Thorium Fuel Cycles â€“ AECL Experience - media.cns-snc.ca

UNRESTRICTED / ILLIMITÉ

Thorium Fuel Cycles &

Heavy Water Reactors

AECL Experience

Energy From Thorium Event – CNS – UOIT

B. P. Bromley

Advanced Reactor Systems

Computational Reactor Physics

AECL - Chalk River Laboratories

March 22, 2013

Opening Remarks

• There’s nothing magical or mysterious about thorium except:

–3 times abundant as uranium in the earth’s crust – a large resource.

–U-233 (bred from Th-232) has a high 2.2, in both thermal and fast

neutron energy spectrum; can be used for a breeder reactor.

–Pu, Am, Cm, etc. production with Th-based fuels will much lower.

• Any reactor (fast or thermal) can be adapted to use thorium.

• Thermal reactors can operate with lower fissile wt%.

• For thermal spectrum reactor, we want:

–Minimal parasitic neutron absorption; maximum neutron economy.

–Maximum burnup for Th-based fuel for a given fissile content.

– OTT (Once Thru Thorium) Cycle

– SSET (Self-sustaining Equilibrium Thorium) Cycle

– Topping fuel cycles: Th (new + recycled) + (U/Pu) (new + recycled)

UNRESTRICTED / ILLIMITÉ 2

Fundamental Advantage of Heavy Water

• Heavy water has the highest moderating ratio (s/a).

–Slows down neutrons with minimal absorption.

– Better than H (in H2O), better than C (in graphite).

–Can maximize neutron economy, in a thermal-spectrum reactor.

– Save neutrons for fission and breeding new fuel.

• HWR can run on natural uranium and achieve good burnup.

– ~7,500 MWd/t in a CANDU PT-HWR


Pressure Tube Heavy Water Reactors

(PT-HWR) - Advantages

• Pathway Canada Chose – AECL Pursued.

• Excellent neutron economy.

– High conversion ratios (C.R.>0.8).

– Can operate on natural uranium (NU).

– High fuel utilization; conservation of resources.

• Continuous On-line refuelling.

– Low excess reactivity.

– Higher fuel burnup for a given enrichment.

– 30% more burnup than 3-batch refuelling.

– Maximize uranium utilization (kWh/kg-U-mined).

– High capacity factors (0.8 to 0.95).

– Flexibility in fuel loading – one or more fuel types can be used.

• Modular construction.

– Short, relatively simple fuel bundle design.

– Pressure tubes; replaceable; reactor can be refurbished.

– Local fabrication (do not need heavy forgings).


PT-HWR

• Operational

Technology.

• Future HWR

variants.

• Potential for

further

improvements.

• Use R&D to

find them.


PT-HWR / CANDU Reactors

• Designed to maximizes neutron economy.

• Flexible in use of fuel types.

• An existing, proven, and operational technology.

• Supply chain in place.

• Design naturally lends itself to implementation of Th-based fuels.

–Thorium-based fuels have been tested in PT-HWR (NPD-2).

–Thorium bundles have been used in India (in their PT-HWRs).

– Power flattening for start-up cores; alternative to DU.

• Practical implementation time should be relatively short.

• AECL / CRL has helped develop and prove this technology, and

is continually exploring technology improvements to facilitate

implementation and expansion of thorium-based fuel cycles.

– Emphasis on use of solid fuel forms.



Overview of AECL Experience

• AECL has over 50 years of extensive experience with

Thoria-based fuels – Investments made in thorium fuel cycle R&D since the late 1950’s

– First irradiation conducted in 1962 and the most recent in 2005

• Experience includes

– Fuel Fabrication.

– Irradiation testing.

– Post Irradiation Examination.

– Thorium fuel reprocessing.

– Waste management.

– Critical experiments.

– Reactor physics.

– Conceptual design studies.

– Economic analyses.

– System studies.

