Robert Worth

NGRGINGSM-14 Seattle, USA

Robert WorthNuclear Graphite Research Group

University of Manchester, UK

Lorraine McDermott, Greg Black, Abbie Jones,Paul Mummery, Barry Marsden, Anthony Wickham

Characterisation and Thermal Treatment of Irradiated PGA Graphite with Investigation into 3H and 14C Behaviour

14th International Nuclear Graphite Specialists MeetingSeattle, USA

15th-18th September, 2013


OVERVIEW

Irradiated Graphite in the UK Thermal Treatment at Manchester Future Research Conclusions


IRRADIATED GRAPHITE IN THE UK


UK GRAPHITE LEGACY• Graphite has been used in nuclear power plants

worldwide

• Historically, the UK has constructed many graphite-moderated reactors

• These include power production, plutonium production and research reactors

• Some still operational

• Graphite contributes to a significant UK waste legacy • The majority of this graphite waste is ILW

• Consequently, dismantling and management of radioactive graphite waste is an important issue in the UK


WHY TREAT GRAPHITE?• There is no current disposal route for irradiated

graphite in the UK• Geological Disposal Facility (GDF)?

• Treatment of irradiated graphite could allow reduction in the volume of ILW (cost-saving)

• Utilise GDF space • Allow disposal in current near-surface facilities

• This could be achieved by preferential removal of radioisotopes, such as tritium and carbon-14

• Goal: Maximise radioisotope removal with minimal weight loss


CARBON-14 FORMATION• There are two dominant mechanisms by which 14C is

produced in irradiated graphite in a reactor environment:

(1) 13C (n,γ) 14C(2) 14N (n,p) 14C


Historically difficult to determine nitrogen content of graphite

~10ppm

~50ppm

NITROGEN SENSITIVITY


THERMAL TREATMENT AT MANCHESTER


THERMAL TREATMENT• A program of thermal treatment work has been

conducted at the University of Manchester as part of the collaborative European project ‘CARBOWASTE’

• My own research is a continuation of this thermal treatment research:

• Investigation of dependent variables, including temperature, time and oxygen • Investigation of 14C and 3H behaviour• Comparison of current world data to UK-

irradiated graphite• Optimisation of the process

— Using pre- and post-treatment characterisation techniques


ISOTOPIC INVENTORY DETERMINATION Thermal oxidation has been used as a method for 3H and 14C

determination Graphite samples are placed in a ceramic combustion boat in a

Carbolite® MTT Furnace A suitable cover gas flows past the sample and the temperature is raised A copper oxide catalyst promotes further oxidation of any gasified 3H

and 14C .


ISOTOPIC INVENTORY DETERMINATION

HTO and 14CO2 are subsequently trapped in the bubbler system for analysis using liquid scintillation counting (LSC) Bubblers have a trapping efficiency of 98%


ISOTOPIC INVENTORY DETERMINATION

Typical determined radioisotope content in Oldbury Magnox installed graphite:

Isotope Activity (Bq/g)3H ~3740014C ~63700


ISOTOPIC VALIDATION How do we know we are capturing all of the 3H and 14C? Regular recovery checks are performed – a known quantity of

3H and 14C labelled sucrose standards are put through the furnace 3H recovery in the range of 88 – 98 % 14C recovery in the range of 85 – 94 % LSC quenched standard analysis to ensure LSC efficiency


THERMAL TREATMENT EXPERIMENTAL PROGRAMME

A thermal treatment programme has been designed to determine the effects of time, temperature and oxygen on 3H and 14C release

The following experimental conditions have been applied to samples machined from installed sets retrieved from the Oldbury Magnox power station:

Time

Temperature

4 5 6 7 8

600oC -- 700oC 800oC -- 900oC -- -- -- -- --

Time

Temperature

4 5 6 7 8

600oC 700oC 800oC 900oC -- -- -- -- --

Argon 1% Oxygen in Argon


THERMAL TREATMENT EXPERIMENTAL PROGRAMME

Issues with the integrity of the samples post-treatment:

A B C D

A = 800°C in 1% O2/Ar for 5 hoursB = 700°C in 1% O2/Ar for 5 hoursC = 700°C in Argon gas for 5 hoursD = untreated sample


AUTORADIOGRAPHY

Hotspots


TIME DEPENDENCETritium, 3H Carbon-14, 14C

3 4 5 6 7 8 90

10

20

30

40

50

60

70

80

90

600°C Ar600°C 1% O2/Ar700°C Ar700°C 1% O2/Ar800°C Ar800°C 1% O2/Ar

Time (h)

Estim

ated

3H

Rele

ase

(%)

3 4 5 6 7 8 90

10

20

30

40

50

60

70

80

90

100


Time (h)Es

timat

ed 1

4C R

elea

se (%

)


WEIGHT LOSSTritium, 3H Carbon-14, 14C

0 10 20 30 40 50 60 700

10

20

30

40

50

60

70

80

90


Weight Loss (%)

Estim

ated

3H

Rele

ase

(%)

0 10 20 30 40 50 60 700

10

20

30

40

50

60

70

80

90

100


Weight Loss (%)Es

timat

ed 1

4C R

elea

se (%

)


GAMMA SPECTROMETRY

SamplePre-

Treatment Activity (kBq/g)

Post-Treatment

Activity (kBq/g)

Percent Loss

OM1 8.824 8.473 3.99%

OM14 6.055 5.132 15.26%

OM18 10.093 9.561 5.26%

OM21 6.951 4.689 32.53%

Cobalt-60:


FUTURE RESEARCH


FUTURE WORK - PHD

• Full optimisation of thermal treatment of irradiated Oldbury Magnox reactor graphite with respect to the sensitivity of:

• Goal: Maximise radioisotope removal with minimal sample weight loss

‘Characterisation and Thermal Treatment of Irradiated PGA Graphite with Investigation into 3H and 14C Behaviour’

Temperature 600 - 900oC

Time 3 - 9 hours

Oxygen content of gas 0.5 - 2% oxygen in argon


PRE- & POST- TREATMENT ANALYSIS

Porosity• Helium-pycnometry

Weight Loss• 4 d.p. Balance

Surface Area• Tristar BET• Laser Confocal Microscopy

To try and determine: Amount of weight loss during treatment The typical location of the radioisotopes before removal

Metrology• Digital Micrometer


PRE- & POST- TREATMENT ANALYSIS

To determine: Amount of radioisotope loss during treatment Identification of ‘hotspots’ of radioactivity, which might

influence the results

Radioactive Content• Liquid Scintillation Counting• Gamma-spectrometry• Autoradiography


CONCLUSIONS It has been demonstrated that thermal treatment in an

oxidising atmosphere is a potential means of removing 3H and 14C radioisotopes from irradiated graphite

The current data suggests that this treatment technique may be suitable for removing up to ~80% 3H and ~55% 14C from Oldbury Magnox reactor graphite

Further work will be required to optimise this thermal treatment process and to determine the mobility and origin of these radioisotopes


ACKNOWLEDGMENTS

The authors are pleased to acknowledge EPSRC funding under agreement EP/P113315

A portion of this work was carried out as part of the CARBOWASTE Program: Treatment and Disposal of Irradiated Graphite and Other Carbonaceous Waste, Grant Agreement Number FP7-211333


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Robert Worth