Heavy W ate r , Gas and Liquid Metal Cooled Reactors

Jacopo Buongiorno

Associa t e Pr of essor of Nuc l ear Science and Engineering

22.06 : Engineering of Nuclear Systems

Heavy W ater Cooled Reactors (CANDU)

Key C ANDU Features

CAN ada D euterium U ranium

D es i gne d f or natura l uran i um f ue l ( no enr i c h ment needed)

Heavy w ater (D 2 O) moderated

Pressure tube reactor (no pressure vessel) Moderator & coolant separated

Pressurized coolant and steam generators (similar to PWR)

On - power refuelling

High resource utilization (150 tons mined uranium per GW e yr, compared to 200 tons mined U per GW e yr for a t yp i ca l PWR)

Source: Jeremy Whitlock, AECL Chalk River Labs,

4/16/07

Pic ker in g , O n t a r i o Wolsong, South Korea

NPD , Ontario (1962 )

(197 1 - 7 3 , 1983 -8 6 )

( 1 9 82, 1 9 97- 99)

Douglas Point, Ontario ( 1 9 66)

Qinshan, China ( 2 0 02- 03)

Pt. Lepreau,

Ne w Bruns w i c k (1983 )

Rajas t ha n, India (197 3 , 198 2 )

Kanupp, Pakis t a n (1972 )

G entilly 1 and 2 , Q uebec

( 1 9 71, 1 9 83)

Bruce, Ontario (197 7 - 7 9 , 1985 -8 7 )

Darlington, Ontario ( 1 9 90- 93)

Cernav oda , Romania (199 6 , 200 7 , …?)

Embalse , Ar g entina ( 1984 )

20 in Canada 12 offshore

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CANDU STATION OVERVIEW

Power cycle s imilar t o P WR and BWR

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CANDU P RIMARY SYSTEM

Natural uranium fuel and D 2 O moderator

Fuel contained in individu al fuel channels (pressure tubes) filled with high pressure (>10 MPa) and high temperatur e (~300  C) D 2 O coolant

Pressure tubes contained in a large cylindrical tank (calandr ia) filled with low pressure ( < 1 MP a ) an d l ow temper atur e (<80  C) D 2 O moder ator

Fuel clad and pressure tubes are made of Zr alloys

Fuelling machines connect to individual pressure tubes for refuellin g

Conventional turbine/generator and auxiliary systems

1 M ain Steam Pipes

2 P ressu riz e r

3 S team Gene ra to rs

4 H eat Transp o rt Pumps

5 H ead e r s

6 C alan d r ia

7 F uel

8 M odera t o r Pumps

9 M odera t o r Heat Exchan g e rs

10 Fuelling Machines

CANDU F UEL B UNDLES

UO 2 pellets in Zircaloy cladding (0.38 mm thick)

28 or 37 pins form a fuel bundle (pins have a 13.08 mm outside diameter) Pins held together by end plates.

Pins separated by spacers. Outer pins have bearing pads. Bundle s : 495 . 3 m m l ong and 102 . 5 m m i n d iameter Average burnup : 7500-85 00 MWd/ton)

Public domain image from Wikipedia.

Image courtesy of Atomic Energy of Canada Limited.

PRESSURE TUBES (OR F UEL CHANNELS)

Each fuel channel consists of a pressure tube and two end-fittings (primary pressure boundary) , plus a calandr ia tube

Pressure tube - c alandr ia tube separated by a gas - filled annulus; gap maintained by “garter” springs

Low neutron cross section

Total c hannel length: 11 . 56 m fuelled)

( ~6 m

Calandria T ube

Fuel

Pressure T ube

.

CALANDRIA ASSEMBLY

Holds the heavy water moderator

Penetrated horizontally by pressure tubes, an d vert i ca ll y b y react i vi ty d ev i ces

380-480 horizontal pressure tubes

12 or 13 fuel bundle s

p er

p ressure tube

Not a pressure vessel

REPRESENTATIVE PARAMETERS FOR A DVANCED CANDU (ACR-700)

P a r a m e t e r V a l u e

N u m b e r o f f u e l e l e m e n t s p e r c h a n n e l P r e s s u r e t u b e l a t t i c e p i t c h ( m m )

4 3

2 2 0

Image by MIT OpenCourseWare.