Thorium in CANDU / PT-HWR Evolution

AECL - OFFICIAL USE ONLY / À USAGE EXCLUSIF - EACL 8

Once-through

LEU/Th

With U-233

Recycle Build U-233

resource U-233 + Pu

Years

Innova

tion

Once-through

Pu/Th

Once-through

Pu/Th

With U-233

Recycle

U-233

+

Pu

PT-HWR Canadian SCWR


Long-term Impact

• Decay heat in spent fuel is a main parameter in

determining the capacity of a long term disposal facility

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.6

0 200 400 600 800 1000

De

cay

he

at (

GW

)

Years since end of scenario

Current global cycle, LWRs + HWRs

Transition to once-through thorium in CANDU

Transition to fast reactors

Transition to Th with

U-233 recycle in

CANDU Once-through thorium gives 50%

reduction over current cycle

Once-through thorium gives the

same reduction as fast reactors

Thorium with U-233 gives a 75%

reduction over the current cycle


Fabrication

• Generally, AECL has targeted a solid solution

of Thoria and the fissile additive.

• Many techniques are capable of achieving

this and they fall into two main categories:

1. Solution blending – sol gel, co-precipitation – Mixing at the atomic level

2. Mechanical mixing – co-milling, high-intensity mixing – Often not a “perfect” solid solution

– Must achieve mixing on the scale of individual particles


Thorium Pellet Structure

Granular

Homogeneous


Thoria Irradiation Experience at AECL

Irradiation

Facility

# of experiments Irradiation Time

frame

NPD 1 1976

NRX 20 1962-1992

NRU 28 1966-2005

WR1 18 1970-1980

• Thoria irradiations ongoing since early 1960s

• Irradiations in NRX, NRU and WR-1 research reactors.

• Irradiations in NPD-2

–~20 MWe prototype PT-HWR.

• Pure ThO2, (U,Th)O2 , and (Pu,Th)O2

• NRU still operational.

NRX, NRU, WR-1, NPD

• NPD-2

• WP-1

• NRU

• NRX


Critical Experiments

• AECL: long history of critical experiments involving

thorium-based fuels.

• Three sets of experiments,

dating back to the 1960’s

– HEU/Th (1966-1968)

– Pu/Th (1986)

– U-233/Th (1990s)

• Performed in the ZED-2

(Zero Energy Deuterium)

critical facility at Chalk River

Laboratories.

–Reaction rate / foil data.

–Reactivity changes due to X

– X = coolant density, temperature, etc.

–Verifies physics; validate computer codes.



