Nuclear Power Reactors

P1: GPA Final Pages Qu: 00, 00, 00, 00

Encyclopedia of Physical Science and Technology EN010M-495 July 18, 2001 2:59

Nuclear Power ReactorsRonald A. KniefXE Corporation

I. OverviewII. Design and Operating Requirements

III. Reactor TypesIV. Safety FeaturesV. Regulations

GLOSSARY

Blanket Region surrounding the fuel core of a breederreactor that contains fertile material to increase pro-duction of new fuel.

Breeder Reactor that produces new fuel from fertile ma-terial at a faster rate than it burns fuel for energyproduction.

Converter Reactor that produces less new fuel from fer-tile material than it burns for energy production.

Coolant Liquid or gaseous medium used to remove fis-sion heat energy from reactor fuel.

Core Region within a reactor occupied by the nuclear fuelthat supports the fission chain reaction.

Critical Condition where a fission chain reaction is stablewith neutron production balancing losses at a nonzerolevel.

Fast neutrons Neutrons of high energy, particularly thoseproduced directly by the fission reaction.

Fertile Material, not itself fissile, capable of being con-verted to fissile material following absorption of aneutron.

Fissile Material capable of sustaining a fission chainreaction.

Fission Process in which a heavy nucleus splits into

two or more large fragments and releases kineticenergy.

Moderator Material of low atomic mass included ina reactor for the purpose of reducing the energy ofneutrons.

Multiplication Ratio of neutron production rate to neu-tron loss rate; value is unity for a critical system.

Reactivity Fractional change in neutron multiplicationreferenced to the critical condition; value is zero fora critical system.

Reactor Combination of fissile and other materials in ageometric arrangement designed to support a neutronchain reaction.

Steam cycle Method used to convert fission heat energy tosteam and hence electricity [often described in terms ofa primary coolant loop and, as appropriate, secondaryheat transfer loop(s)].

Thermal neutrons Low-energy neutrons in thermalequilibrium with their surroundings, produced by slow-ing down or moderating the fast neutrons from nuclearreactions such as fission.

COMMERCIAL POWER production with nuclear en-ergy relies on a sustained neutron chain reaction from the

739

P1: GPA Final Pages


740 Nuclear Power Reactors

fission process. Reactors produce electricity from fission,employing a variety of fuel forms, coolants, moderators,and other materials.

I. OVERVIEW

Nuclear power reactors have many similarities to conven-tional fossil-powered systems. All of their unique designfeatures and operating modes result directly or indirectlyfrom the nature of the fission chain reaction that producesthe energy. These characteristics lead to several classifica-tions appropriate for reactors. (The other major reactionfor nuclear energy production, fusion, offers the prospectas a future energy source.)

A. Fission Process

When a neutron strikes a nucleus of 235U, a fission reactionmay occur in which the nucleus splits into two or morefission fragments, releases radiation and kinetic energy,and emits neutrons. The energy release, over 50 milliontimes as great as from the reaction involved in “burning”a carbon atom with oxygen, is one major advantage offission as an energy source. Another is the presence ofextra neutrons, which present the possibility of a sustainedchain reaction and steady energy production.

The disadvantages of the fission reaction are the partic-ulate and electromagnetic radiations emitted at the timeof fission and the radioactivity (i.e., emission of radia-tions over time) of the fission fragments and their prod-ucts. These features lead to requirements for shielding andcontainment, respectively.

When the chain reaction exactly balances the rates ofneutron production from fission with absorption and leak-age, the system is steady and said to be critical. Whenproduction exceeds losses, it is supercritical and increasesin power. When losses exceed production, it is subcriti-cal and decreases in power, up to and including being shutdown. All three states of criticality are necessary to nuclearpower reactor operation. This status is often quantified interms of the multiplication factor k, defined as

k = production

absorption + leakage,

or by reactivity ρ defined as

ρ = (k − 1)/k.

Thus, k = 1 or ρ = 0 constitutes the critical condition.A material capable of sustaining a chain reaction by

itself is said to be fissile. Alternatively, fissile materialcan be fissioned by neutrons of any energy. Fissionableand fertile materials can contribute to the chain reaction.

Nuclei that are fissionable can be fissioned by neutrons, butnot necessarily neutrons of any energy (particularly, somecannot be fissioned by low-energy or thermal neutrons).Fertile materials on absorbing a neutron are converted tofissile nuclei.

B. Reactor Classifications

Nuclear reactors are designed to achieve a self-sustainedchain reaction with a combination of fissile, fertile, andother materials. Common characteristics useful for clas-sification purposes are

1. Coolant—principle heat removal medium.2. Steam cycle—number of separate coolant “loops.”3. Moderator—material (if any) used to “slow down”

the neutrons produced by fission.4. Neutron energy—general energy range for the neu-

trons that cause most of the fissions.5. Fuel production—system is referred to as a breeder

if it produces (i.e., changes from fertile to fissile) more fuelthan it consumes; it is said to be a converter otherwise.

The first two features relate to the current practice of con-verting fission energy first to heat and then to electricalenergy by employing a steam cycle. Coolants include wa-ter, heavy water, gases, and liquid metal. The steam cyclesmay employ from one to three separate loops, includingone for primary coolant circulation and one (not necessar-ily separate) for steam generation.

Neutrons are emitted from fission at high energy. How-ever, very-low-energy neutrons have a higher likelihood ofcausing additional fission reactions. Thus, many systemsemploy a moderator to “slow down” these neutrons. Thebest moderators are of low mass, allowing maximum en-ergy transfer through neutron collisions (e.g., the limitingcase of potentially total energy transfer between a mov-ing cue ball and a stationary billiard ball of equal mass).Typical materials used for this purpose are hydrogen, deu-terium (heavy hydrogen), and carbon. The moderator andcoolant may be the same (e.g., water) or may be separatematerials (e.g., gaseous coolant and solid graphite mod-erator). Neutrons with low enough energies to be roughlyin thermal equilibrium with the surrounding materials aresaid to be thermal neutrons. Neutrons at or near fissionenergies are fast neutrons. Fast reactors avoid the use ofmoderators, such as with a metal coolant like sodium,instead of one of the moderating materials identifiedabove.

Any reactor that contains fertile species 232Th, 238U, or240Pu produces some amount of new fissile fuel. Breederreactors actually produce more fuel than they consume.Converter reactors produce lesser amounts of new fuel.