Connection of CANDU Core Desi g n to Neutronics

What enables a CANDU reactor to operate with natural uranium?

What determines the pressure tube spacing?

Is the po w er densit y in a CANDU core < , = or > than a PWR? What would happen if the calandria tank were drained?

What happens to reactivity if some voiding (boiling) occurs in a CANDU pressure tube?

F U E L L I N G M A C H I N E S

Two fuelling machines operate simultaneously accepting or loading fuel

Remotely operated from control room

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TWO FAST - ACTING SHUTDOWN SYSTEMS

High Temperature G as Reactors (HTGR)

HTGR Overview

Small modular units: 125-300 MWe H e li um coo l e d , 8 5 0 - 900  C out l et T ,

<9 MPa pressure

Thermal efficiency >40% Gra p hite moderated Microsphere UO 2 or UCO fuel Electricity and process heat Passive decay heat removal

Two “flavors”: block core or pebble bed

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Block Core HTGR

TRISO fuel particle

Py roly tic Carbon Sili con C ar bid e Porous Carbon Buffer UO 2 (or UCO) Kernel

TRISO Coated fuel particl es (left) are formed into cy lindrical fuel compacts (cen ter) and inserted into hexagon al graphite fuel elements (right).

TRISO P ARTICLES CYLINDR I CAL

COM P ACTS

HEXAGONAL FUEL ELEMENTS

L - 0 29( 5) 4-14 -9 4

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Block C ore H TGR ( 2)

U 17-19% enriched Requires mixed

enrichment, burnable Bl o c k 80-cm tall blocks stacked

poisons Core 10 blocks high

P a r t i c le

102 columns o f f uel

R e p l ac ea bl e c e n t r al r e f l e c t o r

1m m R e p l ac ea b l e s i d e r ef l e c t o r

P e r m an ent si de r e f l e c t o r M e t a l l i c c o r e su pp or t ( b a r r e l )

102 f f u e l co l u m n s

Co m p a c t ( 1 0 b l ock s h i gh)

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36 op e r at i n g co nt r o l ro d s 1 2 st a r t - u p co n t r o l r o d s

1 8 r e s e r v e s h u t d ow n c han n e l s

17

Block Core HTGR ( 3 )

Being developed by AREVA, General Atomics and Japan. E xper i ence i n US (Ft . S a i n t V ra i n ) an d J apan (HTTR)

330 MW e

Operated from 1979 to 1989 U/Th fuel

Poor performance, mechanical problems, decommissioned

40 MWth T e st Reactor at JAERI First Critical 1999

Intermediate Heat Exchanger

Currently in T e sting for Power Ascension

18

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Pebble B ed HTGR

F U E L E L E M EN T D E S I G N F O R P B M R

5mm Graphit e l a yer

Coat ed p a r t ic l e s im bedded i n Gr aphite M a t r ix

Dia. 60mm Py ro ly t i c C a r b on 40 / 1 0 0 0 m m

Fu e l Sphe r e

Ha l f S e ct io n

S ilic o n Ca rb it e Ba rr ie r C o at in g 35 / 1 0 0 0 m m

I n ne r Py ro l y t i c Car b on 4 0 /1 00 0m m

Po ro us Ca rb on B u f f e r 9 5 /1 00 0m m

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Di a. 0, 92m m

Coa t e d P a r t i c le

Fu el 19

Pebble Bed HTGR ( 2 )

• 400,000 pebbles in core

• O nline refueling, about 3,000 pebbles handled each day

• about 350 discarded daily

• one p ebble dischar g ed ever y 30 seconds

• a verage pebble cycles through core 6 times

• Enrichment 8 - 9 % - c onstant

- no burnable poisons

• Low excess reactivity - l ower peak operating temperatures

Pebble bed

(200  C l ower )

20

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P ebble Bed HTGR ( 3 )

Being developed by PBMR Ltd. and China.