Alternative Fuel Bundle and

Core Design Options

Row\Col 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 Row\Col

A 0 0 0 0 0 0 0 0 B B B B B B 0 0 0 0 0 0 0 0 A

B 0 0 0 0 0 B B B S S S S S S B B B 0 0 0 0 0 B

C 0 0 0 0 B S S S S S S S S S S S S B 0 0 0 0 C

D 0 0 0 B S S S S S S S S S S S S S S B 0 0 0 D

E 0 0 B S S S S S S S S S S S S S S S S B 0 0 E

F 0 0 B S S S S S S S S S S S S S S S S B 0 0 F

G 0 B S S S S S S S S S S S S S S S S S S B 0 G

H 0 B S S S S S S S S S S S S S S S S S S B 0 H

J B S S S S S S S S S S S S S S S S S S S S B J

K B S S S S S S S S S S S S S S S S S S S S B K

L B S S S S S S S S S S S S S S S S S S S S B L

M B S S S S S S S S S S S S S S S S S S S S B M

N B S S S S S S S S S S S S S S S S S S S S B N

O B S S S S S S S S S S S S S S S S S S S S B O

P 0 B S S S S S S S S S S S S S S S S S S B 0 P

Q 0 B S S S S S S S S S S S S S S S S S S B 0 Q

R 0 0 B S S S S S S S S S S S S S S S S B 0 0 R

S 0 0 B S S S S S S S S S S S S S S S S B 0 0 S

T 0 0 0 B S S S S S S S S S S S S S S B 0 0 0 T

U 0 0 0 0 B S S S S S S S S S S S S B 0 0 0 0 U

V 0 0 0 0 0 B B B S S S S S S B B B 0 0 0 0 0 V

W 0 0 0 0 0 0 0 0 B B B B B B 0 0 0 0 0 0 0 0 W

Row\Col 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 Row\Col

Row\Col 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 Row\Col

A 0 0 0 0 0 0 0 0 B B B B B B 0 0 0 0 0 0 0 0 A

B 0 0 0 0 0 B B B S S S S S S B B B 0 0 0 0 0 B

C 0 0 0 0 B S S B S S B B S S B S S B 0 0 0 0 C

D 0 0 0 B S S B B S S B B S S B B S S B 0 0 0 D

E 0 0 B S S B S S B B S S B B S S B S S B 0 0 E

F 0 0 B S B B S S B B S S B B S S B B S B 0 0 F

G 0 B S B S S B B S S B B S S B B S S B S B 0 G

H 0 B B B S S B B S S B B S S B B S S B B B 0 H

J B S S S B B S S B B S S B B S S B B S S S B J

K B S S S B B S S B B S S B B S S B B S S S B K

L B S B B S S B B S S S S S S B B S S B B S B L

M B S B B S S B B S S S S S S B B S S B B S B M

N B S S S B B S S B B S S B B S S B B S S S B N

O B S S S B B S S B B S S B B S S B B S S S B O

P 0 B B B S S B B S S B B S S B B S S B B B 0 P

Q 0 B S B S S B B S S B B S S B B S S B S B 0 Q

R 0 0 B S B B S S B B S S B B S S B B S B 0 0 R

S 0 0 B S S B S S B B S S B B S S B S S B 0 0 S

T 0 0 0 B S S B B S S B B S S B B S S B 0 0 0 T

U 0 0 0 0 B S S B S S B B S S B S S B 0 0 0 0 U

V 0 0 0 0 0 B B B S S S S S S B B B 0 0 0 0 0 V

W 0 0 0 0 0 0 0 0 B B B B B B 0 0 0 0 0 0 0 0 W

Row\Col 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 Row\Col

Homogeneous

Bundle

Heterogeneous,

Mixed Bundle

Checkerboard

Seed- Blanket

Cores

Annular

Seed-Blanket

Cores 0

0.2

0.4

0.6

0.8

1

1.2

1.4

1 2 3 4 5 6 7 8 9

Fiss

ile U

tiliz

atio

n, R

ela

tive

to

Nat

ura

l Ura

niu

m

CA

ND

U

Checkerboard Core Designs

Annular Core Designs

NU

3.8% Pu

96.2% Th

Hafnium Tube

3 mm thick

Zr Rod

Moderator

CT

PT

Potential to Increase Utilization

• Achieve ~ 20% to 100% higher utilization of fissile fuel

than PT-HWR with NU fuel in an OTT cycle.


0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.6

1.8

2.0

2.2

0.000 0.005 0.010 0.015 0.020 0.025 0.030 0.035 0.040 0.045 0.050 0.055 0.060

Fiss

ile

Uti

lizat

ion

Rel

ativ

e to

PT-

HW

R-N

U

Volume Fraction of Initial Fissile in Bundle IHM (LEUO2, PuO2, ThO2)

35-LEU/Th - 8-Th

35-LEU/Th

35 LEU/Th - ZrO2 Rod

21-LEU/Th

21 LEU/Th - ZrO2 Rod

35-Pu/Th-8-Th

35-Pu/Th

35-Pu/Th - ZrO2 Rod

21-Pu/Th

21-Pu/Th - ZrO2 Rod

PT-HWR NU

Thorium in China

Evolutionary Approach with HWRs

• Uranium resources limited.

– Use of Canadian know-how - PT-HWRs.

• Use NUE in CANDU-6 (2012-2014).

– RU~0.9 wt%; DU~ 0.25 wt% U-235/U

– NUE ~ 70% RU + 30% DU.

– Behaves the ~same as NU in CANDU.

• Use RU in dedicated EC6 (by ~2019).

• Thorium-based fuels in EC6 ( 2020).

– Collaborate/cooperate w/ Canada.

– Simple, evolutionary design first, based on

43-element bundle carrier.

– LEU in smaller outer 35 pins.

– Th in larger inner 8 pins.

– Core could be mix of NU, RU and Th-based

bundles.

– Build up inventory of U-233 in spent fuel.

– Recycle in future.


Summary

• Thorium has a great potential benefit for sustainability,

safety, and waste management.

–Goals can be achieved by current commercial reactor designs.

– We don’t need to wait, at least not long.

–PT-HWR’s are operational today, and can be adapted for thorium.

– Small design changes can be implemented quickly.

– More R&D to enable more substantial design changes.

– R&D that will enable practical engineering solutions.

• AECL has 50 years experience in thorium fuel cycles:

–Reactor and fuel design; alternative concepts.