P1: GPA Final Pages


Nuclear Power Reactors 741

The world’s six major reactor types are

1. boiling water reactor (BWR)2. pressurized water reactor (PWR), including several

similar western designs and the unique Russian VVERPWR

3. heavy-water-moderated reactor (HWR), includingthe pressurized heavy-water reactor (PHWR)

4. gas-cooled reactor (GCR), including the high-temperature gas-cooled reactor (HTGR)

5. light-water cooled graphite moderated reactor(LGR), including the Russian RBMK pressure-tubegraphite reactor (PTGR)

6. breeder reactor, including the liquid-metal fast-breeder reactor (LMFBR)

Data for representative nuclear steam supply systems(NSSS) (i.e., the portions related specifically to the use ofnuclear fission as the energy source) for the six of these re-actor types (including two PWR—a Westinghouse systemrepresentative of the western units and a Russian VVER)are provided in Table I. The section labeled “general” de-scribes the reactor types in terms of the five classificationsidentified at the beginning of this section (e.g., the PTGRis a single-loop, light-water-cooled, graphite-moderated,thermal, converter reactor.)

The world-wide nuclear electric generating capacity foreach country by reactor type is shown in Table II.

II. DESIGN AND OPERATINGREQUIREMENTS

Nuclear power reactors are complex systems whose de-sign represents a balance among conflicting requirements.Principal among these requirements are nuclear design,materials, thermal hydraulics, economics, and control andsafety.

The nuclear design seeks to match fissile and fertileconstituents with appropriate coolants and moderator (ifany) to optimize the neutron economy of the chain reac-tion and production of new fuel. Materials concerns focuson chemical compatibility of components, thermal andradiation stability, and overall mechanical strength. Oneespecially important requirement is that the fuel main-tain its structural integrity throughout 4 years or more ofin-place fission chain reaction, since unlike other energyproduction cycles, the fuel is not literally “burned up.”

Thermal-hydraulic goals include spatially uniformpower density distributions and appropriate match ofcoolant conditions to energy generation. Economics fo-cus on minimizing overall costs (i.e., initial capital out-lay, operating and maintenance costs, and fuel charges),

including attention to reliability and thermal conversionefficiency.

Control and safety considerations include some inter-action with each of the previous areas. Power reactorsmust maintain the critical condition, increase and decreasepower, and adjust to long-term changes such as the con-flicting effects from breeding new fuel, depleting existingfuel, and building in waste products. The desired neutronbalance is maintained predominately by adjusting neutronabsorption, (by using materials designed to remove neu-trons from, or “poison,” the chain reaction), although somedesigns also change neutron production by on-line fuel ex-change. Absorption may depend on a combination of solidmoveable control rods, soluble poisons in the coolant ormoderator, and fixed burnable poisons designed to depleteor be “burned out” by the continuing neutron population.

Routine control strives to make the power density asuniform as possible, while allowing for power changes. Inmost designs, control-rod movement is used with groupsselected for symmetry to maintain uniform power distribu-tion. Measures are instituted to restrict the speed of move-ment and reactivity worth of individual rods or groupsof rods to prevent excessively rapid power increase. Sim-ilarly, the design intends to minimize the likelihood ofinadvertent control rod withdrawal.

Safety concerns are addressed through a protective sys-tem whereby the control rods may be inserted quickly; thatis, they scram or trip through gravity drop or gas pressure,when certain predetermined parameter limits (e.g., onpressure, temperature, flow, or power levels) are exceeded.Overall design with negative feedback mechanisms, sothat power increases tend to be self-terminating, is anotherimportant goal. Fuel temperature and coolant/moderatortemperature effects are examples when a power increasedrives up temperatures and the temperatures in turn causethe reaction to slow somewhat.

Another important safety feature is multiple-barriercontainment of fission products. As may be observed foreach reactor type described in the remainder of this arti-cle, these barriers include the fuel particles, surroundingcladding, the coolant system boundary, and a containmentstructure.

One important example of tradeoffs among the designgoals is seen in thermal-reactor fuel assemblies whose pinarrangement determines the characteristics of the chain re-action, economics, and heat removal. The chain reactionis enhanced by optimum spacing of the fuel in “lumps”with moderator interspersed so that neutrons from fissionwill undergo a number of scattering collisions for slowingdown prior to reentering the fuel; too little and too muchspacing can both be detrimental. The extent of slowingdown also determines the amount of conversion of fer-tile material and the overall energy production possible

P1: GPA Final Pages



TABLE I Characteristics for Seven Representative Nuclear Steam Supply Systemsa

VVERBWR PWR PWR PHWR HTGRb PTGR LMFBR

Reference design

Manufacturer General Electric Westinghouse (Former Soviet Atomic Energy of General Atomic (Former Soviet NovatomeUnion) Canada, Ltd. Union)

System (station) BWR/6 (Sequoyah/ VVER-1000 CANDU-600 (Fulton) RBMK-1000 (Superphenix)SNUPPS)

General

Steam cycle

Loops 1 2 2 2 2 1 3

Primary coolant H2O H2O H2O D2O He H2O Liquid Na

Secondary — H2O H2O H2O H2O — Liquid Na/H2Ocoolant

Moderator H2O H2O H2O D2O Graphite Graphite —

Neutron energy Thermal Thermal Thermal Thermal Thermal Thermal Fast

Fuel production Converter Converter Converter Converter Converter Converter Breeder

Energy conversion

Gross thermal 3579 3411 3200 2180 3000 3200 3000power, MW(th)

Net electrical 1178 1150 953 638 1160 1000 1200power, MW(e)

Efficiency, % 32.9 33.7 33.3 29.3 38.7 31.2 40

Heat transport

Primary loops and 2 4 4 2 6 2/8 4pumps

Intermediate loops — — 4 — — — 8

Steam generators — 4 — 4 6 — 8

Steam gen. type — ∪-tube Horizontal ∪-tube Helical coil — Helical coil

Fuel

Particles Short, cyl. Short, cyl. Short, cyl. Short, cyl. Coated Short, cyl. Short, cyl.pellets pellets pellets pellets micro-spheres pellets pellets

Chemical form UO2 UO2 UO2 UO2 UC/ThC UO2 Mixed UO2/PUO2

Fissile 2–5 wt. % 235U 2–5 wt. % 235U 2–5 wt. % 235U Natural uranium 93 wt. 235U % 1.1–2.4 wt.% 235U 15–18 wt. % Pu(core)

Fertile 238U 238U 238U 238U Thorium 238U 238U (core +blanket)

Pins Pellet stacks in Pellet stacks in Pellet stacks in Pellet stacks in Microspheres in Pellet stacks in Pellet stacks inZr–alloy tubes Zr–alloy tubes Zr–alloy tubes Zr–alloy tubes graphite sticks Zr–alloy tubes stainless steel

tubes

Assembly 8 × 8 Square 17 × 17 Square 331 Hexagonal 37-Pin Hexagonal 2 × 18 pin 271-Pin hexagonalpin array pin array pin array cylindrical array graphite cylindrical array array

blockb

Core

Axis Vertical Vertical Vertical Horizontal Vertical Vertical Vertical

Assemblies on 1 1 1 12 8 2 1axis

Assemblies 748 193 151 380 493 1661 364 (core), 233radially (blanket)

