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 31.0

Fuel UO2 UO2 UC,ThO2 PuO2, UO2 PuO2, UO2 UO2

Cool H2O H2O He Na He D2O

Moder. H2O H2O Graphite – – D2O

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



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 operation in India and are being built in Romania and South Korea.


CANDU reactors


LGR (RBMK). The light-water-cooled, graphite-moderated reactor uses water as a coolant and graphite for moderation. The world’s only currently operating LGRs are the RBMK reactors in the former Soviet Union. There were four such units at the Chernobyl plant at the time of the accident there, two of which are still in operation.

Ignalina



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).


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.


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 (France – Phenix 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.



Monju Reactor


LIGHT WATER REACTORS PWRs 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.


PWR


BWR


Typical conditions in a PWR, temperatures of the cooling water into and out of the reactor vessel are about 292oC and 325oC, 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 BWRs were under construction at the end of 1994, compared to only 1 PWR.


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 to withstand pressures of up to 170 atmospheres. A major component, or set of components, is the system for converting the reactor’s 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. The secondary 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.


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 very successfully 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.


Reactor CoresPWR

BWR


BWR-fuel bundleSVEA-95



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.


Some CAD-pictures of BWRs (FO3)


Steam Separator in BWR


BURNERS, CONVERTERS, AND BREEDERS

Achievement of High Conversion Ratios in Thermal ReactorsDifficulty of Reaching a Conversion Ratio of Unity with 235UThe limiting condition for a breeder reactor is that the conversion ratio C be at least 1. This means that η, 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 not practical. Nonetheless, the production of 239Pu is significant, increasing the overall energy output from the reactor fuel beyond that gained from the U alone.


Thermal Breeders• η233 = 2.296; η235 = 1.65• Eugene Wigner’s group in 1945

235U+232Th→233U; then one uses a self sustaining cycle 233U+232Th;σc(232Th) > σc(238U)

• Conclusion: 233U breeders are possible• Thermal Breeders were abandoned in favor

of Fast Breeders• It is conceivable: interest may revive; only

few breeder programs nowadays (all fast)


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 H2O:

– 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


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

(bar

ns)

238 U

capture

fission


Fast Breeder Reactors• Fission in Pu by fast (1 Mev) neutrons

– No loss while slowing down– High ν and η

– Ratio σc/σf < 0.03 almost all absorptions result in fission ν = η = 3 (235U η = 2.3)

• Detailed analysis: η = 3 = 1(F)+1(B)+1(L)

• Thermalization undesireable → coolant with large A (23Na, Pb+Bi)

• Fuel: UO2(80%)+PuO2(20%); DU is used


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 obtained from 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 are developed, they could be turned to either purpose.


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 been intermittent 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 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 Argonne National 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 not have 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 - InitiativeNeutron Spectrum

FuelCycle Size Applications R&D

Sodium Fast Reactor (SFR)

Fast Closed Med toLarge

Electricity,Actinide Mgmt.

Advanced Recycle

Lead-alloy Fast Reactor (LFR)

Fast Closed Small toLarge

Electricity,Hydrogen Production

Fuels, Materials compatibility

Gas-Cooled FastReactor (GFR)

Fast Closed Med Electricity,Hydrogen, AM

Fuels, Materials,Safety

Very 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


The END


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 energy decrease) 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 are physically different entities with non-uniform and anisotropic composition.


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 homogenous reactor 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 of alternative liquid-fuel projects.


One alternative was the molten-salt reactor. The fluid was a mixture of fluoride compounds, including the fissile component 235UF4 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 of thermal 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 a test 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.



Superphenix


The END


Fuels

Few ”natural” nuclides that can be used as reactor fuelsUranium (Z=92). This is the main fuel in actual use, especially 235U which is fissile. In addition, 238U is important in reactors, primarily as a fertile fuel for 239Pu 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 ( 231Pa) has a half-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 for production of fissile 233U.


Moderators 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.


Boron (Z=5). The very large neutron-absorption cross section in 10B (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 elements that 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 is common practice to use heavy water or graphite, which have particularly high moderating ratios.


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 reactor operation, 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 of coolant 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).


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).


Two commonly used control materials are cadmium and boron. The isotope 113Cd, which is 12.2% abundant in natural cadmium, has a thermal – neutron capture cross section of 20,000 b. The isotope 10B, 19.9% abundant in natural boron, has a thermal absorption cross 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 control rod, 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 silver – indium 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, 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 are produced. 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 are particularly 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.



Potential of 233U for a Thermal Breeder Reactor The number of neutrons produced is significantly higher for 233U (=2.296) than for 235U. In early 1945 Eugene Wigner’s 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 higher for 233U than for 235U, but the capture cross section is significantly higher for 232Th than for 238U at thermal energies, making the conversion ratio higher than in the 235U- 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 breeders could 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 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 than water, 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 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 light water 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 been suggestions that the 232Th – 233U cycle be used, to further increase conversion and extend the life of the uranium fuel, even without breeding.


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 235U is 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 in 239Pu is much more important than fission in 238U. The most probable fast-neutron reactions in U are inelastic scattering, which produces lower energy neutrons, and capture, which produces 239Pu.


Fast Breeder ReactorsPlutonium 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 relatively high energies. If 239Pu 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 and ). For 1.0-MeV neutrons on Pu, f =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 (==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 chain reaction, 1+ for breeding, and 1- for losses. It has been shown that it is quite possible to achieve this.

Documents

HT2005: Reactor Physics T13: Reactor Types1 Types of Nuclear Reactors Generations of Reactors Principles of Classification Survey of Reactor Types Light