–Fuel fabrication.

– Irradiations + Critical Experiments.

–Reprocessing / Recycling, Waste management.

–Development, Testing & Validation of Analysis Tools.

–Economics and system analyses.


More Info

• Visit:

• http://www.aecl.ca/site3.aspx

• https://canteach.candu.org/Pages/Welcome.aspx

• http://cns-snc.ca/home


Acknowledgements

• Bronwyn Hyland

• Jeremy Pencer

• Holly Hamilton

• Laurence Leung

• CRL Library and Report Centres

• Various staff in

–Fuel Development Branch

–Computational Reactor Physics Branch.

–Applied Physics Branch (ZED-2 Facility)

–Thermal-hydraulics Branch

–Fuel Channel and Fuel Channel Safety Branch


Fast Fission in Fertile Isotopes

• U-238, Th-232


Th-232

U-238

AECL Work on ADS / HFFR for

U-233 Production from Th-232

• 1962-1982 - Design studies, watching briefs, economic assessments.

–Accelerator-Drive Systems (ADS)

– Spallation fast neutron source driving U or Th blanket.

–Hybrid Fusion Fission Reactors (HFFRs)

– 14-MeV D-T fusion neutrons driving U, U/Th and Th blankets.

–Alternative to reactor-based breeders; high support ratio (10:1).

– Complement existing fleet of high converter PT-HWRs


Neutron Production by Spallation

• 1 GeV protons or deuterons on Pb/Bi or U target

–~20 neutrons per proton (Pb), ~ 40 neutrons per proton (U)


Neutron Production in Fissile

Isotopes

• Variation of neutron production per neutron absorted.


Isotope

Thermal Spectrum**

Fast Spectrum***

U-233 2.28 - 2.30 2.31 – 2.42

U-235 2.03 - 2.07 1.93 - 2.17

Pu-239 1.80 - 2.11 2.49 - 2.68

Pu-241 2.14 - 2.15 2.72

Spectrum-Averaged Neutron Production () for Various Fissile Isotopes ** Approximate range of values in a thermal-spectrum reactor.

*** Approximate range of values in fast-spectrum reactor.


Overview

• AECL has over 50 years of extensive experience with

Thoria-based fuels – Investments made in Thoria fuel cycle R&D since the late 1950’s

– First irradiation conducted in 1962 and the most recent in 2005

• Experience includes

– Irradiation in Nuclear Power Demonstration reactor (NPD) and 3

experimental reactors

– Manufacturing fuels with a wide range of compositions and pellet

geometries using both novel and traditional fabrication techniques

– Post Irradiation Examination (PIE) studies

• Knowledge gained from various experiments has been

fed into new experiments

• Results indicate that Thoria fuel has always performed

comparably with UO2 and in some cases demonstrated

superior performance


Fabrication – Fissile Additive

• Generally, AECL has targeted a solid

solution of Thoria and the fissile additive

• Many techniques are capable of achieving

this and they fall into two main categories:

1. Solution blending – sol gel, co-precipitation – Mixing at the atomic level

2. Mechanical mixing – co-milling, high-intensity

mixing – Often not a “perfect” solid solution

– Must achieve mixing on the scale of individual particles


Fabrication – Sol-Gel

• Microspheres from 15 to > 1000 μm


Experiment Summary Table

Irradiation

Facility

# of

experiments

Irradiation

Time frame

NPD 1 1976

NRX 20 1962-1992

NRU 28 1966-2005

WR1 18 1970-1980

Each experiment consisted of a series of irradiations

Other irradiations were done and reported in the literature


Granular Pellet Structure

95% Dense


Thoria Irradiation Experience at AECL

• Thoria irradiations have been ongoing

since early 1960s

• Irradiations in NRX, NRU and WR-1

research reactors as well as NPD a 20 MWe

power reactor

• Pure Thoria, Thoria-UO2 & Thoria-PuO2


Post Irradiation Examination

• PIE conducted for most of the individual irradiation Thoria fuels

• Typical PIE data depends on the nature of testing and program objectives and can include:

– Visual exam

– Element profilometry

– Axial gamma scanning

– Element gas puncture and fission gas analysis

– Burnup analysis

– Sheath metallographic exam

– Fuel pellet ceramographic exam

– α -ß-γ autoradiography


Six Thoria-PuO2 Bundles (1)