Performance

Equil. burnup, 27,500 27,500 25–41,000 7500 95,000 18,500 100,000MWD/T

Refueling sequence 14 /yr 1

3 /yrc — Continuous, online 14 /yrb on-line Variable

Thermal hydraulics

Primary system

Pressure, MPa 7.17 15.5 16.5 10.0 4.90 7.2 0.1

Inlet temp., ◦C 278 292 290 267 318 270 395

Average outlet 288 325 322 310 741 284 545temp., ◦C

continues

P1: GPA Final Pages



TABLE I (Continued)

VVERBWR PWR PWR PHWR HTGRb PTGR LMFBR

Core flow, Mg/sec 13.1 18.0 21.1 7.6 1.42 10.4 16.4

Volume, 1 — 3.36 × 105 1.20 × 105 (9550 kg) (3200 Mg)

Secondary system Na/H2O

Pressure, MPa — 6.89 6.4 4.7 17.2 — 0.1/17.7

Inlet temp., ◦C — 227 289 187 188 — 345/235

Outlet temp., ◦C — 285 322 260 513 — 525/487

Power density

Core ave., kW/l 54.1 105 111 12 8.4 280

Fuel ave., kW/l 54.1 105 60 44 280

Linear heat rate

Core ave., kW/m 19.0 17.8 17.6 25.7 7.87 29

Core max., kW/m 44.0 42.7 44.1 23.0 29 45

Design peaking factors

Radial 1.4 1.21

(Total) (2.5) (2.9) (1.55)

Axial 1.6 1.41

Moderator Same as Same as Same as D2O Graphite blocksb Graphite —primary primary primarycoolant coolant coolant

Volume, 1 2.17 × 106

Inlet temp., ◦C 43

Outlet temp., ◦C 71

Reactivity control

Control rods

Geometry Cruciform Rod clusters Rods Rods Rod pairs Rods Hexagonal pinbundles

Absorber material B4C Ag–In–Cd Boron Various B4C/graphite B4C

Burnable poison Gd in fuel pellets Borosilicate B–Zr — B4C/graphite —glass

Other systems Voids in coolant Soluble boron H2O/Various Reserve shutdown 3-Bundle secondary

Reactor vessel

Inside dimensions, m 6.05D × 21.6H 4.83D × 13.4H 7.6D × 4L 11.3D × 14.4H 0.088 × 8H 21D × 19.5Htubes

Wall thickness, mm 152 224 28.6 (4.72 m minimum) 4 25

Material Stainless-steel- Stainless-steel- Stainless steel Prestressed Zr/Nb alloy Stainless steelclad carbon clad carbon concretesteel steel

Other features Pressure tubes Steel liners Pressure tubes Pool-type

a Data summarized from Knief, R. A. (1992). “Nuclear Engineering,” Hemisphere, Washington, DC.b Parameters are for a large conceptual design; the smaller German THTR, or “pebble-bed” reactor, uses fuel microspheres in 6-cm-diameter graphite

spheres with on-line refueling strategy.c Design values; most reactors currently moving from 12 mo. to 1B or 24 mo. refueling cycles.

from a given amount of fuel. Spacing and coolant flowrate establish heat-removal characteristics (including tem-perature effects on the fuel’s multiplication factor). Finaldimensions generally represent a best-estimate balanceamong these and other competing concerns.

III. REACTOR TYPES

The major reactor types are identified in Section I withrepresentative data provided in Table I. General descrip-tions of these systems follow. The principal focus is on

the steam cycle, fuel assemblies, reactivity control, andthe protective system. General safety-related functions aresummarized separately in the next section.

A. Light-Water Reactors

Two light-water reactor (LWR) systems—boiling-waterreactor (BWR) and pressurized-water reactor (PWR)—employ ordinary (“light”) water as coolant and moder-ator. The former design produces steam through a di-rect cycle (Fig. 1), while the latter uses an intermediatesteam-generator heat exchanger to maintain an all-liquid

P1: GPA Final Pages



TABLE II Worldwide Nuclear Generating Capacity by Reactor Type and Summary of Reactor Electrical Performancea,b,c

1990 reactor 1998 reactorperformance performance

Percent of Percent ofTWe-hr electrical TWe-hr electrical

Reactors under Reactors reasonablyReactors operable construction firmly planned

Country No. MWe No. MWe No. MWe

Argentina 7.3 16.9 6.9 10.0

PHWR 2 1005 1 745

Armenia 211.5d 12.2d 1.42 24.7

VVER 1 408

Belgium 40.4 60.2 43.9 55.2

PWR 7 5836

Brazil 2.1 1.0 3.3 1.1

PWR 1 657 1 1309

Bulgaria 13.5 35.7 15.5 41.5

VVER 6 3760

Canada 67.1 14.4 67.5 12.4

PHWR 14 10,915

China — — 13.5 1.2

PWR 3 2268 4 3200

PHWR 2 1400

VVER 2 2000

6 4600

Cuba

PWR 2 880

Czech Republic 24.6e 28.5e 28.5 20.5

VVER 4 1752 2 1962

Finland 18.1 35.0 21.0 27.4

VVER 2 1020

BWR 2 1630

4 2650

France 298.0 75.0 368.4 75.8

PWR 57 64080 1 1516

FBR 1 250

58 64330

Germany 139.1 33.1 145.2 28.3

PWR 13 15,426

BWR 6 6,643

19 22,069

Hungary 13.7 50.0 13.1 35.6

VVER 4 1840

India 6.2 2.4 10.2 2.5

BWR 2 320

PHWR 8 1520 6 1880 6 1880

VVER 2 2000 2 2000

10 1840 8 3880 8 3880

Iran

VVER 2 2000

Japan 186.4 27.1 306.9 35.9

PWR 23 19,366

BWR 28 25,551 2 1925 4 4875

continues

P1: GPA Final Pages



TABLE II (Continued )





FBR 1 280

Other 1 165

53 35,362

Kazakhstan 211.5d 12.2d 0.1 0.2

FBR 1 150

Korea, North

PWR 2 2000

Korea, South 52.9 49.1 85.2 41.4

PWR 11 9995 3 3050 10 11,200

HWR 3 2094 1 700

Other/unknown 2 1,000

14 12,089 4 3750 12 12,200

Lithuania 211.5d 12.2d 12.3 77.2

RBMK 2 2600

Mexico 2.3 4.1 8.8 5.4

BWR 2 1329

Netherlands 3.3 4.9 3.6 4.1

PWR 1 481

BWR 1 58

2 539

Pakistan 0.4 1.1 0.3 0.7

HWR 1 137

PWR 1 325

Poland ?? ??