Experience in Germany (AVR, THTR) and China (HTR-10)

15 MW e research reacto r UO 2 fuel

Operated for 22 years

300 MW e demo plant at Hamm-Uentrop

U/Th fuel

10 MWth - 4 MW e Electric

First c riticality D ec 1 , 2 000 Intermediate Heat Exchanger

- Steam Cycle

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21

HTGR La y outs – Direct C y cle

Tu rb i n e

P r e- co o l e r

Ge n e r a t o r

In t e r -c o o l e r

Shu t -o f f D i s k

C o n t am i n at ed O il L u b e S y s t e m

U n - c on t a m i n a t e d O il L u b e S y s t e m

General Atomics, block cor e ,

PBMR Ltd., pebble bed cor e , horizontal turbo-generator

vertical turbo generator 22

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HTGR La y outs – Indirect C y cle

Re ac tor V e s s el

ARE V A , block cor e , combined cycle

RE A C T O R CAV I T Y S E C ONDAR Y

IHX Vessel

High P r essure Tu rbine

Low P r essure T u rbi n e

Compres s or ( 4 )

Power Tur b i n e

C OOLI NG S Y STE M ( R C C S ) T ANKS

M ODULE FUEL S T OR AGE AR EA

FUEL TR ANS F E R TUNNEL

G A S BY PA SS GAS TUR B I N E

COM P RE SS OR

HEAT REC OVER Y ST EAM GE N E R A T O R (H R S G )

R e cu pera tor V es se l

R C C S HEADE RS A ND S T AN DPI P ES

REA C TOR VES S EL

I N TER M E DI ATE HEAT EXC H ANGE R ( I HX)

SE CONDA R Y GAS I S OLA T I ON VALV E S (T Y P ICA L )

C ONDENS ER C OOLI N G WATER

H. P. / I . P . TUR B I N E

MA I N

NS F O R M ER

GEN E RATOR

L . P . T U RBIN E C OND E N S ER

M I T , p e b bl e b e d c o r e , 3-shaft turbo-generator

23

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HTGR Safety

No fission product release from TRISO fuel at up to 1600  C

Low power density (due to use of graphite moderator) makes it possible to remove decay heat by radial conduction and radiation

In case of unprotected Loss of Coolant Accident (LOCA) with loss of on-site and off-site power (a very serious event) t here i s n o f uel melting

Concerns for air ingress (graphite “burns” at high temperature)

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Liquid M etal (Sodium) C ooled Fast Reactors

Fast Reactors

– the concept

Fast reactor is a system in which neutrons are not moderated

The number of neutrons emitted p er neutron absorbed is hi g her for fast fissions, so the extra neutrons can be absorbed in a

U-238 blanket to produce Pu-239, thus “breeding” new fuel

If properly designed , fast reactors can actually breed more fuel than they consume (multiple fuel recycles become possible)

Needs a coolant that does not moderate neutrons, typically a liquid metal such as sodium

Interestingly, the first nuclear reactor to produce electricity was a fast reactor

in Idaho.

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Sodium-Cooled Fast Reactor ( SFR )

Characteristics

• Sodium coolant

• >500°C Outlet Temp

• 150 to 1300 MWe

• Metal or oxide fuel possible

Benefits

• Efficient fissile material generation (breeding)

• Sodium is excellent heat transfer f luid and has high boiling point (880  C)

• Relatively high temperature (good for efficiency  40%), but low pressure system (good for safety)

Drawbacks

• Sodium is reactive with air and steam, hence the intermediate loop and special fire protection measures , which add to cost and complexity

• Requires higher initial enrichment to get started (why?)

• Has positive void reactivity feedback (why?)

• Generates weapons-grade Pu (proliferation concern)

SFR Core

Uses U-238 blankets to absorb neutrons that escape from the driver fuel core region

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