• Irradiation performed in NRU

• 36 element Bruce type fuel bundle 86.05

wt% Th and 1.53 wt% Pu in (Th, Pu)O2

• Objectives

–To verify the ability of (Th, Pu)O2 fuel to operate at

significant power outputs to burnups of 42

MWd/kgHE

–To examine the power-ramp performance of

Zircaloy clad (Th, Pu)O2 fuel with ES-242 and

siloxane sheath (CANLUB) coatings

–To determine fission-gas release

–To examine micro-structural changes in the fuel


Six Thoria-PuO2 Bundles (2)

• Maximum sustained powers from 49-75 kW/m

• Burnups to 45 MWd/kgHE (also maximum power bundle)

• Fission products accumulate

in fuel grain boundary which

limit fuel performance

• Low gas release

• Low sheath strain

• Significant % of PuO2

present as agglomerates

Outer element in Bundle ADC-1


Thoria Demonstration Irradiation PIE

• Higher than expected gas release due to

granular structure of pellets (WR1-1007 tests)

Granules


High-Density, Homogeneous Thoria

• 1.5 % U-235,

• 35 MWd/kgHE

• 48 kW/m Max

• Low gas release

• Low sheath strain

• Irradiation is

ongoing


Conclusions

• AECL has over forty-seven years of experience with

Thoria-based fuel irradiations, with burnups up to 47

MWd/kgHE

• AECL has extensive experience with Thoria fuels having a

wide range of fuel compositions and pellet geometries

• Successful fabrication technology has been developed and

proven in-reactor tests

• Thoria fuel has always performed comparably with UO2,

with some experiments demonstrating superior

performance


Conclusions

• Thoria-based fuels can achieve superior

performance characteristics to that of UO2

fuels, provided pellet fabrication technologies

are used to achieve a high quality non-granular

microstructure

Critical Experiments at AECL


Critical Experiments

• AECL has a long history of critical experiments

involving thorium fuels

• Three sets of experiments, dating back to the

1960’s

–HEU/Th

–Pu/Th

–U-233/Th

• Performed in the ZED-2 (Zero Energy Deuterium)

reactor at Chalk River Laboratories


• HEU/Th (1966)

–98.5% ThO2, 1.5% HEU (93% U-235)

–19-element bundles

–7 test channels

• U-233/Th (1991)

–98.6% ThO2, 1.4%UO2, (97.6% U-233)


–7 test channels

• Pu/Th (1986)


–97.8% ThO2, 2.2% PuO2 (1.8% fissile)

–7 test channels



• Requires only 35 bundles

substituted in a reference

lattice compared to about 275

bundles for a critical core

• Can measure void-reactivity

and lattice reactivity for

fuel/coolant temperatures in

the range 25 to 300oC

Substituted Channels

28-Element Reference Lattice

Substitution Experiments


Physics Experiments in ZED-2

• Substitution Experiments – determine fuel properties (buckling/reactivity) when only a limited amount of fuel (typically seven assemblies) is available

• Flux Maps – copper foils are irradiated to measure the flux shape and derive extrapolation distances

• Reaction Rate (Fine Structure) - provide detailed information about neutron distributions (in space and energy) in and around a fuel channel, as well as fission-rate and conversion ratio data within the fuel.

• Used for qualification of reactor physics codes– program ongoing

Conclusions

• AECL has a long history of critical experiments in the

ZED-2 facility

• These are substitution experiments, with 7 channels of

the test fuel

• Wide variety of experiments have been performed

–Different lattice pitches

–Different coolants

–Heated channels, etc

• These experiments are currently being analysed as

part of a program to qualify physics codes for design of

thorium fuel cycles



Homogeneous Thorium Fuel

Cycles in CANDU Reactors

Bronwyn Hyland

Global 2009

September 10, 2009


Overview

• Motivation

• Calculation

• Fuel Design

• Results –Low and high burnup Pu driven once-through

–Low and high burnup Pu driven with U-233 recycle

–Low and high burnup LEU driven once-through

• Conclusions


Thorium Fuel Configurations


Thorium Fuel Cycles

• The simplest implementation of a thorium-based

fuel is in a “homogeneous” thorium fuel cycle.