VVER [4] [1860]

Romania ?? ?? 4.9 10.4

HWR 1 706

Russian Federation 211.5d 12.2d 95.4 13.1

VVER 13 9,594 3 2630 2 1260

RBMK 11 11,000 1 1000

FBR 4 48 2 1600

Other/unknown 1 600 1 70

29 21,242 4 3630 5 2930

Slovakia 24.6e 28.5e 11.4 43.8

VVER 5 2200 1 440

Slovenia 4.4 f 5.4 f 4.8 38.3

PWR 1 652

South Africa 8.5 12.4 13.6 7.25

PWR 2 1930

Spain 54.3 35.7 56.7 31.7

PWR 7 5950

BWR 2 1450

9 7400

continues

P1: GPA Final Pages



TABLE II (Continued )





Sweden 65.3 46.0 70.0 45.8

PWR 3 2,835

BWR 9 7,603

12 10,438

Switzerland 22.3 42.6 24.4 41.1

PWR 3 1772

BWR 2 1507

Unknown 5 3279

Taiwan 31.5 38.3 ?? ??

PWR 2 1902

BWR 4 3242 2 2700

6 5144

Turkey

Unknown 2 2000

Ukraine 211.5d 12.2d 70.6 45.4

VVER 13 11,808 2 2000

RBMK 1 1,000

14 12,808

United Kingdom 60.8 20.0 91.1 27.1

PWR 1 1,258

Magnox 20 3,786

AGR 14 9,164

35 15,208

United States 576.8 20.6 673.7 18.7

PWR 69 68,577

BWR 35 33,156

104 101,733

Totals

Reactor type

PWR 204 202,985 12 11,400 10 11,200

VVER PWR 48 32,382 10 9,032 6 5,260

BWR 92 82,431 4 4,625 4 4,875

PHWR 29 16,377 11 5,431 6 1,880

Magnox 20 3,786

AGR 14 9,164

RBMK 14 14,600 1 1,000

FBR 4 1,280 2 1,600

Other/unknown 5 213 5 3,070

Total 430 363,218 38 31,488 33 27,885

a From Nuclear Engineering International World Nuclear Industry Handbook 1999, November 1998.b Operable status as of end of 1997, under-construction and planned status as of end of 1998.c Key: BWR, boiling water reactor; FBR, fast-breeder reactor; Magnox and AGR, gas-cooled reactors; [P]HWR, heavy-water reactor; RBMK,

light-water-cooled, graphite moderated reactor; and PWR and VVER, pressurized water reactors.d Values for Russia, which in 1990 included Armenia, Kazakhstan, Lithuania, Russia, and the Ukraine.e Value for Yugoslavia, which in 1990 included Slovenia.f Value for Czechoslavakia, which in 1990 included the Czech Republic and Slovakia.

P1: GPA Final Pages



FIGURE 1 Steam cycle for boiling water reactor (BWR). [Courtesy of Atomic Industrial Forum.]

primary loop and produce steam in a separate secondaryloop (Fig. 2).

The nature of the water coolant/moderator results insimilarities between the two LWR designs. The fuel is 2–5 wt. % enriched 235U in uranium dioxide fuel pellets cladin sealed zirconium-alloy tubes. Fuel assemblies consistof rectangular arrays of fuel pins with regular spacing.

Since the LWR designs rely on liquid water for mod-erating neutrons, maximum operating temperatures mustremain well below the 706◦F (374◦C) critical temperatureat which pressure increases dramatically and liquid cannot

FIGURE 2 Steam cycle for pressurized water reactor (PWR). [Courtesy of Atomic Industrial Forum.]

exist. “Modern” steam conditions nominally at 1000◦F(540◦C), typical of fossil-fueled plants, thus are notavailable; special “wetsteam” turbines must be employed.

1. Boiling-Water Reactors

The direct-cycle boiling-water reactors (BWR) are man-ufactured by General Electric Company in the UnitedStates. ABB-Atom in Sweden, AEG in the UnitedKingdom, Kraftwerk Union in West Germany, and Hitachiand Toshiba in Japan. Employing the cycle shown in

P1: GPA Final Pages



FIGURE 3 Typical steam cycle schematic diagram for boiling water reactor (BWR). [Courtesy of General ElectricCompany.]

concept by Fig. 1 and in more detail in Fig. 3, feedwa-ter enters the steel reactor vessel, is heated by the fissionchain reaction occurring in the fuel pins, and leaves thevessel as steam. The high- and low-pressure turbine stagesare employed in concert with the multiple heaters and con-denser to enhance energy-conversion efficiency. The morerecent BWR designs use jet pumps to recirculate a fractionof the feedwater flow for better control.

Fuel bundles for the BWR appear as shown in Fig. 4.The 7 × 7 to 9 × 9 square array of fuel pins is surroundedby a metal fuel channel, which prevents the water/steammixture from moving between bundles (and potentiallyresulting in too little liquid in some). Fuel bundles maycontain pins of several different enrichments (Fig. 5). Thereactor fuel core consists of up to 800 fuel bundles.

Reactivity control for routine operation is implementedthrough a combination of control rods and coolant flowadjustment. The bottom-mounted control rods (indicatedbelow the reactor vessel in Fig. 3) are made of long boroncarbide filled pins in a cruciform (“cross”) shape that fitsbetween four fuel assemblies as shown in Fig. 5.

Flow adjustment can provide another effective controlmethod, since the water changes density with tempera-ture. At low temperature, the dense water is very effec-tive at moderating neutrons, and thereby encourages fis-sion. With increased temperature, the density decreases(or, equivalently, void content increases as steam is being

produced), causing a reduction in moderation and fissionrate. Thus, if flow rate is increased, energy removal canbe increased without a net change in coolant temperaturewith a resulting increase in power generation. In practice,power-level changes of up to 40% may be accomplishedby flow control.

Longer-term reactivity control is accomplished usingburnable poisons (e.g., “curtains” of boron between fuelassemblies or gadolinium poisons fabricated into the fuelitself) and gradual withdrawal of inserted control rods overcore lifetime. Reactor scram or trip is accomplished byusing gas pressure to insert all of the bottom-mountedcontrol rods into the core.

2. Pressurized-Water Reactor

The two-loop pressurized-water reactors (PWR) havebeen manufactured by Westinghouse, ABB CombustionEngineering, and Babcock & Wilcox in the United States;Framatome in France; Brown Boveri, Kraftwerk Union,and Seimens in Germany: Mitsubishi in Japan; and Atom-mash in the USSR (now Russia). Water in the primary loop(Fig. 2) is maintained as liquid by using high pressure.Water enters the reactor vessel (Fig. 6) at the inlet nozzle,flows downward along the inner vessel wall, is distributedat the lower vessel plate, flows up through the fuel assem-blies gaining heat energy, and exits at the outlet nozzle.

P1: GPA Final Pages



FIGURE 4 Typical fuel assembly for boiling water reactor. [Courtesy of General Electric Company.]

Energy from the primary loop is extracted and convertedto steam by two to four U-tube (Fig. 7), once-through(B&W), or horizontal (Russian VVER) steam genera-tors. Multiple turbine stages, heaters, and a condenser areemployed as for the BWR (Fig. 3). A pressurizer witha steam–water interface is used to maintain the sensi-tive pressure/temperature balance in the primary systemby using heaters to make more steam and increase pres-sure or spraying cool water to condense steam and reducepressure.

Fuel assemblies for the PWR are of 14 × 14 to 17 × 17square fuel pin arrays (Fig. 8) or a hexagonal array of up to331 pins (Russian VVER). They are not enclosed in a fuelchannel (in part because the single-phase primary fluid isbetter behaved than the BWR’s boiling coolant). Theseassemblies also have unoccupied pin locations, which can

accommodate control rods, burnable poisons, or instru-ments. The large PWR reactor cores consist of from 150up to 200 or more fuel bundles.

Reactivity control is accomplished mainly with solu-ble poison in the form of boric acid assisted by controlrods. The boron concentration is adjusted to match generalchanges from fuel burnup, conversion of fertile material,and depletion of burnable poisons. Control-rod assembliesconsist of five to 24 “fingers,” made of boron carbide or ofa silver–indium–cadmium mixture, which move in chan-nels within the fuel assemblies (e.g., as shown in Fig. 8).A small symmetric group of rods is generally inserted ashort distance into the fuel and then moved as needed tocompensate for routine power fluctuations.

Scram or trip is accomplished by dropping the top-mounted rods (Fig. 6) into the core under the influence

P1: GPA Final Pages



FIGURE 5 Core fuel module (a) and control rod pattern (b) for typical boiling water reactor. [Courtesy of GeneralElectric Company.]

of gravity. The rods are mounted to their drives byelectromagnets so that interruption of the current (frompower failure or a designated indication that parametersare outside of accepted limits) causes the rods to fall.

B. Heavy-Water Reactors

Ordinary hydrogen in the form of water is the most ef-fective material for reducing neutron energy, but it alsoabsorbs some of the neutrons that could otherwise par-ticipate in the chain reaction. Thus, deuterium as heavywater, which requires more collisions for a given energychange but exhibits much less absorption, is also a usefulreactor moderator. Deuterium, existing in nature in a ratioof 1:400 with ordinary hydrogen, requires isotopic enrich-ment prior to use (as does 235U in uranium for many of thereactor applications).

Heavy-water reactors (HWR) have been manufacturedby Atomic Energy of Canada Limited (AECL) and inthe United Kingdom, West Germany, India and Japan.Pressure-vessel designs (similar to the PWR) employ thesame heavy water as coolant and moderator. Pressure-tube designs use heavy water in a moderating volumewith a separate coolant, which could be heavy water, or-dinary water, or an organic liquid. The pressurized heavy-water reactor (PHWR) in the form of the popular Cana-dian deuterium uranium (CANDU) system is consideredbelow.

The steam cycle is two-loop (Fig. 9), like the PWR, withthe primary pressurized heavy-water loop transferring heatenergy to a loop of ordinary water for steam production.A major difference, however, is that the primary fluid is

distributed among several hundred pressure tubes whichpass through a large calandria vessel (Fig. 10) containingseparate heavy-water moderator. The coolant is actuallycollected in two separate loops, with adjacent tubes beingpart of different loops.

The fuel assemblies consist of natural (i.e., 0.711 wt.%235U) uranium dioxide fuel pellets in zirconium clad, sim-ilar to LWR fuel. However, short, cylindrical bundles offuel pins (Fig. 11) allow a unique on-line fueling schemewhereby a machine attaches to each end of a single coolanttube and inserts one fuel bundle while removing another.

A major portion of the reactivity control is accompl-ished by on-line fueling, which is required to compen-sate for the low reactivity inherent in the natural uraniumfuel. Routine operating adjustments and power shapingare accomplished with poison control rods or introduc-tion of ordinary water (which absorbs more neutrons thanheavy water) into special chambers. Other control rodsare available for reactor trip. The separation of the coolantand moderator volumes also provides the possibility formoderator “dumping” as an emergency shutdown method.

C. Gas-Cooled Reactors

The world’s first research reactor used natural uranium,graphite moderator, and natural-circulation air cooling.Subsequent systems have also used graphite with naturalor enriched uranium and with carbon dioxide or heliumcoolant. Various commercial gas-cooled reactors (GCR)have operated in France, the United Kingdom, the UnitedStates, and West Germany. The U.K. still has a numberof CO2-based gas-cooled reactors, split between Magnox

P1: GPA Final Pages



FIGURE 6 Reactor pressure vessel for a typical pressurized-water reactor. [Courtesy of Westinghouse ElectricCompany.]

(natural uranium fuel) and Advanced Gas Reactor (AGR)(slightly enriched uranium fuel) designs. Two versionsof the helium-cooled high temperature gas-cooled reactor(HTGR) developed by the United States and Germany aredescribed below.

The HTGR steam cycle (Fig. 12) employs a primaryloop of helium, heat exchangers, and pumps containedwithin a prestressed concrete reactor vessel (PCRV) (twoversions of which are shown in Fig. 13), and a steam loop.Since the coolant is single-phase gas, no pressurizer is re-quired (in contrast to the two-loop water-cooled designs).The nature of the coolant also provides the prospect fordirect conversion through a gas turbine.

Fuel for the HTGR consists of small microspheres ofuranium or thorium carbide (UC/ThC) with coatings ofgraphite and/or silicon carbide [Fig. 14(a)]. The uraniummicrospheres, enriched to 20–93 wt.%, may be mixed withseparate thorium microspheres to an effective fissile en-richment of about 5 wt.%.

In the United States “prismatic” HTGR system, themicrosphere mixture is formed into roughly finger-sized sticks with a carbon-resin binder. The sticks arethen loaded into large hexagonal graphite blocks withinterspersed coolant holes [Fig. 14(b)]. The blocks arestacked several high and in a roughly cylindrical arrange-ment to form the reactor core [Fig. 13 (a)].

P1: GPA Final Pages



FIGURE 7 Four U-tube steam-generator primary loop configura-tion for a pressurized-water reactor. [Courtesy of WestinghouseElectric Company.]

Another version of the HTGR, the German thoriumhigh-temperature reactor (THTR), forms the microspheresinto a spherical shape and coats them with hard, thickgraphite layers [Fig. 14(c)]. The core is then formed byloading the spheres into a hopper in a PCRV [Fig. 13(b)],from which fueling and defueling can be accomplished on-

FIGURE 8 Typical fuel assembly for a pressurized-water reactor. [Courtesy of Combustion Engineering, Inc.]

line. This design feature has led to the designation “pebblebed” reactor.

Reactivity control in the prismatic design depends oncontrol rods for routine and shutdown functions. Burnablepoisons may be used for long-term reactivity control. A re-serve shutdown system consisting of small boron carbideballs backs up the primary systems.