• The CANDU reactor can efficiently exploit thorium

in a homogeneous thorium fuel cycle (a small

amount of fissile material can go a long way)

• The introduction of U-233 recycle can make a

dramatic improvement in fissile utilization


Calculation

• Lattice cell calculations performed with WIMS-AECL

• 6 cases studied:

Fissile Driver Once-

Through/Recycle

Burnup

(MWd/kg)

Pu Once-Through 20

45

Recycle 20

45

LEU Once-Through 20

45


Calculation

• Models developed to maximize the amount of

energy derived from thorium

• Report results here on:

–Exit burnup

–Fuel temperature coefficient

–Maximum linear element rating

–Percentage of energy derived from thorium

–Distribution of U-233 and Pa-233 in the bundle


Fuel Design

• High burnup and recycle cases the fuel was graded

• Reduce size, increase number of fuel pins to

decrease linear element ratings

• Centre pin of zirconia-filled Hf

Centre

Inner

Intermediate

Outer


Fuel Design

Case Burnup Bundle average Pu

wt% or LEU wt%

Bundle average

U-233 wt%

Pu-driven,

OT

Low 3.5 N/A

High 4.9 N/A

Pu-driven,

Recycle

Low 0.8 1.4

High 2.1 1.4

LEU-driven Low 12.2 N/A

High 14.2 N/A


Results

Case Burnup

(MWd/kg)

FTC

(μk/ºC)

Max. LER

(kW/m)

% Energy

from Th

Pu-driven,

OT

19 -3.8 56 19

45 -5.0 61 29

Pu-driven,

Recycle

20 -7.5 49 78

44 -7.3 59 66

LEU-driven 20 -12.7 51 25

44 -10.7 60 41


Pu-Driven, U-233 Recycle

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.6

1.8

2.0

2.2

0 5 10 15 20 25

Burnup (MWd/kg)

Inner Ring Intermediate Ring

Outer Ring Total

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.6

1.8

0 10 20 30 40 50

Burnup (MWd/kg)

U-2

33 +

Pa-2

33

U-2

33 +

Pa-2

33 +

Th

-232

U-2

33 +

Pa-2

33

U-2

33 +

Pa-2

33 +

Th

-232

20 MWd/kg 45 MWd/kg


Pu-Driven, U-233 Recycle

0

10

20

30

40

50

60

70

80

90

100

0 5 10 15 20 25

Burnup (MWd/kg)

% o

f T

ota

l F

iss

ion

s

Fissions from Pu239 and Pu241

Fissions from Th232, U233, and U238

0

10

20

30

40

50

60

70

80

90

100

0 10 20 30 40

Burnup (MWd/kg)%

of

To

tal F

iss

ion

s

20 MWd/kg 45 MWd/kg


Conclusions

• CANDU reactor can exploit homogeneous thorium fuel cycles

• Low BU Pu-driven case gives the best result for energy from thorium, more energy proportionally required from driver fuel for higher burnup

• For once through cases higher burnup gives higher energy from thorium

• Maximum energy from thorium corresponds to minimum poison in the centre pin

–Results in grading of fissile

–Constrained by LER


On-power

Fuelling

Heavy Water

Moderator –

Good neutron

economy

CANDU fuel channel

Simple fuel bundle

Calandria Tube

Pressure Tube

CANDU Reactor


ZED-2 Reactor

• ZED-2 : Zero Energy Deuterium, successor to ZEEP

• Low-power (200 w), heavy-water moderated reactor

• Tank-type (3.36 meter diameter, 3.35 meter high)

• Peak flux of 1x109 n/cm2/sec

• Designed for CANDU reactor support

• First criticality in September 1960

• Reactor control is via moderator level adjustment

• Primary research activity is support of reactor

physics code development for CANDU reactors


Cross-Section of ZED-2

Top Shield Doors

Moveable Beam

Experimental

Fuel Rods

Heavy Water

Moderator

Side Shield Doors

Graphite Reflector

Air Duct

Hoist

Heavy Water

Dump Tanks

Heavy Water Pump

Aluminum Tank

(Calandria) Shielding Control Room

Dump Valves Filling Valves Drain Valves

These valves control the heavy water level in the

calandria


Moderator Level Control

Beam Chain Fuel Rods Aluminum Calandria

Gap

Graphite Reflector Heavy Water

Shut-Off and Drain Valve

Fill Pump

Dump Valve

Reactor Vessel Approximately to scale 100 cm

Top shields

Dump Tank

(1 of 3)