The THTR has minimal excess reactivity due to its abil-ity to change fuel on-line. Control rods provide the meansfor routine operational adjustments.

D. Light-Water Graphite Reactors

Light-water-cooled graphite-moderated reactors (LGR)were among the first systems used for purposes of re-search, fuel conversion, and power production. A smallSoviet unit of this type is credited with generating the firstcommercial electricity. Current commercial use is lim-ited to the Soviet RBMK pressure-tube graphite reactors(PTGR). The Chernobyl reactor—the site of the serious1986 accident—was of this type.

The PTGR uses a direct steam cycle with boiling-watercoolant like the BWR (Fig. 1). Its pressure-tube design

P1: GPA Final Pages



FIGURE 9 Steam cycle for CANDU pressurized heavy-water reactor.

with separate coolant and moderator, however, also hassimilarities to the CANDU (Fig. 9).

Cylindrical fuel assemblies (Fig. 15) each contain twostacked sub-assemblies of 18 zirconium-clad fuel pinsof UO2 enriched to 1.8 wt.% in 235U. Water enters thecoolant inlet, flows downward through a central tube, re-distributes in the lower head, travels upward among thefuel pins, and exits as a steam–water mixture. The reac-tor core (Fig. 16) consists of nearly 1700 fuel assembliesdistributed through a graphite cylinder of roughly 12-mdiameter and 7-m height.

The PTGR fuel assemblies can be exchanged on-line tomaintain the general reactivity balance. Control rods aremoved for routine power adjustment and fully inserted forshutdown.

E. Breeder Reactors

The breeder-reactor design concept is predicated on max-imizing new fuel production in breeding more fuel thanused to sustain the neutron chain reaction. For this purpose,fissile plutonium and fertile 238U fuel with fast neutronshave been found to be the most efficient.

The liquid-metal fast-breeder reactor (LMFBR) keepsneutron energy high by using liquid sodium as a coolant,and thereby specifically avoiding the presence of moderat-ing material. The liquid sodium, although not the heaviestcoolant available, is not too light, has favorable heat-

transfer properties, and is not an excessively strong ab-sorber of neutrons compared to other choices.

Although experimental fast-breeder reactors have beenoperated in the United States since the late 1950s, the mostrecent intense focus on LMFBR systems had been in West-ern Europe, the Russia, Japan, and India. With shutdownof the major western European systems located in Franceand Germany, (which were funded by consortia thatalso include Italy, Belgium, Netherlands, and the UnitedKingdom), the future of this reactor type is in doubt.

The steam cycle is a three-loop system (Fig. 17) withthe first two of sodium and the third of water. The interme-diate loop is present to isolate the primary from possiblecontact with water in the steam generator. The primarysodium becomes radioactive from neutron absorption andalso can pick up fission-product radionuclides from thefuel. If this sodium were to come in contact with water,it would lead to an exothermic reaction that also wouldspread contamination.

Fuel for the LMFBR consists of mixed-oxide fuel pel-lets, which combine about 10–30 wt.% plutonium withnatural or depleted (0.2–0.35 wt.% 235U) uranium—thebyproduct of the enrichment process. The slender pel-lets are loaded into thin stainless-steel cladding tubes, andhence into hexagonal array subassemblies [Fig. 18(a)].Breeding is optimized by surrounding the mixed-oxidecore with a blanket of depleted uranium. The axial blanketsabove and below the core are created by loading pellets

P1: GPA Final Pages



FIGURE 10 Calandria vessel and pressure tube of CANDU pressurized heavy-water reactor. [Courtesy of AtomicEnergy of Canada Limited.]

of natural or depleted uranium at either end of the corefuel pins. The surrounding radial blanket consists of sepa-rate subassemblies [Fig. 18(b)] of natural or depleted ura-nium, where the pins may be of larger diameter becausethe power density in the blanket is much lower than it isin the core.

Reactivity control is accomplished through use of poi-son control rods. Since the breeder produces more fuelthan it uses, however, the multiplication does not decreasewith fissile burnup and fission product buildup as dramat-ically as in the converter reactors described previously.Shutdown, also using control rods, generally depends on

P1: GPA Final Pages



FIGURE 11 Fuel assembly for CANDU pressurized heavy-water reactor. [Courtesy of Atomic Energy of CanadaLimited.]

FIGURE 12 Steam cycle for high-temperature gas-cooled reactor (HTGR). [Courtesy of Atomic Industrial Forum.]

P1: GPA Final Pages



FIGURE 13 Prestressed concrete reactor vessels (PCRV) for (a)prismatic high-temperature gas-cooled reactor and (b) thoriumhigh-temperature (“pebble-bed”) reactor. [Courtesy of Oak RidgeNational Laboratory.]

two redundant sets of rods. Each set operates on a differentphysical principle and either set, by itself, can shut downthe chain reaction.

F. Other Reactor Concepts

Although the previous designs are the most popular, a widevariety of other possibilities have been built for powerproduction or research purposes. Still others have beenresearched “on paper.” The major approach is to look atviable combinations of fuel, coolant, and moderator. Afew examples are identified below.

CANDU reactors could also be operated with fuel as-semblies of enriched uranium, plutonium, 233U/thorium,or a mixture thereof. Potential coolants other than heavywater include light water and organic liquid.

The gas-cooled fast breeder reactor (GCFBR) useshelium coolant and plutonium fuel in a concept similarto the HTGR, except with no graphite moderator. Themolten-salt breeder reactor (MSBR) concept includesliquid fuel that circulates through a graphite-block coreregion in a closed primary loop. The fuel is processedonline to remove fission products and 233U bred fromthorium in the salt.

Several other novel designs that stress enhanced safetyhave also been proposed. Three such concepts are intro-duced in the next section.

IV. SAFETY FEATURES

Each reactor design incorporates features to respond toanticipated system upsets and to unlikely, but not incredi-ble, serious accidents. The major concerns are the poten-

FIGURE 14 Fuel assembly components for high-temperaturegas-cooled reactors: (a) microspheres, (b) prismatic fuel block[courtesy of GA Technologies], and (c) Fuel sphere (“pebble”).[Knief, R. A. (1992). “Nuclear Engineering: Theory and Technol-ogy of Commercial Nuclear Power,” 2nd ed., Taylor & Francis/Hemisphere, New York.]

tial release of radioactive fission and transuranic-elementproducts to the environment.

A. Fundamentals

Since dispersal of radioactive products requires energy,reactor safety is equivalent to reducing and controllingthe energy source. Energy stored in the system’s coolantand fuel as a result of temperatures and pressures duringroutine power operation must be accomodated without

P1: GPA Final Pages



FIGURE 15 Fuel assembly for a Soviet RBMK pressure-tubegraphite reactor (PTGR). [From, NUREG-1250 (1987).]

causing damage. The unique nuclear energy source, thefission chain reaction, must be terminated. Then, thedecay-heat byproduct of fission product radioactive decaymust be controlled to prevent overheating. If these firstthree energy sources are controlled, the next concern—chemical reactions, primarily between the coolant andcladding—will also be accommodated. External events,such as earthquake or tornado, have the potential for initi-ating an accident, and thus must be incorporated into thedesign basis.

Each reactor has design-basis accidents that determineacceptability in terms of potential radioactive product re-lease. These are classified roughly in terms of:

1. Reactivity transients—neutron poison changes thatlead to unplanned power increases.

2. Overcooling—excessive heat removal from steamwithdrawal, perhaps through steam-generator overfeedingor a steamline break.

3. Loss of cooling—failure of core heat removal abilitythrough loss of coolant flow, up to and including loss of thecoolant inventory itself [called, respectively, loss of flowaccident (LOFA) and loss of coolant accident (LOCA)].

4. Steam-generator tube rupture.5. Spent-fuel drops or waste handling spills.6. External events.

More severe accidents, sometimes referred to as beyond-design-basis accidents, where multiple safety systems failto function, are sometimes considered to evaluate overallresponses.

Safety systems, although often highly design-depen-dent (e.g., based on design and operational differencesamong use of water, gaseous, and liquid sodium coolants),have as their goal prevention of overheating, fuel melting,and the subsequent large-scale dispersal of fission prod-ucts. Reliability is enhanced through redundancy in sub-system function and location.

B. Safety Systems

The basic safety systems may be classified according tofunction as:

1. reactor trip (RT)2. emergency core cooling (ECC)3. postaccident heat removal (PAHR)4. postaccident radioactivity removal (PARR)5. containment integrity (CI)

Their basic functions are summarized by Fig. 19 for light-water reactors. The same basic functions apply to all re-actor systems, even if in somewhat different form.

1. Reactor Trip

Each of the reactor types described previously includesneutron poison control rods, which can be inserted rapidlyinto the fuel core to shut down the fission chain reaction.These rods may be supplemented by secondary meanssuch as reserve shutdown spheres (HTGR); a redundant,independent set of rods (LMFBR); or injection of solubleboric acid poison (LWRs).

2. Emergency Core Cooling

Emergency core cooling for the light-water reactors isprimarily based on injection of (borated) water into thecoolant-starved core region following a LOCA event. Mul-tiple trains of separate systems typically can inject water athigh, intermediate, or low pressure to coincide with vari-ous needs during the time-history and/or magnitude of theevent. Recirculation of coolant that collects in the reactorbuilding sump provides a long-term coolant supply afterthe initial inventories have been exhausted.

The CANDU system also has injection capabilities, al-though grouping the pressure tubes (Fig. 10) into two sep-arate flow circuits means that a given break will removecooling capability from only half of the fuel. The large sep-arate moderator volume in the calandria vessel providesadditional sink for energy removal.

Emergency cooling in the HTGR design depends pri-marily on helium retention by the concrete vessel and theheat capacity of graphite. The LMFBR uses natural cir-culation of the low-pressure liquid sodium coolant, which

P1: GPA Final Pages



FIGURE 16 Sectional view of a Soviet pressure-tube graphite reactor (PTGR). [From IAEA Bull. 25(2).]

due to its high boiling temperature would not automati-cally leave the primary system if a leak were to occur.

3. Postaccident Heat Removal

Removal of postaccident energy consists of two aspects—coolant temperature reduction, and containment-buildingpressure control. The first is accomplished through heatexchangers for ECC water recirculation in the water reac-tors. For the gas or liquid-sodium reactors, continued useof the steam generators can serve a similar function in theprimary coolant loops.

Containment pressure control may be accomplished byusing air coolers or, in water reactors, through water spraysto condense steam.

4. Postaccident Radioactivity Removal

Radioactive fission products, primarily chemically ac-tive iodine and aerosol/particulate constituents, may

be removed by filtration. Noble gases can only becontained.

Water reactors have provision for containment spraysto remove radioactivity. Although the water sprays usedfor pressure reduction naturally remove some radioactivematerial, additives such as sodium hydroxide or thiosulfateincrease removal, especially of elemental iodine.

5. Containment Integrity

The last line of defense against fission product release isthe integrity of the containment structure or building. Ifthe other systems have functioned as intended, pressurebuildup should not threaten the containment.

A common denominator of containments is a leak-tightsteel liner. In several of the designs, the liner is surroundedby thick reinforced concrete [including, for example, thatfor the pressurized water reactor in Fig. 20 and the struc-ture of the HTGR’s reactor vessel (Fig. 13)]. The lack of

P1: GPA Final Pages



FIGURE 17 Steam cycle for liquid metal fast breeder reactor (LMFBR). [Courtesy of Atomic Industrial Forum.]

such a leak-tight containment structure at the ChernobylPTGR was a major contributor to the serious consequencesof the 1986 accident there.

The other major element of containment integrity isthe ability to isolate penetrations using remotely operatedvalves or other means. These typically actuate on prede-

FIGURE 18 Fuel assembly for a typical liquid-metal fast-breederreactor. [Courtesy Nuclear Engineering International.] (a) Fuel as-sembly: 1, pin cladding; 2, slugs of depleted uranium; 3, fuel pel-lets; 4, wire-wrapped pin; 5, fuel-assembly head; 6, fuel-pin as-sembly; 7, stem. (b) Radial blanket assembly: 1, pin cladding; 2,wire-wrapped fin; 3, depleted uranium; 4, blanket assembly stem;5, blanket pin assembly; 6, blanket assembly head.

termined indication of excessive pressure, radiation level,or other related parameter.

C. Advanced Reactors

Interest in reducing the risk of serious reactor accidents hasled to consideration of several advanced reactor designs.Each of the concepts described next includes enhancednegative power feedback mechanisms (to cause inherentshutdown) and passive postaccident/postshutdown cool-ing mechanisms.

The Westinghouse AP600 PWR, featuring large watertanks above the core, is capable of providing emergency

FIGURE 19 Functions of safety systems for light-water reactors.[Adapted from WASH-1400, courtesy of U.S. Nuclear RegulatoryCommission.]

P1: GPA Final Pages



FIGURE 20 Containment structure for typical pressurized water reactor. [Courtesy of Westinghouse ElectricCompany.]

cooling water without pumps or electric power. Heat canbe removed from the steel containment shell by a gravity-fed water spray and natural circulation of air.

The process-inherent ultimate safety (PIUS) reactorconcept from Sweden’s ASEA-Atom is essentially a PWRsubmerged in a large pool of borated water and surroundedby a massive concrete and steel containment structure. Thehydraulic connection between the primary-loop and poolcoolant volumes prevents intermixing during normal op-eration. Under accident conditions, however, the boratedpool water enters the primary to assure both shutdown ofthe chain reaction and long-term cooling.