Typical ZED-2 Fuel Channel

Zircoloy-4

Sheath

Zircoloy-2

Calandria Tube

Zr-2.5%Nb

Pressure Tube

Fuel Calandria Tube

Pressure Tube

Fuel Support

Plate Zr-4

Channel End

Plate Zr-2 Plug (in for

void, out for

cooled)

ZED-2 Calandria floor


Pu-Driven Once-Through

0.0

0.2

0.4

0.6

0.8

1.0

1.2

0 5 10 15 20

Burnup (MWd/kg)

Inner RingIntermediate RingOuter RingTotal

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.6

0 20 40 60Burnup (MWd/kg)

20 MWd/kg 45 MWd/kg

U-2

33 +

Pa-2

33

U-2

33 +

Pa-2

33 +

Th

-232

U-2

33 +

Pa-2

33

U-2

33 +

Pa-2

33 +

Th

-232

Pu-Driven Recycle, FTC


-10

-9

-8

-7

-6

-5

-4

-3

-2

-1

0

0 10 20 30

Fu

el Tem

pera

ture

Coeffic

ient

(μk/º

C)

Burnup (MWd/kg) -8

-7

-6

-5

-4

-3

-2

-1

0

0 20 40 60

Fu

el Tem

pera

ture

Co

effic

ient

(μk/º

C)

Burnup (MWd/kg)

20 MWd/kg 45 MWd/kg


Pu-Driven Once-Through

0

10

20

30

40

50

60

70

80

90

100

0 5 10 15 20

Burnup (MWd/kg)

% o

f T

ota

l F

iss

ion

s

Fissions from Pu239 and Pu241

Fissions from Th232, U233, and U238

0

10

20

30

40

50

60

70

80

90

100

0 10 20 30 40 50

Burnup (MWd/kg)%

of

To

tal F

iss

ion

s

20 MWd/kg 45 MWd/kg


LEU-Driven, Once-Through

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

0 10 20 30

Burnup (MWd/kg)

Inner Ring Intermediate Ring

Outer Ring Total

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.6

0 10 20 30 40 50

Burnup (MWd/kg)

U-2

33 +

Pa-2

33

U-2

33 +

Pa-2

33 +

Th

-232

U-2

33 +

Pa-2

33

U-2

33 +

Pa-2

33 +

Th

-232

20 MWd/kg 45 MWd/kg


LEU-Driven, Once-Through

0

10

20

30

40

50

60

70

80

90

100

0 5 10 15 20 25

Burnup (MWd/kg)

% o

f T

ota

l F

iss

ion

s

Fissions from U235, U238, Pu239, and Pu241

Fissions from U233 and Th232

0

10

20

30

40

50

60

70

80

90

100

0 10 20 30 40 50

Burnup (MWd/kg)

% o

f T

ota

l F

issio

ns

20 MWd/kg 45 MWd/kg

Pu-Driven Recycle, LER


0

10

20

30

40

50

60

0 5 10 15 20 25

Lin

ear

Ele

men

t R

ating (

W/c

m)

Burnup (MWd/kg)

Inner

Intermediate

Outer

20 MWd/kg 45 MWd/kg

0

10

20

30

40

50

60

70

0 10 20 30 40 50 Lin

ear

Ele

men

t R

ating

(W

/cm

)

Burnup (MWd/kg)

Inner

Intermediate

Outer


Other Solution-Based Methods

Solution

impregnation

Sol-gel derived clay

extrusions


Mechanical Mixing

• Wet and dry processes have been evaluated

• Wet processes aid in the dispersion of the different powders amongst each other but require drying of the slurry – danger of residual granules in pellet structure

• Dry processes – due to the cohesive nature of ceramic-grade powders, the degree of homogeneity achieved is related to the intensity of the process used. Dusty, but no drying stage


Mechanical Mixing Methods

Homogenizer (Wet)

Turbula (Dry)

Vibratory Mill (Dry)

Attrition Mill (Wet)

Thorium Fuel Cycles â€“ AECL Experience - media.cns-snc.ca

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