GA Technologies’ modular high-temperature gas-cooled reactor [MHTGR] has a low-power-density graph-

ite core that provides a large inherent heat sink and isvery slow to overheat. The steel vessel and steam genera-tor are enclosed in an underground silo cooled by naturalair circulation and, if necessary, direct heat loss to theground.

The power reactor inherently safe module [PRISM] is asmall LMFBR being developed by General Electric. Fuelassemblies are made of a pyrometallurgical alloy used inthe integrated fast reactor (IFR) concept. The IFR fuelhas been tested at the Experimental Breeder Reactor 2(EBR-2) in Idaho, where feedback alone shut down thecore and natural circulation of the liquid-sodium coolantprovided sufficient decay-heat removal. The PRISMsystem is placed in an underground concrete silo where

P1: GPA Final Pages



air can circulate freely to remove core decay heat, ifnecessary.

V. REGULATIONS

The inherent hazard associated with the radioactive ma-terial in reactor systems has led to the industry beingthe most regulated in the world. Through regulatorybodies such as the U.S. Nuclear Regulatory Commis-sion (NRC), French Commissariat a l’Energie Atomique(CEA), Atomic Energy Control Board (AECB) of Canada,and the United Kingdom Nuclear Installations Inspec-torate (NII), requirements for reactor design and opera-tion are established and implementation is evaluated andmonitored. Regulations with the force of law, licenses, orother methods are developed for reactor operation by suchbodies, often with significant political and/or public inputto the process.

One increasingly important element of the regulatoryprocess is quality assurance (QA), focusing on methodsand procedures to assure proper design, construction, andoperation of safety-related components and subsystems.Another important feature is establishment of acceptableradiation exposures and identification of design-basis ac-cidents whose analysis must show them to have conse-quences within the pre-established limits.

The accident at the Three Mile Island Unit-2 (TMI-2)PWR in 1979 has led to many changes in the regula-tory process in the United States and elsewhere in theworld. Some of these changes relate to the design, qual-ity assurance, and inspection of modifications to plantsafety systems; development and use of preapprovedprocedures for operation, maintenance, and other activ-ities; administration, including staffing, training, and doc-umentation; emergency planning; technical support, in-cluding accident and root-cause analysis; and supportservices such as radiological controls, chemistry, andmaintenance.

The extremely serious 1986 accident at the ChernobylUnit 4 occurred in a system not used elsewhere in theworld. However, it did serve to reinforce many of the de-sign, operations, and management lessons from the earlierTMI-2 accident. It also provided unprecedented insightsinto severe-accident behavior and served as a catalyst forsignificantly enhanced international cooperation and col-laboration in research initiatives and nuclear-power-plantoperation and management.

SEE ALSO THE FOLLOWING ARTICLES

ENERGY RESOURCES AND RESERVES • FISSION REAC-TOR PHYSICS • NUCLEAR ENERGY, RISK ANALYSIS • NU-

CLEAR FUEL CYCLES • NUCLEAR FUSION POWER • NU-CLEAR REACTOR MATERIALS AND FUELS • NUCLEAR RE-ACTOR THEORY • NUCLEAR SAFEGUARDS • RADIOAC-TIVE WASTES • THERMIONIC ENERGY CONVERSION •URANIUM

BIBLIOGRAPHY

Agnew, H. M. (1981). “Gas-cooled nuclear power reactors,” Sci. Am.244, 55–63.

Catron, J. (1989). “New interest in passive reactor designs,” EPRI Journal14(3), 4–13.

Douglas, J. (1994). “Reopening the nuclear option,” EPRI J. 19, 6–17.Fishetti, M. A. (1987). “Inherently safe reactors: They’d work if we’d let

them,” IEEE Spectrum 24(4), 28–33.Framatome (1992). P4: The 1300 MWe PWR Series, Paris, France, April.Glasstone, S., and Sesonske, A. (1994). “Nuclear Reactor Engineering

(Volume 1: Reactor Design Basics & Volume 2: Reactor SystemsEngineering),” 4th ed., Chapman & Hall, New York.

Golay, M. W., and Todreas, N. E. (1990). “Advanced light water reactors,”Sci. Am. 262, 82–89.

Knief, R. A. (1992). “Nuclear Engineering: Theory and Technology ofCommercial Nuclear Power,” 2nd ed., Taylor & Francis/Hemisphere,New York.

Lewis, H. W. (1980). “The safety of fission reactors,” Sci. Am. 242, 53–65.

Lish, K. C. (1972). “Nuclear Power Plant Systems and Equipment,”Industrial Press, New York.

Marshall, W. (ed.) (1983). “Nuclear Power Technology,” Volumes 1, 2,and 3. Clarendon Press, Oxford, England.

McIntyre, M. C. (1975). “Natural uranium heavy-water reactors,” Sci.Am. 233, 17–27.

Murray, R. L. (1993). “Nuclear Energy,” 4th ed., Pergamon Press,Elmsford, NY.

Nero, A. V., Jr. (1979). “A Guidebook to Nuclear Reactors,” Universityof California Press, Berkeley, CA.

Nuclear Engineering International (published monthly in the UnitedKingdom; includes annual “World Nuclear Handbook”).

Nuclear Engineering International Special Publications (1989). FrenchPWR Technology.

(1992). “Special section: The new reactors,” Nuclear News 35, 65–90.Rahn, F. J., et al. (1984). “A Guide to Nuclear Power Technology,” Wiley-

Interscience, New York.Reactor Safety Study: An Assessment of Risks in U.S. Commercial

Nuclear Power Plants. (1975). U.S. Nuclear Regulatory Commission,WASH-1400 (NUREG-74/014), Washington, DC.

Sesonske, A. (1973). “Nuclear Power Plant Design Analysis,” U.S.Atomic Energy Commission, TID-26241, Washington, DC.

Severe Accident Risks: An Assessment for Five U.S. Nuclear Plants(1989). U.S. Nuclear Regulatory Commission, NUREG-1150,Washington, DC.

Todreas, N. E., and Kazimi, M. S. (1990). “Nuclear Systems (Vol. 1:Thermal Hydraulic Fundamentals and Vol. 2: Elements of ThermalHydraulic Design),” Taylor & Francis/Hemisphere, New York.

U.S. Department of Energy (1987). Overall Plant design descriptionsVVER Water-Cooled, Water-Moderated Energy Reactor, DOE/NE-0084, Rev. 1, October.

U.S. Nuclear Regulatory Commission (1987). Report on the Accident atthe Chernobyl Nuclear Power Station, NUREG-1250, Rev. 1.

Vendryes, G. A. (1977). “Superphenix: A full-scale breeder reactor,” Sci.Am. 236, 26–35.

Documents

Nuclear Power Reactors